First vapor explosion calculations performed with MC3D thermal-hydraulic code
Energy Technology Data Exchange (ETDEWEB)
Brayer, C.; Berthoud, G. [CEA Centre d`Etudes de Grenoble, 38 (France). Direction des Reacteurs Nucleaires
1998-01-01
This paper presents the first calculations performed with the `explosion` module of the multiphase computer code MC3D, which is devoted to the fine fragmentation and explosion phase of a fuel coolant interaction. A complete description of the physical laws included in this module is given. The fragmentation models, taking into account two fragmentation mechanisms, a thermal one and an hydrodynamic one, are also developed here. Results to some calculations to test the numerical behavior of MC3D and to test the explosion models in 1D or 2D are also presented. (author)
Energy Technology Data Exchange (ETDEWEB)
Berthoud, G.; Crecy, F. de; Meignen, R.; Valette, M. [CEA-G, DRN/DTP/SMTH, 17 rue des Martyrs, 38054 Grenoble Cedex 9 (France)
1998-01-01
The premixing phase of a molten fuel-coolant interaction is studied by the way of mechanistic multidimensional calculation. Beside water and steam, corium droplet flow and continuous corium jet flow are calculated independent. The 4-field MC3D code and a detailed hot jet fragmentation model are presented. MC3D calculations are compared to the FARO L14 experiment results and are found to give satisfactory results; heat transfer and jet fragmentation models are still to be improved to predict better final debris size values. (author)
Directory of Open Access Journals (Sweden)
Renaud Meignen
2010-01-01
Full Text Available In the course of a postulated severe accident in an NPP, Direct Containment Heating (DCH may occur after an eventual failure of the vessel. DCH is related to dynamical, thermal, and chemical phenomena involved by the eventual fine fragmentation and dispersal of the corium melt out of the vessel pit. It may threaten the integrity of the containment by pressurization of its atmosphere. Several simplified modellings have been proposed in the past but they require a very strong fitting which renders any extrapolation regarding geometry, material, and scales rather doubtful. With the development of multidimensional multiphase flow computer codes, it is now possible to investigate the phenomenon numerically with more details. We present an analysis of the potential of the MC3D code to support the analysis of this phenomenon, restricting our discussion to the dynamical processes. The analysis is applied to the case of French 1300 MWe PWR reactors for which we derive a correlation for the corium dispersal rate for application in a Probabilistic Safety Analysis (PSA level 2 study.
Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code
Energy Technology Data Exchange (ETDEWEB)
Thakre, S.; Ma, W. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)
2013-08-15
Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different
MC3D modelling of stratified explosion
International Nuclear Information System (INIS)
It is known that a steam explosion can occur in a stratified geometry and that the observed yields are lower than in the case of explosion in a premixture configuration. However, very few models are available to quantify the amount of melt which can be involved and the pressure peak that can be developed. In the stratified application of the MC3D code, mixing and fragmentation of the melt are explained by the growth of Kelvin Helmholtz instabilities due to the shear flow of the two phase coolant above the melt. Such a model is then used to recalculate the Frost-Ciccarelli tin-water experiment. Pressure peak, speed of propagation, bubble shape and erosion height are well reproduced as well as the influence of the inertial constraint (height of the water pool). (author)
Energy Technology Data Exchange (ETDEWEB)
Berthoud, G.; Crecy, F. de; Duplat, F.; Meignen, R.; Valette, M. [CEA/Grenoble, DRN/DTP, 17 Avenue des Martyrs, 38054 Grenoble Cedex 9 (France)
1998-01-01
This paper presents the <
Sihver, L.; Mancusi, D.; Niita, K.; Sato, T.; Townsend, L.; Farmer, C.; Pinsky, L.; Ferrari, A.; Cerutti, F.; Gomes, I.
Particles and heavy ions are used in various fields of nuclear physics, medical physics, and material science, and their interactions with different media, including human tissue and critical organs, have therefore carefully been investigated both experimentally and theoretically since the 1930s. However, heavy-ion transport includes many complex processes and measurements for all possible systems, including critical organs, would be impractical or too expensive; e.g. direct measurements of dose equivalents to critical organs in humans cannot be performed. A reliable and accurate particle and heavy-ion transport code is therefore an essential tool in the design study of accelerator facilities as well as for other various applications. Recently, new applications have also arisen within transmutation and reactor science, space and medicine, especially radiotherapy, and several accelerator facilities are operating or planned for construction. Accurate knowledge of the physics of interaction of particles and heavy ions is also necessary for estimating radiation damage to equipment used on space vehicles, to calculate the transport of the heavy ions in the galactic cosmic ray (GCR) through the interstellar medium, and the evolution of the heavier elements after the Big Bang. Concerns about the biological effect of space radiation and space dosimetry are increasing rapidly due to the perspective of long-duration astronaut missions, both in relation to the International Space Station and to manned interplanetary missions in near future. Radiation protection studies for crews of international flights at high altitude have also received considerable attention in recent years. There is therefore a need to develop accurate and reliable particle and heavy-ion transport codes. To be able to calculate complex geometries, including production and transport of protons, neutrons, and alpha particles, 3-dimensional transport using Monte Carlo (MC) technique must be used. Today
Very Low Bit-Rate Video Coding Using Motion ompensated 3-D Wavelet Transform
Institute of Scientific and Technical Information of China (English)
无
1999-01-01
A new motion-compensated 3-D wavelet transform (MC-3DWT) video coding scheme is presented in thispaper. The new coding scheme has a good performance in average PSNR, compression ratio and visual quality of re-constructions compared with the existing 3-D wavelet transform (3DWT) coding methods and motion-compensated2-D wavelet transform (MC-WT) coding method. The new MC-3DWT coding scheme is suitable for very low bit-rate video coding.
The reactor dynamics code DYN3D
International Nuclear Information System (INIS)
The article provides an overview on the code DYN3D which is a three-dimensional core model for steady-state, dynamic and depletion calculations in reactor cores with quadratic or hexagonal fuel assembly geometry being developed by the Helmholtz-Zentrum Dresden-Rossendorf for more than 20 years. The current paper gives an overview on the basic DYN3D models and the available code couplings. The verification and validation status is shortly outlined. The paper concludes with the current developments of the DYN3D code. For more detailed information the reader is referred to the publications cited in the corresponding chapters.
On applicability of the 3D nodal code DYN3D for the analysis of SFR cores
International Nuclear Information System (INIS)
DYN3D is an advanced multi-group nodal diffusion code originally developed for the 3D steady-state and transient analysis of the Light Water Reactor (LWR) systems with square and hexagonal fuel assembly geometries. The main objective of this work is to demonstrate the feasibility of using DYN3D for the modeling of Sodium cooled Fast Reactors (SFRs). In this study a prototypic European Sodium Fast Reactor (ESFR) core is simulated by DYN3D using homogenized multi-group cross sections produced with Monte Carlo (MC) reactor physics code Serpent. The results of the full core DYN3D calculations are in a very good agreement with the reference full core Serpent MC solution. (author)
Steady state analysis of SFR cores using DYN3D-Serpent codes sequence
International Nuclear Information System (INIS)
A few-group cross section generation methodology for the deterministic analysis of SFR cores with DYN3D code has been proposed. The full core DYN3D results obtained using the few-group constants produced by Serpent agreed very well with that of the reference full core MC simulations. Such an agreement demonstrates the feasibility of the proposed few-group cross section generation procedure. In summary, this study showed that the Serpent-DYN3D code sequence can be successfully used for modeling fast spectrum reactor systems. (orig.)
Validation of multipoint kinetics model against 3D Trikin Code
International Nuclear Information System (INIS)
Validation of multipoint kinetics formulation for RELAP5 code has been carried out against 3D TRIKIN code. Core behavior of an asymmetric reactivity transient has been simulated through artificial tuning of lattice constants in 3D code. Individual node normalized reactivity has been conserved and power estimates from multipoint model have been compared with 3D simulation. It has been observed that localized peak power estimates from multipoint simulation are on higher side and therefore are conservative in nature. Improvements in multipoint formulation in regards to evolving coupling coefficients and involving more number of nodes can help in improving its accuracy to some extent. (author)
Analysis of the KROTOS KFC test by coupling X-Ray image analysis and MC3D calculations
International Nuclear Information System (INIS)
During a hypothetical severe accident sequence in a Pressurized Water Reactor (PWR), the hot molten materials (corium) issuing from the degraded reactor core may generate a steam explosion if they come in contact with water and may damage the structures and threaten the reactor integrity. The SERENA program is an international OECD project that aims at helping the understanding of this phenomenon also called Fuel Coolant Interaction (FCI) by providing data. CEA takes part in this program by performing tests in its KROTOS facility where steam explosions using prototypic corium can be triggered. Data about the different phases in the premixing are extracted from the KROTOS X-Ray radioscopy images by using KIWI software (KROTOS Image analysis of Water-corium Interaction) currently developed by CEA. The MC3D code, developed by IRSN, is a thermal-hydraulic multiphase code mainly dedicated to FCI studies. It is composed of two applications: premixing and explosion. An overall FCI calculation with MC3D requires a premixing calculation followed by an explosion calculation. The present paper proposes an alternative approach in which all the features of the premixing are extracted from the X-Ray pictures using the KIWI software and transferred to an MC3D dataset for a direct simulation of the explosion. The main hypothesis are discussed as well as the first explosion results obtained with MC3D for the KROTOS KFC test. These results are rather encouraging and are analyzed on the basis of comparisons with the experimental data. (authors)
Analysis of the KROTOS KFC test by coupling X-Ray image analysis and MC3D calculations
Energy Technology Data Exchange (ETDEWEB)
Brayer, C.; Charton, A.; Grishchenko, D.; Fouquart, P.; Bullado, Y.; Compagnon, F.; Correggio, P.; Cassiaut-Louis, N.; Piluso, P. [Commissariat a l' Energie Atomique et Aux Energies Alternatives, CEA Cadarache, DEN, F-13108 Saint-Paul-Les-Durance (France)
2012-07-01
During a hypothetical severe accident sequence in a Pressurized Water Reactor (PWR), the hot molten materials (corium) issuing from the degraded reactor core may generate a steam explosion if they come in contact with water and may damage the structures and threaten the reactor integrity. The SERENA program is an international OECD project that aims at helping the understanding of this phenomenon also called Fuel Coolant Interaction (FCI) by providing data. CEA takes part in this program by performing tests in its KROTOS facility where steam explosions using prototypic corium can be triggered. Data about the different phases in the premixing are extracted from the KROTOS X-Ray radioscopy images by using KIWI software (KROTOS Image analysis of Water-corium Interaction) currently developed by CEA. The MC3D code, developed by IRSN, is a thermal-hydraulic multiphase code mainly dedicated to FCI studies. It is composed of two applications: premixing and explosion. An overall FCI calculation with MC3D requires a premixing calculation followed by an explosion calculation. The present paper proposes an alternative approach in which all the features of the premixing are extracted from the X-Ray pictures using the KIWI software and transferred to an MC3D dataset for a direct simulation of the explosion. The main hypothesis are discussed as well as the first explosion results obtained with MC3D for the KROTOS KFC test. These results are rather encouraging and are analyzed on the basis of comparisons with the experimental data. (authors)
NEBU_3D afast pseudo-3D photoionization code for aspherical planetary nebulae and HII regions
Morisset, C; Peña, M
2005-01-01
We describe a pseudo-3D photoionization code, NEBU_3D and its associated visualization tool, VIS_NEB3D, which are able to easily and rapidly treat a wide variety of nebular geometries, by combining models obtained with a 1D photoionization code. We also present a tool, VELNEB_3D, which can be applied to the results of 1D or 3D photoionization codes to generate emission line profiles, position-velocity maps and 3D maps in any emission line by assuming an arbitrary velocity field. As examples of the capabilities of these new tools, we consider three very different theoretical cases. The first one is a blister HII region, for which we have also constructed a spherical model (the spherical impostor) which has exactly the same Hbeta surface brightness distribution as the blister model and the same ionizing star. The second example shows how complex line profiles can be obtained even with a simple expansion law if the nebula is bipolar and the slit slightly off-center. The third example shows different ways to prod...
Implatation of MC2 computer code
International Nuclear Information System (INIS)
The implantation of MC2 computer code in the CDC system is presented. The MC2 computer code calculates multigroup cross sections for tipical compositions of fast reactors. The multigroup constants are calculated using solutions of PI or BI approximations for determined buckling value as weighting function. (M.C.K.)
Fast neutron fluence calculation benchmark analysis based on 3D MC-SN bidirectional coupling method
International Nuclear Information System (INIS)
The Monte Carlo (MC)-discrete ordinates (SN) bidirectional coupling method is an efficient approach to solve shielding calculation of the large complex nuclear facility. The test calculation was taken by the application of the MC-SN bidirectional coupling method on the shielding calculation of the large PWR nuclear facility. Based on the characteristics of NUREG/CR-6115 PWR benchmark model issued by the NRC, 3D Monte Carlo code was employed to accurately simulate the structure from the core to the thermal shield and the dedicated model of the calculation parts locating in the pressure vessel, while the TORT was used for the calculation from the thermal shield to the second down-comer region. The transform between particle probability distribution of MC and angular flux density of SN was realized by the interface program to achieve the coupling calculation. The calculation results were compared with MCNP and DORT solutions of benchmark report and satisfactory agreements were obtained. The preliminary validity of feasibility by using the method to solve shielding problem of a large complex nuclear device was proved. (authors)
Periodic boundary conditions in a 3D hydro code
Energy Technology Data Exchange (ETDEWEB)
Morgan, D L; Neely, J R; Vantine, H C
1998-09-18
We have modified a 3D hydrodynamics code so that it has the capability to impose periodic boundary conditions on the problem being considered. This capability allows it to treat only a basic symmetry unit of the problem when translational or rotational periodic symmetries are present. The code has been run successfully for two test problems involving rotational symmetries.
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.
Exvessel Explosion Load Calculation by Using MC3D and TROI Experiments
International Nuclear Information System (INIS)
When the molten core material is poured into the water pool, the steam explosion might occur at the reactor severe accident. The steam explosion at the severe accident might occur in the vessel or in the reactor cavity. The former is called invessel explosion and the latter is called exvessel explosion. In the invessel explosion, the reactor vessel is under the relatively high pressure and the water in the lower pressure vessel hemisphere might be nearly saturated. The high pressure and the low subcooled condition are not good environment for strong steam explosions, and the invessel explosion issue was concluded in not damaging the pressure vessel integrity. However, the reactor cavity during the exvessel explosion might not maintain its integrity and this might resulted in the breakage of the reactor components. The impulse per unit area which is a mainly used physical amount for explaining the steam explosion work, could be calculated by a TNT equivalent method or a computational code based upon the conservation equations. The conversion ratio and the melt mass in the mixture are required for the TNT equivalent method: the latter is evaluated for assuming the breach diameter and the triggering time and the former is measured through the smaller scale experiments. The computational code should be verified and validated by comparing the smaller scale experiments. Thus the smaller scale steam explosions are essential to evaluate the steam explosion loads at reactor severe accidents. The steam explosion kinetic energy should be measured for the TNT equivalent method, otherwise time dependent pressure waves for the computational code method during steam explosion experiments. In this paper, the computational code method is adapted to evaluate steam explosion loads, i.e., impulses per unit area. The evaluation of the computational code was done against TROI experiments and the code was adapted to a PWR condition. All this calculations were done by using MC3D code
Interface requirements to couple thermal-hydraulic codes to 3D neutronic codes
Energy Technology Data Exchange (ETDEWEB)
Langenbuch, S.; Austregesilo, H.; Velkov, K. [GRS, Garching (Germany)] [and others
1997-07-01
The present situation of thermalhydraulics codes and 3D neutronics codes is briefly described and general considerations for coupling of these codes are discussed. Two different basic approaches of coupling are identified and their relative advantages and disadvantages are discussed. The implementation of the coupling for 3D neutronics codes in the system ATHLET is presented. Meanwhile, this interface is used for coupling three different 3D neutronics codes.
PB3D: a new code for 3D ideal linear peeling-ballooning stability
Weyens, Toon; Sanchez, Raul; Huijsmans, Guido; Garcia, Luis; Loarte, Alberto
2015-11-01
Ideal peeling-ballooning modes are important for magnetic nuclear fusion devices, but most of the theoretical and computational work that has been performed over the years to gain insight into their inner workings and consequences has been limited to axisymmetric (so-called 2D) cases which limits the range of applicability of the results. For example, the proposed use of perturbation coils in tokamaks to destabilize ELMS before they have a chance to grow dangerous has an inherently non-axisymmetric (3D) nature. Furthermore, many devices, such as stellarators, are intrinsically not axisymmetric. In this contribution we present a new code, PB3D (Peeling-Ballooning in 3D), that implements the equations of a previously developed theory that overcomes these stringent limitations by making no use of an axisymmetric approximation. The first benchmarking results of PB3D, dealing with the investigation of the stability properties of magnetic equilibria with nested flux surfaces obtained numerically from multiple equilibrium codes, such as HELENA and VMEC, are presented here. The results will be compared with those of various axisymmetric stability codes.
International Nuclear Information System (INIS)
The Monte Carlo (MC)-discrete ordinates (SN) coupled method is an efficient approach to solve shielding calculations of nuclear device with complex geometries and deep penetration. The 3D MC-SN coupled method has been used for PWR shielding calculation for the first time. According to characteristics of NUREG/CR-6115 PWR model, the thermal shield is specified as the common surface to link the Monte Carlo complex geometrical model and the deep penetration SN model. 3D Monte Carlo code is employed to accurately simulate the structure from core to thermal shield. The neutron tracks crossing the thermal shield inner surface are recorded by MC code. The SN boundary source is generated by the interface program and used by the 3D SN code to treat the calculation from thermal shield to pressure vessel. The calculation results include the circular distributions of fast neutron flux at pressure vessel inner wall, pressure vessel T/4 and lower weld locations. The calculation results are performed with comparison to MCNP and DORT solutions of benchmark report and satisfactory agreements are obtained. The validity of the method and the correctness of the programs are proved. (authors)
The 3D-SEEP computer code user's manual
International Nuclear Information System (INIS)
This report describes the 3D-SEEP computer code and presents the direction to use the code effectively. 3D-SEEP calculates the saturated-unsaturated time dependent or steady state flow of groundwater in permeable geologic media for the safety evaluation of nuclear waste disposal. 3D-SEEP is based on the 3-dimensional Galerkin finite element method. This allows the modeling of complex geometrical shapes and complicated patterns of geologic media. The flow is modeled by single phase flow governed by Darcy's law, and the simplified double porosity model is introduced to consider fractured media. This code can handle non-uniform flow regions having irregular boundaries and arbitrary degree of local anisotropy. (author)
Local stabilizer codes in 3D without string logical operators
Haah, Jeongwan
2011-01-01
We suggest concrete models for self-correcting quantum memory by reporting examples of local stabilizer codes in 3D that have no string logical operators. Previously known local stabilizer codes in 3D all have string-like logical operators, which make the codes non-self-correcting. We introduce an algebraic definition of "logical string segments" to avoid difficulties in defining one dimensional objects in discrete lattices. We prove that every string-like logical operator of our code can be deformed to a disjoint union of short segments, and each segment is in the stabilizer group. The code has surface-like logical operators whose partial implementation has unsatisfied stabilizers along its boundary.
Validation of OPERA3D PCMI Analysis Code
International Nuclear Information System (INIS)
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
AVS 3D Video Coding Technology and System
Institute of Scientific and Technical Information of China (English)
Siwei Ma; Shiqi Wang; Wen Gao
2012-01-01
Following the success of the audio video standard （AVS） for 2D video coding, in 2008, the China AVS workgroup started developing 3D video （3DV） coding techniques. In this paper, we discuss the background, technical features, and applications of AVS 3DV coding technology. We introduce two core techniques used in AVS 3DV coding： inter-view prediction and enhanced stereo packing coding. We elaborate on these techniques, which are used in the AVS real-time 3DV encoder. An application of the AVS 3DV coding system is presented to show the great practical value of this system. Simulation results show that the advanced techniques used in AVS 3DV coding provide remarkable coding gain compared with techniques used in a simulcast scheme.
Development of visualization software for McCARD code
International Nuclear Information System (INIS)
The McCARD (Monte Carlo Code for Advanced Reactor Design and analysis) is a Monte Carlo neutron transport code using continuous energy nuclear data library. Arbitrary system geometry can be handled by dividing it into three-dimensional unit cells that are defined using surfaces.In order to confirm geometrical modeling of input file of the McCARD code, a 2D visualization program with a 3D modeling has been developed. It requires lots of mathematical operations and advanced technologies for design graphical program of complicated geometries. It also provides a display function for the flux and power of the core. The software is coded with the visual C++ language and run under the Windows PC environment
International Nuclear Information System (INIS)
Highlights: • McCad – software tool developed at KIT for the automatic conversion of CAD models into the geometry representation of Monte Carlo particle transport codes. • Open source software running under the Linux operating system and utilizing Open Cascade CAD kernel with the Qt4 libraries for the graphical user interface (GUI). • Converted geometry models can be output in the syntax of MCNP and TRIPOLI of the Monte Carlo codes. • Related visualization capabilities, based on coupling of McCad with the ParaView software, allow to overlay mesh tally distributions to the CAD geometry. • McCad applied to solve fusion neutronics problems of ITER and the IFMIF neutron source. -- Abstract: The McCad geometry conversion tool has been developed at KIT to enable the automatic conversion of CAD models into the semi-algebraic geometry representation as utilized in Monte Carlo particle transport simulations. McCad is entirely based on open source software, it is running under the Linux operating system and utilizes the Open Cascade CAD kernel with the Qt4 libraries for the graphical user interface (GUI). The converted geometry models can be output in the syntax of the Monte Carlo codes MCNP and TRIPOLI. Related visualization capabilities are based on the coupling of McCad with the ParaView software and allow to overlay mesh tally distributions to the CAD geometry. This enables perspective 3D representations or animations on the CAD geometry. The paper presents the current status of the McCad approach and its implementation, and discusses its capabilities, limitations as well as future development needs. The use of McCad for fusion neutronics applications is illustrated on the examples of the MCNP model generation for ITER Test Blanket Modules (TBM) and the test cell facility of the IFMIF neutron source including Monte Carlo shielding calculations using the converted models
Energy Technology Data Exchange (ETDEWEB)
Große, D.; Fischer, U., E-mail: ulrich.fischer@kit.edu; Kondo, K.; Leichtle, D.; Pereslavtsev, P.; Serikov, A.
2013-10-15
Highlights: • McCad – software tool developed at KIT for the automatic conversion of CAD models into the geometry representation of Monte Carlo particle transport codes. • Open source software running under the Linux operating system and utilizing Open Cascade CAD kernel with the Qt4 libraries for the graphical user interface (GUI). • Converted geometry models can be output in the syntax of MCNP and TRIPOLI of the Monte Carlo codes. • Related visualization capabilities, based on coupling of McCad with the ParaView software, allow to overlay mesh tally distributions to the CAD geometry. • McCad applied to solve fusion neutronics problems of ITER and the IFMIF neutron source. -- Abstract: The McCad geometry conversion tool has been developed at KIT to enable the automatic conversion of CAD models into the semi-algebraic geometry representation as utilized in Monte Carlo particle transport simulations. McCad is entirely based on open source software, it is running under the Linux operating system and utilizes the Open Cascade CAD kernel with the Qt4 libraries for the graphical user interface (GUI). The converted geometry models can be output in the syntax of the Monte Carlo codes MCNP and TRIPOLI. Related visualization capabilities are based on the coupling of McCad with the ParaView software and allow to overlay mesh tally distributions to the CAD geometry. This enables perspective 3D representations or animations on the CAD geometry. The paper presents the current status of the McCad approach and its implementation, and discusses its capabilities, limitations as well as future development needs. The use of McCad for fusion neutronics applications is illustrated on the examples of the MCNP model generation for ITER Test Blanket Modules (TBM) and the test cell facility of the IFMIF neutron source including Monte Carlo shielding calculations using the converted models.
SNAP-3D: a three-dimensional neutron diffusion code
International Nuclear Information System (INIS)
A preliminary report is presented describing the data requirements of a one- two- or three-dimensional multi-group diffusion code, SNAP-3D. This code is primarily intended for neutron diffusion calculations but it can also carry out gamma calculations if the diffuse approximation is accurate enough. It is suitable for fast and thermal reactor core calculations and for shield calculations. It is assumed the reader is familiar with the older, two-dimensional code SNAP and can refer to the report [TRG-Report-1990], describing it. The present report concentrates on the enhancements to SNAP that have been made to produce the three-dimensional version, SNAP-3D, and is intended to act a a guide on data preparation until a single, comprehensive report can be published. (author)
The coupled code TRAB-3D-SMABRE for 3D transient and accident analyses
International Nuclear Information System (INIS)
The three-dimensional TRAB-3D core dynamics code is being internally coupled to the thermal hydraulics system code SMABRE. The codes have previously been coupled with a parallel coupling scheme. VTT's reactor dynamics codes have performed well in all the situations that they have originally been designed for. The most important limitation of the present code models is their inability to handle coolant flow reversal in the core channel, a phenomenon that can be encountered in e.g. BWR ATWS cases or VVER power excursions. The new coupling of the two codes is realized on the level of each node of each channel in the core, with each fuel bundle described with its own channel. Necessary interfaces have been created, an improved version of SMABRE's thermal hydraulics solution method developed, and a steady state procedure developed. A satisfactorily working steady state solution has been achieved. The next step in the development will be testing of the transient calculation. Besides solving the flow reversal limitation of the present dynamics models, a successful coupling will allow expanding into more realistic modelling of an open core. (orig.)
Analysis of ex-vessel steam explosion with MC3D
International Nuclear Information System (INIS)
An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In the paper, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which was developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed varying the location of the melt release (central, left and right side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. The performed analysis shows that for some ex-vessel steam explosion scenarios significantly higher pressure loads are predicted than obtained in the OECD programme SERENA Phase 1. (author)
FARGO3D: A new GPU-oriented MHD code
Benítez-Llambay, Pablo
2016-01-01
We present the FARGO3D code, recently publicly released. It is a magnetohydrodynamics code developed with special emphasis on protoplanetary disks physics and planet-disk interactions, and parallelized with MPI. The hydrodynamics algorithms are based on finite difference upwind, dimensionally split methods. The magnetohydrodynamics algorithms consist of the constrained transport method to preserve the divergence-free property of the magnetic field to machine accuracy, coupled to a method of characteristics for the evaluation of electromotive forces and Lorentz forces. Orbital advection is implemented, and an N-body solver is included to simulate planets or stars interacting with the gas. We present our implementation in detail and present a number of widely known tests for comparison purposes. One strength of FARGO3D is that it can run on both "Graphical Processing Units" (GPUs) or "Central Processing unit" (CPUs), achieving large speed up with respect to CPU cores. We describe our implementation choices, whi...
Streamlining of the RELAP5-3D Code
Energy Technology Data Exchange (ETDEWEB)
Mesina, George L; Hykes, Joshua; Guillen, Donna Post
2007-11-01
RELAP5-3D is widely used by the nuclear community to simulate general thermal hydraulic systems and has proven to be so versatile that the spectrum of transient two-phase problems that can be analyzed has increased substantially over time. To accommodate the many new types of problems that are analyzed by RELAP5-3D, both the physics and numerical methods of the code have been continuously improved. In the area of computational methods and mathematical techniques, many upgrades and improvements have been made decrease code run time and increase solution accuracy. These include vectorization, parallelization, use of improved equation solvers for thermal hydraulics and neutron kinetics, and incorporation of improved library utilities. In the area of applied nuclear engineering, expanded capabilities include boron and level tracking models, radiation/conduction enclosure model, feedwater heater and compressor components, fluids and corresponding correlations for modeling Generation IV reactor designs, and coupling to computational fluid dynamics solvers. Ongoing and proposed future developments include improvements to the two-phase pump model, conversion to FORTRAN 90, and coupling to more computer programs. This paper summarizes the general improvements made to RELAP5-3D, with an emphasis on streamlining the code infrastructure for improved maintenance and development. With all these past, present and planned developments, it is necessary to modify the code infrastructure to incorporate modifications in a consistent and maintainable manner. Modifying a complex code such as RELAP5-3D to incorporate new models, upgrade numerics, and optimize existing code becomes more difficult as the code grows larger. The difficulty of this as well as the chance of introducing errors is significantly reduced when the code is structured. To streamline the code into a structured program, a commercial restructuring tool, FOR_STRUCT, was applied to the RELAP5-3D source files. The
SCRAM calculations with the KIKO3D code
International Nuclear Information System (INIS)
Discrepancies between calculated static reactivities and measured reactivities evaluated with reactivity meters led to investigating SCRAM with the KIKO3D dynamic code. The spatial effects near to the position of ionisation chambers were studied. As was expected, reactivities calculated from the flux curves of different nodes using inverse point kinetics are in a wide range. The dynamic and static reactivities in case of asymmetric SCRAM differ considerably as a result of the slow flux shape redistribution. The effect of source neutrons from spontaneous fission and the node-wise delayed neutron fraction on the results is also presented. (Authors)
Code portability and data management considerations in the SAS3D LMFBR accident-analysis code
International Nuclear Information System (INIS)
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
FARGO3D: A New GPU-oriented MHD Code
Benítez-Llambay, Pablo; Masset, Frédéric S.
2016-03-01
We present the FARGO3D code, recently publicly released. It is a magnetohydrodynamics code developed with special emphasis on the physics of protoplanetary disks and planet-disk interactions, and parallelized with MPI. The hydrodynamics algorithms are based on finite-difference upwind, dimensionally split methods. The magnetohydrodynamics algorithms consist of the constrained transport method to preserve the divergence-free property of the magnetic field to machine accuracy, coupled to a method of characteristics for the evaluation of electromotive forces and Lorentz forces. Orbital advection is implemented, and an N-body solver is included to simulate planets or stars interacting with the gas. We present our implementation in detail and present a number of widely known tests for comparison purposes. One strength of FARGO3D is that it can run on either graphical processing units (GPUs) or central processing units (CPUs), achieving large speed-up with respect to CPU cores. We describe our implementation choices, which allow a user with no prior knowledge of GPU programming to develop new routines for CPUs, and have them translated automatically for GPUs.
Solute transport benchmark studies for TRACR3D code verification
International Nuclear Information System (INIS)
A three-dimensional code called TRACR3D, which is applicable to solute transport in both unsaturated and saturated media, is being used to model hypothetical transport of radioactive and nonradioactive constituents from calcined high-level radioactive waste (HLW) at the Idaho Chemical Processing Plant (ICPP). The modeling studies are part of a documentation process which will be required for evaluation of onsite disposal in a near-surface facility as a possible alternative strategy for the long-term management of ICPP HLW. This report discusses the results of a benchmark study for code verification. The problems modeled were: (1) A one-dimensional problem involving the transport of the pertechnetate ion (TcO4-) through a 5-cm diameter by 30-cm-long soil column at ICPP. (2) A one-dimensional problem involving the transport of the iodide ion (I-) through a large caisson (3-m diameter by 6-m depth) at LANL. (3) A three-dimensional problem involving the transport of radioactive ruthenium (Ru-106) from a single-shell tank leak into the vadose zone at the Hanford site. For the three benchmark studies performed, it was concluded that the predicted results from TRACR3D were in agreement with documented and reported solute transport problems, that the input data files were properly configured, and that the code correctly performed the mathematical operations specified in the numerical models. These results will provide a greater degree of confidence in results obtained for planned modeling studies at ICPP. 6 refs., 8 figs., 2 tabs
Prototype coupling of the CFD code ANSYS CFX with the 3D neutron kinetic core model DYN3D
International Nuclear Information System (INIS)
Analyses of postulated reactivity initiated accidents in nuclear reactors are carried out using 3D neutron kinetic core models. The feedback is usually calculated using 1D thermal hydraulic models for channel flow, partly with the possibility of cross flow between these channels. A different possibility is the use of subchannel codes for the determination of the feedback. The code DYN3D developed at Forschungszentrum Dresden-Rossendorf is an example for a 3D neutron kinetic core model. In its basic version, the code contains models for the solution of the 3D neutron diffusion equation in two energy groups for fuel assemblies with rectangular and hexagonal cross section. Recently the code was extended to an arbitrary number of energy groups. Further, a simplified transport approximation for the flux calculation was implemented for fuel assemblies with quadratic cross section. The CFD code ANSYS CFX is the reference CFD code of the German CFD Network in Nuclear Reactor Safety. One of the goals of the co-operation inside this network is the development of CFD software for the simulation of multi-dimensional flows in reactor cooling systems. This includes the coupling of the CFD code ANSYS CFX with the 3D neutron kinetic core model DYN3D. (orig.)
Quasi-3d aerodynamic code for analyzing dynamic flap response
DEFF Research Database (Denmark)
Ramos García, Néstor
is modeled using a panel method whereas the viscous part is modeled by using the integral form of the the laminar and turbulent boundary layer equations and with extensions for 3-D rotational effects. Laminar to turbulent transition can be forced with a boundary layer trip or computed with a modified...... 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......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...
3D CFD CONV code: validation and verification
International Nuclear Information System (INIS)
During some years in IBRAE a set of 3D CFD modules (CONV code) for safety analysis of the operated Nuclear Power Plants (NPPs) is developing. These modules are based on the developed algorithms with small scheme diffusion, for which the discrete approximations are constructed with use of finite-volume methods and fully staggered grids. For solving of convection problem the regularized nonlinear monotonic operator-splitting scheme is developed. The Richardson iterative method with Chebyshev's set of parameters using FFT solver for Laplace's operator as pre-conditioner is applied for solving pressure equation. Such approach for solving of the elliptical equations with variable coefficients gives multiple acceleration in a comparison with a usual method of conjugate gradients. For modeling of 3D turbulent single-phase flows LES approach (commutative filters) is used. The CONV code is fully parallelized and highly effective at the high performance computers. The developed modules were validated on a series of the well known tests in a wide range of Rayleigh numbers from a range 106-1016 and Reynolds numbers from a range 103-105. The developed software has been applied to the simulation of the experiment on RASPLAV facility and of large-scale RCW test conducted in the frames of MASCA Project. As a result of numerical modeling of aforementioned experiments qualitative and quantitative agreement with experimental data was obtained including amount of the molten corium and form of the molten pool, distribution of temperature in corium, fluxes and temperatures in a test-wall. The software has been applied also to the analysis results of test L1 and joint analyses on transient molten pool thermal hydraulics in the LIVE facility in the framework of ISTC project. In this paper the examples of use of the developed software for modeling of a fuel assembly, namely, for research of a hydraulic resistance factor of a spacer are demonstrated. The calculations are carried out on a
International Nuclear Information System (INIS)
In conformity with the protocol of the Workshop under Contract open-quotes Assessment of RBMK reactor safety using modern Western Codesclose quotes VNIIEF performed a neutronics computation series to compare western and VNIIEF codes and assess whether VNIIEF codes are suitable for RBMK type reactor safety assessment computation. The work was carried out in close collaboration with M.I. Rozhdestvensky and L.M. Podlazov, NIKIET employees. The effort involved: (1) cell computations with the WIMS, EKRAN codes (improved modification of the LOMA code) and the S-90 code (VNIIEF Monte Carlo). Cell, polycell, burnup computation; (2) 3D computation of static states with the KORAT-3D and NEU codes and comparison with results of computation with the NESTLE code (USA). The computations were performed in the geometry and using the neutron constants presented by the American party; (3) 3D computation of neutron kinetics with the KORAT-3D and NEU codes. These computations were performed in two formulations, both being developed in collaboration with NIKIET. Formulation of the first problem maximally possibly agrees with one of NESTLE problems and imitates gas bubble travel through a core. The second problem is a model of the RBMK as a whole with imitation of control and protection system controls (CPS) movement in a core
PIXIE3D: An efficient, fully implicit, parallel, 3D extended MHD code for fusion plasma modeling
International Nuclear Information System (INIS)
PIXIE3D is a modern, parallel, state-of-the-art extended MHD code that employs fully implicit methods for efficiency and accuracy. It features a general geometry formulation, and is therefore suitable for the study of many magnetic fusion configurations of interest. PIXIE3D advances the state of the art in extended MHD modeling in two fundamental ways. Firstly, it employs a novel conservative finite volume scheme which is remarkably robust and stable, and demands very small physical and/or numerical dissipation. This is a fundamental requirement when one wants to study fusion plasmas with realistic conductivities. Secondly, PIXIE3D features fully-implicit time stepping, employing Newton-Krylov methods for inverting the associated nonlinear systems. These methods have been shown to be scalable and efficient when preconditioned properly. Novel preconditioned ideas (so-called physics based), which were prototypes in the context of reduced MHD, have been adapted for 3D primitive-variable resistive MHD in PIXIE3D, and are currently being extended to Hall MHD. PIXIE3D is fully parallel, employing PETSc for parallelism. PIXIE3D has been thoroughly benchmarked against linear theory and against other available extended MHD codes on nonlinear test problems (such as the GEM reconnection challenge). We are currently in the process of extending such comparisons to fusion-relevant problems in realistic geometries. In this talk, we will describe both the spatial discretization approach and the preconditioning strategy employed for extended MHD in PIXIE3D. We will report on recent benchmarking studies between PIXIE3D and other 3D extended MHD codes, and will demonstrate its usefulness in a variety of fusion-relevant configurations such as Tokamaks and Reversed Field Pinches. (Author)
McIDAS-V: Advanced Visualization for 3D Remote Sensing Data
Rink, T.; Achtor, T. H.
2010-12-01
McIDAS-V is a Java-based, open-source, freely available software package for analysis and visualization of geophysical data. Its advanced capabilities provide very interactive 4-D displays, including 3D volumetric rendering and fast sub-manifold slicing, linked to an abstract mathematical data model with built-in metadata for units, coordinate system transforms and sampling topology. A Jython interface provides user defined analysis and computation in terms of the internal data model. These powerful capabilities to integrate data, analysis and visualization are being applied to hyper-spectral sounding retrievals, eg. AIRS and IASI, of moisture and cloud density to interrogate and analyze their 3D structure, as well as, validate with instruments such as CALIPSO, CloudSat and MODIS. The object oriented framework design allows for specialized extensions for novel displays and new sources of data. Community defined CF-conventions for gridded data are understood by the software, and can be immediately imported into the application. This presentation will show examples how McIDAS-V is used in 3-dimensional data analysis, display and evaluation.
Coupling of the advanced thermohydraulic code ATHLET with the 3D-core model DYN3D
International Nuclear Information System (INIS)
Two strategies of coupling are described: (i) the use of only the neutron kinetic part of DYN3D integrated into the heat transfer and heat conduction model of ATHLET; (ii) complete modeling of the core by DYN3D. Implementation of the coupling is described and the advantages and disadvantages of the two ways of coupling are discussed. Test calculations were carried out for both versions of the coupled codes and compared with pure ATHLET calculations. After validation the code complex will be a powerful instrument for safety analyses of WWER type reactors. (J.B.) 2 figs., 6 refs
International Nuclear Information System (INIS)
This paper reports on the recent model additions to the 3D field code GASFLOW and on validation and application analyses for steam/hydrogen transport with inclusion of mitigation measures. The results of the 3D field simulation of the HDR test E11.2 are summarized. Results from scoping analyses that simulate different modes of CO2 inertization for conditions from the HDR test T31.5 are presented. The last part discusses different ways of recombiner modeling during 3D distribution simulations and gives the results from validation calculations for the HDR recombiner test E11.8.1 and the Battelle test MC3. The results demonstrate that field code simulations with computer codes like GASFLOW are feasible today for complex containment geometries and that they are necessary for a reliable prediction of hydrogen/steam distribution and mitigation effects. (author)
International Nuclear Information System (INIS)
The computational fluid dynamics code ANSYS CFX has been coupled with the neutron-kinetic core model DYN3D. ANSYS CFX calculates the fluid dynamics and related transport phenomena in the reactors coolant and provides the corresponding data to DYN3D. In the fluid flow simulation of the coolant, the core itself is modeled within the porous body approach. DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the prototype that is currently available, the coupling is restricted to single-phase flow problems. In the time domain an explicit coupling of the codes has been implemented so far. Steady-state and transient verification calculations for two small-size test problems confirm the correctness of the implementation of the prototype coupling. The first test problem was a mini-core consisting of nine real-size fuel assemblies with quadratic cross section. Comparison was performed with the DYN3D stand-alone code. In the steady state, the effective multiplication factor obtained by the DYN3D/ANSYS CFX codes hows a deviation of 9.8 pcm from the DYN3D stand-alone solution. This difference can be attributed to the use of different water property packages in the two codes. The transient test case simulated the withdrawal of the control rod from the central fuel assembly at hot zero power in the same mini-core. Power increase during the introduction of positive reactivity and power reduction due to fuel temperature increase are calculated in the same manner by the coupled and the stand-alone codes. The maximum values reached during the power rise differ by about 1 MW at a power level of 50 MW. Beside the different water property packages, these differences are caused by the use of different flow solvers. The same calculations were carried for a mini-core with seven real-size fuel assemblies with hexagonal cross section in
The progress of the RELAPS/DYN3D coupled code development
International Nuclear Information System (INIS)
The paper describes situation about continuing process of the development of the coupled code RELAP5/DYN3D. The basic structure of coupled code is divided into three parts - RELDYN, sDYN3D, sRELAP5. RELDYN is general interface for exchange of thermohydraulic, kinetic data, current time and current time step of calculation. The thermohydraulic code RELAP5/MOD3 and the kinetic code DYN3D/H1.1 were adjusted to subroutines (Authors)
Comparison: RELAP5-3D systems analysis code and fluent CFD code momentum equation formulations
International Nuclear Information System (INIS)
Recently the Idaho National Engineering and Environmental Laboratory (INEEL), in conjunction with Fluent Corporation, have developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, two- or three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D. Fluent and RELAP5-3D have strengths that complement one another. CFD codes, such as Fluent, are commonly used to analyze the flow behavior in regions of a system where complex flow patterns are expected or present. On the other hand, RELAP5-3D was developed to analyze the behavior of two-phase systems that could be modeled in one-dimension. Empirical relationships were used where first-principle physics were not well developed. Both Fluent and RELAP5-3D are exemplary in their areas of specialization. The differences between Fluent and RELAP5 fundamentally stem from their field equations. This study focuses on the differences between the momentum equation representations in the two codes (the continuity equation formulations are equivalent for single phase flow). First the differences between the momentum equations are summarized. Next the effect of the differences in the momentum equations are examined by comparing the results obtained using both codes to study the same problem, i.e., fully-developed turbulent pipe flow. Finally, conclusions regarding the significance of the differences are given. (author)
3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors
Energy Technology Data Exchange (ETDEWEB)
Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others
1997-07-01
This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.
The gradual development steps of the external coupled RELAP5-DYN3D code
International Nuclear Information System (INIS)
This paper describes the on-going and finished parts of project: he external coupled RELAP5 DYN3D code. The RELAP5 is thermo-hydraulics code used for analysis of the thermohydraulics problems the nuclear facilities. The DYN3D is three-dimensional dynamic code used to calculate the dynamics processes the nuclear core. (author)
Depth-based Multi-View 3D Video Coding
DEFF Research Database (Denmark)
Zamarin, Marco
improved, both in terms of objective and visual evaluations. Depth coding based on standard H.264/AVC is explored for multi-view plus depth image coding. A single depth map is used to disparity-compensate multiple views and allow more efficient coding than H.264 MVC at low bit rates. Lossless coding of...... number of standard solutions for lossless coding. New approaches for distributed video-plus-depth coding are also presented in this thesis. Motion correlation between the two signals is exploited at the decoder side to improve the performance of the side information generation algorithm. In addition...... on edge-preserving solutions. In a proposed scheme, texture-depth correlation is exploited to predict surface shapes in the depth signal. In this way depth coding performance can be improved in terms of both compression gain and edge-preservation. Another solution proposes a new intra coding mode...
Analyses of the OECD MSLB Benchmark with the Codes DYN3D and DYN3D/ATHLET
International Nuclear Information System (INIS)
The code DYN3D coupled with ATHLET was used for the analysis of the OECD Main Steam Line Break (MSLB) Benchmark, which is based on real plant design and operational data of the TMI-1 pressurized water reactor. ATHLET is a thermal hydraulic system code with point or one-dimensional neutron kinetic models. DYN3D consists of three-dimensional nodal kinetic models and a thermal hydraulic part with parallel coolant channels of the reactor core. The following conclusions were reached: (1) If the number of coolant channels for the simulation of the reactor core is chosen too small, the maximum values of local parameters as fuel temperatures are not conservative. (2) If coolant mixing in the reactor vessel is modeled, the accident consequences are mitigated
Coupled cavity 3-D codes for linac tolerance simulations
International Nuclear Information System (INIS)
Three developing codes, based on previous work at Los Alamos (1) are now being used in the Fermilab Linac Upgrade project, to survey system tolerance requirements. Both linear and non-linear beam dynamics of a pi-mode side coupled 805 MHz linac are simulated. Particular attention to the z-plane synchronism has been integrated into the code group, so that klystron drive boundaries can be monitored in the dynamics. All system length errors that contribute to desynchronization, and other data set failures are recognized by the particle PIC code (CAVDYN) to enhance design coherence. Some current results of the CAVDYN code simulations will be discussed
Interlayer Simplified Depth Coding for Quality Scalability on 3D High Efficiency Video Coding
Directory of Open Access Journals (Sweden)
Mengmeng Zhang
2014-01-01
Full Text Available A quality scalable extension design is proposed for the upcoming 3D video on the emerging standard for High Efficiency Video Coding (HEVC. A novel interlayer simplified depth coding (SDC prediction tool is added to reduce the amount of bits for depth maps representation by exploiting the correlation between coding layers. To further improve the coding performance, the coded prediction quadtree and texture data from corresponding SDC-coded blocks in the base layer can be used in interlayer simplified depth coding. In the proposed design, the multiloop decoder solution is also extended into the proposed scalable scenario for texture views and depth maps, and will be achieved by the interlayer texture prediction method. The experimental results indicate that the average Bjøntegaard Delta bitrate decrease of 54.4% can be gained in interlayer simplified depth coding prediction tool on multiloop decoder solution compared with simulcast. Consequently, significant rate savings confirm that the proposed method achieves better performance.
Quasi-3d aerodynamic code for analyzing dynamic flap response
Ramos García, Néstor; Sørensen, Jens Nørkær; Shen, Wen Zhong
2011-01-01
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 is modeled using a panel method whereas the viscous part is modeled by using the integral form of the the laminar and turbulent boundary layer equations and with extensions for 3-D rotational effect...
Modeling of SFR cores with Serpent–DYN3D codes sequence
International Nuclear Information System (INIS)
Highlights: ► Serpent–DYN3D sequence was used for the analysis of an SFR core. ► Homogenized cross sections were generated using Monte-Carlo code Serpent. ► The full core analysis was performed with the nodal diffusion code DYN3D. ► The DYN3D results were compared with those of ERANOS and full core Monte-Carlo solution. - Abstract: DYN3D reactor dynamics nodal diffusion code was originally developed for the analysis of Light Water Reactors. In this paper, we demonstrate the feasibility of using DYN3D for modeling of fast spectrum reactors. A homogenized cross sections data library was generated using continuous energy Monte-Carlo code Serpent which provides significant modeling flexibility compared with traditional deterministic lattice transport codes and tolerable execution time. A representative sodium cooled fast reactor core was modeled with the Serpent–DYN3D code sequence and the results were compared with those produced by ERANOS code and with a 3D full core Monte-Carlo solution. Very good agreement between the codes was observed for the core integral parameters and power distribution suggesting that the DYN3D code with cross section library generated using Serpent can be reliably used for the analysis of fast reactors
The gradual development steps of the external coupled RELAP5 - DYN3D code
International Nuclear Information System (INIS)
This paper describes the on-going and finished parts of project: 'The external coupled RELAP5-DYN3D code'. The development progress was divided into four steps. In present time, second and third steps are performed and four step is started. The two parameters of coolant was selected and are exchanged between codes RELAP5 and DYN3D. (authors)
On Analyzing LDPC Codes over Multiantenna MC-CDMA System
Directory of Open Access Journals (Sweden)
S. Suresh Kumar
2014-01-01
Full Text Available Multiantenna multicarrier code-division multiple access (MC-CDMA technique has been attracting much attention for designing future broadband wireless systems. In addition, low-density parity-check (LDPC code, a promising near-optimal error correction code, is also being widely considered in next generation communication systems. In this paper, we propose a simple method to construct a regular quasicyclic low-density parity-check (QC-LDPC code to improve the transmission performance over the precoded MC-CDMA system with limited feedback. Simulation results show that the coding gain of the proposed QC-LDPC codes is larger than that of the Reed-Solomon codes, and the performance of the multiantenna MC-CDMA system can be greatly improved by these QC-LDPC codes when the data rate is high.
3D unstructured-mesh radiation transport codes
Energy Technology Data Exchange (ETDEWEB)
Morel, J. [Los Alamos National Lab., NM (United States)
1997-12-31
Three unstructured-mesh radiation transport codes are currently being developed at Los Alamos National Laboratory. The first code is ATTILA, which uses an unstructured tetrahedral mesh in conjunction with standard Sn (discrete-ordinates) angular discretization, standard multigroup energy discretization, and linear-discontinuous spatial differencing. ATTILA solves the standard first-order form of the transport equation using source iteration in conjunction with diffusion-synthetic acceleration of the within-group source iterations. DANTE is designed to run primarily on workstations. The second code is DANTE, which uses a hybrid finite-element mesh consisting of arbitrary combinations of hexahedra, wedges, pyramids, and tetrahedra. DANTE solves several second-order self-adjoint forms of the transport equation including the even-parity equation, the odd-parity equation, and a new equation called the self-adjoint angular flux equation. DANTE also offers three angular discretization options: $S{_}n$ (discrete-ordinates), $P{_}n$ (spherical harmonics), and $SP{_}n$ (simplified spherical harmonics). DANTE is designed to run primarily on massively parallel message-passing machines, such as the ASCI-Blue machines at LANL and LLNL. The third code is PERICLES, which uses the same hybrid finite-element mesh as DANTE, but solves the standard first-order form of the transport equation rather than a second-order self-adjoint form. DANTE uses a standard $S{_}n$ discretization in angle in conjunction with trilinear-discontinuous spatial differencing, and diffusion-synthetic acceleration of the within-group source iterations. PERICLES was initially designed to run on workstations, but a version for massively parallel message-passing machines will be built. The three codes will be described in detail and computational results will be presented.
Numerical modelling of gravel unconstrained flow experiments with the DAN3D and RASH3D codes
Sauthier, Claire; Pirulli, Marina; Pisani, Gabriele; Scavia, Claudio; Labiouse, Vincent
2015-12-01
Landslide continuum dynamic models have improved considerably in the last years, but a consensus on the best method of calibrating the input resistance parameter values for predictive analyses has not yet emerged. In the present paper, numerical simulations of a series of laboratory experiments performed at the Laboratory for Rock Mechanics of the EPF Lausanne were undertaken with the RASH3D and DAN3D numerical codes. They aimed at analysing the possibility to use calibrated ranges of parameters (1) in a code different from that they were obtained from and (2) to simulate potential-events made of a material with the same characteristics as back-analysed past-events, but involving a different volume and propagation path. For this purpose, one of the four benchmark laboratory tests was used as past-event to calibrate the dynamic basal friction angle assuming a Coulomb-type behaviour of the sliding mass, and this back-analysed value was then used to simulate the three other experiments, assumed as potential-events. The computational findings show good correspondence with experimental results in terms of characteristics of the final deposits (i.e., runout, length and width). Furthermore, the obtained best fit values of the dynamic basal friction angle for the two codes turn out to be close to each other and within the range of values measured with pseudo-dynamic tilting tests.
ERSN-OpenMC, a Java-based GUI for OpenMC Monte Carlo code
Directory of Open Access Journals (Sweden)
Jaafar EL Bakkali
2016-07-01
Full Text Available OpenMC is a new Monte Carlo transport particle simulation code focused on solving two types of neutronic problems mainly the k-eigenvalue criticality fission source problems and external fixed fission source problems. OpenMC does not have any Graphical User Interface and the creation of one is provided by our java-based application named ERSN-OpenMC. The main feature of this application is to provide to the users an easy-to-use and flexible graphical interface to build better and faster simulations, with less effort and great reliability. Additionally, this graphical tool was developed with several features, as the ability to automate the building process of OpenMC code and related libraries as well as the users are given the freedom to customize their installation of this Monte Carlo code. A full description of the ERSN-OpenMC application is presented in this paper.
FISH: A 3D parallel MHD code for astrophysical applications
Kaeppeli, R; Scheidegger, S; Pen, U -L; Liebendörfer, M
2009-01-01
FISH is a fast and simple ideal magneto-hydrodynamics code that scales to ~10 000 processes for a Cartesian computational domain of ~1000^3 cells. The simplicity of FISH has been achieved by the rigorous application of the operator splitting technique, while second order accuracy is maintained by the symmetric ordering of the operators. Between directional sweeps, the three-dimensional data is rotated in memory so that the sweep is always performed in a cache-efficient way along the direction of contiguous memory. Hence, the code only requires a one-dimensional description of the conservation equations to be solved. This approach also enable an elegant novel parallelisation of the code that is based on persistent communications with MPI for cubic domain decomposition on machines with distributed memory. This scheme is then combined with an additional OpenMP parallelisation of different sweeps that can take advantage of clusters of shared memory. We document the detailed implementation of a second order TVD ad...
Current status of the WHAMS-3D code
International Nuclear Information System (INIS)
The program WHAMS-3D is an explicit time integration program which employs a finite element format, so that it possesses considerable versatility in modeling complex shapes and boundary conditions. The element library consists of the following: Quadrilateral and triangular plate-shell elements, a beam element, a spring element and a hexahedral continuum element. In addition, a rigid linkage is included which permits the efficient modeling of very stiff portions of a structure, such as the bottom ring of a core barrel. In a rigid linkage, the motion of a master node defines the motion of all slave nodes linked to the master node. This option is also useful for eccentrically connected elements where the midlines of the connected elements do not coincide, as for example, in stiffeners. Time integration is performed by the central difference method. The mass matrix is diagonal (lumped), so no equations need be solved. Different time steps can be used in different parts of the mesh. (orig./GL)
Emerging technologies for 3D video creation, coding, transmission and rendering
Dufaux, Frederic; Cagnazzo, Marco
2013-01-01
With the expectation of greatly enhanced user experience, 3D video is widely perceived as the next major advancement in video technology. In order to fulfil the expectation of enhanced user experience, 3D video calls for new technologies addressing efficient content creation, representation/coding, transmission and display. Emerging Technologies for 3D Video will deal with all aspects involved in 3D video systems and services, including content acquisition and creation, data representation and coding, transmission, view synthesis, rendering, display technologies, human percepti
Simulations of RUTA-70 reactor with CERMET fuel using DYN3D/ATHLET and DYN3D/RELAP5 coupled codes
International Nuclear Information System (INIS)
RUTA-70 model for simulations with the internally coupled codes DYN3D/ATHLET and DYN3D/RELAP5 was developed. A 3-D power distribution in the core is calculated by DYN3D with thermal-hydraulic feedback from the system codes. A steady-state corresponding to the full reactor power and an accident scenario initiated by failure of all primary coolant pumps were simulated with the DYN3D/ATHLET and DYN3D/RELAP5 coupled code systems to verify these codes. The compared coupled codes give close predictions for the initial and final states of the simulated accident but not for the transition between them. The observed deviations are explained by differences in the subcooled boiling models of the employed versions of ATHLET and RELAP5. Nevertheless, both simulations confirm a high level of the reactor inherent safety. The allowed safety margins were not reached. (orig.)
Simulations of RUTA-70 reactor with CERMET fuel using DYN3D/ATHLET and DYN3D/RELAP5 coupled codes
Energy Technology Data Exchange (ETDEWEB)
Kozmenkov, Y.; Rohde, U. [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Baranaev, Y.; Glebov, A. [State Scientific Center of the Russian Federation, Obninsk, Kaluga Region (Russian Federation). Inst. for Physics and Power Engineering
2012-08-15
RUTA-70 model for simulations with the internally coupled codes DYN3D/ATHLET and DYN3D/RELAP5 was developed. A 3-D power distribution in the core is calculated by DYN3D with thermal-hydraulic feedback from the system codes. A steady-state corresponding to the full reactor power and an accident scenario initiated by failure of all primary coolant pumps were simulated with the DYN3D/ATHLET and DYN3D/RELAP5 coupled code systems to verify these codes. The compared coupled codes give close predictions for the initial and final states of the simulated accident but not for the transition between them. The observed deviations are explained by differences in the subcooled boiling models of the employed versions of ATHLET and RELAP5. Nevertheless, both simulations confirm a high level of the reactor inherent safety. The allowed safety margins were not reached. (orig.)
Application of COREMELT-3D code at analysis of severe fast reactor accidents
International Nuclear Information System (INIS)
The code COREMELT for calculations of initial and transition stages of severe accident is considered. It is used to conduct connected calculations of nonstationary neutronic and thermohydraulic processes in sodium fast reactor core. The code has some versions depending on dimensions of solving problem and consists of thermohydraulic module COREMELT and neutronic module RADAR. Using the code COREMELT-3D connected calculations of core disassembly accidents of ULOF and UTOP type have been conducted for sodium fast reactors safety analysis. The main problem of code COREMELT-3D use is duration of calculation, speeding of the code is possible when calculating algorithms are parallelized
A joint multi-view plus depth image coding scheme based on 3D-warping
DEFF Research Database (Denmark)
Zamarin, Marco; Zanuttigh, Pietro; Milani, Simone;
2011-01-01
scene structure that can be effectively exploited to improve the performance of multi-view coding schemes. In this paper we introduce a novel coding architecture that replaces the inter-view motion prediction operation with a 3D warping approach based on depth information to improve the coding...
Simulation of some plant transients by the coupled code system ATHLET/KIKO3D
International Nuclear Information System (INIS)
The assessment of coupled reactor physics and thermal-hydraulics computations with the coupled KIKO3D-ATHLET code system is provided, from two stand-alone codes. The details of data flow in the coupling are reviewed and some selected results of the validation are described. The validated coupled system code is used in the safety analysis for VVER reactors. (author)
Low-Complexity Multiple Description Coding of Video Based on 3D Block Transforms
Directory of Open Access Journals (Sweden)
Andrey Norkin
2007-02-01
Full Text Available The paper presents a multiple description (MD video coder based on three-dimensional (3D transforms. Two balanced descriptions are created from a video sequence. In the encoder, video sequence is represented in a form of coarse sequence approximation (shaper included in both descriptions and residual sequence (details which is split between two descriptions. The shaper is obtained by block-wise pruned 3D-DCT. The residual sequence is coded by 3D-DCT or hybrid, LOT+DCT, 3D-transform. The coding scheme is targeted to mobile devices. It has low computational complexity and improved robustness of transmission over unreliable networks. The coder is able to work at very low redundancies. The coding scheme is simple, yet it outperforms some MD coders based on motion-compensated prediction, especially in the low-redundancy region. The margin is up to 3 dB for reconstruction from one description.
Wall-touching kink mode calculations with the M3D code
International Nuclear Information System (INIS)
This paper seeks to address a controversy regarding the applicability of the 3D nonlinear extended MHD code M3D [W. Park et al., Phys. Plasmas 6, 1796 (1999)] and similar codes to calculations of the electromagnetic interaction of a disrupting tokamak plasma with the surrounding vessel structures. M3D is applied to a simple test problem involving an external kink mode in an ideal cylindrical plasma, used also by the Disruption Simulation Code (DSC) as a model case for illustrating the nature of transient vessel currents during a major disruption. While comparison of the results with those of the DSC is complicated by effects arising from the higher dimensionality and complexity of M3D, we verify that M3D is capable of reproducing both the correct saturation behavior of the free boundary kink and the “Hiro” currents arising when the kink interacts with a conducting tile surface interior to the ideal wall
Wall-touching kink mode calculations with the M3D code
Energy Technology Data Exchange (ETDEWEB)
Breslau, J. A., E-mail: jbreslau@pppl.gov; Bhattacharjee, A. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08542 (United States)
2015-06-15
This paper seeks to address a controversy regarding the applicability of the 3D nonlinear extended MHD code M3D [W. Park et al., Phys. Plasmas 6, 1796 (1999)] and similar codes to calculations of the electromagnetic interaction of a disrupting tokamak plasma with the surrounding vessel structures. M3D is applied to a simple test problem involving an external kink mode in an ideal cylindrical plasma, used also by the Disruption Simulation Code (DSC) as a model case for illustrating the nature of transient vessel currents during a major disruption. While comparison of the results with those of the DSC is complicated by effects arising from the higher dimensionality and complexity of M3D, we verify that M3D is capable of reproducing both the correct saturation behavior of the free boundary kink and the “Hiro” currents arising when the kink interacts with a conducting tile surface interior to the ideal wall.
MDPC-McEliece: New McEliece Variants from Moderate Density Parity-Check Codes
Misoczki, Rafael; Tillich, Jean-Pierre; Sendrier, Nicolas; Barreto, Paulo S.L.M.
2013-01-01
In this work, we propose two McEliece variants: one from Moderate Density Parity-Check (MDPC) codes and another from quasi-cyclic MDPC codes. MDPC codes are LDPC codes of higher density (and worse error-correction capability) than what is usually adopted for telecommunication applications. However, in cryptography we are not necessarily interested in correcting many errors, but only a number which ensures an adequate security level. By this approach, we reduce under certain hypotheses the sec...
Studies of spherical tori, stellarators and anisotropic pressure with the M3D code
International Nuclear Information System (INIS)
The Multi-level 3D (M3D) project simulates plasmas using multiple levels of physics, geometry and grid models in one code package. The M3D code has been extended to fundamentally non-axisymmetric and small aspect ratio, R/a > or ∼ 1, configurations. Applications include the non-linear stability of the NSTX spherical torus and of the spherical pinch, as well as the relaxation of stellarator equilibria. The fluid level physics model has been extended to evolve the anisotropic pressures pjparallel and pjperpendicular for the ion and electron species and has been applied to magnetic island evolution. (author)
Solution of the AER6 benchmark problem by KIKO3D/ATHLET code system
International Nuclear Information System (INIS)
The realistic analysis of accident conditions requires the extension of thermohydraulic plant system codes with 3D neutronics models. Recently many activities have been performed to develop, verify and validate such coupled codes. During the last years the thermohydraulic system code ATHLET developed by GRS was coupled to the 3D neutronic code KIKO3D developed by KFKI AEKI in order to simulate time dependent behavior of the VVER NPP. The 6-loop ATHLET input model worked out by AEKI assures the more precise characterization of the primary system. As an example of application, results are presented for the AER6 Benchmark, which is a VVER specific Main Steam Line Break (MSLB) transient. Emphasis is given to one of the basic problems of coupled codes, namely the effect of the slightly different nodalization in the core vessel.(author)
Simulation of some plant transients by the coupled code system ATHLET/KIKO3D
International Nuclear Information System (INIS)
The assessment of coupled reactor physics and thermal-hydraulic computation with the KIKO3D-ATHLET code is provided. The details of data flow in the coupling are reviewed and some selected results of the validation are described. The validated coupled system code is used in the safety analysis for WWER reactor. (Authors)
3-D field computation: The near-triumph of commerical codes
Energy Technology Data Exchange (ETDEWEB)
Turner, L.R.
1995-07-01
In recent years, more and more of those who design and analyze magnets and other devices are using commercial codes rather than developing their own. This paper considers the commercial codes and the features available with them. Other recent trends with 3-D field computation include parallel computation and visualization methods such as virtual reality systems.
Further validation and development of the 3-dimensional dynamics code TRAB-3D
International Nuclear Information System (INIS)
TRAB-3D, the newest member of VTT's code system for LWR dynamics calculations, is a coupled neutronics-thermal hydraulics code for transient and accident analyses of BWR reactors. The code is largely based on the ID code TRAB and the 3D hexagonal code HEXTRAN which have long been used in safety analyses of Finnish and foreign reactors. In TRAB-3D the two-group neutron diffusion equations are solved in three dimensions in a rectangular fuel assembly geometry by a new method which is similar to the nodal expansion method developed earlier at VTT for hexagonal geometry. The accuracy of the method is shown by comparison with 2D fine-mesh calculations and with 3D calculation for the Olkiluoto reactor with the POLCA-4 code. Capabilities of the code in dynamic analyses is validated with the OECD/NEA LWR benchmark problems and with transient calculations for the Olkiluoto reactor. Further development of the code includes a pin power reconstruction method which makes use of precomputed power distributions within fuel assemblies
The integrated code system CASCADE-3D for advanced core design and safety analysis
International Nuclear Information System (INIS)
The new program system CASCADE-3D (Core Analysis and Safety Codes for Advanced Design Evaluation) links some of Siemens advanced code packages for in-core fuel management and accident analysis: SAV95, PANBOX/COBRA and RELAP5. Consequently by using CASCADE-3D the potential of modern fuel assemblies and in-core fuel management strategies can be much better utilized because safety margins which had been reduced due to conservative methods are now predicted more accurately. By this innovative code system the customers can now take full advantage of the recent progress in fuel assembly design and in-core fuel management.(author)
SYN3D: a single-channel, spatial flux synthesis code for diffusion theory calculations
Energy Technology Data Exchange (ETDEWEB)
Adams, C. H.
1976-07-01
This report is a user's manual for SYN3D, a computer code which uses single-channel, spatial flux synthesis to calculate approximate solutions to two- and three-dimensional, finite-difference, multigroup neutron diffusion theory equations. SYN3D is designed to run in conjunction with any one of several one- and two-dimensional, finite-difference codes (required to generate the synthesis expansion functions) currently being used in the fast reactor community. The report describes the theory and equations, the use of the code, and the implementation on the IBM 370/195 and CDC 7600 of the version of SYN3D available through the Argonne Code Center.
Novel video coding algorithm based on 3D-binDCT
Institute of Scientific and Technical Information of China (English)
NI Wei; GUO Bao-long; YANG Liu
2005-01-01
In this paper we propose a three dimensional multiplierless discrete cosine transform(DCT) with lifting scheme called 3D-binDCT.Based on 3D-binDCT,a novel video coding algorithm without motion estimation/compensation is proposed.It uses the 3D-binDCT to exploit spatial or temporal redundancy.The computation of binDCT only needs shift and addition operations,thus the computational complexity is minimized.DC coefficient prediction,modified scan mode and arithmetic coding techniques are also adopted.Extensive simulation results show that the proposed coding scheme provides higher coding efficiency and improves visual quality, and it is easy to be realized by software and hardware.
3D video coding: an overview of present and upcoming standards
Merkle, Philipp; Müller, Karsten; Wiegand, Thomas
2010-07-01
An overview of existing and upcoming 3D video coding standards is given. Various different 3D video formats are available, each with individual pros and cons. The 3D video formats can be separated into two classes: video-only formats (such as stereo and multiview video) and depth-enhanced formats (such as video plus depth and multiview video plus depth). Since all these formats exist of at least two video sequences and possibly additional depth data, efficient compression is essential for the success of 3D video applications and technologies. For the video-only formats the H.264 family of coding standards already provides efficient and widely established compression algorithms: H.264/AVC simulcast, H.264/AVC stereo SEI message, and H.264/MVC. For the depth-enhanced formats standardized coding algorithms are currently being developed. New and specially adapted coding approaches are necessary, as the depth or disparity information included in these formats has significantly different characteristics than video and is not displayed directly, but used for rendering. Motivated by evolving market needs, MPEG has started an activity to develop a generic 3D video standard within the 3DVC ad-hoc group. Key features of the standard are efficient and flexible compression of depth-enhanced 3D video representations and decoupling of content creation and display requirements.
Benchmark of the 3-dimensional plasma transport codes E3D and BoRiS
International Nuclear Information System (INIS)
The next generation of experiments - both for tokamaks and stellarators - require the development of appropriate theoretical models. One important aspect here is the plasma edge physics description. Fluid transport codes extending beyond the standard 2-D code packages like B2-Eirene or UEDGE are under development. In the case of tokamaks, an interesting alternative line is the concept of an ergodic edge (necessary e.g. for ergodic divertors in TORE SUPRA or TEXTOR-94) creating a 3-D edge structure. To study this effects, a 3-D code E3D based upon Multiple Coordinate Systems Approach is being developed. Presently, we extend the program towards stellarator applications. A few new options are made available: single-island geometry and new formulation of boundary conditions. For the new stellarator W7-X a 3-D finite volume code BoRiS is developed using magnetic (Boozer) coordinates. In this paper, we present a benchmark of both codes for a test geometry (single magnetic island in W7-X) accounting for full 3-D metric variations for strongly anisotropic electron heat conduction equation. (orig.)
Depth-based coding of MVD data for 3D video extension of H.264/AVC
Rusanovskyy, Dmytro; Hannuksela, Miska M.; Su, Wenyi
2013-06-01
This paper describes a novel approach of using depth information for advanced coding of associated video data in Multiview Video plus Depth (MVD)-based 3D video systems. As a possible implementation of this conception, we describe two coding tools that have been developed for H.264/AVC based 3D Video Codec as response to Moving Picture Experts Group (MPEG) Call for Proposals (CfP). These tools are Depth-based Motion Vector Prediction (DMVP) and Backward View Synthesis Prediction (BVSP). Simulation results conducted under JCT-3V/MPEG 3DV Common Test Conditions show, that proposed in this paper tools reduce bit rate of coded video data by 15% of average delta bit rate reduction, which results in 13% of bit rate savings on total for the MVD data over the state-of-the-art MVC+D coding. Moreover, presented in this paper conception of depth-based coding of video has been further developed by MPEG 3DV and JCT-3V and this work resulted in even higher compression efficiency, bringing about 20% of delta bit rate reduction on total for coded MVD data over the reference MVC+D coding. Considering significant gains, proposed in this paper coding approach can be beneficial for development of new 3D video coding standards. [Figure not available: see fulltext.
Development of integrated transport code, TASK3D, and its applications to LHD experiment
International Nuclear Information System (INIS)
The integrated transport code for helical plasmas, TASK3D, has been developed both by modifying modules in TASK to be applicable to three-dimensional magnetic configurations, and by adding new modules for stellarator-heliotron specific physics and incorporating three-dimensional equilibria. In this paper, these module developments so far are collectively introduced, and recent progress on the applications of TASK3D to heat transport analyses of LHD plasmas is introduced. (author)
Non-linear correction for accuracy improvement of the neutron calculations with HEXAB-3D Code
International Nuclear Information System (INIS)
A differential approach of application of the Improved Coarse Mesh Method in the 3D hexagonal geometry diffusion problem is presented. A non-linear nodal model of improvement based on the solution of the local balance equation in a triangular sub-region of the node with triple decreased mesh step and a presentation of the spatial distribution of the neutron flux by linear combination of trigonometric al hyperbolic functions are presented. A principal program realisation of the differential nonlinear correction in the hexagonal geometry diffusion code HEXA-B-3D is described. Benchmark results for a 3D WWER-1000 benchmark problem are presented
International Nuclear Information System (INIS)
Comprehensive safety studies of high temperature gas cooled reactors (HTR) require full three dimensional coupled treatments of both neutron kinetics and thermal-hydraulics. In a common effort, GRS and IKE developed the coupled code system TORT-TD/ATTICA3D for pebble bed type HTR that connects the 3-D transient discrete-ordinates transport code TORT-TD with the 3-D porous medium thermal-hydraulics code ATTICA3D. In this paper, the physical models and calculation capabilities of TORT-TD and ATTICA3D are presented, focusing on model improvements in ATTICA3D and extensions made in TORT-TD related to HTR application. For first applications, the OECD/NEA/NSC PBMR-400 benchmark has been chosen. Results obtained with TORT-TD/ATTICA3D will be shown for transient exercises, e.g. control rod withdrawal and a control rod ejection. Results are compared to other benchmark participants' solutions with special focus on fuel temperature modelling features of ATTICA3D. The provided “grey-curtain” nuclear cross section libraries have been used. First results on 3-D effects during a control rod withdrawal transient will be presented. (author)
New 3D nodal method HEXNEM for improving the accuracy of the hexagonal version of the code DYN3D
International Nuclear Information System (INIS)
The nodal expansion method (NEM) used in the hexagonal version of the code DYN3D is based on the node averaged values of fluxes in the node volume and averaged values of fluxes and currents at the interfaces of the nodes. The 3-dimensional diffusion equation is split into a 2-dimensional equation in the hexagonal plane solved with the help of Bessel functions and a 1-dimensional equation in axial direction solved by polynomial expansion. The two equations are coupled by the transversal bucklings. The accuracy of this method is sufficient for the VVER-440 where the assembly pitch is 14.7 cm. The assemblies of the VVER-1000 have a larger pitch of 24.1 cm. Comparisons with mathematical benachmarks for the VVER-1000 show a maximal deviation of powers in the order of 5%. The new nodal expansion method HEXNEM presented here uses a different flux expansion in the nodes. In addition to the averaged values at the interfaces of the hexagon the values at the corner points are included. It is shown that the accuracy is improved particularly for the VVER-1000 problems. (orig.)
Coupling of the thermohydraulic code ATHLET with the neutron kinetic core model DYN3D
International Nuclear Information System (INIS)
The coupling of advanced thermohydraulic codes with 3-dimensional neutron kinetic codes corresponds to the effort to replace conservative estimations by best estimate calculations. ATHLET is an advanced thermohydraulic code, developed by the German Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS). Up to now only point kinetics and 1-dimensional neutron kinetics have been included. The DYN3D code, developed in the Research Centre Rossendorf (RCR) for the improvement of the simulation of reactivity initiated accidents in nuclear reactors with hexagonal fuel elements comprises 3-dimensional neutron kinetics, models for the thermohydraulics of the core including heat transfer from the fuel to the coolant and a fuel rod behavior model. The reactor core model DYN3D was coupled with the ATHLET code on two basically different ways. The first way of coupling uses only the neutron kinetics part of the DYN3D code (internal coupling). This coupling along the core is very close and demands an high effort of programming due to the high number of coupling parameters. In the second way the whole core is cut out from the ATHLET plant model. The core is completely modeled by the DYN3D code (external coupling). In this case the interfaces are located at the bottom and at the top of the core. At this interfaces the pressures, mass flow rates, enthalpies and concentrations of boron acid have to be transferred. This way of coupling is easy to realize by interconnection of an interface routine. It is effectively supported by the General Control and Simulation Modul (GCSM) of the ATHLET code. Almost no changes of the single programs are necessary. Another advantage of this coupling is that the complete DYN3D model can be used. The disadvantage of this method is the splitting of the thermohydraulics
Three-dimensional parallel UNIPIC-3D code for simulations of high-power microwave devices
Wang, Jianguo; Chen, Zaigao; Wang, Yue; Zhang, Dianhui; Liu, Chunliang; Li, Yongdong; Wang, Hongguang; Qiao, Hailiang; Fu, Meiyan; Yuan, Yuan
2010-07-01
This paper introduces a self-developed, three-dimensional parallel fully electromagnetic particle simulation code UNIPIC-3D. In this code, the electromagnetic fields are updated using the second-order, finite-difference time-domain method, and the particles are moved using the relativistic Newton-Lorentz force equation. The electromagnetic field and particles are coupled through the current term in Maxwell's equations. Two numerical examples are used to verify the algorithms adopted in this code, numerical results agree well with theoretical ones. This code can be used to simulate the high-power microwave (HPM) devices, such as the relativistic backward wave oscillator, coaxial vircator, and magnetically insulated line oscillator, etc. UNIPIC-3D is written in the object-oriented C++ language and can be run on a variety of platforms including WINDOWS, LINUX, and UNIX. Users can use the graphical user's interface to create the complex geometric structures of the simulated HPM devices, which can be automatically meshed by UNIPIC-3D code. This code has a powerful postprocessor which can display the electric field, magnetic field, current, voltage, power, spectrum, momentum of particles, etc. For the sake of comparison, the results computed by using the two-and-a-half-dimensional UNIPIC code are also provided for the same parameters of HPM devices, the numerical results computed from these two codes agree well with each other.
THE McELIECE CRYPTOSYSTEM WITH ARRAY CODES
Directory of Open Access Journals (Sweden)
Vedat Şiap
2011-12-01
Full Text Available Public-key cryptosystems form an important part of cryptography. In these systems, every user has a public and a private key. The public key allows other users to encrypt messages, which can only be decoded using the secret private key. In that way, public-key cryptosystems allow easy and secure communication between all users without the need to actually meet and exchange keys. One such system is the McEliece Public-Key cryptosystem, sometimes also called McEliece Scheme. However, as we live in the information age, coding is used in order to protecet or correct the messages in the transferring or the storing processes. So, linear codes are important in the transferring or the storing. Due to richness of their structure array codes which are linear are also an important codes. However, the information is then transferred into the source more securely by increasing the error correction capability with array codes. In this paper, we combine two interesting topics, McEliece cryptosystem and array codes.
Impact of packet losses in scalable 3D holoscopic video coding
Conti, Caroline; Nunes, Paulo; Ducla Soares, Luís.
2014-05-01
Holoscopic imaging became a prospective glassless 3D technology to provide more natural 3D viewing experiences to the end user. Additionally, holoscopic systems also allow new post-production degrees of freedom, such as controlling the plane of focus or the viewing angle presented to the user. However, to successfully introduce this technology into the consumer market, a display scalable coding approach is essential to achieve backward compatibility with legacy 2D and 3D displays. Moreover, to effectively transmit 3D holoscopic content over error-prone networks, e.g., wireless networks or the Internet, error resilience techniques are required to mitigate the impact of data impairments in the user quality perception. Therefore, it is essential to deeply understand the impact of packet losses in terms of decoding video quality for the specific case of 3D holoscopic content, notably when a scalable approach is used. In this context, this paper studies the impact of packet losses when using a three-layer display scalable 3D holoscopic video coding architecture previously proposed, where each layer represents a different level of display scalability (i.e., L0 - 2D, L1 - stereo or multiview, and L2 - full 3D holoscopic). For this, a simple error concealment algorithm is used, which makes use of inter-layer redundancy between multiview and 3D holoscopic content and the inherent correlation of the 3D holoscopic content to estimate lost data. Furthermore, a study of the influence of 2D views generation parameters used in lower layers on the performance of the used error concealment algorithm is also presented.
RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors
International Nuclear Information System (INIS)
The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point
RELAP5-3D code for supercritical-pressure, light-water-cooled reactors
International Nuclear Information System (INIS)
The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point. (author)
Multitasking the INS3D-LU code on the Cray Y-MP
Fatoohi, Rod; Yoon, Seokkwan
1991-01-01
This paper presents the results of multitasking the INS3D-LU code on eight processors. The code is a full Navier-Stokes solver for incompressible fluid in three dimensional generalized coordinates using a lower-upper symmetric-Gauss-Seidel implicit scheme. This code has been fully vectorized on oblique planes of sweep and parallelized using autotasking with some directives and minor modifications. The timing results for five grid sizes are presented and analyzed. The code has achieved a processing rate of over one Gflops.
Multitasking the INS3D-LU code on the Cray Y-MP
International Nuclear Information System (INIS)
This paper presents the results of multitasking the INS3D-LU code on eight processors. The code is a full Navier-Stokes solver for incompressible fluid in three dimensional generalized coordinates using a lower-upper symmetric-Gauss-Seidel implicit scheme. This code has been fully vectorized on oblique planes of sweep and parallelized using autotasking with some directives and minor modifications. The timing results for five grid sizes are presented and analyzed. The code has achieved a processing rate of over one Gflops
Performance Evaluation of Space-Time Turbo Code Concatenated With Block Code MC-CDMA Systems
Lokesh Kumar Bansal; Aditya Trivedi
2011-01-01
In this paper, performance of a space-time turbo code (STTuC) in concatenation with space-time block code (STBC) in multi-carrier code-division multiple-access (MCCDMA) system with multi-path fading channel is considered. The performance in terms of bit error rate (BER) is evaluated through simulations. The corresponding BER of the concatenated STTuC-STBC-MC-CDMA system is compared with STTuC-MC-CDMA system and STBC-MC-CDMA system. The simulation results show that the STTuC-MCCDMA system perf...
Description of a parallel, 3D, finite element, hydrodynamics-diffusion code
International Nuclear Information System (INIS)
We describe a parallel, 3D, unstructured grid finite element, hydrodynamic diffusion code for inertial confinement fusion (ICF) applications and the ancillary software used to run it. The code system is divided into two entities, a controller and a stand-alone physics code. The code system may reside on different computers; the controller on the user s workstation and the physics code on a supercomputer. The physics code is composed of separate hydrodynamic, equation-of-state, laser energy deposition, heat conduction, and radiation transport packages and is parallelized for distributed memory architectures. For parallelization, a SPMD model is adopted; the domain is decomposed into a disjoint collection of sub-domains, one per processing element (PE). The PEs communicate using MPI. The code is used to simulate the hydrodynamic implosion of a spherical bubble
Planet-Disk Interaction on the GPU: The FARGO3D code
Masset, F. S.; Benítez-Llambay, P.
2015-10-01
We present the new code FARGO3D. It is a finite difference code that solves the equations of hydrodynamics or magnetohydrodynamics on a Cartesian, cylindrical or spherical mesh. It features orbital advection, conserves mass and (angular) momentum to machine accuracy. Special emphasis is put on the description of planet disk tidal interactions. It is parallelized with MPI, and it can run indistinctly on CPUs or GPUs, without the need to program in a GPU oriented language.
Study of magnetic island using a 3D MHD equilibrium calculation code
International Nuclear Information System (INIS)
Coupling the magnetic diagnostics and a 3D MHD equilibrium calculation code, the magnetic island is studied in the Large Helical Device (LHD) experiment. In an experiment, the collapse in the plasma core was observed in a configuration, which has large magnetic island produced by external perturbation coils. At the collapse, the temperature profile was flattened. This suggests the magnetic island evolved. The magnetic island was observed by the magnetic diagnostics. The magnetic diagnostics also suggests evolving the magnetic island. A 3D MHD equilibrium is calculated by the 3D MHD equilibrium code then signals of the magnetic diagnostics are simulated. Since the comparison of observed and calculated signals is comparable, the magnetic island in calculated equilibrium is similar to one of the experiment. (author)
Implementation and testing of the CFDS-FLOW3D code
International Nuclear Information System (INIS)
FLOW3D is a multi-purpose, transient fluid dynamics and heat transfer code developed by Computational Fluid Dynamics Services (CFDS), a branch of AEA Technology, based at Harwell. The code is supplied with a SUN-based operating environment consisting of an interactive grid generator SOPHIA and a post-processor JASPER for graphical display of results. Both SOPHIA and JASPER are extensions of the support software originally written for the ASTEC code, also promoted by CFDS. The latest release of FLOW3D contains well-tested turbulence and combustion models and, in a less-developed form, a multi-phase modelling potential. This document describes briefly the modelling capabilities of FLOW3D (Release 3.2) and outlines implementation procedures for the VAX, CRAY and CONVEX computer systems. Additional remarks are made concerning the in-house support programs which have been specially written in order to adapt existing ASTEC input data for use with FLOW3D; these programs operate within a VAX-VMS environment. Three sample calculations have been performed and results compared with those obtained previously using the ASTEC code, and checked against other available data, where appropriate. (author) 35 figs., 3 tabs., 42 refs
Simplified 3D model of a PWR reactor vessel using fluid dynamics code ANSYS CFX computational
International Nuclear Information System (INIS)
This paper presents the results from the calculation of the steady state simulation with model of CFD (computational fluid dynamic) operating under conditions of operation at full power (Hot Full Power). Development and the CFD model results show the usefulness of these codes for calculating 3D of the variable thermohydraulics of these reactors.
Modeling and validation of CFD code KIRAN3D for electron beam melting of zirconium
International Nuclear Information System (INIS)
The validation of the computer code KIRAN3D is carried out with the physical experiments carried out using electron beam melting of zirconium ingot in cold hearth. The measured maximum surface temperature shows good agreement with the predicted results by computational analysis, when the Gaussian beam profile is used. (author)
A new 3-D integral code for computation of accelerator magnets
International Nuclear Information System (INIS)
For computing accelerator magnets, integral codes have several advantages over finite element codes; far-field boundaries are treated automatically, and computed fields in the bore region satisfy Maxwell's equations exactly. A new integral code employing the edge elements rather than nodal elements has overcome the difficulties associated with earlier integral codes. By the use of field integrals (potential differences) as solution variables, the number of unknowns is reduced to one less than the number of nodes. Two examples, a hollow iron sphere and the dipole magnet of Advanced Photon source injector synchrotron, show the capability of the code. The CPU time requirements are comparable to those of three-dimensional (3-D) finite-element codes. Experiments show that in practice it can realize much of the potential CPU time saving that parallel processing makes possible
VizieR Online Data Catalog: ATLAS3D Project. XXX (McDermid+, 2015)
McDermid, R. M.; Alatalo, K.; Blitz, L.; Bournaud, F.; Bureau, M.; Cappellari, M.; Crocker, A. F.; Davies, R. L.; Davis, T. A.; De Zeeuw, P. T.; Duc, P.-A.; Emsellem, E.; Khochfar, S.; Krajnovic, D.; Kuntschner, H.; Morganti, R.; Naab, T.; Oosterloo, T.; Sarzi, M.; Scott, N.; Serra, P.; Weijmans, A.-M.; Young, L. M.
2015-09-01
We present the stellar population content of early-type galaxies from the ATLAS3D survey. Using spectra integrated within apertures covering up to one effective radius, we apply two methods: one based on measuring line-strength indices and applying single stellar population (SSP) models to derive SSP-equivalent values of stellar age, metallicity, and alpha enhancement; and one based on spectral fitting to derive non-parametric star formation histories, mass-weighted average values of age, metallicity, and half-mass formation time-scales. Using homogeneously derived effective radii and dynamically determined galaxy masses, we present the distribution of stellar population parameters on the Mass Plane (MJAM, σe, Rmaje), showing that at fixed mass, compact early-type galaxies are on average older, more metal-rich, and more alpha-enhanced than their larger counterparts. From non-parametric star formation histories, we find that the duration of star formation is systematically more extended in lower mass objects. Assuming that our sample represents most of the stellar content of today's local Universe, approximately 50 percent of all stars formed within the first 2Gyr following the big bang. Most of these stars reside today in the most massive galaxies (>1010.5M⊙), which themselves formed 90 percent of their stars by z~2. The lower mass objects, in contrast, have formed barely half their stars in this time interval. Stellar population properties are independent of environment over two orders of magnitude in local density, varying only with galaxy mass. In the highest density regions of our volume (dominated by the Virgo cluster), galaxies are older, alpha-enhanced, and have shorter star formation histories with respect to lower density regions. (4 data files).
International Nuclear Information System (INIS)
The paper gives a brief survey of the 6th three-dimensional AER dynamic benchmark calculation results received with the codes DYN3D and RELAP5-3D at NRI Rez. This benchmark was defined at the 10th AER Symposium. Its initiating event is a double ended break in the steam line of steam generator No. 1 in a WWER-440/213 plant at the end of the first fuel cycle and in hot full power conditions. Stationary and burnup calculations as well as tuning of initial state before the transient were performed with the code DYN3D. Transient calculations were made with the system code RELAP5-3D. The KASSETA library was used for the generation of reactor core neutronic parameters. The detailed six loops model of NPP Dukovany was adopted for the 6th AER dynamic benchmark purposes. The RELAP5-3D full core neutronic model was connected with 37 coolant channels thermal-hydraulic model of the core, 6-sector nodalization of reactor downcomer, lower and upper plenum was used. Mixing in lower and upper plenum was simulated. The first part of paper contains a brief characteristic of RELAP5 -3D system code and a short description of NPP input deck and reactor core model. The second part shows the time dependencies of important global and local parameters (Authors)
RELAP5-3D Code Includes ATHENA Features and Models
International Nuclear Information System (INIS)
Version 2.3 of the RELAP5-3D computer program includes all features and models previously available only in the ATHENA version of the code. These include the addition of new working fluids (i.e., ammonia, blood, carbon dioxide, glycerol, helium, hydrogen, lead-bismuth, lithium, lithium-lead, nitrogen, potassium, sodium, and sodium-potassium) and a magnetohydrodynamic model that expands the capability of the code to model many more thermal-hydraulic systems. In addition to the new working fluids along with the standard working fluid water, one or more noncondensable gases (e.g., air, argon, carbon dioxide, carbon monoxide, helium, hydrogen, krypton, nitrogen, oxygen, SF6, xenon) can be specified as part of the vapor/gas phase of the working fluid. These noncondensable gases were in previous versions of RELAP5-3D. Recently four molten salts have been added as working fluids to RELAP5-3D Version 2.4, which has had limited release. These molten salts will be in RELAP5-3D Version 2.5, which will have a general release like RELAP5-3D Version 2.3. Applications that use these new features and models are discussed in this paper. (authors)
RELAP5-3D Code Includes Athena Features and Models
Energy Technology Data Exchange (ETDEWEB)
Richard A. Riemke; Cliff B. Davis; Richard R. Schultz
2006-07-01
Version 2.3 of the RELAP5-3D computer program includes all features and models previously available only in the ATHENA version of the code. These include the addition of new working fluids (i.e., ammonia, blood, carbon dioxide, glycerol, helium, hydrogen, lead-bismuth, lithium, lithium-lead, nitrogen, potassium, sodium, and sodium-potassium) and a magnetohydrodynamic model that expands the capability of the code to model many more thermal-hydraulic systems. In addition to the new working fluids along with the standard working fluid water, one or more noncondensable gases (e.g., air, argon, carbon dioxide, carbon monoxide, helium, hydrogen, krypton, nitrogen, oxygen, sf6, xenon) can be specified as part of the vapor/gas phase of the working fluid. These noncondensable gases were in previous versions of RELAP5- 3D. Recently four molten salts have been added as working fluids to RELAP5-3D Version 2.4, which has had limited release. These molten salts will be in RELAP5-3D Version 2.5, which will have a general release like RELAP5-3D Version 2.3. Applications that use these new features and models are discussed in this paper.
Magneto-acoustic waves in sunspots: first results from a new 3D nonlinear magnetohydrodynamic code
Felipe, T; Collados, M
2010-01-01
Waves observed in the photosphere and chromosphere of sunspots show complex dynamics and spatial patterns. The interpretation of high-resolution sunspot wave observations requires modeling of three-dimensional non-linear wave propagation and mode transformation in the sunspot upper layers in realistic spot model atmospheres. Here we present the first results of such modeling. We have developed a 3D non-linear numerical code specially designed to calculate the response of magnetic structures in equilibrium to an arbitrary perturbation. The code solves the 3D nonlinear MHD equations for perturbations; it is stabilized by hyper-diffusivity terms and is fully parallelized. The robustness of the code is demonstrated by a number of standard tests. We analyze several simulations of a sunspot perturbed by pulses of different periods at subphotospheric level, from short periods, introduced for academic purposes, to longer and realistic periods of three and five minutes. We present a detailed description of the three-d...
ATHENA 3D: A finite element code for ultrasonic wave propagation
International Nuclear Information System (INIS)
The understanding of wave propagation phenomena requires use of robust numerical models. 3D finite element (FE) models are generally prohibitively time consuming. However, advances in computing processor speed and memory allow them to be more and more competitive. In this context, EDF R and D developed the 3D version of the well-validated FE code ATHENA2D. The code is dedicated to the simulation of wave propagation in all kinds of elastic media and in particular, heterogeneous and anisotropic materials like welds. It is based on solving elastodynamic equations in the calculation zone expressed in terms of stress and particle velocities. The particularity of the code relies on the fact that the discretization of the calculation domain uses a Cartesian regular 3D mesh while the defect of complex geometry can be described using a separate (2D) mesh using the fictitious domains method. This allows combining the rapidity of regular meshes computation with the capability of modelling arbitrary shaped defects. Furthermore, the calculation domain is discretized with a quasi-explicit time evolution scheme. Thereby only local linear systems of small size have to be solved. The final step to reduce the computation time relies on the fact that ATHENA3D has been parallelized and adapted to the use of HPC resources. In this paper, the validation of the 3D FE model is discussed. A cross-validation of ATHENA 3D and CIVA is proposed for several inspection configurations. The performances in terms of calculation time are also presented in the cases of both local computer and computation cluster use.
The trigonal nodal SP3 method of the reactor code DYN3D
International Nuclear Information System (INIS)
DYN3D is a 3D nodal diffusion code for steady-state and transient analyses of Light-Water Reactors (LWRs) with square and hexagonal fuel assembly geometries. Currently several versions of the DYN3D code are available including a multi-group diffusion and a simplified P3 (SP3) neutron transport option. In this work, the multi-group SP3 method based on trigonal-z geometry was developed. The method is applicable to the analysis of reactor cores with hexagonal fuel assemblies and allows flexible mesh refinement, which is of particular importance for VVER-type Pressurized Water Reactors (PWRs) as well as for innovative reactor concepts including block type High-Temperature Reactors (HTRs) and Sodium Fast Reactors (SFRs). In this paper, the theoretical background for the trigonal SP3 methodology is outlined and the results of a preliminary verification analysis are presented by means of two VVER-440 single assembly test examples with different material compositions. The accordant cross sections and reference solutions were produced by the Monte Carlo code SERPENT. The DYN3D results are shown for 2 and 8 energy groups, respectively, and are in good agreement with the reference solutions. The deviation in the nodal power distribution is about 1%. (author)
Analysis of the Boiling Water Reactor Turbine Trip Benchmark with the Codes DYN3D and ATHLET/DYN3D
International Nuclear Information System (INIS)
The OECD/NRC Boiling Water Reactor (BWR) Turbine Trip Benchmark was analyzed by the code DYN3D and the coupled code system ATHLET/DYN3D. For the exercise 2 benchmark calculations with given thermal-hydraulic boundary conditions of the core, the analyses were performed with the core model DYN3D. Concerning the modeling of the BWR core in the DYN3D code, several simplifications and their influence on the results were investigated. The standard calculations with DYN3D were performed with 764 coolant channels (one channel per fuel assembly), the assembly discontinuity factors (ADF), and the phase slip model of Molochnikov. Comparisons were performed with the results obtained by calculations with 33 thermal-hydraulic channels, without the ADF and with the slip model of Zuber and Findlay. It is shown that the influence on core-averaged values of the steady state and the transient is small. Considering local parameters, the influence of the ADF or the reduced number of coolant channels is not negligible. For the calculations of exercise 3, the DYN3D model validated during the exercise 2 calculations in combination with the ATHLET system model, developed at Gesellschaft fuer Anlagen- und Reaktorsicherheit for exercise 1, has been used. Calculations were performed for the basic scenario as well as for all specified extreme versions. They were carried out using a modified version of the external coupling of the codes, the 'parallel' coupling. This coupling shows a stable performance at the low time step sizes necessary for an appropriate description of the feedback during the transient. The influence of assumed failures of different relevant safety systems on the plant and the core behavior was investigated in the calculations of the extreme scenarios. The calculations of exercises 2 and 3 contribute to the validation of DYN3D and ATHLET/DYN3D for BWR systems
MCMG: a 3-D multigroup P3 Monte Carlo code and its benchmarks
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In this paper a 3-D Monte Carlo multigroup neutron transport code MCMG has been developed from a coupled neutron and photon transport Monte Carlo code MCNP. The continuous-energy cross section library of the MCNP code is replaced by the multigroup cross section data generated by the transport lattice code, such as the WIMS code. It maintains the strong abilities of MCNP for geometry treatment, counting, variance reduction techniques and plotting. The multigroup neutron scattering cross sections adopt the Pn (n ≤ 3) approximation. The test results are in good agreement with the results of other methods and experiments. The number of energy groups can be varied from few groups to multigroup, and either macroscopic or microscopic cross section can be used. (author)
A 3D transport-based core analysis code for research reactors with unstructured geometry
International Nuclear Information System (INIS)
Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal SN method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of keff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results
Li, Yong Gang; Yang, Yang; Short, Michael P.; Ding, Ze Jun; Zeng, Zhi; Li, Ju
2015-12-01
SRIM-like codes have limitations in describing general 3D geometries, for modeling radiation displacements and damage in nanostructured materials. A universal, computationally efficient and massively parallel 3D Monte Carlo code, IM3D, has been developed with excellent parallel scaling performance. IM3D is based on fast indexing of scattering integrals and the SRIM stopping power database, and allows the user a choice of Constructive Solid Geometry (CSG) or Finite Element Triangle Mesh (FETM) method for constructing 3D shapes and microstructures. For 2D films and multilayers, IM3D perfectly reproduces SRIM results, and can be ∼102 times faster in serial execution and > 104 times faster using parallel computation. For 3D problems, it provides a fast approach for analyzing the spatial distributions of primary displacements and defect generation under ion irradiation. Herein we also provide a detailed discussion of our open-source collision cascade physics engine, revealing the true meaning and limitations of the “Quick Kinchin-Pease” and “Full Cascades” options. The issues of femtosecond to picosecond timescales in defining displacement versus damage, the limitation of the displacements per atom (DPA) unit in quantifying radiation damage (such as inadequacy in quantifying degree of chemical mixing), are discussed.
Equation-of-State Test Suite for the DYNA3D Code
Energy Technology Data Exchange (ETDEWEB)
Benjamin, Russell D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
2015-11-05
This document describes the creation and implementation of a test suite for the Equationof- State models in the DYNA3D code. A customized input deck has been created for each model, as well as a script that extracts the relevant data from the high-speed edit file created by DYNA3D. Each equation-of-state model is broken apart and individual elements of the model are tested, as well as testing the entire model. The input deck for each model is described and the results of the tests are discussed. The intent of this work is to add this test suite to the validation suite presently used for DYNA3D.
International Nuclear Information System (INIS)
We have integrated the electron-cloud code POSINST [1] with WARP [2]--a 3-D parallel Particle-In-Cell accelerator code developed for Heavy Ion Inertial Fusion--so that the two can interoperate. Both codes are run in the same process, communicate through a Python interpreter (already used in WARP), and share certain key arrays (so far, particle positions and velocities). Currently, POSINST provides primary and secondary sources of electrons, beam bunch kicks, a particle mover, and diagnostics. WARP provides the field solvers and diagnostics. Secondary emission routines are provided by the Tech-X package CMEE
Development of a 3D neutron transport code and benchmark tests
International Nuclear Information System (INIS)
Results are reported of NEACRP '3D Neutron Transport Benchmarks' proposed from Osaka UNiversity, and of recent progress in the development of a 3D neutron transport code. Takeda et al. proposed four problems to NEACRP as 3D neutron transport benchmarks, and 22 results from 20 organizations were submitted. A variety of methods have been used, such as the Monte Carlo, Sn, Pn, synthetic, and nodal method. The results for k-eff, control-rod worths, and region-averaged fluxes are summarized with the conclusions that (1) in XYZ geometry the Sn method with n=8 shows a good agreement with the Monte-Carlo method, and gives even better results in some cases, (2) the Pn method has significant spatial mesh effects, and (3) the Sn method is not satisfactory in hexagonal-Z geometry, and improvements in accuracy are desirable. Improvement of a 3D neutron transport code is in progress to resolve the problem in the hexagonal-Z geometry by considering new diamond difference schemes and an improved coarse-mesh method, and also by applying the nodal method. (author)
The trigonal nodal SP3 method of the code DYN3D. Verification on pin level
International Nuclear Information System (INIS)
The neutronics model of the nodal reactor dynamics code DYN3D developed for 3D analyses of steady states and transients in Light- Water Reactors has been extended by a simplified P3 (SP3) neutron transport option - to overcome the limitations of the diffusion approach at regions with significant anisotropy effects. To provide a method being applicable to reactors with hexagonal fuel assemblies and to furthermore allow flexible mesh refinement, the nodal SP3 method has been developed on the basis of a flux expansion in trigonal-z geometry. In this paper, a verification of the methodology on quasi-pin level is performed by means of a single-assembly test example. The corresponding pin-wise few-group cross sections were obtained by the deterministic lattice code HELIOS. The power distributions were calculated using both the trigonal DYN3D diffusion and SP3 solver and compared to the HELIOS reference solutions. Close to regions with non-negligible flux anisotropies, e.g., caused by the presence of a strong absorbing material, the power distribution calculated by DYN3D-SP3 shows a significant improvement in comparison to the diffusion method. (orig.)
Wall touching kink mode calculations with the M3D code
Breslau, J. A.
2014-10-01
In recent years there have been a number of results published concerning the transient vessel currents and forces occurring during a tokamak VDE, as predicted by simulations with the nonlinear MHD code M3D. The nature of the simulations is such that these currents and forces occur at the boundary of the computational domain, making the proper choice of boundary conditions critical to the reliability of the results. The M3D boundary condition includes the prescription that the normal component of the velocity vanish at the wall. It has been argued that this prescription invalidates the calculations because it would seem to rule out the possibility of advection of plasma surface currents into the wall. This claim has been tested by applying M3D to an idealized case - a kink-unstable plasma column - in order to abstract the essential physics from the complications involved in the attempt to model real devices. While comparison of the results is complicated by effects arising from the higher dimensionality and complexity of M3D, we have verified that M3D is capable of reproducing both the correct saturation behavior of the free boundary kink and the ``Hiro'' currents arising when the kink interacts with a conducting tile surface interior to the ideal wall.
Development of multidimensional two-fluid model code ACE-3D for evaluation of constitutive equations
International Nuclear Information System (INIS)
In order to perform design calculations for a passive safety reactor with good accuracy by a multidimensional two-fluid model, we developed an analysis code, ACE-3D, which can apply for evaluation of constitutive equations. The developed code has the following features: 1. The basic equations are based on 3-dimensional two-fluid model and the orthogonal or the cylindrical coordinate system can be selected. The fluid system is air-water or steam-water. 2. The basic equations are formulated by the finite-difference scheme of staggered mesh. The convection term is formulated by an upwind scheme and the diffusion term by a center-difference scheme. 3. Semi-implicit numerical scheme is adopted and the mass and the energy equations are treated equally in convergent steps for Jacobi equations. 4. The interfacial stress term consists of drag force, life force, turbulent dispersion force, wall force and virtual mass force. 5. A κ-ε turbulent model for bubbly flow is incorporated as the turbulent model. The predictive capability of ACE-3D has been verified using a data-base for bubbly flow in a small-scale vertical pipe. In future, the constitutive equations will be improved with a data-base in a large vertical pipe developed in our laboratory and we have a plan to construct a reliable analytical tool through the improvement work, the progress of calculational speed with vector and parallel processing, the assessments for phase change terms and so on. This report describes the outline for the basic equations and the finite-difference equations in ACE-3D code and also the outline for the program structure. Besides, the results for the assessments of ACE-3D code for the small-scale pipe are summarized. (author)
V1000CT-1 benchmark analyses with the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems
International Nuclear Information System (INIS)
Full text of publication follows:Plant-measured data provided within the specification of the OECD/NEA VVER-1000 coolant transient benchmark (V1000CT) were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to the MCP (main coolant pump) switching on experiment conducted in the frame of the plant-commissioning activities at the Kozloduy NPP Unit 6 in Bulgaria. The experiment was started at the beginning of cycle (BOC) with average core expose of 30.7 effective full power days (EFPD), when the reactor power was at 27.5% of the nominal level and 3 of 4 MCPs were operating. The transient is characterized by a rapid increase in the primary coolant flow through the core and, as a consequence, a decrease of the space-dependent core inlet temperature. Control rods were not changing their positions during the transient. Both DYN3D/RELAP5 and DYN3D/ATHLET analyses were based on the same reactor model, including identical nodalization schemes, MCP characteristics, boundary conditions and the benchmark-specified nuclear data library. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has been performed to evaluate relevant thermohydraulic models of the system codes RELAP5 and ATHLET. (authors)
International Nuclear Information System (INIS)
Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET
Energy Technology Data Exchange (ETDEWEB)
Kozmenkov, Y. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Kliem, S. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)]. E-mail: S.Kliem@fzd.de; Grundmann, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Rohde, U. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany); Weiss, F.-P. [Forschungszentrum Dresden-Rossendorf, Institute of Safety Research, P.O.B. 510119, D-01314 Dresden (Germany)
2007-09-15
Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. The transient is characterized by a rapid increase in the primary coolant flow through the core, and as a consequence, a decrease of the space-dependent core inlet temperature. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET.
Code-Generierung für einen neuen 3D-Druckprozess auf Tropfenbasis
Schwaiger, Johannes
2014-01-01
Es wird die automatische Code-Generierung für einen neuen 3D-Druckprozess beschrieben, der es ermöglicht durch sequentielles Austragen von Tropfen aus Kunststoff ein 3-dimensionales Bauteil Schicht für Schicht aufzubauen. Ziel der Arbeit ist der Einsatz der gedruckten Teile als Funktionsteile. Wesentlicher Bestandteil dieser Arbeit war dabei die Entwicklung einer prozessspezifischen Software, welche in der Lage ist aus CAD-Daten im STL-Format, automatisch Maschinencode (G-Code) zu generieren,...
A 3D multigroup transport kinetics code in hexagonal geometry for fast reactor transient analysis
International Nuclear Information System (INIS)
A description of the 3D multigroup diffusion/transport kinetics code HEXNODYN is given and numerical results are reported. HEXNODYN couples time integration by the quasi-static method with space integration by HEXNOD's analytic (diffusion option) or discrete ordinates (transport option) nodal method. An equivalent hexagonal version of the KfK rod ejection problem has been set up to validate the diffusion option by comparison with available 2D diffusion codes. The transport option has been validated by comparison with the diffusion option. Numerical results indicate that the diffusion option may be considered as fully validated while the transport version is at least internally consistent
Modelling of aspherical nebulae. I. A quick pseudo-3D photoionization code
Morisset, C; Peña, M
2005-01-01
We describe a pseudo-3D photoionization code, NEBU_3D and its associated visualization tool, VIS_NEB3D, which are able to easily and rapidly treat a wide variety of nebular geometries, by combining models obtained with a 1D photoionization code. The only requirement for the code to work is that the ionization source is uniqu e and not extended. It is applicable as long as the diffuse ionizing radiation f ield is not dominant and strongly inhomogeneous. As examples of the capabilities of these new tools, we consider two very differ ent theoretical cases. One is that of a high excitation planetary nebula that ha s an ellipsoidal shape with two polar density knots. The other one is that of a blister HII region, for which we have also constructed a spherical model (the sp herical impostor) which has exactly the same Hbeta surface brightness distrib ution as the blister model and the same ionizing star. These two examples warn against preconceived ideas when interpreting spectroscop ic and imaging data of HII regi...
Agent code: Neutron transport benchmark example and extension to 3D lattice geometry
Directory of Open Access Journals (Sweden)
Hursin Mathieu
2005-01-01
Full Text Available The general methodology be hind 2D arbitrary geometry neutron transport AGENT code is the theory of R-functions, which al lows for simple modeling of complex geometries, and the method of characteristics, which solves the integral transport equation along characteristic neutron trajectories. This paper focuses on the extension of the methodology to ac count for 3D lattice geometries. Since the direct application of method of characteristics to 3D non-homogenized core con figuration may re quire a tremendous amount of memory and computing time, an alternative approximate solution based on coupling 2D method of characteristics and 1D diffusion solution is developed. The planar 2D method of characteristics and axial 1D diffusion solutions are coupled through the trans verse leak age. The use of a lower order 1D solution in the axial direction is justified by the fact that more heterogeneity in current PWR and BWR reactor cores occurs in the radial direction than in the axial one. In order to demonstrate the versatility and accuracy of the AGENT code, a 2D heterogeneous lattice problem, C5G7 is described in details. A theoretical description of the coupling methodology for 3D method of characteristics solution is followed by preliminary validation in comparison to the DeCART code.
International Nuclear Information System (INIS)
A comprehensive analysis of the double ended Main Steam Line Break (MSLB) accident assumed to occur in the Babcock and Wilcox nuclear power plant of Three Miles Island Unit 1 (TMI-1) has been carried out of the University of Pisa in co-operation with the University of Zagreb and the Texas A and M University. The overall activity has been completed within the framework of the participation in the OECD-CSNI/NSC (Committee on the Safety of Nuclear Installations - Nuclear Science Committee) 'PWR MSLB Benchmark'. Different code versions have been adopted in the analysis. Results from the following codes (or code versions) are described in this paper: RELAP5/MOD3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code; RELAP5/MOD3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code; RELAP5/3D, coupled with the 3-D neutron kinetics Nestle code. Boundary and initial conditions of the system including those relevant to the fuel status, have been supplied by Pensilvania State University that had a co-operation GPU (the utility, owner of TMI) and NRC (US Nuclear Regulatory Commission). The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some 're-criticality' or 'return-to-power' whose magnitude is largely affected by boundary and initial conditions. The comparison among the results obtained by adopting the same thermalhydraulic nodalization and the different 'coupled' code version is discussed in the present document. (author)
International Nuclear Information System (INIS)
Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the 'user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP 2005 (Scaling, Uncertainty and 3D COuPled code calculations) seminar has been organized by University of Pisa and University of Zagreb as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users (D'Auria, 1998). It was recognized that such a course represented both a source of continuing education for current code users and a means for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The seminar-training was successfully held with the participation of 19 persons coming from 9 countries and 14 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 15 scientists were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and holding the training and the final examination. A certificate (LA Code User grade) was released
SUSD3D, 1-, 2-, 3-Dimensional Cross Section Sensitivity and Uncertainty Code
International Nuclear Information System (INIS)
1 - Description of program or function: SUSD3D calculates sensitivity coefficients and standard deviation in the calculated detector responses or design parameters of interest due to input cross sections and their uncertainties. One-, two- and three-dimensional transport problems can be studied. Several types of uncertainties can be considered, i.e. those due to: (1) neutron/gamma multigroup cross sections, (2) energy-dependent response functions, (3) secondary angular distribution (SAD) or secondary energy distribution (SED) uncertainties. SUSD3D development was started from the SUSD [6] code. Besides several minor modifications and extensions SUSD3D differs from SUSD in particular that: - Three-dimensional analysis is possible, - Flux moment files are used to evaluate the sensitivity profiles, instead of angular flux files; substantially reducing in this was the computer space requirements. SUSD3D can use the flux moment files produced by the DORT, TORT [7], ONEDANT, TWODANT and THREEDANT [8] discrete ordinates codes. The method used in the SUSD code based on the angular flux files from the ANISN [9] and DOT-III codes was kept for comparison, - Processing codes were updated to the ENDF-6 Format, - Processing of SAD covariance matrices was programmed, - Complete SAD covariance matrices can be taken into account in SUSD3D to calculate the variance. NEA-1628/03: This version differs from the previous one in the following points: Modifications are relevant for the sensitivity calculations of the critical systems and include: - Correction of the sensitivity calculation for prompt fission and number of delayed neutrons per fission (MT=18 and MT=455). - An option allows the re-normalisation of the prompt fission spectra covariance matrices to be applied via the 'normalisation' of the sensitivity profiles. This option is useful in case if the fission spectra covariances (MF=35) used do not comply with the ENDF-6 Format Manual rules. - For the criticality calculations the
RAVE code system for 3-D core non-LOCA accident analysis
International Nuclear Information System (INIS)
Full text of publication follows: This paper provides an overview of the application of the Westinghouse updated RAVE three dimensional (3-D) core transient analysis code system for PWR non-LOCA accident analysis. The RAVE code system consists of a linkage of the following USNRC-approved codes: the EPRI RETRAN-02 (RETRAN) system transient analysis code, the Westinghouse SPNOVA (also referred to as ANC-K) reactor core neutron kinetic nodal code, and the EPRI VIPRE-01 (VIPRE) reactor core thermal-hydraulic (T/H) code. The RETRAN code is used for calculating transient conditions in the reactor coolant system (RCS), including reactor vessel, RCS loops, pressurizer and steam generators. RETRAN also models reactor trips, engineering safety feature (ESF) functions, and the control systems. The SPNOVA code is used to perform 3-D core neutronic calculations for core average power and power distributions in the core. Its reactivity feedback calculation is based on transient fluid conditions and fuel temperatures obtained from the VIPRE code. Based on core inlet temperature, inlet flow and core exit pressure from RETRAN, and the nodal nuclear power from SPNOVA, VIPRE provides back to RETRAN transient nodal heat flux in the reactor core region. An effective 3-D analysis requires RETRAN, SPNOVA and VIPRE calculations to be closely linked for the entire reactor core. The linking architecture uses a standard external communication interface protocol for communication among the running programs on the same or different computers. The RAVE code system currently uses the Parallel Virtual Machine (PVM) software for the data transfer. Besides the necessary changes for data transfer, no other changes were made to RETRAN, SPNOVA or VIPRE fundamental code algorithms or solution methods. The RETRAN model in the RAVE system uses the same detailed reactor vessel, RCS loops, pressurizer, and steam generator, and control and protection models as has been licensed for current plant Safety
User Guide for the R5EXEC Coupling Interface in the RELAP5-3D Code
Energy Technology Data Exchange (ETDEWEB)
Forsmann, J. Hope [Idaho National Lab. (INL), Idaho Falls, ID (United States); Weaver, Walter L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)
2015-04-01
This report describes the R5EXEC coupling interface in the RELAP5-3D computer code from the users perspective. The information in the report is intended for users who want to couple RELAP5-3D to other thermal-hydraulic, neutron kinetics, or control system simulation codes.
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.
The future of 3D and video coding in mobile and the internet
Bivolarski, Lazar
2013-09-01
The current Internet success has already changed our social and economic world and is still continuing to revolutionize the information exchange. The exponential increase of amount and types of data that is currently exchanged on the Internet represents significant challenge for the design of future architectures and solutions. This paper reviews the current status and trends in the design of solutions and research activities in the future Internet from point of view of managing the growth of bandwidth requirements and complexity of the multimedia that is being created and shared. Outlines the challenges that are present before the video coding and approaches to the design of standardized media formats and protocols while considering the expected convergence of multimedia formats and exchange interfaces. The rapid growth of connected mobile devices adds to the current and the future challenges in combination with the expected, in near future, arrival of multitude of connected devices. The new Internet technologies connecting the Internet of Things with wireless visual sensor networks and 3D virtual worlds requires conceptually new approaches of media content handling from acquisition to presentation in the 3D Media Internet. Accounting for the entire transmission system properties and enabling adaptation in real-time to context and content throughout the media proceeding path will be paramount in enabling the new media architectures as well as the new applications and services. The common video coding formats will need to be conceptually redesigned to allow for the implementation of the necessary 3D Media Internet features.
Feasibility of the integration of CRONOS, a 3-D neutronics code, into real-time simulators
International Nuclear Information System (INIS)
In its effort to contribute to nuclear power plant safety, CEA proposes the integration of an engineering grade 3-D neutronics code into a real-time plant analyser. This paper describes the capabilities of the neutronics code CRONOS to achieve a fast running performance. First, we will present current core models in simulators and explain their drawbacks. Secondly, the mean features of CRONOS's spatial-kinetics methods will be reviewed. We will then present an optimum core representation with respect to mesh size, choice of finite elements (FE) basis and execution time, for accurate results as well as the multi 1-D thermal-hydraulics (T/H) model developed to take into account 3-D effects in updating the cross-sections. A Main Steam Line Break (MSLB) End-of-Life (EOL) Hot-Zero-Power (HZP) accident will be used as an example, before we conclude with the perspectives of integrating CRONOS's 3-D core model into real-time simulators. (author)
Feasibility of the integration of CRONOS, a 3-D neutronics code, into real-time simulators
Energy Technology Data Exchange (ETDEWEB)
Ragusa, J.C. [CEA Saclay, Dept. de Mecanique et de Technologie, 91 - Gif-sur-Yvette (France)
2001-07-01
In its effort to contribute to nuclear power plant safety, CEA proposes the integration of an engineering grade 3-D neutronics code into a real-time plant analyser. This paper describes the capabilities of the neutronics code CRONOS to achieve a fast running performance. First, we will present current core models in simulators and explain their drawbacks. Secondly, the mean features of CRONOS's spatial-kinetics methods will be reviewed. We will then present an optimum core representation with respect to mesh size, choice of finite elements (FE) basis and execution time, for accurate results as well as the multi 1-D thermal-hydraulics (T/H) model developed to take into account 3-D effects in updating the cross-sections. A Main Steam Line Break (MSLB) End-of-Life (EOL) Hot-Zero-Power (HZP) accident will be used as an example, before we conclude with the perspectives of integrating CRONOS's 3-D core model into real-time simulators. (author)
GPU-accelerated 3D neutron diffusion code based on finite difference method
International Nuclear Information System (INIS)
Finite difference method, as a traditional numerical solution to neutron diffusion equation, although considered simpler and more precise than the coarse mesh nodal methods, has a bottle neck to be widely applied caused by the huge memory and unendurable computation time it requires. In recent years, the concept of General-Purpose computation on GPUs has provided us with a powerful computational engine for scientific research. In this study, a GPU-Accelerated multi-group 3D neutron diffusion code based on finite difference method was developed. First, a clean-sheet neutron diffusion code (3DFD-CPU) was written in C++ on the CPU architecture, and later ported to GPUs under NVIDIA's CUDA platform (3DFD-GPU). The IAEA 3D PWR benchmark problem was calculated in the numerical test, where three different codes, including the original CPU-based sequential code, the HYPRE (High Performance Pre-conditioners)-based diffusion code and CITATION, were used as counterpoints to test the efficiency and accuracy of the GPU-based program. The results demonstrate both high efficiency and adequate accuracy of the GPU implementation for neutron diffusion equation. A speedup factor of about 46 times was obtained, using NVIDIA's Geforce GTX470 GPU card against a 2.50 GHz Intel Quad Q9300 CPU processor. Compared with the HYPRE-based code performing in parallel on an 8-core tower server, the speedup of about 2 still could be observed. More encouragingly, without any mathematical acceleration technology, the GPU implementation ran about 5 times faster than CITATION which was speeded up by using the SOR method and Chebyshev extrapolation technique. (authors)
DYN3D/M2 - a Code for Calculation of Reactivity Transients in Cores with Hexagonal Geometry
Rohde, Ulrich; Grundmann, Ulrich
2010-01-01
The code DYN3D/M2 consists of a the 3-dimensional neutron kinetic model of the code HEXDYN3D and the thermohydraulic model of the code FLOCAL. The neutron kinetics of DYN3D/M2 is calculated by using a nodal expansion method (NEM) for hexagonal geometry. The developed method solves the neutron diffusion equation for two energy groups. Stationary state and transient behaviour can be calculated. By help of the code PREPAR-EC parameterizid neutron physical constants of given burnup distribution c...
Adaptation of Zerotrees Using Signed Binary Digit Representations for 3D Image Coding
Directory of Open Access Journals (Sweden)
Mailhes Corinne
2007-01-01
Full Text Available Zerotrees of wavelet coefficients have shown a good adaptability for the compression of three-dimensional images. EZW, the original algorithm using zerotree, shows good performance and was successfully adapted to 3D image compression. This paper focuses on the adaptation of EZW for the compression of hyperspectral images. The subordinate pass is suppressed to remove the necessity to keep the significant pixels in memory. To compensate the loss due to this removal, signed binary digit representations are used to increase the efficiency of zerotrees. Contextual arithmetic coding with very limited contexts is also used. Finally, we show that this simplified version of 3D-EZW performs almost as well as the original one.
International Nuclear Information System (INIS)
Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the 'user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users [1]. Five seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004), at University of Zagreb (2005) and at the School of Industrial Engineering of Barcelona (2006). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The 3D S.UN.COP 2006 was successfully held with the attendance of 33 participants coming from 18 countries and 28 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 30 scientists (coming from 13 countries and 23 different institutions) were
International Nuclear Information System (INIS)
Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysis to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users. Six seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004), at University of Zagreb (2005), at the School of Industrial Engineering of Barcelona (January-February 2006) and in Buenos Aires, Argentina (October 2006), being this last one requested by ARN (Autoridad Regulatoria Nuclear), NA-SA (Nucleoelectrica Argentina S.A) and CNEA (Comision Nacional de Energia Atomica). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The 3D S.UN.COP 2006 in Barcelona was successfully held with the attendance of 33
FURN3D: A computer code for radiative heat transfer in pulverized coal furnaces
Energy Technology Data Exchange (ETDEWEB)
Ahluwalia, R.K.; Im, K.H.
1992-08-01
A computer code FURN3D has been developed for assessing the impact of burning different coals on heat absorption pattern in pulverized coal furnaces. The code is unique in its ability to conduct detailed spectral calculations of radiation transport in furnaces fully accounting for the size distributions of char, soot and ash particles, ash content, and ash composition. The code uses a hybrid technique of solving the three-dimensional radiation transport equation for absorbing, emitting and anisotropically scattering media. The technique achieves an optimal mix of computational speed and accuracy by combining the discrete ordinate method (S{sub 4}), modified differential approximation (MDA) and P, approximation in different range of optical thicknesses. The code uses spectroscopic data for estimating the absorption coefficients of participating gases C0{sub 2}, H{sub 2}0 and CO. It invokes Mie theory for determining the extinction and scattering coefficients of combustion particulates. The optical constants of char, soot and ash are obtained from dispersion relations derived from reflectivity, transmissivity and extinction measurements. A control-volume formulation is adopted for determining the temperature field inside the furnace. A simple char burnout model is employed for estimating heat release and evolution of particle size distribution. The code is written in Fortran 77, has modular form, and is machine-independent. The computer memory required by the code depends upon the number of grid points specified and whether the transport calculations are performed on spectral or gray basis.
FURN3D: A computer code for radiative heat transfer in pulverized coal furnaces
Energy Technology Data Exchange (ETDEWEB)
Ahluwalia, R.K.; Im, K.H.
1992-08-01
A computer code FURN3D has been developed for assessing the impact of burning different coals on heat absorption pattern in pulverized coal furnaces. The code is unique in its ability to conduct detailed spectral calculations of radiation transport in furnaces fully accounting for the size distributions of char, soot and ash particles, ash content, and ash composition. The code uses a hybrid technique of solving the three-dimensional radiation transport equation for absorbing, emitting and anisotropically scattering media. The technique achieves an optimal mix of computational speed and accuracy by combining the discrete ordinate method (S[sub 4]), modified differential approximation (MDA) and P, approximation in different range of optical thicknesses. The code uses spectroscopic data for estimating the absorption coefficients of participating gases C0[sub 2], H[sub 2]0 and CO. It invokes Mie theory for determining the extinction and scattering coefficients of combustion particulates. The optical constants of char, soot and ash are obtained from dispersion relations derived from reflectivity, transmissivity and extinction measurements. A control-volume formulation is adopted for determining the temperature field inside the furnace. A simple char burnout model is employed for estimating heat release and evolution of particle size distribution. The code is written in Fortran 77, has modular form, and is machine-independent. The computer memory required by the code depends upon the number of grid points specified and whether the transport calculations are performed on spectral or gray basis.
Calculation of SPERT Reactor benchmarks using 3D diffusion code DIREN
International Nuclear Information System (INIS)
The three dimensional diffusion code DIREN was developed at Institute for Nuclear Research (INR) Pitesti for reactor physics calculations for natural uranium and advanced CANDU reactors. Cell codes used are WIMS (from NEA library) and DRAGON (available in open source system). The latter is used also for super cell modeling of reactor control devices. These codes and the auxiliary programs were linked together in a calculation system. In order to apply WIMS-DRAGON-DIREN system to LWR, first the reactor SPERT benchmarks problems were calculated. The core including the control rods was modeled in three dimensional geometry. Following the calculations of the critical height (Hcrit), three dimensional power and flux distributions were obtained. The standard procedure used for CANDU reactor calculations (incremental cross sections for reactivity devices) underestimated the worth of control rods. A simple procedure to obtain the internal boundary conditions was developed using the super cell code DRAGON. Also the DIREN 3D diffusion code was modified to apply inner boundary conditions at control rods assigned volumes. Applying the inner boundary conditions yielded results closer to the measured values (e.g. the measured Hcrit was 49.53 cm as compared to 53.15 cm, the calculated one on 7 groups for nominal temperature). The reactivity coefficients for temperature and density required in transient's simulations were also calculated. The sample test problem T83 (hot stand-by, fast transient) was simulated using the RELAP code. (authors)
International Nuclear Information System (INIS)
The Light Water Reactor (LWR) dynamics code DYN3D is extended and adopted for the application to block-type High Temperature Gas-Cooled Reactor (HTGR). A procedure for the cross section generation for the HTGR core calculations was developed. The modified Reactivity-Equivalent Physical Transformation (RPT) approach is applied in order to eliminate the double-heterogeneity of HTGR fuel elements in the deterministic lattice calculations. A full core analysis of the reference simplified HTGR core is performed with DYN3D using macroscopic nodal cross sections provided by HELIOS. The SP3 transport approximation is integrated into the multi-group DYN3D code to take anisotropy of the neutron flux and heterogeneity of the core more precisely into account. The SP3 method was developed for hexagonal geometry of the graphite blocks, where the hexagons are subdivided into triangular nodes. A 3D heat conduction module coupled with a channel-type coolant flow model is implemented into the code. It is shown that there is significant redistribution of the produced heat by heat conduction between the graphite blocks. (orig.)
New adaptive differencing strategy in the PENTRAN 3-d parallel Sn code
International Nuclear Information System (INIS)
It is known that three-dimensional (3-D) discrete ordinates (Sn) transport problems require an immense amount of storage and computational effort to solve. For this reason, parallel codes that offer a capability to completely decompose the angular, energy, and spatial domains among a distributed network of processors are required. One such code recently developed is PENTRAN, which iteratively solves 3-D multi-group, anisotropic Sn problems on distributed-memory platforms, such as the IBM-SP2. Because large problems typically contain several different material zones with various properties, available differencing schemes should automatically adapt to the transport physics in each material zone. To minimize the memory and message-passing overhead required for massively parallel Sn applications, available differencing schemes in an adaptive strategy should also offer reasonable accuracy and positivity, yet require only the zeroth spatial moment of the transport equation; differencing schemes based on higher spatial moments, in spite of their greater accuracy, require at least twice the amount of storage and communication cost for implementation in a massively parallel transport code. This paper discusses a new adaptive differencing strategy that uses increasingly accurate schemes with low parallel memory and communication overhead. This strategy, implemented in PENTRAN, includes a new scheme, exponential directional averaged (EDA) differencing
Depth map coding using residual segmentation for 3D video system
Lee, Cheon; Ho, Yo-Sung
2013-06-01
Advanced 3D video systems employ multi-view video-plus-depth data to support the free-viewpoint navigation and comfortable 3D viewing; thus efficient depth map coding becomes an important issue. Unlike the color image, the depth map has a property that depth values of the inner part of an object are monotonic, but those of object boundaries change abruptly. Therefore, residual data generated by prediction errors around object boundaries consume many bits in depth map coding. Representing them with segment data can be better than the use of the conventional transformation around the boundary regions. In this paper, we propose an efficient depth map coding method using a residual segmentation instead of using transformation. The proposed residual segmentation divides residual data into two regions with a segment map and two mean values. If the encoder selects the proposed method in terms of rates, two quantized mean values and an index of the segment map are transmitted. Simulation results show significant gains of up to 10 dB compared to the state-of-the-art coders, such as JPEG2000 and H.264/AVC. [Figure not available: see fulltext.
International Nuclear Information System (INIS)
Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the 'user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users (D'Auria, 1998). Four seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004) and at University of Zagreb (2005). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The 3D S.UN.COP 2005 was successfully held with the participation of 19 persons coming from 9 countries and 14 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 15 scientists were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and
Spatial Parallelism of a 3D Finite Difference, Velocity-Stress Elastic Wave Propagation Code
Energy Technology Data Exchange (ETDEWEB)
MINKOFF,SUSAN E.
1999-12-09
Finite difference methods for solving the wave equation more accurately capture the physics of waves propagating through the earth than asymptotic solution methods. Unfortunately. finite difference simulations for 3D elastic wave propagation are expensive. We model waves in a 3D isotropic elastic earth. The wave equation solution consists of three velocity components and six stresses. The partial derivatives are discretized using 2nd-order in time and 4th-order in space staggered finite difference operators. Staggered schemes allow one to obtain additional accuracy (via centered finite differences) without requiring additional storage. The serial code is most unique in its ability to model a number of different types of seismic sources. The parallel implementation uses the MP1 library, thus allowing for portability between platforms. Spatial parallelism provides a highly efficient strategy for parallelizing finite difference simulations. In this implementation, one can decompose the global problem domain into one-, two-, and three-dimensional processor decompositions with 3D decompositions generally producing the best parallel speed up. Because i/o is handled largely outside of the time-step loop (the most expensive part of the simulation) we have opted for straight-forward broadcast and reduce operations to handle i/o. The majority of the communication in the code consists of passing subdomain face information to neighboring processors for use as ''ghost cells''. When this communication is balanced against computation by allocating subdomains of reasonable size, we observe excellent scaled speed up. Allocating subdomains of size 25 x 25 x 25 on each node, we achieve efficiencies of 94% on 128 processors. Numerical examples for both a layered earth model and a homogeneous medium with a high-velocity blocky inclusion illustrate the accuracy of the parallel code.
Steam generator experiment for 3-D computer code qualification - CLOTAIRE international program
International Nuclear Information System (INIS)
The current 1988/89 test program does focus on the production of accurate data sets dedicated to the qualifications of both 3-D thermalhydraulic codes and flow induced vibration predictive tools. In order to meet these challenging objectives the test program includes: detailed measurements of two-phase flow distributions relying on advanced optical probe techniques, throughout the bundle straight part; investigations at the same time of flow distributions and of the tubes' vibratory responses, in the U-band region; for a limited number of preselected positions, measurements of the emulsion's changing characteristics during transient sequences similar to those in an actual plant. (orig./DG)
Lebreton, Pierre; Raake, Alexander; Barkowsky, Marcus; Le Callet, Patrick
2011-01-01
This paper describes the results of a subjective test to assess current technology used for 3DTV broadcasting. As a first aspect, the performance of the currently deployed coding schemes was compared to state of the art algorithms. Our results show that downsampling and packing 3D stereoscopic videos according to the so called Side-By-Side format gives the highest perceived quality for a given bitrate. The second aspect of the study was to investigate how common 2D error concealment algorithm...
World's first ABWR start-up test analysis with 3-D transient computational code
International Nuclear Information System (INIS)
The Kashiwazaki-Kariwa Nuclear Power Station Unit 6, the world's first Advanced BWR (ABWR), began commercial operation from November 1996 following one year of start-up tests. A large number of variables which may be used to validate the advanced design features were obtained from transient tests. These test data are now being used for the qualification of TRACG, a BWR 3-D transient analysis code. Calculated results show that TRACG is fully capable of accurately predicting ABWR transient response and will be useful for application to future plant designs
3-D localization of gamma ray sources with coded apertures for medical applications
Kaissas, I.; Papadimitropoulos, C.; Karafasoulis, K.; Potiriadis, C.; Lambropoulos, C. P.
2015-09-01
Several small gamma cameras for radioguided surgery using CdTe or CdZnTe have parallel or pinhole collimators. Coded aperture imaging is a well-known method for gamma ray source directional identification, applied in astrophysics mainly. The increase in efficiency due to the substitution of the collimators by the coded masks renders the method attractive for gamma probes used in radioguided surgery. We have constructed and operationally verified a setup consisting of two CdTe gamma cameras with Modified Uniform Redundant Array (MURA) coded aperture masks of rank 7 and 19 and a video camera. The 3-D position of point-like radioactive sources is estimated via triangulation using decoded images acquired by the gamma cameras. We have also developed code for both fast and detailed simulations and we have verified the agreement between experimental results and simulations. In this paper we present a simulation study for the spatial localization of two point sources using coded aperture masks with rank 7 and 19.
Automated design of coupled RF cavities using 2-D and 3-D codes
International Nuclear Information System (INIS)
Coupled RF cavities in the Accelerator Production of Tritium Project have been designed using a procedure in which a 2-D code (CCT) searches for a design that meets frequency and coupling requirements, while a 3-D code (HFSS) is used to obtain empirical factors used by CCT to characterize the coupling slot between cavities. Using assumed values of the empirical factors, CCT runs the Superfish code iteratively to solve for a trial cavity design that has a specified frequency and coupling. The frequency shifts and the coupling constant k of the slot are modeled in CCT using a perturbation theory, the results of which are adjusted using the empirical factors. Given a trial design, HFSS is run using periodic boundary conditions to obtain a mode spectrum. The mode spectrum is processed using the DISPER code to obtain values of the coupling and the frequencies with slots. These results are used to calculate a new set of empirical factors, which are fed back into CCT for another design iteration. Cold models have been fabricated and tested to validate the codes, and results will be presented.
Assessment of void fraction prediction using the RETRAN-3d and CORETRAN-01/VIPRE-02 codes
International Nuclear Information System (INIS)
A review of wide-range void fraction correlations against an extensive database has been undertaken to identify the correlations best suited for nuclear safety applications. Only those based on the drift-flux model have been considered. The survey confirmed the application range of the Chexal-Lellouche correlation, and the database was also used to obtain new parameters for the Inoue drift-flux correlation, which was also found suitable. A void fraction validation study has also been undertaken for the codes RETRAN-3D and CORETRAN-01/VIPRE-02 at the assembly and sub-assembly levels. The study showed the impact of the RETRAN-03 user options on the predicted void fraction, and the RETRAN-3D limitation at very low fluid velocity. At the sub-assembly level, CORETRAN-01/VIPRE-02 substantially underestimates the void in regions with low power-to-flow ratios. Otherwise, a generally good predictive performance was obtained with both RETRAN-3D and CORETRAN-01/VIPRE-02. (authors)
Development of 3D multi-group neutron diffusion code for hexagonal geometry
International Nuclear Information System (INIS)
Based on the theory of new flux expansion nodal method to solve the neutron diffusion equations, the intra-nodal fluence rate distribution was expanded in a series of analytic basic functions for each group. In order to improve the accuracy of calculation result, continuities of neutron fluence rate and current were utilized across the nodal surfaces. According to the boundary conditions, the iteration method was adopted to solve the diffusion equation, where inner iteration speedup method is Gauss-Seidel method and outer is Lyusternik-Wagner. A new speedup method (one-outer-iteration and multi-inner-iteration method) was proposed according to the characteristic that the convergence speed of multiplication factor is faster than that of neutron fluence rate and the update of inner iteration matrix is slow. Based on the proposed model, the code HANDF-D was developed and tested by 3D two-group vver440 benchmark, experiment 2 of HFETR, 3D four-group thermal reactor benchmark, and 3D seven-group fast reactor benchmark. The numerical results show that HANDF-D can predict accurately the multiplication factor and nodal powers. (authors)
Assessment of void fraction prediction using the RETRAN-3d and CORETRAN-01/VIPRE-02 codes
Energy Technology Data Exchange (ETDEWEB)
Aounallah, Y.; Coddington, P.; Gantner, U
2000-07-01
A review of wide-range void fraction correlations against an extensive database has been undertaken to identify the correlations best suited for nuclear safety applications. Only those based on the drift-flux model have been considered. The survey confirmed the application range of the Chexal-Lellouche correlation, and the database was also used to obtain new parameters for the Inoue drift-flux correlation, which was also found suitable. A void fraction validation study has also been undertaken for the codes RETRAN-3D and CORETRAN-01/VIPRE-02 at the assembly and sub-assembly levels. The study showed the impact of the RETRAN-03 user options on the predicted void fraction, and the RETRAN-3D limitation at very low fluid velocity. At the sub-assembly level, CORETRAN-01/VIPRE-02 substantially underestimates the void in regions with low power-to-flow ratios. Otherwise, a generally good predictive performance was obtained with both RETRAN-3D and CORETRAN-01/VIPRE-02. (authors)
FERM3D: A finite element R-matrix electron molecule scattering code
Tonzani, S
2006-01-01
FERM3D is a three-dimensional finite element program, for the elastic scattering of a low energy electron from a general polyatomic molecule, which is converted to a potential scattering problem. The code is based on tricubic polynomials in spherical coordinates. The electron-molecule interaction is treated as a sum of three terms: electrostatic, exchange. and polarisation. The electrostatic term can be extracted directly from ab initio codes ({\\sc{GAUSSIAN 98}} in the work described here), while the exchange term is approximated using a local density functional. A local polarisation potential based on density functional theory [C. Lee, W. Yang and R. G. Parr, {Phys. Rev. B} {37}, (1988) 785] describes the long range attraction to the molecular target induced by the scattering electron. Photoionisation calculations are also possible and illustrated in the present work. The generality and simplicity of the approach is important in extending electron-scattering calculations to more complex targets than it is po...
Modelling magnetic fields diagnostic coils using a 3D free-boundary equilibrium code
International Nuclear Information System (INIS)
A project to interpret the magnetic field diagnostics of the W VII-X stellarator is summarized. The NEMEC free-boundary equilibrium code is used to calculate 3D ideal-MHD equilibria which are consistent with the fields due to external currents. The signals of the diagnostic coils are related to the plasma equilibrium by combining the NEMEC code with a technique for calculating the magnetic field outside the plasma due to the plasma currents alone. These techniques will be used to design the diagnostic coils on the W VII-X device. Test run results are shown. The arrow plot for a beta value of 0.9 % shows the characteristic dipole-like field of the plasma currents. The signals of three flux loops as a function of beta produce curves which are quite smooth for moderate beta values
A 3D multi-block structured version of the KIVA 2 code
Habachi, C.; Torres, A.
A numerical procedure is developed in the KIVA 2 code for calculating flows in complex geometries. Those geometries consist of an arbitrary number of 3D secondary domains which are connected with any angle to a main region. In this procedure, the governing equations are discretized on a system of partial overlapping structured grids. Calculations are performed in the different meshes of the computation domain which are linked by a fully conservative algorithm. By this numerical technique, calculations in those geometries are possible with a reasonable number of inactive cells involved by a structured code like KIVA 2. This algorithm was validated on an 1D analytical case and a 2D experimental case. It was then used for modeling an industrial problem, a two stroke engine with ports and moving boundaries.
The multigroup neutronics model of NuStar's 3D core code EGRET
International Nuclear Information System (INIS)
As a key component of NuStar's core analysis system for PWR application, EGRET is designed to perform steady-state coupled neutronic/hydraulic analysis of PWRs. This paper presents EGRET's unique 3D nodal diffusion model and 2D pin power reconstruction (PPR) model. Unlike the practice in most of today's production codes that iteratively solves the global 3D coarse-mesh problem and the local axially 1D fine-mesh problem to handle the axial heterogeneity within a node caused by fuel grid and partially-inserted control rod, EGRET resolves the issue by inventing a new nodal technology and introducing the adaptive meshing technique to follow the movement of control rod tip. The new nodal method employs fine-mesh heterogeneous calculation with coarse-mesh transverse coupling such that the axial heterogeneous nodes can be explicitly modeled in exact geometry and directly incorporated into the scheme of transversely coupled coarse-mesh nodal methods. Each axial channel can have its own fine-mesh division without the need of dividing the whole core into radially coupled fine-meshes. There is no need to do 1D fine-mesh and 3D coarse-mesh iteration either. While for the PPR model, EGRET adopts a group-decoupled direct fitting method, which avoids both the complication of constructing 2D analytic multigroup flux solution and any group-coupled iteration. Another unique feature of the PPR model is that it fully utilizes all the information available from 3D core calculation into the downstream PPR process. Particularly, for the first time, the 1D profiles of transversely-integrated fluxes are utilized as the additional conditions to reconstruct pin power. Numerical results of series of benchmark problems verify the good performance of EGRET's unique multi-group neutronics model. (author)
Radiation Coupling with the FUN3D Unstructured-Grid CFD Code
Wood, William A.
2012-01-01
The HARA radiation code is fully-coupled to the FUN3D unstructured-grid CFD code for the purpose of simulating high-energy hypersonic flows. The radiation energy source terms and surface heat transfer, under the tangent slab approximation, are included within the fluid dynamic ow solver. The Fire II flight test, at the Mach-31 1643-second trajectory point, is used as a demonstration case. Comparisons are made with an existing structured-grid capability, the LAURA/HARA coupling. The radiative surface heat transfer rates from the present approach match the benchmark values within 6%. Although radiation coupling is the focus of the present work, convective surface heat transfer rates are also reported, and are seen to vary depending upon the choice of mesh connectivity and FUN3D ux reconstruction algorithm. On a tetrahedral-element mesh the convective heating matches the benchmark at the stagnation point, but under-predicts by 15% on the Fire II shoulder. Conversely, on a mixed-element mesh the convective heating over-predicts at the stagnation point by 20%, but matches the benchmark away from the stagnation region.
The MICHELLE 2D/3D ES PIC Code Advances and Applications
Petillo, John; De Ford, John F; Dionne, Norman J; Eppley, Kenneth; Held, Ben; Levush, Baruch; Nelson, Eric M; Panagos, Dimitrios; Zhai, Xiaoling
2005-01-01
MICHELLE is a new 2D/3D steady-state and time-domain particle-in-cell (PIC) code* that employs electrostatic and now magnetostatic finite-element field solvers. The code has been used to design and analyze a wide variety of devices that includes multistage depressed collectors, gridded guns, multibeam guns, annular-beam guns, sheet-beam guns, beam-transport sections, and ion thrusters. Latest additions to the MICHELLE/Voyager tool are as follows: 1) a prototype 3D self magnetic field solver using the curl-curl finite-element formulation for the magnetic vector potential, employing edge basis functions and accumulating current with MICHELLE's new unstructured grid particle tracker, 2) the electrostatic field solver now accommodates dielectric media, 3) periodic boundary conditions are now functional on all grids, not just structured grids, 4) the addition of a global optimization module to the user interface where both electrical parameters (such as electrode voltages)can be optimized, and 5) adaptive mesh ref...
Inogamov, Nail A.; Zhakhovsky, Vasily V.
2016-02-01
There are many important applications in which the ultrashort diffraction-limited and therefore tightly focused laser pulses irradiates metal films mounted on dielectric substrate. Here we present the detailed picture of laser peeling and 3D structure formation of the thin (relative to a depth of a heat affected zone in the bulk targets) gold films on glass substrate. The underlying physics of such diffraction-limited laser peeling was not well understood previously. Our approach is based on a physical model which takes into consideration the new calculations of the two-temperature (2T) equation of state (2T EoS) and the two-temperature transport coefficients together with the coupling parameter between electron and ion subsystems. The usage of the 2T EoS and the kinetic coefficients is required because absorption of an ultrashort pulse with duration of 10-1000 fs excites electron subsystem of metal and transfers substance into the 2T state with hot electrons (typical electron temperatures 1-3 eV) and much colder ions. It is shown that formation of submicrometer-sized 3D structures is a result of the electron-ion energy transfer, melting, and delamination of film from substrate under combined action of electron and ion pressures, capillary deceleration of the delaminated liquid metal or semiconductor, and ultrafast freezing of molten material. We found that the freezing is going in non-equilibrium regime with strongly overcooled liquid phase. In this case the Stefan approximation is non-applicable because the solidification front speed is limited by the diffusion rate of atoms in the molten material. To solve the problem we have developed the 2T Lagrangian code including all this reach physics in. We also used the high-performance combined Monte- Carlo and molecular dynamics code for simulation of surface 3D nanostructuring at later times after completion of electron-ion relaxation.
Energy Technology Data Exchange (ETDEWEB)
Rohde, Ulrich; Apanasevich, Pavel; Baier, Silvio; Duerigen, Susan; Fridman, Emil; Grahn, Alexander; Kliem, Soeren; Merk, Bruno
2012-07-15
Based on the reactor dynamics code DYN3D for the simulation of transient processes in Light Water Reactors, a code version DYN3D-HTR for application to graphitemoderated, gas-cooled block-type high temperature reactors has been developed. This development comprises: - the methodical improvement of the 3D steady-state neutron flux calculation for the hexagonal geometry of the HTR fuel element blocks - the development of methods for the generation of homogenised cross section data taking into account the double heterogeneity of the fuel element block structure - the implementation of a 3D model for heat conduction and heat transport in the graphite matrix. The nodal method for neutron flux calculation based on SP3 transport approximation was extended to hexagonal fuel element geometry, where the hexagons are subdivided into triangles, thus the method had finally to be derived for triangular geometry. In triangular geometry, a subsequent subdivision of the hexagonal elements can be considered, and therefore, the effect of systematic mesh refinement can be studied. The algorithm was verified by comparison with Monte Carlo reference solutions, on the node-wise level, as well as also on the pin-wise level. New procedures were developed for the homogenization of the double-heterogeneous fuel element structures. One the one hand, the so-called Reactivity equivalent Physical Transformation (RPT), the two-step homogenization method based on 2D deterministic lattice calculations, was extended to cells with different temperatures of the materials. On the other hand, the progress in development of Monte Carlo methods for spectral calculations, in particular the development of the code SERPENT, opened a new, fully consistent 3D approach, where all details of the structures on fuel particle, fuel compact and fuel block level can be taken into account within one step. Moreover, a 3D heat conduction and heat transport model was integrated into DYN3D to be able to simulate radial
3-D TECATE/BREW: Thermal, stress, and birefringent ray-tracing codes for solid-state laser design
International Nuclear Information System (INIS)
This report describes the physics, code formulations, and numerics that are used in the TECATE (totally Eulerian code for anisotropic thermo-elasticity) and BREW (birefringent ray-tracing of electromagnetic waves) codes for laser design. These codes resolve thermal, stress, and birefringent optical effects in 3-D stationary solid-state systems. This suite of three constituent codes is a package referred to as LASRPAK
Development and preliminary verification of the 3D core neutronic code: COCO
International Nuclear Information System (INIS)
As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code, the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)
RBMK fuel channel blockage analysis by MCNP5, DRAGON and RELAP5-3D codes
International Nuclear Information System (INIS)
The aim of this work was to perform precise criticality analyses by Monte-Carlo code MCNP5 for a Fuel Channel (FC) flow blockage accident, considering as calculation domain a single FC and a 3x3 lattice of RBMK cells. Boundary conditions for MCNP5 input were derived by a previous transient calculation by state-of-the-art codes HELIOS/RELAP5-3D. In a preliminary phase, suitable MCNP5 models of a single cell and of a small lattice of RBMK cells were set-up; criticality analyses were performed at reference conditions for 2.0% and 2.4% enriched fuel. These analyses were compared with results obtained by University of Pisa (UNIPI) using deterministic transport code DRAGON and with results obtained by NIKIET Institute using MCNP4C. Then, the changes of the main physical parameters (e.g. fuel and water/steam temperature, water density, graphite temperature) at different time intervals of the FC blockage transient were evaluated by a RELAP5-3D calculation. This information was used to set up further MCNP5 inputs. Criticality analyses were performed for different systems (single channel and lattice) at those transient' states, obtaining global criticality versus transient time. Finally the weight of each parameter's change (fuel overheating and channel voiding) on global criticality was assessed. The results showed that reactivity of a blocked FC is always negative; nevertheless, when considering the effect of neighboring channels, the global reactivity trend reverts, becoming slightly positive or not changing at all, depending in inverse relation to the fuel enrichment. (author)
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The paper describes the successfully performed verification tests with the ATLAS simulator environment and the coupled QUABOX/CUBBOX-ATHLET code system with enhanced option of switching from point kinetics (PK) to 3D calculations. (authors)
Calibration of the CAFE-3D fire code with controlled indoor fire data
International Nuclear Information System (INIS)
The Container Analysis Fire Environment (CAFE) code contains a computational fluid dynamics (CFD) based fire model that has been successfully coupled to standard finite element computer codes. This coupling of CFD and finite element codes allows for a more realistic modeling of the thermal performance of objects engulfed in fire, which aids in the design and risk analysis of radioactive material packages. The CAFE fire model is based on a three-dimensional finite volume formulation of basic fire chemistry and fluid dynamics. This fire model includes a variable-density primitive-variable formulation of mass, momentum, energy and species equations. Multiple chemical species and soot formation are included in the combustion model. Thermal radiation is modeled as diffusive radiation transport inside the flame zone and as view-factor radiation outside the flame zone. Turbulence is modeled with an eddy diffusivity model. The soot model is coupled to the diffusive radiation formulation using the Rosseland approximation and the optical properties of soot. In order to verify and improve the accuracy of computers codes, they should be benchmarked against test data. This paper describes a set of experiments that were performed at the Fire Laboratory for Accreditation of Modeling by Experiment (FLAME) fire facility of Sandia National Laboratories in Albuquerque, New Mexico, USA. The paper also describes how the data collected from the experiments was used to calibrate and benchmark the CAFE-3D fire code. Detailed description of the tests performed and comparisons between the calculated results and the collected data from the experiments are provided
Conti, Caroline; Nunes, Paulo; Ducla Soares, Luís.
2013-09-01
Holoscopic imaging, also known as integral imaging, has been recently attracting the attention of the research community, as a promising glassless 3D technology due to its ability to create a more realistic depth illusion than the current stereoscopic or multiview solutions. However, in order to gradually introduce this technology into the consumer market and to efficiently deliver 3D holoscopic content to end-users, backward compatibility with legacy displays is essential. Consequently, to enable 3D holoscopic content to be delivered and presented on legacy displays, a display scalable 3D holoscopic coding approach is required. Hence, this paper presents a display scalable architecture for 3D holoscopic video coding with a three-layer approach, where each layer represents a different level of display scalability: Layer 0 - a single 2D view; Layer 1 - 3D stereo or multiview; and Layer 2 - the full 3D holoscopic content. In this context, a prediction method is proposed, which combines inter-layer prediction, aiming to exploit the existing redundancy between the multiview and the 3D holoscopic layers, with self-similarity compensated prediction (previously proposed by the authors for non-scalable 3D holoscopic video coding), aiming to exploit the spatial redundancy inherent to the 3D holoscopic enhancement layer. Experimental results show that the proposed combined prediction can improve significantly the rate-distortion performance of scalable 3D holoscopic video coding with respect to the authors' previously proposed solutions, where only inter-layer or only self-similarity prediction is used.
International Nuclear Information System (INIS)
The gas-cooled high temperature reactor is a concept to produce energy at high temperatures with a high level of inherent safety. It gets special attraction due to e.g. high thermal efficiency and the possibility of hydrogen production. In addition to the PBMR (Pebble Bed Modular Reactor) the (V)HTR (Very high temperature reactor) concept has been established. The basic design of a prismatic HTR consists of the following elements. The fuel is coated with four layers of isotropic materials. These so-called TRISO particles are dispersed into compacts which are placed in a graphite block matrix. The graphite matrix additionally contains holes for the coolant gas. A one-dimensional model is sufficient to describe (the radial) heat transfer in LWRs. But temperature gradients in a prismatic HTR can occur in axial as well as in radial direction, since regions with different heat source release and with different coolant temperature heat up are coupled through the graphite matrix elements. Furthermore heat transfer into reflector elements is possible. DYN3D is a code system for coupled neutron and thermal hydraulics core calculations developed at the Helmholtzzentrum Dresden-Rossendorf. Concerning neutronics DYN3D consists of a two-group and multi-group diffusion approach based on nodal expansion methods. Furthermore a 1D thermal-hydraulics model for parallel coolant flow channels is included. The DYN3D code was extensively verified and validated via numerous numerical and experimental benchmark problems. That includes the NEA CRP benchmarks for PWR and BWR, the Three-Miles-Island-1 main steam line break and the Peach Bottom Turbine Trip benchmarks, as well as measurements carried out in an original-size VVER-1000 mock-up. An overview of the verification and validation activities can be found. Presently a DYN3D-HTR version is under development. It involves a 3D heat conduction model to deal with higher-(than one)-dimensional effects of heat transfer and heat conduction in
Introducing ZEUS-MP A 3D, Parallel, Multiphysics Code for Astrophysical Fluid Dynamics
Norman, M L
2000-01-01
We describe ZEUS-MP: a Multi-Physics, Massively-Parallel, Message-Passing code for astrophysical fluid dynamics simulations in 3 dimensions. ZEUS-MP is a follow-on to the sequential ZEUS-2D and ZEUS-3D codes developed and disseminated by the Laboratory for Computational Astrophysics (lca.ncsa.uiuc.edu) at NCSA. V1.0 released 1/1/2000 includes the following physics modules: ideal hydrodynamics, ideal MHD, and self-gravity. Future releases will include flux-limited radiation diffusion, thermal heat conduction, two-temperature plasma, and heating and cooling functions. The covariant equations are cast on a moving Eulerian grid with Cartesian, cylindrical, and spherical polar coordinates currently supported. Parallelization is done by domain decomposition and implemented in F77 and MPI. The code is portable across a wide range of platforms from networks of workstations to massively parallel processors. Some parallel performance results are presented as well as an application to turbulent star formation.
A novel code for numerical 3-D MHD studies of CME expansion
Directory of Open Access Journals (Sweden)
J. Kleimann
2009-03-01
Full Text Available A recent third-order, essentially non-oscillatory central scheme to advance the equations of single-fluid magnetohydrodynamics (MHD in time has been implemented into a new numerical code. This code operates on a 3-D Cartesian, non-staggered grid, and is able to handle shock-like gradients without producing spurious oscillations.
To demonstrate the suitability of our code for the simulation of coronal mass ejections (CMEs and similar heliospheric transients, we present selected results from test cases and perform studies of the solar wind expansion during phases of minimum solar activity. We can demonstrate convergence of the system into a stable Parker-like steady state for both hydrodynamic and MHD winds. The model is subsequently applied to expansion studies of CME-like plasma bubbles, and their evolution is monitored until a stationary state similar to the initial one is achieved. In spite of the model's (current simplicity, we can confirm the CME's nearly self-similar evolution close to the Sun, thus highlighting the importance of detailed modelling especially at small heliospheric radii.
Additionally, alternative methods to implement boundary conditions at the coronal base, as well as strategies to ensure a solenoidal magnetic field, are discussed and evaluated.
Status and future of the 3D MAFIA group of codes
International Nuclear Information System (INIS)
This paper reports on the group of fully three dimensional computer codes for solving Maxwell's equations for a wide range of applications, MAFIA. Extensive comparisons with measurement have demonstrated the accuracy of the computations. A large number of components have been designed for accelerators, such as kicker magnets, non cylindrical cavities, ferrite loaded cavities, vacuum chambers with slots and transitions, etc. The latest additions to the system include a new static solver that can calculate 3D magneto- and electrostatic fields, and a self consistent version of the 2D-BCI that solves the field equations and the equations of motion in parallel. Work on new eddy current modules has started, which will allow treatment of laminated and/or solid iron cores excited by low frequency currents
Developments in the analysis of 3D piping and shells by means of PAULA code
International Nuclear Information System (INIS)
Non linear analyses of three dimensional piping and shells are becoming more and more common, in the safety analysis of nuclear power plants. The pipe whip accident, the Hypothetic core Distruptive Accident (HCDA) for LMFBR represent, two significative examples, where non linear analyses of the pressure boundary have been used with considerable success. Seismic analysis and other extreme loading of conditions are other cases, where non linear analyses have been used even if not extensively due to cost reasons. The authors have presented a code, named PAULA to deal with the 3D non linear analysis of piping; it is the aim of this paper to briefly describe the basic library of PAULA and to describe the new shell elements in some more detail. (orig./GL)
International Nuclear Information System (INIS)
Current work presents a new methodology which uses Serpent Monte-Carlo (MC) code for generating multi-group beginning-of-life (BOL) cross section (XS) database file that is compatible with PARCS 3D reactor core simulator and allows simulation of transients with the FAST code system. The applicability of the methodology was tested on European Sodium-cooled Fast Reactor (ESFR) design with an oxide fuel proposed by CEA (France). The k-effective, power peaking factors and safety parameters (such as Doppler constant, coolant density coefficient, fuel axial expansion coefficient, diagrid expansion coefficients and control rod worth) calculated by PARCS/TRACE were compared with the results of the Serpent MC code. The comparison indicates overall reasonable agreement between conceptually different (deterministic and stochastic) codes. The new development makes it in principle possible to use the Serpent MC code for cross section generation for the PARCS code to perform transient analyses for fast reactors. The advantages and limitations of this methodology are discussed in the paper. (author)
Calculation of an accident with delayed scram at NPP Greifswald using the coupled code DYN3D/ATHLET
Energy Technology Data Exchange (ETDEWEB)
Kliem, S.
1998-10-01
Complex computer codes modeling the whole reactor system including 3D neutron kinetics in combination with advanced thermohydraulic plant models become more and more important for the safety assessment of nuclear reactors. Transients or experiments with both neutron kinetic and thermalhydraulic data are needed for the validation of such coupled codes like DYN3D/ATHLET. First of all measured results from nuclear power plant (NPP) transients should be used, because the experimental thermalhydraulic facilities do not offer the possibility to model space-dependent neutron kinetic effects and research reactors with reliably measured 3D neutron kinetic data do not allow to study thermalhydraulic feedback effects. In this paper, an accident with delayed scram which occurred in 1989 at the NPP Greifswald is analyzed. Calculations of this accident were carried out with the goal to validate the coupled code DYN3D/ATHLET. (orig.)
International Nuclear Information System (INIS)
This document is a user's manual for the Rn3D finite element code. Rn3D was developed to simulate gas flow and radon transport in variably saturated, nonisothermal porous media. The Rn3D model is applicable to a wide range of problems involving radon transport in soil because it can simulate either steady-state or transient flow and transport in one-, two- or three-dimensions (including radially symmetric two-dimensional problems). The porous materials may be heterogeneous and anisotropic. This manual describes all pertinent mathematics related to the governing, boundary, and constitutive equations of the model, as well as the development of the finite element equations used in the code. Instructions are given for constructing Rn3D input files and executing the code, as well as a description of all output files generated by the code. Five verification problems are given that test various aspects of code operation, complete with example input files, FORTRAN programs for the respective analytical solutions, and plots of model results. An example simulation is presented to illustrate the type of problem Rn3D is designed to solve. Finally, instructions are given on how to convert Rn3D to simulate systems other than radon, air, and water
Codes complex for quick transport 3D neutron calculations of WWER
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: more numbers of groups, more high approximations for angle distributions. The code is used for carrying out 3D coarse mesh calculation. The 6F approximation of the Surface Harmonics Method is sufficient for WWER calculations. The complex Surface Values System is convenient for calculations of the core with the monitoring and control system and it is verificated for work with shell KRUIS, that used on some AES with the reactors WWER (Authors)
Status and future of the 3D MAFIA group of codes
Energy Technology Data Exchange (ETDEWEB)
Ebeling, F.; Klatt, R.; Krawzcyk, F.; Lawinsky, E.; Weiland, T.; Wipf, S.G.; Steffen, B.; Barts, T.; Browman, J.; Cooper, R.K.; and others
1988-12-01
The group of fully three dimensional computer codes for solving Maxwell's equations for a wide range of applications, MAFIA, is already well established. Extensive comparisons with measurements have demonstrated the accuracy of the computations. A large numer of components have been designed for accelerators, such as kicker magnets, non cyclindrical cavities, ferrite loaded cavities, vacuum chambers with slots and transitions, etc. The latest additions to the system include a new static solver that can calculate 3D magneto- and electrostatic fields, and a self consistent version of the 2D-BCI that solves the field equations and the equations of motion in parallel. Work on new eddy current modules has started, which will allow treatment of laminated and/or solid iron cores excited by low frequency currents. Based on our experience with the present releases 1 and 2, we have started a complete revision of the whole user interface and data structure, which will make the codes even more user-friendly and flexible.
3D-MAPTOR Code for Computation of Magnetic Fields in Tokamaks
International Nuclear Information System (INIS)
Full text: A 3D code has been developed in order to simulate the magnetic field lines in tokamaks, in two versions. In the first one, the toroidal magnetic field can be obtained from the individual fields of circular coils arranged around the torus, or alternatively, as a ripple-less field. The poloidal field is provided by a given toroidal current density profile. In an upgraded version, rectangular toroidal field coils and D-shaped plasma cross sections have been included, in order to aid in the design of spherical tokamaks. Proposing initial conditions for magnetic filed lines, they are integrated along the toroidal angle coordinate, and Poincare maps can be obtained at any desired cross section plane along the toroidal coordinate. The evolution of the field lines is also monitored from above, so the ripple due to the toroidal field coils can be appreciated. The effects of loss of axisymmetry, either originated by ripples, or by additional external coils, such as an inner coil with tilted circular loops, can therefore be studied. This is useful for the study of breaking-up of external surfaces, as in the case of ergodic divertors. The code can also be used in order to reconstruct the evolution of the plasma column, using the experimental signals of tokamak discharges. In the latter case, the results have been compared with tomographic results of the ISTTOK tokamak. (author)
Studies of Coupled Cavity Linac (CCL) Accelerating Structures with 3-D Codes
Spalek, G; Smith, P D; Greninger, P T; Charman, C M
2000-01-01
The cw CCL being designed for the Accelerator Production of Tritium (APT) project accelerates protons from 96MeV to 211MeV. It consists of 99 segments each containing up to seven accelerating cavities. Segments are coupled by intersegment coupling cavities and grouped into supermodules. The design method needs to address not only basic cavity sizing for a given coupling and pi/2 mode frequency, but also the effects of high power densities on the cavity frequency, mechanical stresses, and the structure's stop band during operation. On the APT project, 3-D RF (Ansoft Corp.'s HFSS) and coupled RF/structural (Ansys Inc.'s ANSYS) codes are being used to develop tools to address the above issues and guide cooling channel design. The code's predictions are being checked against available low power Aluminum models. Stop band behavior under power will be checked once the tools are extended to CCDTL structures that have been tested at high power. A summary of calculations made to date and agreement with measured result...
3D PiC code investigations of Auroral Kilometric Radiation mechanisms
International Nuclear Information System (INIS)
Efficient (∼1%) electron cyclotron radio emissions are known to originate in the X mode from regions of locally depleted plasma in the Earths polar magnetosphere. These emissions are commonly referred to as the Auroral Kilometric Radiation (AKR). AKR occurs naturally in these polar regions where electrons are accelerated by electric fields into the increasing planetary magnetic dipole. Here conservation of the magnetic moment converts axial to rotational momentum forming a horseshoe distribution in velocity phase space. This distribution is unstable to cyclotron emission with radiation emitted in the X-mode. Initial studies were conducted in the form of 2D PiC code simulations [1] and a scaled laboratory experiment that was constructed to reproduce the mechanism of AKR. As studies progressed, 3D PiC code simulations were conducted to enable complete investigation of the complex interaction dimensions. A maximum efficiency of 1.25% is predicted from these simulations in the same mode and frequency as measured in the experiment. This is also consistent with geophysical observations and the predictions of theory.
3D Measurement Technology by Structured Light Using Stripe-Edge-Based Gray Code
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The key problem of 3D vision measurement using triangle method based on structured light is to acquiring projecting angle of projecting light accurately. In order to acquire projecting angle thereby determine the corresponding relationship between sampling point and image point, method for encoding and decoding structured light based on stripe edge of Gray code is presented. The method encoded with Gray code stripe and decoded with stripe edge acquired by sub-pixel technology instead of pixel centre, so latter one-bit decoding error was removed. Accuracy of image sampling point location and correspondence between image sampling point and object sampling point achieved sub-pixel degree. In addition, measurement error caused by dividing projecting angle irregularly by even-width encoding stripe was analysed and corrected. Encoding and decoding principle and decoding equations were described. Finally, 3dsmax and Matlab software were used to simulate measurement system and reconstruct measured surface. Indicated by experimental results, measurement error is about 0.05%
Quantum Self-Correction in the 3D Cubic Code Model
Bravyi, Sergey; Haah, Jeongwan
2013-11-01
A big open question in the quantum information theory concerns the feasibility of a self-correcting quantum memory. A quantum state recorded in such memory can be stored reliably for a macroscopic time without need for active error correction, if the memory is in contact with a cold enough thermal bath. Here we report analytic and numerical evidence for self-correcting behavior in the quantum spin lattice model known as the 3D cubic code. We prove that its memory time is at least Lcβ, where L is the lattice size, β is the inverse temperature of the bath, and c>0 is a constant coefficient. However, this bound applies only if the lattice size L does not exceed a critical value which grows exponentially with β. In that sense, the model can be called a partially self-correcting memory. We also report a Monte Carlo simulation indicating that our analytic bounds on the memory time are tight up to constant coefficients. To model the readout step we introduce a new decoding algorithm, which can be implemented efficiently for any topological stabilizer code. A longer version of this work can be found in Bravyi and Haah, arXiv:1112.3252.
Spectral history modeling in the reactor dynamics code DYN3D
International Nuclear Information System (INIS)
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.
Studies of coupled cavity LINAC (CCL) accelerating structures with 3-D codes
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The cw CCL being designed for the Accelerator Production of Tritium (APT) project accelerates protons from 96 MeV to 211 MeV. It consists of 99 segments each containing up to seven accelerating cavities. Segments are coupled by intersegment coupling cavities and grouped into supermodules. The design method needs to address not only basic cavity sizing for a given coupling and pi/2 mode frequency, but also the effects of high power densities on the cavity frequency, mechanical stresses, and the structure's stop band during operation. On the APT project, 3-D RF (Ansoft Corp.'s HFSS) and coupled RF/structural (Ansys Inc.'s ANSYS) codes are being used. to develop tools to address the above issues and guide cooling channel design. The code's predictions are being checked against available low power Aluminum models. Stop band behavior under power will be checked once the tools are extended to CCDTL structures that have been tested at high power. A summary of calculations made to date and agreement with measured results will be presented
ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®
Damian, F.; Brun, E.
2014-06-01
ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14 MW. ORPHEE core is 90 cm height and has a cross section of 27x27 cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® [1,2,3]. A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 [4]. The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.
FERM3D: A finite element R-matrix electron molecule scattering code
Tonzani, Stefano
2007-01-01
FERM3D is a three-dimensional finite element program, for the elastic scattering of a low energy electron from a general polyatomic molecule, which is converted to a potential scattering problem. The code is based on tricubic polynomials in spherical coordinates. The electron-molecule interaction is treated as a sum of three terms: electrostatic, exchange, and polarization. The electrostatic term can be extracted directly from ab initio codes ( GAUSSIAN 98 in the work described here), while the exchange term is approximated using a local density functional. A local polarization potential based on density functional theory [C. Lee, W. Yang, R.G. Parr, Phys. Rev. B 37 (1988) 785] describes the long range attraction to the molecular target induced by the scattering electron. Photoionization calculations are also possible and illustrated in the present work. The generality and simplicity of the approach is important in extending electron-scattering calculations to more complex targets than it is possible with other methods. Program summaryTitle of program:FERM3D Catalogue identifier:ADYL_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/ADYL_v1_0 Program obtainable from: CPC Program Library, Queen's University of Belfast, N. Ireland Computer for which the program is designed and others on which it has been tested:Intel Xeon, AMD Opteron 64 bit, Compaq Alpha Operating systems or monitors under which the program has been tested:HP Tru64 Unix v5.1, Red Hat Linux Enterprise 3 Programming language used:Fortran 90 Memory required to execute with typical data:900 MB (neutral CO 2), 2.3 GB (ionic CO 2), 1.4 GB (benzene) No. of bits in a word:32 No. of processors used:1 Has the code been vectorized?:No No. of lines in distributed program, including test data, etc.:58 383 No. of bytes in distributed program, including test data, etc.:561 653 Distribution format:tar.gzip file CPC Program library subprograms used:ADDA, ACDP Nature of physical problem:Scattering of an
Zhang, Yujia; Yilmaz, Alper
2016-06-01
Surface reconstruction using coded structured light is considered one of the most reliable techniques for high-quality 3D scanning. With a calibrated projector-camera stereo system, a light pattern is projected onto the scene and imaged by the camera. Correspondences between projected and recovered patterns are computed in the decoding process, which is used to generate 3D point cloud of the surface. However, the indirect illumination effects on the surface, such as subsurface scattering and interreflections, will raise the difficulties in reconstruction. In this paper, we apply maximum min-SW gray code to reduce the indirect illumination effects of the specular surface. We also analysis the errors when comparing the maximum min-SW gray code and the conventional gray code, which justifies that the maximum min-SW gray code has significant superiority to reduce the indirect illumination effects. To achieve sub-pixel accuracy, we project high frequency sinusoidal patterns onto the scene simultaneously. But for specular surface, the high frequency patterns are susceptible to decoding errors. Incorrect decoding of high frequency patterns will result in a loss of depth resolution. Our method to resolve this problem is combining the low frequency maximum min-SW gray code and the high frequency phase shifting code, which achieves dense 3D reconstruction for specular surface. Our contributions include: (i) A complete setup of the structured light based 3D scanning system; (ii) A novel combination technique of the maximum min-SW gray code and phase shifting code. First, phase shifting decoding with sub-pixel accuracy. Then, the maximum min-SW gray code is used to resolve the ambiguity resolution. According to the experimental results and data analysis, our structured light based 3D scanning system enables high quality dense reconstruction of scenes with a small number of images. Qualitative and quantitative comparisons are performed to extract the advantages of our new
Description of FEL3D: A three dimensional simulation code for TOK and FEL
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FEL3D is a three dimensional simulation code, written for the purpose of calculating the parameters of coherent radiation emitted by electrons in an undulator. The program was written predominantly for simulating the coherent super-radiant harmonic frequency emission of electrons which are being bunched by an external laser beam while propagating in an undulator magnet. This super-radiant emission is to be studied in the TOK (transverse optical klystron) experiment, which is under construction in the NSLS department at Brookhaven National Laboratory. The program can also calculate the stimulated emission radiometric properties of a free electron laser (FEL) taking into account three dimensional effects. While this application is presently limited to the small gain operation regime of FEL's, extension to the high gain regime is expected to be relatively easy. The code is based on a semi-analytical concept. Instead of a full numerical solution of the Maxwell-Lorentz equations, the trajectories of the electron in the wiggler field are calculated analytically, and the radiation fields are expanded in terms of free space eigen-modes. This approach permits efficient computation, with a computation time of about 0.1 sec/electron on the BNL IBM 3090. The code reflects the important three dimensional features of the electron beam, the modulating laser beam, and the emitted radiation field. The statistical approach is based on averaging over the electron initial conditions according to a given distribution function in phase space, rather than via Monte-Carlo simulation. The present version of the program is written for uniform periodic wiggler field, but extension to nonuniform fields is straightforward. 4 figs., 5 tabs
Skála, J.; Baruffa, F.; Büchner, J.; Rampp, M.
2015-08-01
Context. The numerical simulation of turbulence and flows in almost ideal astrophysical plasmas with large Reynolds numbers motivates the implementation of magnetohydrodynamical (MHD) computer codes with low resistivity. They need to be computationally efficient and scale well with large numbers of CPU cores, allow obtaining a high grid resolution over large simulation domains, and be easily and modularly extensible, for instance, to new initial and boundary conditions. Aims: Our aims are the implementation, optimization, and verification of a computationally efficient, highly scalable, and easily extensible low-dissipative MHD simulation code for the numerical investigation of the dynamics of astrophysical plasmas with large Reynolds numbers in three dimensions (3D). Methods: The new GOEMHD3 code discretizes the ideal part of the MHD equations using a fast and efficient leap-frog scheme that is second-order accurate in space and time and whose initial and boundary conditions can easily be modified. For the investigation of diffusive and dissipative processes the corresponding terms are discretized by a DuFort-Frankel scheme. To always fulfill the Courant-Friedrichs-Lewy stability criterion, the time step of the code is adapted dynamically. Numerically induced local oscillations are suppressed by explicit, externally controlled diffusion terms. Non-equidistant grids are implemented, which enhance the spatial resolution, where needed. GOEMHD3 is parallelized based on the hybrid MPI-OpenMP programing paradigm, adopting a standard two-dimensional domain-decomposition approach. Results: The ideal part of the equation solver is verified by performing numerical tests of the evolution of the well-understood Kelvin-Helmholtz instability and of Orszag-Tang vortices. The accuracy of solving the (resistive) induction equation is tested by simulating the decay of a cylindrical current column. Furthermore, we show that the computational performance of the code scales very
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The single particle orbit code, TIBRO, has been modified extensively to improve the interpolation methods used and to allow use of vector potential fields in the simulation of charged particle orbits on a 3D domain. A 3D cubic B-spline algorithm is used to generate spline coefficients used in the interpolation. Smooth and accurate field representations are obtained. When vector potential fields are used, the 3D cubic spline interpolation formula analytically generates the magnetic field used to push the particles. This field has del.BETA = 0 to computer roundoff. When magnetic induction is used the interpolation allows del.BETA does not equal 0, which can lead to significant nonphysical results. Presently the code assumes quadrupole symmetry, but this is not an essential feature of the code and could be easily removed for other applications. Many details pertaining to this code are given on microfiche accompanying this report
Security and complexity of the McEliece cryptosystem based on QC-LDPC codes
Baldi, Marco; Bianchi, Marco; Chiaraluce, Franco
2011-01-01
In the context of public key cryptography, the McEliece cryptosystem represents a very smart solution based on the hardness of the decoding problem, which is believed to be able to resist the advent of quantum computers. Despite this, the original McEliece cryptosystem, based on Goppa codes, has encountered limited interest in practical applications, partly because of some constraints imposed by this very special class of codes. We have recently introduced a variant of the McEliece cryptosyst...
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A multigroup diffusion theory code, TRIHEX-3D, has been developed for hexagonal lattice core analyses. For 2-D problems one can use hexagonal or triangular centre-mesh finite difference (FD) schemes. The geometrical description of the problem is for hexagonal geometry only. Subdivision of each hexagon into uniform triangles is facilitated by a built-in auto-triangularisati on procedure. One can analyse any symmetric part of the core or the whole core as well. Reflective (30deg, 60deg, 90deg, 120deg and 180deg) and rotational (60deg, 120deg and 180deg) symmetry boundary conditions are allowed. For 3-D problems one can use a direct 3-D FDM or an axial flux synthesis method. TRIHEX-3D can be used for the core design problems of VVER type of hexagonal lattice cores. The code has been validated against a LMFBR SNR-300 benchmark problem. (author). 8 tabs., 9 figs., 9 refs., 5 appendixes
Bazarov, Ivan V.; Dunham, Bruce M.; Gulliford, Colwyn; Li, Yulin; Liu, Xianghong; Sinclair, Charles K.; Soong, Ken; Hannon, Fay
2008-01-01
We present a comparison between space charge calculations and direct measurements of the transverse phase space for space charge dominated electron bunches after a high voltage photoemission DC gun followed by an emittance compensation solenoid magnet. The measurements were performed using a double-slit setup for a set of parameters such as charge per bunch and the solenoid current. The data is compared with detailed simulations using 3D space charge codes GPT and Parmela3D with initial parti...
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This paper contains a description and evaluation of the thermal-hydraulic calculation of the transient connected with steam dump to atmosphere (SDA) opening during decreased reactor power to 20 % of nominal power (Nnom). The calculation was performed with the thermal-hydraulic system program ATHLET coupled with 3-D reactor dynamic code DYN3D. A comparison with the experiment was performed on the base of measured values during the SDA project function test on the VVER-1000 Temelin NPP Unit 2. Results obtained from calculated vs. experimental values could contribute to the validation of DYN3D/ATHLET coupling. (author)
DANTE. A 3-D unstructured-mesh finite-element transport code
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The DANTE code solves the standard multigroup Sn equations on 3-D unstructured finite-element meshes composed of arbitrary combinations of hexahedra and degenerate hexahedra (wedges, pyramids and tetrahedra). DANTE solves several second-order self-adjoint forms of the transport equation. This includes the standard even-parity and odd-parity equations, but also includes a new form of the transport equation that has the standard angular flux as its unknown rather than the even-parity or odd-parity component of the angular flux. DANTE also offers options for three different types of angular discretization: SPn, Sn and Pn. DANTE is written in FORTRAN 90 but uses a special MPI-based library, called PGSLIB, to define data array layouts over the processors and perform data communications between processors. This approach enables DANTE to be written in a SIMD fashion, yet be compatible with both SIMD and MIMD architectures. Computational results are given comparing solutions to the self-adjoint angular flux equation using the same spatial discretization scheme for both S-n and P-n angular discretizations. (R.P.)
Analytic and numerical demonstration of quantum self-correction in the 3D Cubic Code
Bravyi, Sergey
2011-01-01
A big open question in the quantum information theory concerns feasibility of a self-correcting quantum memory. A quantum state recorded in such memory can be stored reliably for a macroscopic time without need for active error correction if the memory is put in contact with a cold enough thermal bath. In this paper we derive a rigorous lower bound on the memory time $T_{mem}$ of the 3D Cubic Code model which was recently conjectured to have a self-correcting behavior. Assuming that dynamics of the memory system can be described by a Markovian master equation of Davies form, we prove that $T_{mem}\\ge L^{c\\beta}$ for some constant $c>0$, where $L$ is the lattice size and $\\beta$ is the inverse temperature of the bath. However, this bound applies only if the lattice size does not exceed certain critical value $L^*\\sim e^{\\beta/3}$. We also report a numerical Monte Carlo simulation of the studied memory indicating that our analytic bounds on $T_{mem}$ are tight up to constant coefficients. In order to model the ...
Science version 2: the most recent capabilities of the Framatome 3-D nuclear code package
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The Framatome nuclear code package SCIENCE developed in the 1990's has been fully operational for nuclear design since 1997. Results obtained using the package demonstrate the high accuracy of its physical models. Nevertheless, since the first release of the SCIENCE package, continuous improvement work has been carried out at Framatome, which leads today to Version 2 of the package. The intensive use of the package by Framatome teams, for example, while performing reload calculations and the associated core follow, is a permanent opportunity to point out any trend or scattering in the results, even the smaller they are. Thus the main objective of improvements was to take advantage of the progress in computer performances in using more sophisticated calculation schemes conducting to more accurate results. Besides the implementation of more accurate physical models, SCIENCE Version 2 also exploits developments conducted in other fields, mainly for transient calculations using 3-D kinetics or coupling with open-channel core thermal-hydraulics and the plant simulator. These developments allow Framatome to perform accident analyses with advanced methodologies using the SCIENCE package. (author)
Analyses of the OECD Main Steam Line Break Benchmark with the DYN3D and ATHLET Codes
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The Organization for Economic Cooperation and Development (OECD) Main Steam Line Break (MSLB) Benchmark was defined to validate the thermal-hydraulic system codes coupled with three-dimensional (3-D) neutron kinetic codes. The reference problem is an MSLB in a pressurized water reactor at end of cycle. The analyses were performed with the 3-D core model DYN3D, the thermal-hydraulic system code ATHLET, and the coupled code DYN3D/ATHLET. The results of the DYN3D and ATHLET simulations based on the specification are compared with the results of other participants in the final OECD reports. The effect of the thermal-hydraulic nodalization of the core, i.e., the number of coolant channels, and the influence of the coolant mixing inside the pressure vessel are studied in the paper. Calculations with a reduced number of coolant channels are performed often in coupled calculations for saving computational time. Results of a 25-channel model were compared with the 177-channel calculation (1 channel per assembly). The results for global parameters like nuclear power show only small differences for the two models; however, the prediction of local parameters such as maximum fuel temperatures requires a detailed thermal-hydraulic modeling. The effect of different coolant mixing within the reactor pressure vessel is investigated. It is shown that the influence of coolant mixing mitigates the accident consequences when 3-D neutron kinetics is applied. In case of point kinetics, coolant mixing leads to an opposite effect. To profit from the 3-D core model, a realistic description of the coolant mixing in the coupled codes is a topic of further investigations
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The reactor code DYN3D was developed at the Helmholtz-Zentrum Dresden-Rossendorf to study steady state and transient behavior of Light Water Reactors. Concerning the neutronics part, the multigroup diffusion or SP3 transport equation based on nodal expansion methods is solved both for hexagonal and square fuel element geometry. To deal with Block-type High Temperature Reactor cores DYN3D was extended to a version DYN3D-HTR. A 3D heat conduction model was introduced to include 3D effects of heat transfer and heat conduction and the detailed structure of the fuel element. Homogenized neutronic cross sections were generated by applying a Monte Carlo approach with resolution of each individual TRISO fuel particle. Results of coupled steady state and transient calculations with 12 energy groups are presented. Transient case studies are control rod insertion, a change of the inlet coolant temperature and a change of the coolant gas mass flow rate. It is shown that DYN3D-HTR is an appropriate code system to simulate steady states and short time transients. Furthermore the necessity of the 3D heat conduction model is demonstrated
Meléndez, A.; Korenaga, J.; Sallares, V.; Ranero, C. R.
2012-12-01
We present the development state of tomo3d, a code for three-dimensional refraction and reflection travel-time tomography of wide-angle seismic data based on the previous two-dimensional version of the code, tomo2d. The core of both forward and inverse problems is inherited from the 2-D version. The ray tracing is performed by a hybrid method combining the graph and bending methods. The graph method finds an ordered array of discrete model nodes, which satisfies Fermat's principle, that is, whose corresponding travel time is a global minimum within the space of discrete nodal connections. The bending method is then applied to produce a more accurate ray path by using the nodes as support points for an interpolation with beta-splines. Travel time tomography is formulated as an iterative linearized inversion, and each step is solved using an LSQR algorithm. In order to avoid the singularity of the sensitivity kernel and to reduce the instability of inversion, regularization parameters are introduced in the inversion in the form of smoothing and damping constraints. Velocity models are built as 3-D meshes, and velocity values at intermediate locations are obtained by trilinear interpolation within the corresponding pseudo-cubic cell. Meshes are sheared to account for topographic relief. A floating reflector is represented by a 2-D grid, and depths at intermediate locations are calculated by bilinear interpolation within the corresponding square cell. The trade-off between the resolution of the final model and the associated computational cost is controlled by the relation between the selected forward star for the graph method (i.e. the number of nodes that each node considers as its neighbors) and the refinement of the velocity mesh. Including reflected phases is advantageous because it provides a better coverage and allows us to define the geometry of those geological interfaces with velocity contrasts sharp enough to be observed on record sections. The code also
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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
Pin cell discontinuity factors in the transient 3-D discrete ordinates code TORT-TD
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Even with the rapid increase of computing power, whole core transient and accident analyses based on the direct solution of the 3-D neutron transport equation with a large number of energy groups and a detailed heterogeneous description of the core still remain computationally challenging. Current industrial methods for reactor core calculations therefore involve a two step approach in which lattice (assembly) depletion transport methods are used to prepare energy collapsed and fuel assembly or pin cell homogenized cross sections for subsequent whole core transport calculations. Spatial homogenization, in principal, requires the knowledge of both the actual boundary condition (local core environment) of the fuel assembly and the exact heterogeneous flux distribution (reference solution) of the whole core problem within that fuel assembly. Since, in particular, the latter is not known a priori, an infinite medium (zero net current) condition is used in the lattice calculations. It is well known that this approximation may lead to undesirable errors in cores in which large flux gradients are present across the fuel assemblies. This is the case in cores that have high heterogeneity and/or strong local absorbers, e.g. PWRs with partial MOX loading and inserted control rod clusters. There are two major approaches to mitigate spatial homogenization errors, superhomogenization (SPH) factors, and discontinuity factors within the scope of equivalence theory (ET) and generalized equivalence theory (GET). Although discontinuity factors are usually applied at the level of fuel assembly node size (assembly discontinuity factors, ADF), the methodology can be extended to pin cell homogenized whole core calculations involving pin cell discontinuity factors (PDF). There are also further developments for both the diffusion and the simplified transport (SP3) equation. In this paper, PDFs are introduced into the time-dependent 3-D discrete ordinates code TORT-TD in order to reduce the
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A deterministic particle-tracking code (TRACK3D) has been developed to compute convective flow paths of conservative (nonreactive) contaminants through porous geological media. TRACK3D requires the groundwater velocity distribution, which, in our applications, results from flow simulations using AECL's MOTIF code. The MOTIF finite-element code solves the transient and steady-state coupled equations of groundwater flow, solute transport and heat transport in fractured/porous media. With few modifications, TRACK3D can be used to analyse the velocity distributions calculated by other finite-element or finite-difference flow codes. TRACK3D temporarily integrates the velocity distribution, in conjunction with the model geometry, to calculate convective flow paths, exit locations and travel times of as many as 1000 water-coincident particles released in the flow domain. Both steady-state and time-varying velocity distributions can be handled. TRACK3D requires the flow domain to be discretized by a finite-element mesh containing as many as 25 000 elements. The mesh can contain three-dimensional (3-D) eight-noded hexahedral elements representing a solid region, or two-dimensional four-noded quadrilateral elements representing a plane, which can be oriented arbitrarily in 3-D space. TRACK3D has been verified by comparison with analytical and numerical solutions, and in an independent confirmation by Ontario Hydro Research. This report describes the assumptions, limitations, organization, operation and applications of the TRACK3D code, and provides a comprehensive user's manual. TRACK3D has been applied by AECL Research in the concept assessment phase of the Canadian Nuclear Fuel Waste Management Program to analyse convective radionuclide pathways and travel times from a hypothetical vault containing (for example) CANDU reactor fuel waste, through the surrounding geologic formations to discharge locations in the biosphere. The program has been used to examine the sensitivity
Impact of Different Spreading Codes Using FEC on DWT Based MC-CDMA System
Masum, Saleh; Kabir, M. Hasnat; Islam, Md. Matiqul; Shams, Rifat Ara; Ullah, Shaikh Enayet
2012-01-01
The effect of different spreading codes in DWT based MC-CDMA wireless communication system is investigated. In this paper, we present the Bit Error Rate (BER) performance of different spreading codes (Walsh-Hadamard code, Orthogonal gold code and Golay complementary sequences) using Forward Error Correction (FEC) of the proposed system. The data is analyzed and is compared among different spreading codes in both coded and uncoded cases. It is found via computer simulation that the performance...
Unequal-period combination approach of gray code and phase-shifting for 3-D visual measurement
Yu, Shuang; Zhang, Jing; Yu, Xiaoyang; Sun, Xiaoming; Wu, Haibin
2016-09-01
Combination of Gray code and phase-shifting is the most practical and advanced approach for the structured light 3-D measurement so far, which is able to measure objects with complex and discontinuous surface. However, for the traditional combination of the Gray code and phase-shifting, the captured Gray code images are not always sharp cut-off in the black-white conversion boundaries, which may lead to wrong decoding analog code orders. Moreover, during the actual measurement, there also exists local decoding error for the wrapped analog code obtained with the phase-shifting approach. Therefore, for the traditional approach, the wrong analog code orders and the local decoding errors will consequently introduce the errors which are equivalent to a fringe period when the analog code is unwrapped. In order to avoid one-fringe period errors, we propose an approach which combines Gray code with phase-shifting according to unequal period. With theoretical analysis, we build the measurement model of the proposed approach, determine the applicable condition and optimize the Gray code encoding period and phase-shifting fringe period. The experimental results verify that the proposed approach can offer a reliable unwrapped analog code, which can be used in 3-D shape measurement.
Meyer, Michael J; Lapcevic, Ryan; Romero, Alfonso E; Yoon, Mark; Das, Jishnu; Beltrán, Juan Felipe; Mort, Matthew; Stenson, Peter D; Cooper, David N; Paccanaro, Alberto; Yu, Haiyuan
2016-05-01
A new algorithm and Web server, mutation3D (http://mutation3d.org), proposes driver genes in cancer by identifying clusters of amino acid substitutions within tertiary protein structures. We demonstrate the feasibility of using a 3D clustering approach to implicate proteins in cancer based on explorations of single proteins using the mutation3D Web interface. On a large scale, we show that clustering with mutation3D is able to separate functional from nonfunctional mutations by analyzing a combination of 8,869 known inherited disease mutations and 2,004 SNPs overlaid together upon the same sets of crystal structures and homology models. Further, we present a systematic analysis of whole-genome and whole-exome cancer datasets to demonstrate that mutation3D identifies many known cancer genes as well as previously underexplored target genes. The mutation3D Web interface allows users to analyze their own mutation data in a variety of popular formats and provides seamless access to explore mutation clusters derived from over 975,000 somatic mutations reported by 6,811 cancer sequencing studies. The mutation3D Web interface is freely available with all major browsers supported. PMID:26841357
Development of Three-dimensional Reactor Analysis Code System for Accelerator-Driven System, ADS3D
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To investigate an Accelerator-Driven System (ADS) with sub-criticality control mechanism such as control rods or burnable poison, the ADS3D code has been developed on MARBLE which is a next generation reactor analysis code system developed by JAEA. In the past neutronics calculation for the ADS, JAEA employed RZ calculation models to realize efficient investigations. However, it was very difficult to model sub-criticality control mechanisms in RZ calculation models. The ADS3D code system is able to calculate the transportation of protons and neutrons, the burn-up calculation and the fuel exchange in three-dimensional calculation models. It means this code system can treat ADS concepts with sub-criticality control mechanism and makes it possible to investigate a new concept of ADS. (author)
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The main objective of the PHARE project SRR1/95 is the validation of coupled thermal-hydraulics - neutron kinetics codes, that are currently used for modelling the behaviour of the Russian pressurized water reactors VVER. Short descriptions of two transients, measured in the Loviisa-1 VVER-440 and the Balakovo-4 VVER-1000, respectively, have been presented in. DYN3D burnup and steady-state calculations for the states before the transients and their comparison with measurements have also been. The present report contains the main results of the simulation of both measured transients by the coupled code DYN3D-ATHLET. Two different versions of coupling the three-dimensional core model DYN3D to the thermal-hydraulic system code ATHLET are available and have been used for the calculations. In the external coupling, the whole core model DYN3D (neutron kinetics, thermal-hydraulics, and fuel rod model) is coupled to ATHLET by interfaces at the core bottom and top. In the internal coupling, only the neutron kinetics of DYN3D is implemented into ATHLET. (orig.)
Bartels, Robert E.
2012-01-01
This paper presents the implementation of gust modeling capability in the CFD code FUN3D. The gust capability is verified by computing the response of an airfoil to a sharp edged gust. This result is compared with the theoretical result. The present simulations will be compared with other CFD gust simulations. This paper also serves as a users manual for FUN3D gust analyses using a variety of gust profiles. Finally, the development of an Auto-Regressive Moving-Average (ARMA) reduced order gust model using a gust with a Gaussian profile in the FUN3D code is presented. ARMA simulated results of a sequence of one-minus-cosine gusts is shown to compare well with the same gust profile computed with FUN3D. Proper Orthogonal Decomposition (POD) is combined with the ARMA modeling technique to predict the time varying pressure coefficient increment distribution due to a novel gust profile. The aeroelastic response of a pitch/plunge airfoil to a gust environment is computed with a reduced order model, and compared with a direct simulation of the system in the FUN3D code. The two results are found to agree very well.
A study of flow mixing in a PWR vessel in asymmetric cooldown faults using the FLOW3D code
International Nuclear Information System (INIS)
The Harwell computational fluid dynamics code, FLOW3D has been used to simulate a flow mixing test in the Oconee-1 reactor. The object was to test the ability of FLOW3D to describe thermal mixing in a PWR pressure vessel for the conditions of an over-cooling fault. The code produced reasonable estimates of the thermal diffusion observed in the Oconee test, with a tendency to underpredict mixing. However, the test exhibited gross swirl and asymmetric mixing which was not predicted by FLOW3D. Sensitivity studies to investigate the effects of downcomer ovality, inlet flow vorticity and flow imbalance between loops, have not revealed the source of the observed asymmetries
TRIO a general computer code for reactor 3-D flows analysis. Application to a LMFBR hot plenum
International Nuclear Information System (INIS)
TRIO is a code developed at CEA to investigate general incompressible 2D and 3D viscous flows. Two calculations are presented: the lid driven cubic cavity at Re=400; steady state (velocity and temperature field) of a LMFBR hot plenum, carried out in order to prepare the calculation of a cold shock consecutive to a reactor scram. 8 refs., 26 figs.
Rotated Walsh-Hadamard Spreading with Robust Channel Estimation for a Coded MC-CDMA System
Directory of Open Access Journals (Sweden)
Raulefs Ronald
2004-01-01
Full Text Available We investigate rotated Walsh-Hadamard spreading matrices for a broadband MC-CDMA system with robust channel estimation in the synchronous downlink. The similarities between rotated spreading and signal space diversity are outlined. In a multiuser MC-CDMA system, possible performance improvements are based on the chosen detector, the channel code, and its Hamming distance. By applying rotated spreading in comparison to a standard Walsh-Hadamard spreading code, a higher throughput can be achieved. As combining the channel code and the spreading code forms a concatenated code, the overall minimum Hamming distance of the concatenated code increases. This asymptotically results in an improvement of the bit error rate for high signal-to-noise ratio. Higher convolutional channel code rates are mostly generated by puncturing good low-rate channel codes. The overall Hamming distance decreases significantly for the punctured channel codes. Higher channel code rates are favorable for MC-CDMA, as MC-CDMA utilizes diversity more efficiently compared to pure OFDMA. The application of rotated spreading in an MC-CDMA system allows exploiting diversity even further. We demonstrate that the rotated spreading gain is still present for a robust pilot-aided channel estimator. In a well-designed system, rotated spreading extends the performance by using a maximum likelihood detector with robust channel estimation at the receiver by about 1 dB.
International Nuclear Information System (INIS)
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
Calibration of 3D Woven Preform Design Code for CMC Materials Project
National Aeronautics and Space Administration — Mechanical and thermal performance of CMC components benefit from low part count, integrally fabricated designs of 3D woven reinforcement. The advantages of these...
Development of a neutronic code broadcasting 2D and 3D stationary by the finite volume method
International Nuclear Information System (INIS)
The objective of this work is the development of a modal neutronic code of diffusion in 2D and 3D steady using the finite volume method, from free codes and can be applied to reactors of any geometry. Currently, numerical methods most commonly used in the broadcasting codes provide good results in structured mesh, but its application to non-structured mesh is not easy and may present problems of convergence and stability of the solution. Regarding the non-structured mesh, its use is justified by their easy adaptation to complex geometries and the development of coupled Thermo-hydraulic-neutronic codes, as well as the development of codes fluid dynamic (CFD) that encourage the development of a neutronic code that has the same mesh as the codes of fluid dynamics, which in general tends to be unstructured. On the other hand, refining the mesh and its adaptation to complex geometries is another stimulus of face to learn more about what is happening at the core of the reactor. Finally, the code has been validated with a homogeneous reactor simulation and other heterogeneous for 2D and 3D. (Author)
Miensopust, Marion P.; Queralt, Pilar; Jones, Alan G.; 3D MT modellers
2013-06-01
Over the last half decade the need for, and importance of, three-dimensional (3-D) modelling of magnetotelluric (MT) data have increased dramatically and various 3-D forward and inversion codes are in use and some have become commonly available. Comparison of forward responses and inversion results is an important step for code testing and validation prior to `production' use. The various codes use different mathematical approximations to the problem (finite differences, finite elements or integral equations), various orientations of the coordinate system, different sign conventions for the time dependence and various inversion strategies. Additionally, the obtained results are dependent on data analysis, selection and correction as well as on the chosen mesh, inversion parameters and regularization adopted, and therefore, a careful and knowledge-based use of the codes is essential. In 2008 and 2011, during two workshops at the Dublin Institute for Advanced Studies over 40 people from academia (scientists and students) and industry from around the world met to discuss 3-D MT inversion. These workshops brought together a mix of code writers as well as code users to assess the current status of 3-D modelling, to compare the results of different codes, and to discuss and think about future improvements and new aims in 3-D modelling. To test the numerical forward solutions, two 3-D models were designed to compare the responses obtained by different codes and/or users. Furthermore, inversion results of these two data sets and two additional data sets obtained from unknown models (secret models) were also compared. In this manuscript the test models and data sets are described (supplementary files are available) and comparisons of the results are shown. Details regarding the used data, forward and inversion parameters as well as computational power are summarized for each case, and the main discussion points of the workshops are reviewed. In general, the responses
International Nuclear Information System (INIS)
Highlights: • General description of the coupled code DYN3D/ATHLET developed in HZDR (Helmholtz-Zentrum Dresden-Rossendorf) and used to simulate steady-state and transient behaviour of LWRs is given. • Nearly 20 real plant transients and dynamic benchmarks for 5 LWR designs were simulated to evaluate the coupling interface between two integrated codes and interaction of DYN3D model of 3-D neutron kinetics with plant components modelled by ATHLET. • An overview of DYN3D/ATHLET validation and verification activities is given with references to published results. • The results of two most recently performed validation (Kalinin-3) and verification (AER-7) tasks are presented in a separate chapter in more detail. • The validation/verification status of DYN3D/ATHLET was noticeably increased. - Abstract: One of the most intensively developing areas in the LWR multi-physics is a coupling of different best estimate 3-D neutron kinetic (BIPR, DYN3D, KIKO3D, NEM, PARCS, etc.) and thermal hydraulic (ATHLET, CATHARE, RELAP5, etc.) codes. Resulting coupled code systems have advanced capabilities of modeling both steady-state spatial distributions of the core power and their evolutions during different kinds of reactor transients. They are also highly useful in the analyses of possible reactor instabilities. Initial steady-state core power distributions can be disturbed by changes in the reactor loop mass flow rates and/or temperatures, by relocations of the low-temperature/diluted-boron water slugs within the primary system or by movements of control rods. The coupled code used for LWR simulations in HZDR is DYN3D/ATHLET, which includes the 3-D core neutron kinetic and thermal hydraulic model of own development – DYN3D. The paper reports major capabilities of DYN3D as well as different ways of its coupling with the thermal hydraulic code ATHLET (external, internal and parallel), but mainly focuses on the validation and verification of the coupled code DYN3D
International Nuclear Information System (INIS)
This report constitutes the user's manual for DCM3D. DCM3D is a computer code for solving three-dimensional, ground-water flow problems in variably saturated, fractured porous media. The code is based on a dual-continuum model with porous media comprising one continuum and fractures comprising the other. The continua are connected by a transfer term that depends on the unsaturated permeability of the porous medium. An integrated finite-difference scheme is used to discretize the governing equations in space. The time-dependent term is allowed to remain continuous. The resulting set of ordinary differential equations (ODE's) is solved with a general ODE solver, LSODES. The code is capable of handling transient, spatially dependent source terms and boundary conditions. The boundary conditions can either prescribed head or prescribed flux. 24 refs., 22 figs., 5 tabs
BARS - a heterogeneous code for 3D pin-by-pin LWR steady-state and transient calculation
International Nuclear Information System (INIS)
A 3D pin-by-pin dynamic model for LWR detailed calculation was developed. The model is based on a coupling of the BARS neutronic code with the RELAP5/MOD3.2 thermal hydraulic code. This model is intended to calculate a fuel cycle, a xenon transient, and a wide range of reactivity initiated accidents in a WWER and a PWR. Galanin-Feinberg heterogeneous method was realized in the BARS code. Some results for a validation of the heterogeneous method are presented for reactivity coefficients, a pin-by-pin power distribution, and a fast pulse transient. (Authors)
International Nuclear Information System (INIS)
The codes ATHLET and QUABOX/CUBBOX were developed by German company GRS for light water reactors. For RBMK-1500 NPP a model for the simulation of transients was developed, using a coupled version of the thermal - hydraulic system code ATHLET and the 3D core model QUABOX/CUBBOX. The coupled code system was applied to the analysis of an ATWS event with 'loss of feedwater'. There are local differences, which are influencing the transient behaviour and can be taken into account only by the 3D core model. The LAC system can only compensate the strong reactivity feedback by the fuel temperature rise for a short time. The results of 3D -kinetics and point kinetics models were compared. (author)
Directory of Open Access Journals (Sweden)
Jin Qi
Full Text Available Real-time human activity recognition is essential for human-robot interactions for assisted healthy independent living. Most previous work in this area is performed on traditional two-dimensional (2D videos and both global and local methods have been used. Since 2D videos are sensitive to changes of lighting condition, view angle, and scale, researchers begun to explore applications of 3D information in human activity understanding in recently years. Unfortunately, features that work well on 2D videos usually don't perform well on 3D videos and there is no consensus on what 3D features should be used. Here we propose a model of human activity recognition based on 3D movements of body joints. Our method has three steps, learning dictionaries of sparse codes of 3D movements of joints, sparse coding, and classification. In the first step, space-time volumes of 3D movements of body joints are obtained via dense sampling and independent component analysis is then performed to construct a dictionary of sparse codes for each activity. In the second step, the space-time volumes are projected to the dictionaries and a set of sparse histograms of the projection coefficients are constructed as feature representations of the activities. Finally, the sparse histograms are used as inputs to a support vector machine to recognize human activities. We tested this model on three databases of human activities and found that it outperforms the state-of-the-art algorithms. Thus, this model can be used for real-time human activity recognition in many applications.
Energy Technology Data Exchange (ETDEWEB)
J. D. Hales; D. M. Perez; R. L. Williamson; S. R. Novascone; B. W. Spencer
2013-03-01
BISON is a modern finite-element based nuclear fuel performance code that has been under development at the Idaho National Laboratory (USA) since 2009. The code is applicable to both steady and transient fuel behaviour and is used to analyse either 2D axisymmetric or 3D geometries. BISON has been applied to a variety of fuel forms including LWR fuel rods, TRISO-coated fuel particles, and metallic fuel in both rod and plate geometries. Code validation is currently in progress, principally by comparison to instrumented LWR fuel rods. Halden IFA experiments constitute a large percentage of the current BISON validation base. The validation emphasis here is centreline temperatures at the beginning of fuel life, with comparisons made to seven rods from the IFA-431 and 432 assemblies. The principal focus is IFA-431 Rod 4, which included concentric and eccentrically located fuel pellets. This experiment provides an opportunity to explore 3D thermomechanical behaviour and assess the 3D simulation capabilities of BISON. Analysis results agree with experimental results showing lower fuel centreline temperatures for eccentric fuel with the peak temperature shifted from the centreline. The comparison confirms with modern 3D analysis tools that the measured temperature difference between concentric and eccentric pellets is not an artefact and provides a quantitative explanation for the difference.
Particle optics in the TIT-RFQ calculated using a 3D particle-in-cell code
International Nuclear Information System (INIS)
Beam dynamics in an RFQ at the Tokyo Institute of Technology was analyzed using a 3D particle-in-cell computer code. In this calculation not only space charge force between each macroparticles but also 3D image charge field were included. Beam transmission performance was calculated for two types of vane-tip design with different tip curvature radii. These results are compared with ones obtained with the idealized linear two-term potential. The old vane tip design with a small tip curvature radius has given very poor beam transmission efficiency which cannot be accepted for the actual machine. (author)
The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes
International Nuclear Information System (INIS)
Highlights: • Design of a heavy liquid thermal-hydraulic loop for CFD/STH code validation. • Description of the loop instrumentation and assessment of measurement error. • Experimental data from forced to natural circulation transient. - Abstract: Application of coupled CFD (Computational Fluid Dynamics) and STH (System Thermal Hydraulics) codes is a prerequisite for computationally affordable and sufficiently accurate prediction of thermal-hydraulics of complex systems. Coupled STH and CFD codes require validation for understanding and quantification of the sources of uncertainties in the code prediction. TALL-3D is a liquid Lead Bismuth Eutectic (LBE) loop developed according to the requirements for the experimental data for validation of coupled STH and CFD codes. The goals of the facility design are to provide (i) mutual feedback between natural circulation in the loop and complex 3D mixing and stratification phenomena in the pool-type test section, (ii) a possibility to validate standalone STH and CFD codes for each subsection of the facility, and (iii) sufficient number of experimental data to separate the process of input model calibration and code validation. Description of the facility design and its main components, approach to estimation of experimental uncertainty and calibration of model input parameters that are not directly measured in the experiment are discussed in the paper. First experimental data from the forced to natural circulation transient is also provided in the paper
Using the RELAP5-3D advanced systems analysis code with commercial and advanced CFD software
International Nuclear Information System (INIS)
The Idaho National Engineering and Environmental Laboratory (INEEL), in conjunction with Fluent Corporation, has developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D/ATHENA advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D/ATHENA. Both steady-state and transient calculations can be performed using many working fluids and also point to three-dimensional neutronics. The Fluent/RELAP5-3D coupled code is intended as a state-of-the-art tool to study the behavior of systems with single-phase working fluids, such as advanced gas-cooled reactors. For systems with two-phase working fluids, particularly during loss-of-coolant accident (LOCA) scenarios where a multitude of flow regimes, heat transfer regimes, and phenomena are present, the Fluent-RELAP5-3D coupling will have less general applicability since Fluent's capabilities to analyze global two-phase problems are limited. Consequently, for two-phase advanced reactor analysis, INEEL plans to employ not only the Fluent-RELAP5-3D coupling, but also to make use of state-of-the-art experimental CFD tools such as CFDLib (available from the Los Alamos National Laboratory). A general description of the techniques used to couple the codes is given. A summary of the process used to checkout the coupled configuration is given. A demonstration calculation is presented. Finally, future tasks and plans are outlined. (author)
Coupling the RELAP5-3d advanced systems analysis code with commercial and advanced CFD software
International Nuclear Information System (INIS)
The Idaho National Engineering and Environmental Laboratory (INEEL), in conjunction with Fluent Corporation, has developed a new analysis tool by coupling the Fluent computational fluid dynamics (CFD) code to the RELAP5-3D/ATHENA advanced thermal-hydraulic analysis code. This tool enables researchers to perform detailed, three-dimensional analyses using Fluent's CFD capability while the boundary conditions required by the Fluent calculation are provided by the balance-of-system model created using RELAP5-3D/ATHENA. Both steady-state and transient calculations can be performed using many working fluids and also point to three-dimensional neutronics. The Fluent/RELAP5-3D coupled code is intended as a state-of-the-art tool to study the behavior of systems with single-phase working fluids, such as advanced gas-cooled reactors. For systems with two-phase working fluids, particularly during loss-of-coolant accident (LOCA) scenarios where a multitude of flow regimes, heat transfer regimes, and phenomena are present, the Fluent-RELAP5-3D coupling will have less general applicability since Fluent's capabilities to analyze global two-phase problems are limited. Consequently, for two-phase advanced reactor analysis, INEEL plans to employ not only the Fluent-RELAP5-3D coupling, but also to make use of state-of-the-art experimental CFD tools such as CFDLib (available from the Los Alamos National Laboratory). A general description of the techniques used to couple the codes is given. A summary of the process used to checkout the coupled configuration is given. Finally, future tasks and plans are outlined. (author)
International Nuclear Information System (INIS)
The best estimate thermal-hydraulic codes used in the area of nuclear reactor safety have reached a marked level of sophistication and they require to be used by competent analysts. The need for user qualification and training is clearly recognized. An effort is being made to develop a proposal for a systematic approach to user training. The estimated duration of training at the course venue, including a set of training seminars, workshops, and practical exercises, is approximately two years. In addition, the specification and assignment of tasks to be performed by the participants at their home institutions, with continuous supervision from the training center, has been foreseen. The 3D S.UN.COP seminars constitute the follow-up of the presented proposal. The seminar is subdivided into three main parts, each of one with a program to be developed in one week: the first week is dedicated to fundamental theoretical aspects, the second week deals with industrial application, coupling methodologies and hands-on training, and the third week focuses on training for transient analysis in the interaction between thermal-hydraulics and fuel behaviour. The responses of the participants during the training have demonstrated an increase in the capabilities to develop and/or modify nodalization and to perform a qualitative and quantitative accuracy evaluation. It is expected that the participants will be able to set up more accurate, reliable and efficient simulation models, applying the procedures for qualifying the thermal-hydraulic system code calculations, and for the evaluation of the uncertainty
What we've learned from 3-D and r,z intense-beam simulations using the WARP code
International Nuclear Information System (INIS)
We describe a multi-dimensional discrete-particle simulation code, WARP, and its application to Heavy Ion Fusion beams. The code's 3-D package combines features of an accelerator code and a particle-in-cell plasma simulation, and can efficiently track beams through many lattice elements and around bends. The code's r, z package allows us to follow beams over very long times and models the accelerating module impedances. A number of applications are presented. These have led to an improved understanding of: Beam equilibria, and the approach to equilibrium; longitudinal beam dynamics and stability; electrostatic quadrupole (ESQ) injector aberrations; bending and recirculation of space-charge-dominated beams; and the drift-compression process. The code is being used for accelerator design, as well as for theoretical investigations
International Nuclear Information System (INIS)
The implementation of a version of the Rutherford Laboratory's magnetostatic computer code GFUN3D on the CDC 7600 at the National Magnetic Fusion Energy Computer Center is reported. A new iteration technique that greatly increases the probability of convergence and reduces computation time by about 30% for calculations with nonlinear, ferromagnetic materials is included. The use of GFUN3D on the NMFE network is discussed, and suggestions for future work are presented. Appendix A consists of revisions to the GFUN3D User Guide (published by Rutherford Laboratory( that are necessary to use this version. Appendix B contains input and output for some sample calculations. Appendix C is a detailed discussion of the old and new iteration techniques
Systematic procedures for core neutronics modelling in the 3-D BWR system transient code RAMONA-3B
International Nuclear Information System (INIS)
RAMONA-3B is a BWR System Transient Code with a full 3-D core neutronics model. The neutronics modelling is based on the PRESTO nodal method extended for kinetics. Established procedures exist for cross-section and kinetic parameter generation. The methods are fully compatible with state-of-the-art static methods for 3-D LWR core simulation and produces results with a high accuracy. The neutronics model and cross-section data generation methods are described. Results from the 3-D calculations are presented in comparison with experimental data, to illustrate the accuray under steady state as well as transient conditions. Procedures for generating consistent 1-D and point kinetics data, also applicable in RAMONA-3B, are being described. (orig.)
Extension of the reactor dynamics code MGT-3D for pebblebed and blocktype high-temperature-reactors
International Nuclear Information System (INIS)
The High Temperature Gas cooled Reactor (HTGR) is an improved, gas cooled nuclear reactor. It was chosen as one of the candidates of generation IV nuclear plants [1]. The reactor can be shut down automatically because of the negative reactivity feedback due to the temperature's increasing in designed accidents. It is graphite moderated and Helium cooled. The residual heat can be transferred out of the reactor core by inactive ways as conduction, convection, and thermal radiation during the accident. In such a way, a fuel temperature does not go beyond a limit at which major fission product release begins. In this thesis, the coupled neutronics and fluid mechanics code MGT-3D used for the steady state and time-dependent simulation of HTGRs, is enhanced and validated [2]. The fluid mechanics part is validated by SANA experiments in steady state cases as well as transient cases. The fuel temperature calculation is optimized by solving the heat conduction equation of the coated particles. It is applied in the steady state and transient simulation of PBMR, and the results are compared to the simulation with the old overheating model. New approaches to calculate the temperature profile of the fuel element of block-type HTGRs, and the calculation of the homogeneous conductivity of composite materials are introduced. With these new developments, MGT-3D is able to simulate block-type HTGRs as well. This extended MGT-3D is used to simulate a cuboid ceramic block heating experiment in the NACOK-II facility. The extended MGT-3D is also applied to LOFC and DLOFC simulation of GT-MHR. It is a fluid mechanics calculation with a given heat source. This calculation result of MGT-3D is verified with the calculation results of other codes. The design of the Japanese HTTR is introduced. The deterministic simulation of the LOFC experiment of HTTR is conducted with the Monte-Carlo code Serpent and MGT-3D, which is the LOFC Project organized by OECD/NEA [3]. With Serpent the burnup
International Nuclear Information System (INIS)
Highlights: • Pu-239 based spectral history method was tested on 3D BWR single assembly case. • Burnup of a BWR fuel assembly was performed with the nodal code DYN3D. • Reference solution was obtained by coupled Monte-Carlo thermal-hydraulic code BGCore. • The proposed method accurately reproduces moderator density history effect for BWR test case. - Abstract: This research focuses on the verification of a recently developed methodology accounting for spectral history effects in 3D full core nodal simulations. The traditional deterministic core simulation procedure includes two stages: (1) generation of homogenized macroscopic cross section sets and (2) application of these sets to obtain a full 3D core solution with nodal codes. The standard approach adopts the branch methodology in which the branches represent all expected combinations of operational conditions as a function of burnup (main branch). The main branch is produced for constant, usually averaged, operating conditions (e.g. coolant density). As a result, the spectral history effects that associated with coolant density variation are not taken into account properly. Number of methods to solve this problem (such as micro-depletion and spectral indexes) were developed and implemented in modern nodal codes. Recently, we proposed a new and robust method to account for history effects. The methodology was implemented in DYN3D and involves modification of the few-group cross section sets. The method utilizes the local Pu-239 concentration as an indicator of spectral history. The method was verified for PWR and VVER applications. However, the spectrum variation in BWR core is more pronounced due to the stronger coolant density change. The purpose of the current work is investigating the applicability of the method to BWR analysis. The proposed methodology was verified against recently developed BGCore system, which couples Monte Carlo neutron transport with depletion and thermal-hydraulic solvers and
Simulations of 3D LPI's relevant to IFE using the PIC code OSIRIS
Tsung, F. S.; Mori, W. B.; Winjum, B. J.
2014-10-01
We will study three dimensional effects of laser plasma instabilities, including backward raman scattering, the high frequency hybrid instability, and the two plasmon instability using OSIRIS in 3D Cartesian geometry and cylindrical 2D OSIRIS with azimuthal mode decompositions. With our new capabilities we hope to demonstrate that we are capable of studying single speckle physics relevant to IFE in an efficent manner.
Recent Hydrodynamics Improvements to the RELAP5-3D Code
Energy Technology Data Exchange (ETDEWEB)
Richard A. Riemke; Cliff B. Davis; Richard.R. Schultz
2009-07-01
The hydrodynamics section of the RELAP5-3D computer program has been recently improved. Changes were made as follows: (1) improved turbine model, (2) spray model for the pressurizer model, (3) feedwater heater model, (4) radiological transport model, (5) improved pump model, and (6) compressor model.
Recent Hydrodynamics Improvements to the RELAP5-3D Code
International Nuclear Information System (INIS)
The hydrodynamics section of the RELAP5-3D computer program has been recently improved. Changes were made as follows: (1) improved turbine model, (2) spray model for the pressurizer model, (3) feedwater heater model, (4) radiological transport model, (5) improved pump model, and (6) compressor model
Depletion methodology in the 3-D whole core transport code DeCART
International Nuclear Information System (INIS)
Three dimensional whole-core transport code DeCART has been developed to include a characteristics of the numerical reactor to replace partly the experiment. This code adopts the deterministic method in simulating the neutron behavior with the least assumption and approximation. This neutronic code is also coupled with the thermal hydraulic code CFD and the thermo mechanical code to simulate the combined effects. Depletion module has been implemented in DeCART code to predict the depleted composition in the fuel. The exponential matrix method of ORIGEN-2 has been used for the depletion calculation. The library of including decay constants, yield matrix and others has been used and greatly simplified for the calculation efficiency. This report summarizes the theoretical backgrounds and includes the verification of the depletion module in DeCART by performing the benchmark calculations
Medley, S. S.; Liu, D.; Gorelenkova, M. V.; Heidbrink, W. W.; Stagner, L.
2016-02-01
A 3D halo neutral code developed at the Princeton Plasma Physics Laboratory and implemented for analysis using the TRANSP code is applied to projected National Spherical Torus eXperiment-Upgrade (NSTX-U plasmas). The legacy TRANSP code did not handle halo neutrals properly since they were distributed over the plasma volume rather than remaining in the vicinity of the neutral beam footprint as is actually the case. The 3D halo neutral code uses a ‘beam-in-a-box’ model that encompasses both injected beam neutrals and resulting halo neutrals. Upon deposition by charge exchange, a subset of the full, one-half and one-third beam energy components produce first generation halo neutrals that are tracked through successive generations until an ionization event occurs or the descendant halos exit the box. The 3D halo neutral model and neutral particle analyzer (NPA) simulator in the TRANSP code have been benchmarked with the Fast-Ion D-Alpha simulation (FIDAsim) code, which provides Monte Carlo simulations of beam neutral injection, attenuation, halo generation, halo spatial diffusion, and photoemission processes. When using the same atomic physics database, TRANSP and FIDAsim simulations achieve excellent agreement on the spatial profile and magnitude of beam and halo neutral densities and the NPA energy spectrum. The simulations show that the halo neutral density can be comparable to the beam neutral density. These halo neutrals can double the NPA flux, but they have minor effects on the NPA energy spectrum shape. The TRANSP and FIDAsim simulations also suggest that the magnitudes of beam and halo neutral densities are relatively sensitive to the choice of the atomic physics databases.
Energy Technology Data Exchange (ETDEWEB)
Medley, S. S. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Liu, D. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Univ. of California, Irvine, CA (United States). Dept. of Physics and Astronomy; Gorelenkova, M. V. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Heidbrink, W. W. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Univ. of California, Irvine, CA (United States). Dept. of Physics and Astronomy; Stagner, L. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Univ. of California, Irvine, CA (United States). Dept. of Physics and Astronomy
2016-01-12
A 3D halo neutral code developed at the Princeton Plasma Physics Laboratory and implemented for analysis using the TRANSP code is applied to projected National Spherical Torus eXperiment-Upgrade (NSTX-U plasmas). The legacy TRANSP code did not handle halo neutrals properly since they were distributed over the plasma volume rather than remaining in the vicinity of the neutral beam footprint as is actually the case. The 3D halo neutral code uses a 'beam-in-a-box' model that encompasses both injected beam neutrals and resulting halo neutrals. Upon deposition by charge exchange, a subset of the full, one-half and one-third beam energy components produce first generation halo neutrals that are tracked through successive generations until an ionization event occurs or the descendant halos exit the box. The 3D halo neutral model and neutral particle analyzer (NPA) simulator in the TRANSP code have been benchmarked with the Fast-Ion D-Alpha simulation (FIDAsim) code, which provides Monte Carlo simulations of beam neutral injection, attenuation, halo generation, halo spatial diffusion, and photoemission processes. When using the same atomic physics database, TRANSP and FIDAsim simulations achieve excellent agreement on the spatial profile and magnitude of beam and halo neutral densities and the NPA energy spectrum. The simulations show that the halo neutral density can be comparable to the beam neutral density. These halo neutrals can double the NPA flux, but they have minor effects on the NPA energy spectrum shape. The TRANSP and FIDAsim simulations also suggest that the magnitudes of beam and halo neutral densities are relatively sensitive to the choice of the atomic physics databases.
The Monte Carlo SRNA-VOX code for 3D proton dose distribution in voxelized geometry using CT data
Energy Technology Data Exchange (ETDEWEB)
Ilic, Radovan D [Laboratory of Physics (010), Vinca Institute of Nuclear Sciences, PO Box 522, 11001 Belgrade (Serbia and Montenegro); Spasic-Jokic, Vesna [Laboratory of Physics (010), Vinca Institute of Nuclear Sciences, PO Box 522, 11001 Belgrade (Serbia and Montenegro); Belicev, Petar [Laboratory of Physics (010), Vinca Institute of Nuclear Sciences, PO Box 522, 11001 Belgrade (Serbia and Montenegro); Dragovic, Milos [Center for Nuclear Medicine MEDICA NUCLEARE, Bulevar Despota Stefana 69, 11000 Belgrade (Serbia and Montenegro)
2005-03-07
This paper describes the application of the SRNA Monte Carlo package for proton transport simulations in complex geometry and different material compositions. The SRNA package was developed for 3D dose distribution calculation in proton therapy and dosimetry and it was based on the theory of multiple scattering. The decay of proton induced compound nuclei was simulated by the Russian MSDM model and our own using ICRU 63 data. The developed package consists of two codes: the SRNA-2KG, which simulates proton transport in combinatorial geometry and the SRNA-VOX, which uses the voxelized geometry using the CT data and conversion of the Hounsfield's data to tissue elemental composition. Transition probabilities for both codes are prepared by the SRNADAT code. The simulation of the proton beam characterization by multi-layer Faraday cup, spatial distribution of positron emitters obtained by the SRNA-2KG code and intercomparison of computational codes in radiation dosimetry, indicate immediate application of the Monte Carlo techniques in clinical practice. In this paper, we briefly present the physical model implemented in the SRNA package, the ISTAR proton dose planning software, as well as the results of the numerical experiments with proton beams to obtain 3D dose distribution in the eye and breast tumour.
The Monte Carlo SRNA-VOX code for 3D proton dose distribution in voxelized geometry using CT data
Ilic, Radovan D.; Spasic-Jokic, Vesna; Belicev, Petar; Dragovic, Milos
2005-03-01
This paper describes the application of the SRNA Monte Carlo package for proton transport simulations in complex geometry and different material compositions. The SRNA package was developed for 3D dose distribution calculation in proton therapy and dosimetry and it was based on the theory of multiple scattering. The decay of proton induced compound nuclei was simulated by the Russian MSDM model and our own using ICRU 63 data. The developed package consists of two codes: the SRNA-2KG, which simulates proton transport in combinatorial geometry and the SRNA-VOX, which uses the voxelized geometry using the CT data and conversion of the Hounsfield's data to tissue elemental composition. Transition probabilities for both codes are prepared by the SRNADAT code. The simulation of the proton beam characterization by multi-layer Faraday cup, spatial distribution of positron emitters obtained by the SRNA-2KG code and intercomparison of computational codes in radiation dosimetry, indicate immediate application of the Monte Carlo techniques in clinical practice. In this paper, we briefly present the physical model implemented in the SRNA package, the ISTAR proton dose planning software, as well as the results of the numerical experiments with proton beams to obtain 3D dose distribution in the eye and breast tumour.
The Monte Carlo SRNA-VOX code for 3D proton dose distribution in voxelized geometry using CT data
International Nuclear Information System (INIS)
This paper describes the application of the SRNA Monte Carlo package for proton transport simulations in complex geometry and different material compositions. The SRNA package was developed for 3D dose distribution calculation in proton therapy and dosimetry and it was based on the theory of multiple scattering. The decay of proton induced compound nuclei was simulated by the Russian MSDM model and our own using ICRU 63 data. The developed package consists of two codes: the SRNA-2KG, which simulates proton transport in combinatorial geometry and the SRNA-VOX, which uses the voxelized geometry using the CT data and conversion of the Hounsfield's data to tissue elemental composition. Transition probabilities for both codes are prepared by the SRNADAT code. The simulation of the proton beam characterization by multi-layer Faraday cup, spatial distribution of positron emitters obtained by the SRNA-2KG code and intercomparison of computational codes in radiation dosimetry, indicate immediate application of the Monte Carlo techniques in clinical practice. In this paper, we briefly present the physical model implemented in the SRNA package, the ISTAR proton dose planning software, as well as the results of the numerical experiments with proton beams to obtain 3D dose distribution in the eye and breast tumour
Heat Transfer Boundary Conditions in the RELAP5-3D Code
International Nuclear Information System (INIS)
The heat transfer boundary conditions used in the RELAP5-3D computer program have evolved over the years. Currently, RELAP5-3D has the following options for the heat transfer boundary conditions: (a) heat transfer correlation package option, (b) non-convective option (from radiation/conduction enclosure model or symmetry/insulated conditions), and (c) other options (setting the surface temperature to a volume fraction averaged fluid temperature of the boundary volume, obtaining the surface temperature from a control variable, obtaining the surface temperature from a time-dependent general table, obtaining the heat flux from a time-dependent general table, or obtaining heat transfer coefficients from either a time- or temperature-dependent general table). These options will be discussed, including the more recent ones
Heat Transfer Boundary Conditions in the RELAP5-3D Code
Energy Technology Data Exchange (ETDEWEB)
Richard A. Riemke; Cliff B. Davis; Richard R. Schultz
2008-05-01
The heat transfer boundary conditions used in the RELAP5-3D computer program have evolved over the years. Currently, RELAP5-3D has the following options for the heat transfer boundary conditions: (a) heat transfer correlation package option, (b) non-convective option (from radiation/conduction enclosure model or symmetry/insulated conditions), and (c) other options (setting the surface temperature to a volume fraction averaged fluid temperature of the boundary volume, obtaining the surface temperature from a control variable, obtaining the surface temperature from a time-dependent general table, obtaining the heat flux from a time-dependent general table, or obtaining heat transfer coefficients from either a time- or temperature-dependent general table). These options will be discussed, including the more recent ones.
Synergia: A hybrid, parallel beam dynamics code with 3D space charge
Energy Technology Data Exchange (ETDEWEB)
James F. Amundson; Panagiotis Spentzouris
2003-07-09
We describe Synergia, a hybrid code developed under the DOE SciDAC-supported Accelerator Simulation Program. The code combines and extends the existing accelerator modeling packages IMPACT and beamline/mxyzptlk. We discuss the design and implementation of Synergia, its performance on different architectures, and its potential applications.
The 3D core thermohydraulics and neutronics solution in the TRAB-SMABRE accident and transient code
International Nuclear Information System (INIS)
The TRAB-SMABRE code is a result of code development efforts carried out at VTT Processes in Finland for calculating transient and accident behaviour in Finnish LWR plants. The operating plants are two 770 MWe BWR units in Olkiluoto and two 500 MWe PWR units of VVER-440 type in Loviisa. In addition a new PWR plant of the 1600 MWe EPR type will be built into Olkiluoto. The TRAB core model, two group neutronics solution with nodal expansion method, has been initially developed as a transient 3D transient code for the BWR plant transients, and HEXTRAN code for the 3D VVER transients having the hexagonal fuel geometry. The SMABRE model thermohydraulic model is a drift flux based LOCA model. HEXTRAN and SMABRE were coupled in parallel for making the ATWS analyses in VVER plants possible and TRAB and SMABRE were coupled in parallel for calculating the transients in the PWR plants with the squared core array. As the new step the TRAB core model was coupled internally with the SMABRE for making possible the BWR analyses with the flow reversal possible and as an optional tool for the PWR plant analyses with the squared core array. The model predicts well the 3D core thermohydraulics with the encapsulated fuel, but not for the open PWR core. To overcome this deficiency a thermohydraulics simulation model for the core was introduced based on the 3D porous media thermohydraulics solution PORFLO. In the paper the basic equations of the 3D neutronics, SMABRE thermohydraulics and PORFLO thermohydraulics are described. For the BWR plant the calculation results using the parallel coupling of neutronics and thermohydraulics and internal coupling will be compared for the two transients, MSIV closure in the steam line and partial load reduction, both compared against the real plant data. The calculation result proves that the internal coupling gives the most extensive possibilities for the core simulation and is recommended for the further BWR analyses for the PWR plant the control
THAC-SIP-3D: a three-dimensional, transient heat analysis code using the strongly implicit procedure
International Nuclear Information System (INIS)
THAC-SIP-3D is a transient heat analysis code designed to use the Strongly Implicit Procedure to calculate temperature distributions for problems that can be modeled in the three-dimensional Cartesian coordinate system. In THAC-SIP-3D, the thermal conductivity, density, specific heat, and heat generation may be dependent on position, temperature, and time. The thermal conductivity may be anisotropic. A model may have boundary conditions on the external surfaces which may be adiabatic, a prescribed temperature, or any combination of prescribed heat flux, forced convection, natural convection, and radiation. The boundary temperatures may vary with time, and other boundary condition parameters may be dependent on position, temperature, and time. The mesh spacing may be variable along each axis. The code will allow a model with a maximum of 200 fine lattice planes along each axis, 100 regions, 100 materials, and 25 boundary conditions. The maximum number of nodes can be easily adjusted to fit the problem and the computer storage requirements. The amount of storage required by the code on an IBM 360 or 370 computer varies from approximately 240K bytes for one node to 1500K bytes for 5000 nodes. The code uses a finite difference scheme which can range from the Crank--Nicolson Procedure to the Classical Implicit Procedure (backwards Euler) to generate the system of equations solved by the Strongly Implicit Procedure to obtain the transient temperature distribution. The size of the time step may be varied as a function of the maximum temperature change or the maximum percent of relative change in temperature throughout the model. This report contains a discussion of the numerical technique used by THAC-SIP-3D, instructions on how to model a typical problem, a description of the input data, a description of the JCL (job control language) necessary to access the code at the Oak Ridge installations, and some sample problems. 14 figures, 16 tables
3D high-efficiency video coding for multi-view video and depth data.
Muller, Karsten; Schwarz, Heiko; Marpe, Detlev; Bartnik, Christian; Bosse, Sebastian; Brust, Heribert; Hinz, Tobias; Lakshman, Haricharan; Merkle, Philipp; Rhee, Franz Hunn; Tech, Gerhard; Winken, Martin; Wiegand, Thomas
2013-09-01
This paper describes an extension of the high efficiency video coding (HEVC) standard for coding of multi-view video and depth data. In addition to the known concept of disparity-compensated prediction, inter-view motion parameter, and inter-view residual prediction for coding of the dependent video views are developed and integrated. Furthermore, for depth coding, new intra coding modes, a modified motion compensation and motion vector coding as well as the concept of motion parameter inheritance are part of the HEVC extension. A novel encoder control uses view synthesis optimization, which guarantees that high quality intermediate views can be generated based on the decoded data. The bitstream format supports the extraction of partial bitstreams, so that conventional 2D video, stereo video, and the full multi-view video plus depth format can be decoded from a single bitstream. Objective and subjective results are presented, demonstrating that the proposed approach provides 50% bit rate savings in comparison with HEVC simulcast and 20% in comparison with a straightforward multi-view extension of HEVC without the newly developed coding tools. PMID:23715605
Security and complexity of the McEliece cryptosystem based on QC-LDPC codes
Baldi, Marco; Chiaraluce, Franco
2011-01-01
In the context of public key cryptography, the McEliece cryptosystem represents a very smart solution based on the hardness of the decoding problem, that is believed to be able to resist the future advent of quantum computers. Despite this, the original McEliece cryptosystem, based on Goppa codes, has encountered limited interest in practical applications, partly because of some constraints imposed by this special class of codes. We have recently introduced a variant of the McEliece cryptosystem adopting low-density parity-check codes, that are state-of-art codes now used in many telecommunication standards and applications. In this paper, we discuss the possible usage of a bit-flipping decoder in such context, that gives a significant advantage in terms of complexity. We also provide theoretical arguments and practical tools for estimating the trade-off between security and complexity, in such a way to give a simple procedure for the system design.
International Nuclear Information System (INIS)
The simulation of complex thermal-hydraulic phenomena is a challenging task. On one hand Computational Fluid Dynamics (CFD) codes allow a fine resolution of 3D phenomena but have a computational cost which is still prohibitive for some applications. On the other hand, System Analysis codes are fast running but cannot account for 3D phenomena. The coupling of these two approaches provides a tool which combines their advantages. In the context of the European THINS Project (7th Framework Program) the Gesellschaft für Anlagen- und Reaktorsicherheit mbH (GRS) developed a coupling between ANSYS CFX and ATHLET. The validation of this coupled code is to be performed with the help of experimental data provided by KTH (Sweden), which has built the TALL-3D facility for this purpose. This facility investigates the transition from forced to natural circulation of the Lead-Bismuth Eutectic (LBE) in a pool connected to a 3-leg primary circuit with two heaters and a heat exchanger. TUM is responsible for the Uncertainty and Sensitivity Analysis (USA) of the coupled ATHLET-CFX simulations in the THINS Project. The influence of modeling uncertainty on the simulation results needs to be assessed because it can significantly impair their accuracy. USA is a powerful tool to assess the model output variability resulting from modeling uncertainty (Uncertainty Analysis) and to identify and rank the influential model input parameters (Sensitivity Analysis). TUM has developed a computational framework to propagate modeling uncertainty through coupled Systems Analysis – Computational Fluid Dynamics (CFD) codes. This framework is being applied to the simulation of the experiments performed on the TALL-3D facility. The uncertainty methodology used is based on the statistical sampling of the uncertain inputs and models used by the two codes, its propagation through coupled calculations, and the final processing of the output sample of variables of interest with non-parametric statistical
A nodal expansion method for solving the multigroup SP3 equations in the reactor code DYN3D
International Nuclear Information System (INIS)
The core model DYN3D which has been developed for three-dimensional analyses of steady states and transients in thermal reactors with quadratic or hexagonal fuel assemblies is based on nodal methods for the solution of the two-group neutron diffusion equation. Loading cores with higher content of MOX fuel, the increase of the fuel cycle length and new types of reactors are challenging for these standard methods. A nodal expansion method for solving the equations of the simplified P3 approximation (SP3) of the multigroup transport equation was developed to improve the accuracy of the DYN3D code. In this paper, the method used in DYN3D-SP3 is described. It is applied for the pin-wise calculation of a steady state of the OECD/NEA and U.S. NRC PWR MOX/UO2 Core Transient Benchmark. The eigenvalue keff, assembly powers and the pin powers are computed. The results calculated with different approaches including diffusion theory are compared with the reference solution obtained from a heterogeneous transport calculation with the code DeCART. Different approaches of the diffusion coefficient used in the SP3 equations are investigated. The SP3 results obtained with the transport cross section of multigroup diffusion theory show the smallest deviations from the reference solution. These deviations are in the same order as the results of the code DORT, whereas the DORT and DYN3D calculations were carried out with the same library of group constants for homogenized pin cells. (authors)
Energy Technology Data Exchange (ETDEWEB)
Vay, J.-L.; Furman, M.A.; Azevedo, A.W.; Cohen, R.H.; Friedman, A.; Grote, D.P.; Stoltz, P.H.
2004-04-19
We have integrated the electron-cloud code POSINST [1] with WARP [2]--a 3-D parallel Particle-In-Cell accelerator code developed for Heavy Ion Inertial Fusion--so that the two can interoperate. Both codes are run in the same process, communicate through a Python interpreter (already used in WARP), and share certain key arrays (so far, particle positions and velocities). Currently, POSINST provides primary and secondary sources of electrons, beam bunch kicks, a particle mover, and diagnostics. WARP provides the field solvers and diagnostics. Secondary emission routines are provided by the Tech-X package CMEE.
Fast wave current drive modeling using the combined RANT3D and PICES codes
International Nuclear Information System (INIS)
Two numerical codes are combined to give a theoretical estimate of the current drive and direct electron heating by fast waves launched from phased antenna arrays on the DIII-D tokamak. Results are compared with experiment
Users manual for CAFE-3D : a computational fluid dynamics fire code.
Energy Technology Data Exchange (ETDEWEB)
Khalil, Imane; Lopez, Carlos; Suo-Anttila, Ahti Jorma (Alion Science and Technology, Albuquerque, NM)
2005-03-01
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.
Development of 3d reactor burnup code based on Monte Carlo method and exponential Euler method
International Nuclear Information System (INIS)
Burnup analysis plays a key role in fuel breeding, transmutation and post-processing in nuclear reactor. Burnup codes based on one-dimensional and two-dimensional transport method have difficulties in meeting the accuracy requirements. A three-dimensional burnup analysis code based on Monte Carlo method and Exponential Euler method has been developed. The coupling code combines advantage of Monte Carlo method in complex geometry neutron transport calculation and FISPACT in fast and precise inventory calculation, meanwhile resonance Self-shielding effect in inventory calculation can also be considered. The IAEA benchmark text problem has been adopted for code validation. Good agreements were shown in the comparison with other participants' results. (authors)
International Nuclear Information System (INIS)
Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers, vendors, and research organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the user effect and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification represent an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. In addition, this paper presents the organization and the main features of the 3D S.UN.COP (scaling, uncertainty, and 3D coupled code calculations) seminars during which particular emphasis is given to the areas of the scaling, uncertainty, and 3D coupled code analysis
Coupling of the thermal-hydraulics TRAC codes with 3-D neutron kinetics code SKETCH-N
International Nuclear Information System (INIS)
Since 1997, the Japan Atomic Energy Research Institute (JAERI) has been developing the three-dimensional transient analysis codes: J-TRAC/SKETCH-N code system for pressurised water reactor (PWR), and the TRAC-BEY SKETCH-N for boiling water reactor (BWR). The J-TRAC code, developed by JAERI, has a framework based on the TRAC-PF1, which is best estimate system transient analysis code having the capability to simulate PWR system components and three-dimensional core in cylindrical R-9-Z coordinates. The TRAC-BF1 is a BWR system analysis code. The neutronics code SKETCH-N can solve the steady-state and kinetics forms of neutron diffusion equations in X-Y-Z geometry. The efficient polynomial and semi-analytic nodal methods based on the nonlinear iteration procedure are used in the code. Coupling with the TRAC codes is performed using the interface module based on the message-passing library Parallel Virtual Machine (PVM). The interface module is responsible for data transfer between the codes, mapping of the data between different spatial meshes and synchronization of the time stepping. Verification of the coupled TRAC/SKETCH code system has been performed against light water reactor (LWR) transient benchmark problems. J-TRAC/SKETCH has been verified by NEACRP PWR rod ejection benchmark and NEA/NSC PWR rod withdrawal benchmark. Verification of the TRAC-BF1/SKETCH-N has been performed using NEACRP BWR cold water injection benchmark. TRAC/SKETCH numerical results and a comparison with the other codes are given in the paper. The second part of the paper presents current status of the out-of-pile experiments at JAERI on BWR stability to assess and improve the TRAC-BF1 code. (authors)
GRILLIX. A 3D turbulence code for magnetic fusion devices based on a field line map
International Nuclear Information System (INIS)
The complex geometry in the scrape-off layer of tokamaks poses problems to existing turbulence codes. The usually employed field aligned coordinates become ill defined at the separatrix. Therefore the parallel code GRILLIX was developed, which is based on a field line map. This allows simulations in additional complex geometries, especially across the separatrix. A new discretisation, based on the support operator method, for the highly anisotropic diffusion was developed and applied to a simple turbulence model (Hasegawa-Wakatani).
Full Core, Heterogeneous, Time Dependent Neutron Transport Calculations with the 3D Code DeCART
Hursin, Mathieu
2010-01-01
The current state of the art in reactor physics methods to assess safety, fuel failure, and operability margins for Design Basis Accidents (DBAs) for Light Water Reactors (LWRs) rely upon the coupling of nodal neutronics and one-dimensional thermal hydraulic system codes. The neutronic calculations use a multi-step approach in which the assembly homogenized macroscopic cross sections and kinetic parameters are first calculated using a lattice code for the range of conditions (temperatures, bu...
Qualification of McCARD/MASTER Code System for Yonggwang Unit 4
International Nuclear Information System (INIS)
Recently, we have developed the new two-step procedure based on the Monte Carlo (MC) methods. In this procedure, one can generate the few group constants including the few-group diffusion constants by the MC method augmented by the critical spectrum, which is provided by the solution to the homogeneous 0-dimensional B1 equation. In order to examine the qualification of the few-group constants generated by MC method, we combine MASTER with McCARD to form McCARD/MASTER code system for two-step core neutronics calculations. In the fictitious PWR system problems, the core design parameters calculated by the two-step McCARD/MASTER analysis agree well with those from the direct MC calculations. In this paper, a neutronic design analysis for the initial core of Yonggwang Nuclear Unit 4 (YGN4) is conducted using McCARD/MASTER two-step procedure to examine the qualification of two group constants from McCARD in terms of a real PWR core problem. To compare with the results, the nuclear design report and measured data are chosen as the reference solutions
International Nuclear Information System (INIS)
In the present Non-LOCA safety analysis of the Pressurized Water Reactor (PWR), plant transient, core response and fuel behavior are independently calculated by different analysis codes to estimate the plant safety. Therefore these results often involve large un-quantified conservativeness due to additional safety margins for initial/boundary conditions of each calculation and simplistic approximations for complicated interactions between the core neutronics and plant thermal-hydraulics during the transient. Recently, best estimate 3-D core transient analysis codes have been widely developed in the area of nuclear reactor accident analysis to understand actual physical phenomena and quantification of conservativeness in the current safety analysis. Evaluating safety margins appropriately contributes to the more safety of the plant design and the efficiency of the plant operation. Mitsubishi Heavy Industries (MHI) has developed the 3-D core kinetics coupled with the thermal-hydraulics code SPARKLE, and has a plan to apply it for the commercial licensing in the near future. This paper presents the feature of the SPARKLE code and the results of the application to representative accident events. (author)
VVER-1000 main steam line break analysis using the coupled code system DYN3D/ATHLET
International Nuclear Information System (INIS)
Calculations using the coupled code system DYN3D/ATHLET were performed in the frame of the OECD/NEA MSLB benchmark for a VVER-1000 reactor. The coolant mixing inside the reactor pressure vessel was treated using a validated empirical mixing model implemented into the DYN3D/ATHLET code. Using very conservative boundary conditions (reduced scram worth, two stuck rods, running MCP throughout the whole transient) a return-to-power was predicted. For the assessment of the empirical mixing model a time dependent calculation using the computational fluid dynamics code CFX-10 was performed. For that analysis, a detailed model of the reactor pressure vessel consisting of the inlets nozzles, downcomer, lower plenum and a part of the core and having 4.67 million unstructured tetra cell elements was used. For the considered case with running main coolant pumps, this calculation shows a sector formation at the core inlet with a certain amount of mixing at the edges of the sector. A core calculation using these CFX results as boundary conditions predicted also a return-to-power with a maximum value being about 200 MW lower than in the coupled code calculation. This variation calculation confirms the applicability of the empirical mixing model. The comparison shows also, that in this way results with a reasonable degree of conservatism can be obtained. (authors)
Velioǧlu, Deniz; Cevdet Yalçıner, Ahmet; Zaytsev, Andrey
2016-04-01
Tsunamis are huge waves with long wave periods and wave lengths that can cause great devastation and loss of life when they strike a coast. The interest in experimental and numerical modeling of tsunami propagation and inundation increased considerably after the 2011 Great East Japan earthquake. In this study, two numerical codes, FLOW 3D and NAMI DANCE, that analyze tsunami propagation and inundation patterns are considered. Flow 3D simulates linear and nonlinear propagating surface waves as well as long waves by solving three-dimensional Navier-Stokes (3D-NS) equations. NAMI DANCE uses finite difference computational method to solve 2D depth-averaged linear and nonlinear forms of shallow water equations (NSWE) in long wave problems, specifically tsunamis. In order to validate these two codes and analyze the differences between 3D-NS and 2D depth-averaged NSWE equations, two benchmark problems are applied. One benchmark problem investigates the runup of long waves over a complex 3D beach. The experimental setup is a 1:400 scale model of Monai Valley located on the west coast of Okushiri Island, Japan. Other benchmark problem is discussed in 2015 National Tsunami Hazard Mitigation Program (NTHMP) Annual meeting in Portland, USA. It is a field dataset, recording the Japan 2011 tsunami in Hilo Harbor, Hawaii. The computed water surface elevation and velocity data are compared with the measured data. The comparisons showed that both codes are in fairly good agreement with each other and benchmark data. The differences between 3D-NS and 2D depth-averaged NSWE equations are highlighted. All results are presented with discussions and comparisons. Acknowledgements: Partial support by Japan-Turkey Joint Research Project by JICA on earthquakes and tsunamis in Marmara Region (JICA SATREPS - MarDiM Project), 603839 ASTARTE Project of EU, UDAP-C-12-14 project of AFAD Turkey, 108Y227, 113M556 and 213M534 projects of TUBITAK Turkey, RAPSODI (CONCERT_Dis-021) of CONCERT
International Nuclear Information System (INIS)
Shutdown dose rate (SDDR) inside and around the diagnostics ports of ITER is performed at PPPL/UCLA using the 3-D, FEM, Discrete Ordinates code, ATTILA, along with its updated FORNAX transmutation/decay gamma library. Other ITER partners assess SDDR using codes based on the Monte Carlo (MC) approach (e.g. MCNP code) for transport calculation and the radioactivity inventory code FISPACT or other equivalent decay data libraries for dose rate assessment. To reveal the range of discrepancies in the results obtained by various analysts, an extensive experimental and calculation benchmarking effort has been undertaken to validate the capability of ATTILA for dose rate assessment. On the experimental validation front, the comparison was performed using the measured data from two SDDR experiments performed at the FNG facility, Italy. Comparison was made to the experimental data and to MC results obtained by other analysts. On the calculation validation front, the ATTILA's predictions were compared to other results at key locations inside a calculation benchmark whose configuration duplicates an upper diagnostics port plug (UPP) in ITER. Both serial and parallel version of ATTILA-7.1.0 are used in the PPPL/UCLA analysis performed with FENDL-2.1/FORNAX databases. In the FNG 1st experimental, it was shown that ATTILA's dose rates are largely over estimated (by ∼30–60%) with the ANSI/ANS-6.1.1 flux-to-dose factors whereas the ICRP-74 factors give better agreement (10–20%) with the experimental data and with the MC results at all cooling times. In the 2nd experiment, there is an under estimation in SDDR calculated by both MCNP and ATTILA based on ANSI/ANS-6.1.1 for cooling times up to ∼4 days after irradiation. Thereafter, an over estimation is observed (∼5–10% with MCNP and ∼10–15% with ATTILA). As for the calculation benchmark, the agreement is much better based on ICRP-74 1996 data. The divergence among all dose rate results at ∼11 days cooling time is no
MC2-2: a code to calculate fast neutron spectra and multigroup cross sections
International Nuclear Information System (INIS)
MC2-2 is a program to solve the neutron slowing down problem using basic neutron data derived from the ENDF/B data files. The spectrum calculated by MC2-2 is used to collapse the basic data to multigroup cross sections for use in standard reactor neutronics codes. Four different slowing down formulations are used by MC2-2: multigroup, continuous slowing down using the Goertzel-Greuling or Improved Goertzel-Greuling moderating parameters, and a hyper-fine-group integral transport calculation. Resolved and unresolved resonance cross sections are calculated accounting for self-shielding, broadening and overlap effects. This document provides a description of the MC2-2 program. The physics and mathematics of the neutron slowing down problem are derived and detailed information is provided to aid the MC2-2 user in preparing input for the program and implementation of the program on IBM 370 or CDC 7600 computers
Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System
Energy Technology Data Exchange (ETDEWEB)
Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong [Kookmin Univ., Seoul (Korea, Republic of)
2007-03-15
The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow.
International Nuclear Information System (INIS)
Validation of the TRACR3D code in a one-dimensional form was obtained for flow of soil water in three experiments. In the first experiment, a pulse of water entered a crushed-tuff soil and initially moved under conditions of saturated flow, quickly followed by unsaturated flow. In the second experiment, steady-state unsaturated flow took place. In the final experiment, two slugs of water entered crushed tuff under field conditions. In all three experiments, experimentally measured data for volumetric water content agreed, within experimental errors, with the volumetric water content predicted by the code simulations. The experiments and simulations indicated the need for accurate knowledge of boundary and initial conditions, amount and duration of moisture input, and relevant material properties as input into the computer code. During the validation experiments, limitations on monitoring of water movement in waste burial sites were also noted. 5 references, 34 figures, 9 tables
Directory of Open Access Journals (Sweden)
Alessandro Petruzzi
2008-01-01
The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the “user effect” and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification represent an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. In addition, this paper presents the organization and the main features of the 3D S.UN.COP (scaling, uncertainty, and 3D coupled code calculations seminars during which particular emphasis is given to the areas of the scaling, uncertainty, and 3D coupled code analysis.
3-D kinetics simulations of the NRU reactor using the DONJON code
International Nuclear Information System (INIS)
The NRU reactor is highly heterogeneous, heavy-water cooled and moderated, with online refuelling capability. It is licensed to operate at a maximum power of 135 MW, with a peak thermal flux of approximately 4.0 x 1018 n.m-2 . s-1. In support of the safe operation of NRU, three-dimensional kinetics calculations for reactor transients have been performed using the DONJON code. The code was initially designed to perform space-time kinetics calculations for the CANDUR power reactors. This paper describes how the DONJON code can be applied to perform neutronic simulations for the analysis of reactor transients in NRU, and presents calculation results for some transients. (authors)
CAVEAT: A 3-D ale code based on a Godunov method
Energy Technology Data Exchange (ETDEWEB)
Hall, M.S.
1987-09-17
CAVEAT is a three-dimensional Lagrangian computer code that solves the Euler equations of motion for a compressible fluid. Through the option of continuous rezoning, the code can be run in an Eulerian, rezoned Lagrangian, or Lagrangian mode. CAVEAT is thus well-suited to solving problems involving severe distortions and interfacial slip. Velocities and pressures on cell faces are found using a noniterative, approximate Riemann solver developed by J. Dukowicz, and the solution is advanced in time using a Godunov method. This shock capturing technique results in good resolution of strong shocks, which are typically spread over two or three computational cells. The following report contains an overview of some of the major features of the code. A derivation of the approximate Riemann solver is given and we present results obtained for a selection of well-known test problems.
Calculation of the RSG-GAS core using computer code citation-3D
International Nuclear Information System (INIS)
Since core reactivity is one of the reactor safety parameters, this R and D has been carried out. To carry out the R and D, the code called WIMSD4 was used respectively for generating cross section and diffusion parameters. The code CITATION was then applied to estimate core reactivity in the RSG-GAS core. To verify the result of the calculation, data and information of the RSG-GAS Typical Working Core Were used. To Prove the codes reliably used, the case of all control elements down in the reactor core and that of all control rods up in the core were applied. The result taking into account those cases showed respectively that Keff are less and greater than unity (Keffeff>1)
Solving linear systems in FLICA-4, thermohydraulic code for 3-D transient computations
International Nuclear Information System (INIS)
FLICA-4 is a computer code, developed at the CEA (France), devoted to steady state and transient thermal-hydraulic analysis of nuclear reactor cores, for small size problems (around 100 mesh cells) as well as for large ones (more than 100000), on, either standard workstations or vector super-computers. As for time implicit codes, the largest time and memory consuming part of FLICA-4 is the routine dedicated to solve the linear system (the size of which is of the order of the number of cells). Therefore, the efficiency of the code is crucially influenced by the optimization of the algorithms used in assembling and solving linear systems: direct methods as the Gauss (or LU) decomposition for moderate size problems, iterative methods as the preconditioned conjugate gradient for large problems. 6 figs., 13 refs
2D and 3D Core-Collapse Supernovae Simulation Results Obtained with the CHIMERA Code
Bruenn, S W; Hix, W R; Blondin, J M; Marronetti, P; Messer, O E B; Dirk, C J; Yoshida, S
2010-01-01
Much progress in realistic modeling of core-collapse supernovae has occurred recently through the availability of multi-teraflop machines and the increasing sophistication of supernova codes. These improvements are enabling simulations with enough realism that the explosion mechanism, long a mystery, may soon be delineated. We briefly describe the CHIMERA code, a supernova code we have developed to simulate core-collapse supernovae in 1, 2, and 3 spatial dimensions. We then describe the results of an ongoing suite of 2D simulations initiated from a 12, 15, 20, and 25 solar mass progenitor. These have all exhibited explosions and are currently in the expanding phase with the shock at between 5,000 and 20,000 km. We also briefly describe an ongoing simulation in 3 spatial dimensions initiated from the 15 solar mass progenitor.
Space-Time Coded MC-CDMA: Blind Channel Estimation, Identifiability, and Receiver Design
Directory of Open Access Journals (Sweden)
Li Hongbin
2002-01-01
Full Text Available Integrating the strengths of multicarrier (MC modulation and code division multiple access (CDMA, MC-CDMA systems are of great interest for future broadband transmissions. This paper considers the problem of channel identification and signal combining/detection schemes for MC-CDMA systems equipped with multiple transmit antennas and space-time (ST coding. In particular, a subspace based blind channel identification algorithm is presented. Identifiability conditions are examined and specified which guarantee unique and perfect (up to a scalar channel estimation when knowledge of the noise subspace is available. Several popular single-user based signal combining schemes, namely the maximum ratio combining (MRC and the equal gain combining (EGC, which are often utilized in conventional single-transmit-antenna based MC-CDMA systems, are extended to the current ST-coded MC-CDMA (STC-MC-CDMA system to perform joint combining and decoding. In addition, a linear multiuser minimum mean-squared error (MMSE detection scheme is also presented, which is shown to outperform the MRC and EGC at some increased computational complexity. Numerical examples are presented to evaluate and compare the proposed channel identification and signal detection/combining techniques.
Bazarov, Ivan V; Gulliford, Colwyn; Li, Yulin; Liu, Xianghong; Sinclair, Charles K; Soong, Ken; Hannon, Fay
2008-01-01
We present a comparison between space charge calculations and direct measurements of the transverse phase space for space charge dominated electron bunches after a high voltage photoemission DC gun followed by an emittance compensation solenoid magnet. The measurements were performed using a double-slit setup for a set of parameters such as charge per bunch and the solenoid current. The data is compared with detailed simulations using 3D space charge codes GPT and Parmela3D with initial particle distributions created from the measured transverse and temporal laser profiles. Beam brightness as a function of beam fraction is calculated for the measured phase space maps and found to approach the theoretical maximum set by the thermal energy and accelerating field at the photocathode.
Development and validation of the 3-D PWR core dynamics SIMTRAN code
International Nuclear Information System (INIS)
We discuss the main features and results of the SIMTRAN development and validation work. Included in the first are the extension of the nodal neutronic solution to account for intranodal shape and spectrum, due to both heterogeneities and flux gradients, the implicit scheme for spatial kinetics with six delayed neutron precursors and the integration of the neutronic and thermohydraulic solutions on an staggered time mesh. Validation results are discussed for the NEACRP 3-D PWR Core Transient Benchmark and an actual transient with sudden increase of core flow occurred in the Vandellos-II 3-loop PWR NPP. Agreement with the reference numerical solution and measured plant data is shown for both problems. (orig./DG)
A Methodology to Validate 3-D Arbitrary Lagrangian Eulerian Codes with Applications to Alegra
Energy Technology Data Exchange (ETDEWEB)
Chhabildas, L.C.; Duggins, B.D.; Konrad, C.H.; Mosher, D.A.; Perry, J.S.; Reinhart, W.D.; Summers, R.M.; Trucano, T.G.
1998-11-04
In this study we provided an experimental test bed for validating features of the Arbitrary Lagrangian Eulerian Grid for Research Applications (ALEGRA) code over a broad range of strain rates with overlapping diagnostics that encompass the multiple responses. A unique feature of the ALEGRA code is that it allows simultaneous computational treatment, within one code, of a wide range of strain-rates varying from hydrodynamic to structural conditions. This range encompasses strain rates characteristic of shock-wave propagation (107/s) and those characteristics of structural response (102/s). Most previous code validation experimental &udies, however, have been restricted to simulating or investigating a single strain-rate regime. What is new and different in this investigation is that we have performed well-controlled and well-instrumented experiments, which capture features relevant to both hydrodynamic and structural response in a single experiment. Aluminum was chosen for use in this study because it is a well-characterized material. The current experiments span strain rate regimes of over 107/s to less than 102/s in a single experiment. The input conditions were extremely well defined. Velocity interferometers were used to record the high' strain-rate response, while low strain rate data were collected using strain gauges. Although the current tests were conducted at a nominal velocity of - 1.5 km/s, it is the test methodology that is being emphasized herein. Results of a three-dimensional experiment are also presented.
Fast wave current drive modeling using the combined RANT3D and PICES Codes
International Nuclear Information System (INIS)
Two numerical codes are combined to give a theoretical estimate of the current drive and direct electron heating by fast waves launched from phased antenna arrays on the DIII-D tokamak. Results are compared with experiment. copyright 1996 American Institute of Physics
A 3D coarse-mesh time dependent code for nuclear reactor kinetic calculations
International Nuclear Information System (INIS)
A course-mesh code for time-dependent multigroup neutron diffusion calculation based on a direct integration scheme for the time dependence and a low order nodal flux expansion approximation for the space variables has been implemented as a fast tool for transient analysis. (Author)
International Nuclear Information System (INIS)
The Supercritical-Water-Cooled Reactor (SCWR) is a high-temperature, high-pressure water cooled reactor that operates above the critical pressure of water. In order to perform efficiently the thermal design of the SCWR, it is important to assess the thermal-hydraulics in rod bundles of the core. The experimental conditions of mockup tests, however, have to be limited because of technical and financial reasons. Therefore, it is required to establish an analytical design technique which can extrapolate experimental data to various design conditions of the reactor. JAEA (Japan Atomic Energy Agency) have been improved the three-dimensional two-fluid model analysis code ACE-3D, which has been developed originally for the two-phase flow thermal hydraulics of light water reactors, to handle the thermal hydraulic properties of water at supercritical region. In the present paper, heat transfer experiments of supercritical water flowing in a vertical annular channel around a heater pin, which simulates the core flow around a fuel rod, were analyzed with the improved ACE-3D to assess the prediction performance of the code. As a result, it was confirmed that the calculated wall surface temperature agreed with the measured results and the code is applicable to prediction of heat transfer of supercritical water in the system that simulates the SCWR core. (author)
International Nuclear Information System (INIS)
This paper describes the time-dependent 3D discrete ordinates transport code TORT-TD. Thermal-hydraulic feedback is considered by coupling TORT-TD with the thermal-hydraulics system code ATHLET. The coupled code TORT-TD/ATHLET allows 3D pin-by-pin analyses of transients in few energy groups and anisotropic scattering by solving the time-dependent transport equation using the unconditionally stable implicit method. The nuclear cross sections are interpolated between pre-calculated table values of fuel temperature, moderator density and boron concentration. For verification of the implementation, selected test cases have been calculated by TORT-TD/ATHLET. They include a control rod ejection transient in a small PWR fuel assembly arrangement and a local boron concentration change in a single PWR fuel assembly. In the latter, special attention has been paid to study the influence of the thermal-hydraulic feedback modelling in ATHLET. The results obtained for a control rod ejection accident in a PWR quarter core demonstrate the applicability of TORT-TD/ATHLET. (authors)
Energy Technology Data Exchange (ETDEWEB)
Seubert, A.; Velkov, K.; Langenbuch, S. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsinstitute, D-85748 Garching (Germany)
2008-07-01
This paper describes the time-dependent 3D discrete ordinates transport code TORT-TD. Thermal-hydraulic feedback is considered by coupling TORT-TD with the thermal-hydraulics system code ATHLET. The coupled code TORT-TD/ATHLET allows 3D pin-by-pin analyses of transients in few energy groups and anisotropic scattering by solving the time-dependent transport equation using the unconditionally stable implicit method. The nuclear cross sections are interpolated between pre-calculated table values of fuel temperature, moderator density and boron concentration. For verification of the implementation, selected test cases have been calculated by TORT-TD/ATHLET. They include a control rod ejection transient in a small PWR fuel assembly arrangement and a local boron concentration change in a single PWR fuel assembly. In the latter, special attention has been paid to study the influence of the thermal-hydraulic feedback modelling in ATHLET. The results obtained for a control rod ejection accident in a PWR quarter core demonstrate the applicability of TORT-TD/ATHLET. (authors)
3D Analysis of Cooling Performance with Loss of Offsite Power Using GOTHIC Code
International Nuclear Information System (INIS)
GOTHIC code enables to analyze one-dimensional or multi-dimensional problems for evaluating the cooling performance of loss of offsite power. The conventional GOTHIC code analysis performs heat transfer between plant containment and the outside of the fan cooler tubes by modeling each of fan cooler part model and component cooling water inside tube each to analyze boiling probability. In this paper, we suggest a way which reduces the multi-procedure of the cooling performance with loss of offsite power or the heat transfer states with complex geometrical structure to a single-procedure and verify the applicability of the heat transfer differences from the containment atmosphere humidity changes by the multi-nodes which component cooling water of tube or air of Reactor Containment Fan Cooler in the containment, otherwise the component model uses only one node
3D Analysis of Cooling Performance with Loss of Offsite Power Using GOTHIC Code
Energy Technology Data Exchange (ETDEWEB)
Oh, Kye Min; Heo, Gyun Young [Kyung Hee University, Yongin (Korea, Republic of); Na, In Sik [Atomic Creative Technology, Daejeon (Korea, Republic of); Choi, Yu Jung [Nuclear Engineering and Technology, Daejeon (Korea, Republic of)
2010-05-15
GOTHIC code enables to analyze one-dimensional or multi-dimensional problems for evaluating the cooling performance of loss of offsite power. The conventional GOTHIC code analysis performs heat transfer between plant containment and the outside of the fan cooler tubes by modeling each of fan cooler part model and component cooling water inside tube each to analyze boiling probability. In this paper, we suggest a way which reduces the multi-procedure of the cooling performance with loss of offsite power or the heat transfer states with complex geometrical structure to a single-procedure and verify the applicability of the heat transfer differences from the containment atmosphere humidity changes by the multi-nodes which component cooling water of tube or air of Reactor Containment Fan Cooler in the containment, otherwise the component model uses only one node
Modelling 3-D mechanical phenomena in a 1-D industrial finite element code: results and perspectives
International Nuclear Information System (INIS)
Assessing fuel rod integrity in PWR reactors must enjoin two opposite goals: a one-dimensional finite element code (axial revolution symmetry) is needed to provide industrial results at the scale of the reactor core, while the main risk of cladding failure [e.g. pellet-cladding interaction (PCI)] is based on fully three-dimensional phenomena. First, parametric three-dimensional elastic calculations were performed to identify the relevant parameters (fragment number, contact pellet-cladding conditions, etc.) as regards PCI. Axial fragment number as well as friction coefficient are shown to play a major role in PCI as opposed to other parameters. Next, the main limitations of the one-dimensional hypothesis of the finite element code CYRANO3 are identified. To overcome these limitations, both two- and three-dimensional emulations of CYRANO3 were developed. These developments are shown to significantly improve the results provided by CYRANO3. (authors)
International Nuclear Information System (INIS)
A model of a gamma sterilizer was built using the ITS/ACCEPT Monte Carlo code and verified through dosimetry. Individual dosimetry measurements in homogeneous material were pooled to represent larger bodies that could be simulated in a reasonable time. With the assumptions and simplifications described, dose predictions were within 2-5% of dosimetry. The model was used to simulate product movement through the sterilizer and to predict information useful for process optimization and facility design
Applications of the 3-D Deterministic Transport Code Attlla for Core Safety Analysis
Energy Technology Data Exchange (ETDEWEB)
D. S. Lucas
2004-10-01
An LDRD (Laboratory Directed Research and Development) project is ongoing at the Idaho National Engineering and Environmental Laboratory (INEEL) for applying the three-dimensional multi-group deterministic neutron transport code (Attila®) to criticality, flux and depletion calculations of the Advanced Test Reactor (ATR). This paper discusses the model development, capabilities of Attila, generation of the cross-section libraries, and comparisons to an ATR MCNP model and future.
Visual numerical steering in 3D AGENT code system for advanced nuclear reactor modeling and design
International Nuclear Information System (INIS)
Highlights: ► Numerical steering framework developed for deterministic neutron transport code AGENT to speed up the solution. ► Resulting speed up is on the order of 50%. ► Use of the steering framework is demonstrated modeling a TRIGA reactor. ► Numerical steering framework showed to be well suited for the deterministic neutron transport methods. - Abstract: The AGENT simulation system is used for detailed three-dimensional modeling of neutron transport and corresponding properties of nuclear reactors of any design. Numerical solution to the neutron transport equation in the AGENT system is based on the Method of Characteristics (MOCs) and the theory of R-functions. The latter of which is used for accurately describing current and future heterogeneous lattices of reactor core configurations. The AGENT code has been extensively verified to assure a high degree of accuracy for predicting neutron three-dimensional point-wise flux spatial distributions, power peaking factors, reaction rates, and eigenvalues. In this paper, a new AGENT code feature, a computational steering, is presented. This new feature provides a novel way for using deterministic codes for fast evaluation of reactor core parameters, at no loss to accuracy. The computational steering framework as developed at the Technische Universität München is smoothly integrated into the AGENT solver. This framework allows for an arbitrary interruption of AGENT simulation, allowing the solver to restart with updated parameters. One possible use of this is to accelerate the convergence of the final values resulting in significantly reduced simulation times. Using this computational steering in the AGENT system, coarse MOC resolution parameters can initially be selected and later update them – while the simulation is actively running – into fine resolution parameters. The utility of the steering framework is demonstrated using the geometry of a research reactor at the University of Utah: this new
I-Simpa, a graphical user interface devoted to host 3D sound propagation numerical codes
PICAUT, Judicaël; Fortin, Nicolas
2012-01-01
Whatever for indoor noise applications (room acoustics, noise in vehicles...) or sound propagation in the environment (open field, urban areas...), many numerical codes have been developed by researchers. Most of them have many common aspects, like the definition of the domain geometry and the materials (boundary conditions, impedance...), the definition of sound sources and of receivers (position, spectrum, directivity...). Moreover, they all have the same objective that is to predict the so...
International Nuclear Information System (INIS)
Thermal hydraulic system codes have been extensively developed by the nuclear industry, research institutes and technical safety organizations with the goal to improve the design and safety of nuclear installations. A large number of these simulation tools are based on the lumped parameter theory. Such programs use networks consisting of 1D cells, where mass, momentum and energy equations are solved for each fluid phase and balanced over each node of the network. System codes are extensively validated against experiments and provide reliable results at low computational cost. Lump parameter programs use simplifications in the mathematical models describing the simulated systems. Balance equations for mass, momentum and energy for two phases are obtained by averaging of the local basic flow equations in the space. As a result, mean values for relevant physical parameters which in reality are spatially distributed fields are calculated. However, since relevant reactor fluid flow and heat transfer phenomena are 3D in nature, 1D system codes have limitations on their application for specific nuclear reactor safety (NRS) problems with pronounced 3D phenomena like boron dilution, pressurized thermal shock and main steam line break. Modern computational fluid dynamics (CFD) codes are capable to predict fluid flow behavior in complex geometries and can provide detailed distribution of the physical parameters in the space. Unfortunately, CFD simulations require very high computation time and hence full representation of the primary circuit of a PWR is currently not feasible. In order to overcome the deficiencies of CFD and system codes, different approaches are used by the scientists dealing with complex fluid flows. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Papukchiev, Angel; Lerchl, Georg [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany)
2010-05-15
Thermal hydraulic system codes have been extensively developed by the nuclear industry, research institutes and technical safety organizations with the goal to improve the design and safety of nuclear installations. A large number of these simulation tools are based on the lumped parameter theory. Such programs use networks consisting of 1D cells, where mass, momentum and energy equations are solved for each fluid phase and balanced over each node of the network. System codes are extensively validated against experiments and provide reliable results at low computational cost. Lump parameter programs use simplifications in the mathematical models describing the simulated systems. Balance equations for mass, momentum and energy for two phases are obtained by averaging of the local basic flow equations in the space. As a result, mean values for relevant physical parameters which in reality are spatially distributed fields are calculated. However, since relevant reactor fluid flow and heat transfer phenomena are 3D in nature, 1D system codes have limitations on their application for specific nuclear reactor safety (NRS) problems with pronounced 3D phenomena like boron dilution, pressurized thermal shock and main steam line break. Modern computational fluid dynamics (CFD) codes are capable to predict fluid flow behavior in complex geometries and can provide detailed distribution of the physical parameters in the space. Unfortunately, CFD simulations require very high computation time and hence full representation of the primary circuit of a PWR is currently not feasible. In order to overcome the deficiencies of CFD and system codes, different approaches are used by the scientists dealing with complex fluid flows. (orig.)
Drug-laden 3D biodegradable label using QR code for anti-counterfeiting of drugs.
Fei, Jie; Liu, Ran
2016-06-01
Wiping out counterfeit drugs is a great task for public health care around the world. The boost of these drugs makes treatment to become potentially harmful or even lethal. In this paper, biodegradable drug-laden QR code label for anti-counterfeiting of drugs is proposed that can provide the non-fluorescence recognition and high capacity. It is fabricated by the laser cutting to achieve the roughness over different surface which causes the difference in the gray levels on the translucent material the QR code pattern, and the micro mold process to obtain the drug-laden biodegradable label. We screened biomaterials presenting the relevant conditions and further requirements of the package. The drug-laden microlabel is on the surface of the troches or the bottom of the capsule and can be read by a simple smartphone QR code reader application. Labeling the pill directly and decoding the information successfully means more convenient and simple operation with non-fluorescence and high capacity in contrast to the traditional methods. PMID:27040262
Mixing effect in the solution of the AER6 Benchmark problem by KIKO3D/ATHLET code system
International Nuclear Information System (INIS)
The former result calculated with the coupled KIKO3D/ATHLET code system for the sixth dynamic benchmark problem is presented and compared with a new one. The only difference between the two calculations is the slightly different nodalization in the core vessel. Though it is a physically plausible fact that the lack of mixing in the upper plenum causes a considerable change in the results a rough nodalization above the core is widely used as it makes easier. The effect of this simplification is investigated (Authors)
Experiences with the coupled code system S3R/RELAP5-3D in training simulators
International Nuclear Information System (INIS)
The paper describes the implementation of S3R (core neutronics) and RELAP5-3D (RCS thermal-hydraulics) in a PWR training simulator for the Grohnde NPP located at the Kraftwerks-Simulator-Gesellschaft (KSG) in Essen, Germany. The models are briefly described as well as the coupling between these two codes and the interface with the rest of the simulator. The paper also describes the procedure that will be used to update the simulator core data after future core reloads. Results from the integrated simulator are presented. (orig.)
Integrating bar-code devices with computerized MC and A systems
International Nuclear Information System (INIS)
Over the past seven years, Los Alamos National Laboratory developed several generations of computerized nuclear materials control and accountability (MC and A) systems for tracking and reporting the storage, movement, and management of nuclear materials at domestic and international facilities. During the same period, Oak Ridge National Laboratory was involved with automated data acquisition (ADA) equipment, including installation of numerous bar-code scanning stations at various facilities to serve as input devices to computerized systems. Bar-code readers, as well as other ADA devices, reduce input errors, provide faster input, and allow the capture of data in remote areas where workstations do not exist. Los Alamos National Laboratory and Oak Ridge National Laboratory teamed together to implement the integration of bar-code hardware technology with computerized MC and A systems. With the expertise of both sites, the two technologies were successfully merged with little difficulty. Bar-code input is now available with several functions of the MC and A systems: material movements within material balance areas (MBAs), material movements between MBAs, and physical inventory verification. This paper describes the various components required for the integration of these MC and A systems with the installed bar-code reader devices and the future directions for these technologies
3-D Monte Carlo neutron-photon transport code JMCT and its algorithms
International Nuclear Information System (INIS)
JMCT Monte Carlo neutron and photon transport code has been developed which is based on the JCOGIN toolbox. JCOGIN includes the geometry operation, tally, the domain decomposition and the parallel computation about particle (MPI) and spatial domain (OpenMP) etc. The viewdata of CAD is equipped in JMCT preprocessor. The full-core pin-mode, which is from Chinese Qinshan-II nuclear power station, is design and simulated by JMCT. The detail pin-power distribution and keff results are shown in this paper. (author)
Predictions of bubbly flows in vertical pipes using two-fluid models in CFDS-FLOW3D code
Energy Technology Data Exchange (ETDEWEB)
Banas, A.O.; Carver, M.B. [Chalk River Laboratories (Canada); Unrau, D. [Univ. of Toronto (Canada)
1995-09-01
This paper reports the results of a preliminary study exploring the performance of two sets of two-fluid closure relationships applied to the simulation of turbulent air-water bubbly upflows through vertical pipes. Predictions obtained with the default CFDS-FLOW3D model for dispersed flows were compared with the predictions of a new model (based on the work of Lee), and with the experimental data of Liu. The new model, implemented in the CFDS-FLOW3D code, included additional source terms in the {open_quotes}standard{close_quotes} {kappa}-{epsilon} transport equations for the liquid phase, as well as modified model coefficients and wall functions. All simulations were carried out in a 2-D axisymmetric format, collapsing the general multifluid framework of CFDS-FLOW3D to the two-fluid (air-water) case. The newly implemented model consistently improved predictions of radial-velocity profiles of both phases, but failed to accurately reproduce the experimental phase-distribution data. This shortcoming was traced to the neglect of anisotropic effects in the modelling of liquid-phase turbulence. In this sense, the present investigation should be considered as the first step toward the ultimate goal of developing a theoretically sound and universal CFD-type two-fluid model for bubbly flows in channels.
OpenMC: A state-of-the-art Monte Carlo code for research and development
International Nuclear Information System (INIS)
Highlights: • OpenMC is an open source Monte Carlo particle transport code. • Solid geometry and continuous-energy physics allow high-fidelity simulations. • Development has focused on high performance and modern I/O techniques. • OpenMC is capable of scaling up to hundreds of thousands of processors. • Other features include plotting, CMFD acceleration, and variance reduction. - Abstract: This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently developed at the Massachusetts Institute of Technology. OpenMC uses continuous-energy cross sections and a constructive solid geometry representation, enabling high-fidelity modeling of nuclear reactors and other systems. Modern, portable input/output file formats are used in OpenMC: XML for input, and HDF5 for output. High performance parallel algorithms in OpenMC have demonstrated near-linear scaling to over 100,000 processors on modern supercomputers. Other topics discussed in this paper include plotting, CMFD acceleration, variance reduction, eigenvalue calculations, and software development processes
Boiling water flows. A local wall heat transfer model for use in an Eulerian 3-D computer code
International Nuclear Information System (INIS)
Electricite de France is currently developing a 3-D computer code for the Eulerian simulation of two-phase flows. This code, named ASTRID, is based on the six-equation two-fluid model. Boiling water flows are among the main applications of ASTRID, especially for nuclear power plant design. In order to provide ASTRID with appropriate closure laws and boundary conditions, Electricite de France and the Institut de Mecanique des Fluides de Toulouse (IMFT) have collaborated since 1991. The analysis of the current knowledge made possible to build a first set of closure laws and boundary conditions for boiling water flows, suitable for ASTRID. This paper is focused on the model used for heat transfer and bubble production at the wall, in a convective boiling situation. This model has been tested for a first comparison with existing experimental data. The results of this comparison are also presented here. (authors). 5 figs., 9 refs
Modeling TRIGA reactor pulses using the STAR 3D nodal kinetics and WIMS-D4 codes
International Nuclear Information System (INIS)
A detailed three-dimensional (3D) time-dependent STAR nodal kinetics model coupled to a one-dimensional (1D) thermal-hydraulics WIGL model has been developed to describe and benchmark the peak power and pulse behavior of the Penn State University (PSU) Breazeale TRIGA reactor. Different core loading patterns were used for several TRIGA pulse tests with different reactivity insertion worths (1.5 dollar, 2.0 dollar, 2.5 dollar). The STAR nodal kinetics code and TRIGA model adequately simulates TRIGA pulses when group constants are generated from physics codes (i.e., WIMS-D4) that can accurately model the TRIGA uranium-zirconium-hydride fuel
International Nuclear Information System (INIS)
Highlights: • Overview of the capabilities and features of the MC21 Monte Carlo code, version 6. • Detailed description of in-line reactor feedback capabilities in MC21. • Discussion of running strategies for Monte Carlo simulations with feedback effects. • Includes representative MC21 results for massively-parallel 3D reactor simulations. - Abstract: MC21 is a continuous-energy Monte Carlo radiation transport code for the calculation of the steady-state spatial distributions of reaction rates in three-dimensional models. The code supports neutron and photon transport in fixed source problems, as well as iterated-fission-source (eigenvalue) neutron transport problems. MC21 has been designed and optimized to support large-scale problems in reactor physics, shielding, and criticality analysis applications. The code also supports many in-line reactor feedback effects, including depletion, thermal feedback, xenon feedback, eigenvalue search, and neutron and photon heating. MC21 uses continuous-energy neutron/nucleus interaction physics over the range from 10−5 eV to 20 MeV. The code treats all common neutron scattering mechanisms, including fast-range elastic and non-elastic scattering, and thermal- and epithermal-range scattering from molecules and crystalline materials. For photon transport, MC21 uses continuous-energy interaction physics over the energy range from 1 keV to 100 GeV. The code treats all common photon interaction mechanisms, including Compton scattering, pair production, and photoelectric interactions. All of the nuclear data required by MC21 is provided by the NDEX system of codes, which extracts and processes data from EPDL-, ENDF-, and ACE-formatted source files. For geometry representation, MC21 employs a flexible constructive solid geometry system that allows users to create spatial cells from first- and second-order surfaces. The system also allows models to be built up as hierarchical collections of previously defined spatial cells
Code and Solution Verification of 3D Numerical Modeling of Flow in the Gust Erosion Chamber
Yuen, A.; Bombardelli, F. A.
2014-12-01
Erosion microcosms are devices commonly used to investigate the erosion and transport characteristics of sediments at the bed of rivers, lakes, or estuaries. In order to understand the results these devices provide, the bed shear stress and flow field need to be accurately described. In this research, the UMCES Gust Erosion Microcosm System (U-GEMS) is numerically modeled using Finite Volume Method. The primary aims are to simulate the bed shear stress distribution at the surface of the sediment core/bottom of the microcosm, and to validate the U-GEMS produces uniform bed shear stress at the bottom of the microcosm. The mathematical model equations are solved by on a Cartesian non-uniform grid. Multiple numerical runs were developed with different input conditions and configurations. Prior to developing the U-GEMS model, the General Moving Objects (GMO) model and different momentum algorithms in the code were verified. Code verification of these solvers was done via simulating the flow inside the top wall driven square cavity on different mesh sizes to obtain order of convergence. The GMO model was used to simulate the top wall in the top wall driven square cavity as well as the rotating disk in the U-GEMS. Components simulated with the GMO model were rigid bodies that could have any type of motion. In addition cross-verification was conducted as results were compared with numerical results by Ghia et al. (1982), and good agreement was found. Next, CFD results were validated by simulating the flow within the conventional microcosm system without suction and injection. Good agreement was found when the experimental results by Khalili et al. (2008) were compared. After the ability of the CFD solver was proved through the above code verification steps. The model was utilized to simulate the U-GEMS. The solution was verified via classic mesh convergence study on four consecutive mesh sizes, in addition to that Grid Convergence Index (GCI) was calculated and based on
International Nuclear Information System (INIS)
In this report we describe theory and 3D full wave code description for the wave excitation, propagation and absorption in 3-dimensional (3D) stellarator equilibrium high beta plasma in ion cyclotron frequency range (ICRF). This theory forms a basis for a 3D code creation, urgently needed for the ICRF heating scenarios development for the operated LHD, constructed W7-X, NCSX and projected CSX3 stellarators, as well for re evaluation of ICRF scenarios in operated tokamaks and in the ITER . The theory solves the 3D Maxwell-Vlasov antenna-plasma-conducting shell boundary value problem in the non-orthogonal flux coordinates (Ψ, θ, (varphi)), Ψ being magnetic flux function, θ and (varphi) being the poloidal and toroidal angles, respectively. All basic physics, like wave refraction, reflection and diffraction are self consistently included, along with the fundamental ion and ion minority cyclotron resonances, two ion hybrid resonance, electron Landau and TTMP absorption. Antenna reactive impedance and loading resistance are also calculated and urgently needed for an antenna -generator matching. This is accomplished in a real confining magnetic field being varying in a plasma major radius direction, in toroidal and poloidal directions, through making use of the hot dense plasma wave induced currents with account to the finite Larmor radius effects. We expand the solution in Fourier series over the toroidal ((varphi)) and poloidal (θ) angles and solve resulting ordinary differential equations in a radial like Ψ-coordinate by finite difference method. The constructed discretization scheme is divergent-free one, thus retaining the basic properties of original equations. The Fourier expansion over the angle coordinates has given to us the possibility to correctly construct the ''parallel'' wave number k//, and thereby to correctly describe the ICRF waves absorption by a hot plasma. The toroidal harmonics are tightly coupled with each other due to magnetic field
OpenMC: a state-of-the-Art Monte Carlo code for research and development
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This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently developed at the Massachusetts Institute of Technology. OpenMC uses continuous-energy cross sections and a constructive solid geometry representation, enabling high-fidelity modeling of nuclear reactors and other systems. Modern, portable input/output file formats are used in OpenMC: XML for input, and HDF5 for output. High performance parallel algorithms in OpenMC have demonstrated near-linear scaling to over 100,000 processors on modern supercomputers. Other topics discussed in this paper include plotting, CMFD acceleration, variance reduction, eigenvalue calculations, and software development processes. (authors)
MC++: A parallel, portable, Monte Carlo neutron transport code in C++
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MC++ is an implicit multi-group Monte Carlo neutron transport code written in C++ and based on the Parallel Object-Oriented Methods and Applications (POOMA) class library. MC++ runs in parallel on and is portable to a wide variety of platforms, including MPPs, SMPs, and clusters of UNIX workstations. MC++ is being developed to provide transport capabilities to the Accelerated Strategic Computing Initiative (ASCI). It is also intended to form the basis of the first transport physics framework (TPF), which is a C++ class library containing appropriate abstractions, objects, and methods for the particle transport problem. The transport problem is briefly described, as well as the current status and algorithms in MC++ for solving the transport equation. The alpha version of the POOMA class library is also discussed, along with the implementation of the transport solution algorithms using POOMA. Finally, a simple test problem is defined and performance and physics results from this problem are discussed on a variety of platforms
DISCO: A 3D Moving-Mesh Magnetohydrodynamics Code Designed for the Study of Astrophysical Disks
Duffell, Paul C
2016-01-01
This work presents the publicly available moving-mesh magnetohydrodynamics code DISCO. DISCO is efficient and accurate at evolving orbital fluid motion in two and three dimensions, especially at high Mach number. DISCO employs a moving-mesh approach utilizing a dynamic cylindrical mesh that can shear azimuthally to follow the orbital motion of the gas. The moving mesh removes diffusive advection errors and allows for longer timesteps than a static grid. Magnetohydrodynamics is implemented in DISCO using an HLLD Riemann solver and a novel constrained transport scheme which is compatible with the mesh motion. DISCO is tested against a wide variety of problems, which are designed to test its stability, accuracy and scalability. In addition, several magnetohydrodynamics tests are performed which demonstrate the accuracy and stability of the new constrained transport approach, including two tests of the magneto-rotational instability (MRI); one testing the linear growth rate and the other following the instability...
Implementation of a 3D plasma particle-in-cell code on a MIMD parallel computer
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A three-dimensional plasma particle-in-cell (PIC) code has been implemented on the Intel Delta MIMD parallel supercomputer using the General Concurrent PIC algorithm. The GCPIC algorithm uses a domain decomposition to divide the computation among the processors: A processor is assigned a subdomain and all the particles in it. Particles must be exchanged between processors as they move. Results are presented comparing the efficiency for 1-, 2- and 3-dimensional partitions of the three dimensional domain. This algorithm has been found to be very efficient even when a large fraction (e.g. 30%) of the particles must be exchanged at every time step. On the 512-node Intel Delta, up to 125 million particles have been pushed with an electrostatic push time of under 500 nsec/particle/time step
Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code
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Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and not-permissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a German BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second - much smaller - maximum that would occur around one second after the first one in the absence of a SCRAM. (author)
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The estimation methodology of the mass and energy (M/E) release in the main steam line break (MSLB) has been developed with the RETRAN-3D code. In the case of equipment qualification (EQ), the over-estimated temperature would exceed the design limits of some cables or valves. In order to have a more flexible EQ profiles from the MSLB M/E release, the methodology with the best-estimated code was used. The major conditions affecting the MSLB M/E were found to be the initial SG level, heat transfer between primary and secondary sides, power level, operable protection system, main or auxiliary feedwater availability, and break conditions. The RETRAN-3D models were developed for the Kori unit 1(KRN-1) which is typical two loop Westinghouse (WH) designed plant. Particularly, a detailed model of the steam generators was developed to estimate a more realistic two-phase heat transfer effect of the steam flow. After the modeling, the methodology has been developed through the sensitivity analyses. The M/E release data generated from the analyses have been used as the input to the inside containment pressure and temperature (P/T) analysis. According to the results at the point of view containment P/T, the Kori unit 1 can have more margin of 5 - 15 kPa in pressure and 8 - 15degC in temperature. (author)
Thermal-hydraulic system study of a high pressure, high temperature helium loop using RELAP5-3D code
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Highlights: ► A thermal-hydraulic system analysis for a high pressure, high temperature helium loop has been investigated. ► The loop belongs to the Helium Loop Karlsruhe (HELOKA) facility, which contains the European Helium Cooled Pebble Beds Test Blanket Module (HCPB TBM) as the test module. ► The loop including all components has been modeled using the system code REALP5-3D, and the main control strategy has been implemented as well. ► With this model, the loop dynamics in conditions relevant for blanket module operation have been demonstrated. - Abstract: The thermal-hydraulic system analysis for the Helium Loop Karlsruhe (HELOKA) facility, a high pressure, high temperature experimental helium loop having the European Helium Cooled Pebble Beds Test Blanket Module (HCPB TBM) as the test module, was investigated. Using the system code REALP5-3D, all components in the loop are modeled as well as the main control strategy. With this model, the loop dynamics in conditions relevant for blanket module operation are simulated and analyzed.
Benchmark of coupling codes (ALOHA, TOPLHA and GRILL3D) with ITER-relevant Lower Hybrid antenna
Energy Technology Data Exchange (ETDEWEB)
Milanesio, D., E-mail: daniele.milanesio@polito.it [Politecnico di Torino, Dipartimento di Elettronica, Torino (Italy); Hillairet, J. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Panaccione, L. [Associazione ENEA-EURATOM sulla Fusione, C.R. Frascati (Italy); Maggiora, R. [Politecnico di Torino, Dipartimento di Elettronica, Torino (Italy); Artaud, J.F. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Bae, Y.S. [Fusion Research Institute, Daejeon (Korea, Republic of); Barbera, A.M.A. [Politecnico di Torino, Dipartimento di Energetica, Torino (Italy); Belo, J. [Euratom-IST, Centro de Fusao Nuclear, Lisboa (Portugal); Berger-By, G.; Bernard, J.M.; Cara, Ph. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Cardinali, A.; Castaldo, C.; Ceccuzzi, S.; Cesario, R. [Associazione ENEA-EURATOM sulla Fusione, C.R. Frascati (Italy); Decker, J.; Delpech, L.; Ekedahl, A.; Garcia, J.; Garibaldi, P. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France)
2011-10-15
In order to assist the design of the future ITER Lower Hybrid launcher, coupling codes ALOHA, from CEA/IRFM, TOPLHA, from Politecnico di Torino, and GRILL3D, developed by Dr. Mikhail Irzak (A.F. Ioffe Physico-Technical Institute, St. Petersburg, Russia) and operated by ENEA Frascati, have been compared with the updated (six modules with four active waveguides per module) Passive-Active Multi-junction (PAM) Lower Hybrid antennas. Both ALOHA and GRILL3D formulate the problem in terms of rectangular waveguides modes, while TOPLHA is based on boundary-value problem with the adoption of a triangular cell-mesh to represent the relevant waveguides surfaces. Several plasma profiles, with varying edge density and density increase, have been adopted to provide a complete description of the simulated launcher in terms of reflection coefficient, computed at the beginning of each LH module, and of power spectra. Good agreement can be observed among codes for all the simulated profiles.
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The Nuclear Steam Supply System (NSSS) thermal-hydraulic model adopted in the Korea Nuclear Plant Education Center (KNPEC)-2 simulator was provided in the early 1980s. The reference plant for KNPEC-2 is the Yong Gwang Nuclear Unit 1, which is a Westinghouse-type 3-loop, 950 MW(electric) pressurized water reactor. Because of the limited computational capability at that time, it uses overly simplified physical models and assumptions for a real-time simulation of NSSS thermal-hydraulic transients. This may entail inaccurate results and thus, the possibility of so-called ''negative training,'' especially for complicated two-phase flows in the reactor coolant system. To resolve the problem, we developed a realistic NSSS thermal-hydraulic program (named ARTS code) based on the best-estimate code RETRAN-3D. The systematic assessment of ARTS has been conducted by both a stand-alone test and an integrated test in the simulator environment. The non-integrated stand-alone test (NIST) results were reasonable in terms of accuracy, real-time simulation capability, and robustness. After successful completion of the NIST, ARTS was integrated with a 3-D reactor kinetics model and other system models. The site acceptance test (SAT) has been completed successively and confirmed to comply with the ANSI/ANS-3.5-1998 simulator software performance criteria. This paper presents our efforts for the ARTS development and some test results of the NIST and SAT
Benchmark of coupling codes (ALOHA, TOPLHA and GRILL3D) with ITER-relevant Lower Hybrid antenna
International Nuclear Information System (INIS)
In order to assist the design of the future ITER Lower Hybrid launcher, coupling codes ALOHA, from CEA/IRFM, TOPLHA, from Politecnico di Torino, and GRILL3D, developed by Dr. Mikhail Irzak (A.F. Ioffe Physico-Technical Institute, St. Petersburg, Russia) and operated by ENEA Frascati, have been compared with the updated (six modules with four active waveguides per module) Passive-Active Multi-junction (PAM) Lower Hybrid antennas. Both ALOHA and GRILL3D formulate the problem in terms of rectangular waveguides modes, while TOPLHA is based on boundary-value problem with the adoption of a triangular cell-mesh to represent the relevant waveguides surfaces. Several plasma profiles, with varying edge density and density increase, have been adopted to provide a complete description of the simulated launcher in terms of reflection coefficient, computed at the beginning of each LH module, and of power spectra. Good agreement can be observed among codes for all the simulated profiles.
Crest Factor Reduction in MC-CDMA Employing Carrier Interferometry Codes
Directory of Open Access Journals (Sweden)
Natarajan Balasubramaniam
2004-01-01
Full Text Available This paper addresses signal compactness issues in MC-CDMA employing carrier interferometry codes using the measure of crest factor (CF. Carrier interferometry codes, applied to N -carrier MC-CDMA systems, enable 2N users to simultaneously share the system bandwidth with minimal degradation in performance (relative to the N -orthogonal-user case. First, for a fully loaded ( K=N and K=2N users MC-CDMA system with practical values of N , it is shown that the CF in downlink transmission demonstrates desirable properties of low mean and low variance. The downlink CF degrades when the number of users in the system decreases. Next, the high CF observed in the uplink is characterized and the poor CF in a partially loaded downlink as well as uplink is effectively combated using Schroeder's analytical CF reduction techniques.
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The coolability of particulate debris beds is a major issue during the course of a severe accident. In case that water is injected into the containment/reactor cavity as an accident mitigation measure there is the potential to stop the accident or at least mitigate its consequences if coolability of the debris bed can be achieved. Assessment of pros and cons of such measures requires sufficient knowledge and thus creates a challenge for the further development of severe accident codes. To give a more realistic description of the coolability during a severe accident, when molten corium is released into the reactor cavity and a debris bed is built up due to fragmentation of the ejected molten corium jet, a three dimensional code for MElt and WAter interaction, MEWA 3D, is being developed at developed at IKE, University of Stuttgart. MEWA 3D simulates the boil-off and quenching behavior of particulate or porous debris beds in multidimensional flow conditions. The aim is to describe the in- and ex-vessel behavior of solidified corium during the late phase of severe accidents in light water reactors, taking into account processes of core heat-up, melting, degradation and relocation either to the lower plenum or to the cavity. A brief description of the major modeling assumptions, governing conservation equations and constitutive laws used in the code will be given. The focus of this contribution is on the investigation of multidimensional aspects of the cooling behavior in order to analyze the coolability under reactor conditions. Taking into account the lateral pouring of the melt into the spreading room, the bed configuration most likely will be truly three-dimensional. For realistic simulations, it is essential to consider the real geometrical conditions, e.g. non-symmetrical debris configurations as result of the non-central core relocation. For the validation of the code for long-term coolability, MEWA 3D is applied to perform calculations of the COOLOCE
3-D Analysis of Natural Circulation in PCCT of PAFS using the CUPID Code
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Lee, Seung Jun; Yoon, Han Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2013-10-15
A three.dimensional thermal-hydraulic code, CUPID, which has been developed for the component scale analysis of a nuclear reactor, is used for the analysis of PAFS (Passive Auxiliary Feedwater System). PAFS is one of the safety features of APR+ (Advanced Power Reactor Plus), which intended to completely replace the conventional active feedwater system. The physical phenomena in PAFS include various thermal hydraulic issues such as flashing, subcoolled boiling, single- and two-phase natural circulation, swelling and boil-off. Moreover, all those thermal hydraulic phenomena happen in multi-dimensional way due to the geometrical features of PAFS. A preliminary study was done for a secondary system of the real scale PAFS with CUPID. A preliminary analysis for the real scale PAFS PCCT was performed. The collapsed water level coincides with the expected results. The temperature and velocity field showed the single-phase natural circulation mode as shown in previous PASCAL simulations. Finally, void fraction is appeared to be asymmetric in flashing so that it may affect strongly the convection form after saturation.
3D relaxation MHD modeling with FOI-PERFECT code for electromagnetically driven HED systems
Wang, Ganghua; Duan, Shuchao; Xie, Weiping; Kan, Mingxian; Institute of Fluid Physics Collaboration
2015-11-01
One of the challenges in numerical simulations of electromagnetically driven high energy density (HED) systems is the existence of vacuum region. The electromagnetic part of the conventional model adopts the magnetic diffusion approximation (magnetic induction model). The vacuum region is approximated by artificially increasing the resistivity. On one hand the phase/group velocity is superluminal and hence non-physical in the vacuum region, on the other hand a diffusion equation with large diffusion coefficient can only be solved by implicit scheme. Implicit method is usually difficult to parallelize and converge. A better alternative is to solve the full electromagnetic equations for the electromagnetic part. Maxwell's equations coupled with the constitutive equation, generalized Ohm's law, constitute a relaxation model. The dispersion relation is given to show its transition from electromagnetic propagation in vacuum to resistive MHD in plasma in a natural way. The phase and group velocities are finite for this system. A better time stepping is adopted to give a 3rd full order convergence in time domain without the stiff relaxation term restriction. Therefore it is convenient for explicit & parallel computations. Some numerical results of FOI-PERFECT code are also given. Project supported by the National Natural Science Foundation of China (Grant No. 11172277,11205145).
3-D Analysis of Natural Circulation in PCCT of PAFS using the CUPID Code
International Nuclear Information System (INIS)
A three.dimensional thermal-hydraulic code, CUPID, which has been developed for the component scale analysis of a nuclear reactor, is used for the analysis of PAFS (Passive Auxiliary Feedwater System). PAFS is one of the safety features of APR+ (Advanced Power Reactor Plus), which intended to completely replace the conventional active feedwater system. The physical phenomena in PAFS include various thermal hydraulic issues such as flashing, subcoolled boiling, single- and two-phase natural circulation, swelling and boil-off. Moreover, all those thermal hydraulic phenomena happen in multi-dimensional way due to the geometrical features of PAFS. A preliminary study was done for a secondary system of the real scale PAFS with CUPID. A preliminary analysis for the real scale PAFS PCCT was performed. The collapsed water level coincides with the expected results. The temperature and velocity field showed the single-phase natural circulation mode as shown in previous PASCAL simulations. Finally, void fraction is appeared to be asymmetric in flashing so that it may affect strongly the convection form after saturation
Modelling transient 3D multi-phase criticality in fluidised granular materials - the FETCH code
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The development and application of a generic model for modelling criticality in fluidised granular materials is described within the Finite Element Transient Criticality (FETCH) code - which models criticality transients in spatial and temporal detail from fundamental principles, as far as is currently possible. The neutronics model in FETCH solves the neutron transport in full phase space with a spherical harmonics angle of travel representation, multi-group in neutron energy, Crank Nicholson based in time stepping, and finite elements in space. The fluids representation coupled with the neutronics model is a two-fluid-granular-temperature model, also finite element fased. A separate fluid is used to represent the liquid/vapour gas and the solid fuel particle phases, respectively. Particle-particle, particle-wall interactions are modelled using a kinetic theory approach on an analogy between the motion of gas molecules subject to binary collisions and granular flows. This model has been extensively validated by comparison with fluidised bed experimental results. Gas-fluidised beds involve particles that are often extremely agitated (measured by granular temperature) and can thus be viewed as a particularly demanding application of the two-fluid model. Liquid fluidised systems are of criticality interest, but these can become demanding with the production of gases (e.g. radiolytic and water vapour) and large fluid/particle velocities in energetic transients. We present results from a test transient model in which fissile material (239Pu) is presented as spherical granules subsiding in water, located in a tank initially at constant temperature and at two alternative over-pressures in order to verify the theoretical model implemented in FETCH. (author)
TDTORT: Time-Dependent, 3-D, Discrete Ordinates, Neutron Transport Code System with Delayed Neutrons
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1 - Description of program or function: TDTORT solves the time-dependent, three-dimensional neutron transport equation with explicit representation of delayed neutrons to estimate the fission yield from fissionable material transients. This release includes a modified version of TORT from the C00650MFMWS01 DOORS3.1 code package plus the time-dependent TDTORT code. GIP is also included for cross-section preparation. TORT calculates the flux or fluence of particles due to particles incident upon the system from extraneous sources or generated internally as a result of interaction with the system in two- or three-dimensional geometric systems. The principle application is to the deep-penetration transport of neutrons and photons. Reactor eigenvalue problems can also be solved. Numerous printed edits of the results are available, and results can be transferred to output files for subsequent analysis. TDTORT reads ANISN-format cross-section libraries, which are not included in the package. Users may choose from several available in RSICC's data library collection which can be identified by the keyword 'ANISN FORMAT'. 2 - Methods:The time-dependent spatial flux is expressed as a product of a space-, energy-, and angle-dependent shape function, which is usually slowly varying in time and a purely time-dependent amplitude function. The shape equation is solved for the shape using TORT; and the result is used to calculate the point kinetics parameters (e.g., reactivity) by using their inner product definitions, which are then used to solve the time-dependent amplitude and precursor equations. The amplitude function is calculated by solving the kinetics equations using the LSODE solver. When a new shape calculation is needed, the flux is calculated using the newly computed amplitude function. The Boltzmann transport equation is solved using the method of discrete ordinates to treat the directional variable and weighted finite-difference methods, in addition to Linear Nodal
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TRAB-3D models the PWR and BWR reactor core using the two-group diffusion equations in homogenized fuel assembly geometry with a sophisticated nodal method. Thermohydraulics is described using four equation formulation. The stand-alone version of the code also describes thermohydraulics of the rest of the BWR circuit with one dimensional components. The SMABRE code models the thermohydraulics of light water reactors. The five equation formulation with the drift flux phase separation is modelling the two-phase behaviour. Conservation equations are solved for the phase mass, mixture momentum and phase energy. Additional equations are for the noncondensable part in gas and boron in liquid. The TRAB-3D and SMABRE codes have been coupled earlier by using the parallel coupling principle, where in the core section the 3-dimensional TRAB core, and the parallel channel coarse SMABRE core are solved in parallel, but the rest of the circulation system is solved with SMABRE. As a new development the internal coupling to meet new requirements for the PWR and BWR transient analyses is being realised. Both the circuit and core thermohydraulics are solved in SMABRE. The core thermohydraulics solution inside the core wide iterations is repeated to allow rapid power changes. These are the fast pressure changes, control rod ejection and ATWS. The numerical solution in SMABRE has been improved to allow full core simulation with separate flow channel for each fuel element of a BWR core. For the PWR plants the method is used as well by simulating the core by one-dimensional parallel channels. New development is needed for the open core calculation. (authors)
Energy Technology Data Exchange (ETDEWEB)
Martinez, M.; Miro, R.; Barrachina, T.; Verdu, G.
2011-07-01
This paper presents the results from the calculation of the steady state simulation with model of CFD (computational fluid dynamic) operating under conditions of operation at full power (Hot Full Power). Development and the CFD model results show the usefulness of these codes for calculating 3D of the variable thermohydraulics of these reactors.
International Nuclear Information System (INIS)
Full text of publication follows: TRAB-3D models the PWR and BWR reactor core using the two-group diffusion equations in homogenized fuel assembly geometry with a sophisticated nodal method. Thermohydraulics is described using four-equation formulation. The stand-alone version of the code also describes thermohydraulics of the rest of the BWR circuit with one-dimensional components. The SMABRE code models the thermohydraulics of light water reactors. The five-equation formulation with the drift flux phase separation is modelling the two-phase behaviour. Conservation equations are solved for the phase mass, mixture momentum and phase energy. Additional equations are for the noncondensable in gas and boron in liquid. The TRAB-3D and SMABRE codes have been coupled earlier by using the parallel coupling principle, where in the core section the 3-dimensional TRAB core, and the parallel channel coarse SMABRE core are solved in parallel, but rest of the circulation system is solved with SMABRE. As a new development the internal coupling to meet new requirements for the PWR and BWR transient analyses is being realised. Both the circuit and core thermohydraulics are solved in SMABRE. The core thermohydraulics solution inside the core wide iterations is repeated to allow rapid power changes. These are the fast pressure changes, control rod ejection and ATWS. The numerical solution in SMABRE has been improved to allow full core simulation with separate flow channel for each fuel element of a BWR core. For the PWR plants the method is used as well by simulating the core by one-dimensional parallel channels. New development is needed for the open core calculation. In general questions could be raised, what advantages are seen with the new internal coupling in comparison with the earlier realised parallel coupling, and which advantages may be seen in building the realtor physical model on the basis of the old code, developed since 1970's. The internal coupling allows modelling
Qualification of the 3D thermal hydraulics model of the code system TRACE based on plant data
International Nuclear Information System (INIS)
In the frame of the VVER-1000 Coolant Transient Benchmark Phase-1 the coupled code RELAP5/PARCS has been extensively assessed. The Phase-2 of this benchmark - currently underway - focuses on both multidimensional thermal hydraulics phenomena within the reactor pressure vessel (RPV) such as coolant mixing and core physics. Hence it is an excellent opportunity to qualify the prediction capability of the new coupled code system TRACE/PARCS taking into account plant data obtained from the Kozloduy nuclear power plant unit 6. In addition a lose coupling of CFX with RELAP5 is applied for the posttest calculation of the coolant mixing experiment. The developed multidimensional models of the VVER-1000 reactor pressure vessel as well as the performed calculations using these models are described in some detail. The predicted results are in good agreement with the data. It was demonstrated that the chosen 3D-nodalization of the RPV is adequate for the description of the coolant mixing phenomena in a VVER-1000 reactor. In addition selected results of the code TRACE/PARCS for a postulated main steam line transient (MSLB) are given. The investigations have shown that the multidimensional neutronics and thermal hydraulic model developed for the RPV of the VVER-1000 reactor are well qualified and consequently they are ready for their integration into a overall plant model so that the exercise 3 of the Phase 2 can be investigated as next. (authors)
Directory of Open Access Journals (Sweden)
Koniges Alice
2013-11-01
Full Text Available The Neutralized Drift Compression Experiment II (NDCX II is an induction accelerator planned for initial commissioning in 2012. The final design calls for a 3 MeV, Li+ ion beam, delivered in a bunch with characteristic pulse duration of 1 ns, and transverse dimension of order 1 mm. The NDCX II will be used in studies of material in the warm dense matter (WDM regime, and ion beam/hydrodynamic coupling experiments relevant to heavy ion based inertial fusion energy. We discuss recent efforts to adapt the 3D ALE-AMR code to model WDM experiments on NDCX II. The code, which combines Arbitrary Lagrangian Eulerian (ALE hydrodynamics with Adaptive Mesh Refinement (AMR, has physics models that include ion deposition, radiation hydrodynamics, thermal diffusion, anisotropic material strength with material time history, and advanced models for fragmentation. Experiments at NDCX-II will explore the process of bubble and droplet formation (two-phase expansion of superheated metal solids using ion beams. Experiments at higher temperatures will explore equation of state and heavy ion fusion beam-to-target energy coupling efficiency. Ion beams allow precise control of local beam energy deposition providing uniform volumetric heating on a timescale shorter than that of hydrodynamic expansion. We also briefly discuss the effects of the move to exascale computing and related computational changes on general modeling codes in fusion.
Qualification of the 3D thermal hydraulics model of the code system TRACE based on plant data
Energy Technology Data Exchange (ETDEWEB)
Sanchez, V.H.; Jager, W. [Forschungzentrum Karlsruhe (FZK), Institute of Reactor Safety (IRS) (Germany); Kozlowski, T. [Royal Institute of Technology (KTH), Stockholm (Sweden)
2007-07-01
In the frame of the VVER-1000 Coolant Transient Benchmark Phase-1 the coupled code RELAP5/PARCS has been extensively assessed. The Phase-2 of this benchmark - currently underway - focuses on both multidimensional thermal hydraulics phenomena within the reactor pressure vessel (RPV) such as coolant mixing and core physics. Hence it is an excellent opportunity to qualify the prediction capability of the new coupled code system TRACE/PARCS taking into account plant data obtained from the Kozloduy nuclear power plant unit 6. In addition a lose coupling of CFX with RELAP5 is applied for the posttest calculation of the coolant mixing experiment. The developed multidimensional models of the VVER-1000 reactor pressure vessel as well as the performed calculations using these models are described in some detail. The predicted results are in good agreement with the data. It was demonstrated that the chosen 3D-nodalization of the RPV is adequate for the description of the coolant mixing phenomena in a VVER-1000 reactor. In addition selected results of the code TRACE/PARCS for a postulated main steam line transient (MSLB) are given. The investigations have shown that the multidimensional neutronics and thermal hydraulic model developed for the RPV of the VVER-1000 reactor are well qualified and consequently they are ready for their integration into a overall plant model so that the exercise 3 of the Phase 2 can be investigated as next. (authors)
Energy Technology Data Exchange (ETDEWEB)
Li, Shengtai [Los Alamos National Laboratory; Li, Hui [Los Alamos National Laboratory
2012-06-14
We develop a 3D simulation code for interaction between the proto-planetary disk and embedded proto-planets. The protoplanetary disk is treated as a three-dimensional (3D), self-gravitating gas whose motion is described by the locally isothermal Navier-Stokes equations in a spherical coordinate centered on the star. The differential equations for the disk are similar to those given in Kley et al. (2009) with a different gravitational potential that is defined in Nelson et al. (2000). The equations are solved by directional split Godunov method for the inviscid Euler equations plus operator-split method for the viscous source terms. We use a sub-cycling technique for the azimuthal sweep to alleviate the time step restriction. We also extend the FARGO scheme of Masset (2000) and modified in Li et al. (2001) to our 3D code to accelerate the transport in the azimuthal direction. Furthermore, we have implemented a reduced 2D (r, {theta}) and a fully 3D self-gravity solver on our uniform disk grid, which extends our 2D method (Li, Buoni, & Li 2008) to 3D. This solver uses a mode cut-off strategy and combines FFT in the azimuthal direction and direct summation in the radial and meridional direction. An initial axis-symmetric equilibrium disk is generated via iteration between the disk density profile and the 2D disk-self-gravity. We do not need any softening in the disk self-gravity calculation as we have used a shifted grid method (Li et al. 2008) to calculate the potential. The motion of the planet is limited on the mid-plane and the equations are the same as given in D'Angelo et al. (2005), which we adapted to the polar coordinates with a fourth-order Runge-Kutta solver. The disk gravitational force on the planet is assumed to evolve linearly with time between two hydrodynamics time steps. The Planetary potential acting on the disk is calculated accurately with a small softening given by a cubic-spline form (Kley et al. 2009). Since the torque is extremely
Yan, X.; Cai, D.; Nishikawa, K.; Lembege, B.
2004-12-01
We made our efforts to parallelize the global 3D HPF Electromagnetic particle model (EMPM) for several years and have also reported our meaningful simulation results that revealed the essential physics involved in interaction of the solar wind with the Earth's magnetosphere using this EMPM (Nishikawa et al., 1995; Nishikawa, 1997, 1998a, b, 2001, 2002) in our PC cluster and supercomputer(D.S. Cai et al., 2001, 2003). Sash patterns and related phenomena have been observed and reported in some satellite observations (Fujumoto et al. 1997; Maynard, 2001), and have motivated 3D MHD simulations (White and al., 1998). We also investigated it with our global 3D parallelized HPF EMPM with dawnward IMF By (K.-I. Nishikawa, 1998) and recently new simulation with dusk-ward IMF By was accomplished in the new VPP5000 supercomputer. In the new simulations performed on the new VPP5000 supercomputer of Tsukuba University, we used larger domain size, 305×205×205, smaller grid size (Δ ), 0.5R E(the radium of the Earth), more total particle number, 220,000,000 (about 8 pairs per cell). At first, we run this code until we get the so-called quasi-stationary status; After the quasi-stationary status was established, we applied a northward IMF (B z=0.2), and then wait until the IMF arrives around the magnetopuase. After the arrival of IMF, we begin to change the IMF from northward to duskward (IMF B y=-0.2). The results revealed that the groove structure at the day-side magnetopause, that causes particle entry into inner magnetosphere and the cross structure or S-structure at near magneto-tail are formed. Moreover, in contrast with MHD simulations, kinetic characteristic of this event is also analyzed self-consistently with this simulation. The new simulation provides new and more detailed insights for the observed sash event.
International Nuclear Information System (INIS)
Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers, vendors, and research organizations. Computer code user represents a source of uncertainty that may significantly affect the results of system code calculations. Code user training and qualification represent an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes the experience in applying a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In addition, this paper presents the organization and the main features of the 3D S.UN.COP (scaling, uncertainty, and 3D coupled code calculations) seminars during which particular emphasis is given to practical applications in connection with the licensing process of best estimate plus uncertainty methodologies, showing the designer, utility and regulatory approaches. (author)
Energy Technology Data Exchange (ETDEWEB)
Bernal Garcia, A.
2014-07-01
The objective of this work is the development of a modal neutronic code of diffusion in 2D and 3D steady using the finite volume method, from free codes and can be applied to reactors of any geometry. Currently, numerical methods most commonly used in the broadcasting codes provide good results in structured mesh, but its application to non-structured mesh is not easy and may present problems of convergence and stability of the solution. Regarding the non-structured mesh, its use is justified by their easy adaptation to complex geometries and the development of coupled Thermo-hydraulic-neutronic codes, as well as the development of codes fluid dynamic (CFD) that encourage the development of a neutronic code that has the same mesh as the codes of fluid dynamics, which in general tends to be unstructured. On the other hand, refining the mesh and its adaptation to complex geometries is another stimulus of face to learn more about what is happening at the core of the reactor. Finally, the code has been validated with a homogeneous reactor simulation and other heterogeneous for 2D and 3D. (Author)
Energy Technology Data Exchange (ETDEWEB)
Petruzzi, Alessandro; D' Auria, Francesco [University of Pisa, San Piero a Grado (Italy). Nuclear Research Group San Piero a Grado (GRNSPG); Galetti, Regina, E-mail: regina@cnen.gov.b [National Commission for Nuclear Energy (CNEN), Rio de Janeiro, RJ (Brazil); Bajs, Tomislav [University of Zagreb (Croatia). Fac. of Electrical Engineering and Computing. Dept. of Power Systems; Reventos, Francesc [Technical University of Catalonia, Barcelona (Spain). Dept. of Physics and Nuclear Engineering
2011-07-01
Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers, vendors, and research organizations. Computer code user represents a source of uncertainty that may significantly affect the results of system code calculations. Code user training and qualification represent an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes the experience in applying a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In addition, this paper presents the organization and the main features of the 3D S.UN.COP (scaling, uncertainty, and 3D coupled code calculations) seminars during which particular emphasis is given to practical applications in connection with the licensing process of best estimate plus uncertainty methodologies, showing the designer, utility and regulatory approaches. (author)
Gillespie, K. M.; Speirs, D. C.; Ronald, K.; McConville, S. L.; Phelps, A. D. R.; Bingham, R.; Cross, A. W.; Robertson, C. W.; Whyte, C. G.; He, W.; Vorgul, I.; Cairns, R. A.; Kellett, B. J.
2008-12-01
Auroral Kilometric Radiation (AKR), occurs naturally in the polar regions of the Earth's magnetosphere where electrons are accelerated by electric fields into the increasing planetary magnetic dipole. Here conservation of the magnetic moment converts axial to rotational momentum forming a horseshoe distribution in velocity phase space. This distribution is unstable to cyclotron emission with radiation emitted in the X-mode. In a scaled laboratory reproduction of this process, a 75-85 keV electron beam of 5-40 A was magnetically compressed by a system of solenoids and emissions were observed for cyclotron frequencies of 4.42 GHz and 11.7 GHz resonating with near cut-off TE0,1 and TE0,3 modes, respectively. Here we compare these measurements with numerical predictions from the 3D PiC code KARAT. The 3D simulations accurately predicted the radiation modes and frequencies produced by the experiment. The predicted conversion efficiency between electron kinetic and wave field energy of around 1% is close to the experimental measurements and broadly consistent with quasi-linear theoretical analysis and geophysical observations.
Energy Technology Data Exchange (ETDEWEB)
Gillespie, K M; Speirs, D C; Ronald, K; McConville, S L; Phelps, A D R; Bingham, R; Cross, A W; Robertson, C W; Whyte, C G; He, W [SUPA Department of Physics, John Anderson Building, 107 Rottenrow, University of Strathclyde, Glasgow, G4 0NG (United Kingdom); Vorgul, I; Cairns, R A [School of Mathematics and Statistics, University of St Andrews, St Andrews, KY16 9SS (United Kingdom); Kellett, B J [Space Science and Technology Department, STFC Rutherford Appleton Laboratory, Didcot, OX11 0QX (United Kingdom)], E-mail: karen.gillespie@strath.ac.uk
2008-12-15
Auroral Kilometric Radiation (AKR), occurs naturally in the polar regions of the Earth's magnetosphere where electrons are accelerated by electric fields into the increasing planetary magnetic dipole. Here conservation of the magnetic moment converts axial to rotational momentum forming a horseshoe distribution in velocity phase space. This distribution is unstable to cyclotron emission with radiation emitted in the X-mode. In a scaled laboratory reproduction of this process, a 75-85 keV electron beam of 5-40 A was magnetically compressed by a system of solenoids and emissions were observed for cyclotron frequencies of 4.42 GHz and 11.7 GHz resonating with near cut-off TE{sub 0,1} and TE{sub 0,3} modes, respectively. Here we compare these measurements with numerical predictions from the 3D PiC code KARAT. The 3D simulations accurately predicted the radiation modes and frequencies produced by the experiment. The predicted conversion efficiency between electron kinetic and wave field energy of around 1% is close to the experimental measurements and broadly consistent with quasi-linear theoretical analysis and geophysical observations.
International Nuclear Information System (INIS)
Auroral Kilometric Radiation (AKR), occurs naturally in the polar regions of the Earth's magnetosphere where electrons are accelerated by electric fields into the increasing planetary magnetic dipole. Here conservation of the magnetic moment converts axial to rotational momentum forming a horseshoe distribution in velocity phase space. This distribution is unstable to cyclotron emission with radiation emitted in the X-mode. In a scaled laboratory reproduction of this process, a 75-85 keV electron beam of 5-40 A was magnetically compressed by a system of solenoids and emissions were observed for cyclotron frequencies of 4.42 GHz and 11.7 GHz resonating with near cut-off TE0,1 and TE0,3 modes, respectively. Here we compare these measurements with numerical predictions from the 3D PiC code KARAT. The 3D simulations accurately predicted the radiation modes and frequencies produced by the experiment. The predicted conversion efficiency between electron kinetic and wave field energy of around 1% is close to the experimental measurements and broadly consistent with quasi-linear theoretical analysis and geophysical observations.
Evaluation of LOCA in a swimming-pool type reactor using the 3D-AIRLOCA code
International Nuclear Information System (INIS)
The 3D-AIRLOCA code was used to calculate core temperature evolution curves in the wake of a full LOCA in a swimming pool type reactor, resulting in complete core exposure and dryout within about 1000 sec of the initiating event. The results show that fuel integrity loss thresholds (450 C for softening and 650 C for melting) are reached and exceeded over large fractions of the core at power levels as low as 2 MW. At 4.5 MW, the softening threshold is reached even when the accident occurs up to 12 hours after reactor shutdown for continuous operation, and up to 2 hrs after shutdown for intermittent (6 hrs/day, 4 days a week) operation. The situation is even more severe in blockage cases, when the air flow through the core is blocked by residual water at the grid plate level. It is concluded that substantial fission product releases are quite likely in this class of accidents. (orig.)
International Nuclear Information System (INIS)
The four group, 2D and 3D hexagonal geometry HTGR benchmark problems and a 2D hexagonal geometry PWR (WWER) benchmark problem have been solved by using the finite element diffusion code DIFGEN. The hexagons (or hexagonal prisms) were subdivided into first order or second order triangles or quadrilaterals (or triangular or quadrilateral prisms). In the 2D HTGR case of the number of the inserted absorber rods was also varied (7, 6, 0 or 37 rods). The calculational results are in a good agreement with the results of other calculations. The larger systematic series of DIFGEN calculations have given a quantitative picture on the convergence properties of various finite element modellings of hexagonal grids in DIFGEN. (orig.)
Evaluation of LOCA in a swimming-pool type reactor using the 3D-AIRLOCA code
International Nuclear Information System (INIS)
The 3D-AIRLOCA code was used to calculate core temperature evolution curves in the wake of a full LOCA in a swimming pool type reactor, resulting in complete core exposure and dryout within about 1000 sec of the initiating event. The results show that fuel integrity loss thresholds (450 C for softening and 650 C for melting) are reached and exceeded over large fractions of the core at powr levels as low as 2 MW. At 4.5 MW, the softening threshold is reached even when the accident occurs up to 12 hours after reactor shutdown for continuous operation, and up to 2 hrs after shutdown for intermittent (6 hrs/day, 4 days a week) operation. The situation is even more severe in blockage cases, when the air flow through the core is blocked by residual water at the grid plate level. It is concluded that substantial fission product releases are quite likely in this class of accidents. (orig.)
An investigation on the capabilities of the PENELOPE MC code in nanodosimetry
International Nuclear Information System (INIS)
The Monte Carlo (MC) method has been widely implemented in studies of radiation effects on human genetic material. Most of these works have used specific-purpose MC codes to simulate radiation transport in condensed media. PENELOPE is one of the general-purpose MC codes that has been used in many applications related to radiation dosimetry. Based on the fact that PENELOPE can carry out event-by-event coupled electron-photon transport simulations following these particles down to energies of the order of few tens of eV, we have decided to investigate the capacities of this code in the field of nanodosimetry. Single and double strand break probabilities due to the direct impact of γ rays originated from Co60 and Cs137 isotopes and characteristic x-rays, from Al and C K-shells, have been determined by use of PENELOPE. Indirect damage has not been accounted for in this study. A human genetic material geometrical model has been developed, taking into account five organizational levels. In an article by Friedland et al. [Radiat. Environ. Biophys. 38, 39-47 (1999)], a specific-purpose MC code and a very sophisticated DNA geometrical model were used. We have chosen that work as a reference to compare our results. Single and double strand-break probabilities obtained here underestimate those reported by Friedland and co-workers by 20%-76% and 50%-60%, respectively. However, we obtain RBE values for Cs137, AlK and CK radiations in agreement with those reported in previous works [Radiat. Environ. Biophys. 38, 39-47 (1999)] and [Phys. Med. Biol. 53, 233-244 (2008)]. Some enhancements can be incorporated into the PENELOPE code to improve its results in the nanodosimetry field.
An investigation on the capabilities of the PENELOPE MC code in nanodosimetry
Energy Technology Data Exchange (ETDEWEB)
Bernal, M. A.; Liendo, J. A. [Departamento de Fisica, Universidad Simon Bolivar, P.O. Box 89000, Caracas (Venezuela, Bolivarian Republic of)
2009-02-15
The Monte Carlo (MC) method has been widely implemented in studies of radiation effects on human genetic material. Most of these works have used specific-purpose MC codes to simulate radiation transport in condensed media. PENELOPE is one of the general-purpose MC codes that has been used in many applications related to radiation dosimetry. Based on the fact that PENELOPE can carry out event-by-event coupled electron-photon transport simulations following these particles down to energies of the order of few tens of eV, we have decided to investigate the capacities of this code in the field of nanodosimetry. Single and double strand break probabilities due to the direct impact of {gamma} rays originated from Co{sup 60} and Cs{sup 137} isotopes and characteristic x-rays, from Al and C K-shells, have been determined by use of PENELOPE. Indirect damage has not been accounted for in this study. A human genetic material geometrical model has been developed, taking into account five organizational levels. In an article by Friedland et al. [Radiat. Environ. Biophys. 38, 39-47 (1999)], a specific-purpose MC code and a very sophisticated DNA geometrical model were used. We have chosen that work as a reference to compare our results. Single and double strand-break probabilities obtained here underestimate those reported by Friedland and co-workers by 20%-76% and 50%-60%, respectively. However, we obtain RBE values for Cs{sup 137}, Al{sub K} and C{sub K} radiations in agreement with those reported in previous works [Radiat. Environ. Biophys. 38, 39-47 (1999)] and [Phys. Med. Biol. 53, 233-244 (2008)]. Some enhancements can be incorporated into the PENELOPE code to improve its results in the nanodosimetry field.
An investigation on the capabilities of the PENELOPE MC code in nanodosimetry.
Bernal, M A; Liendo, J A
2009-02-01
The Monte Carlo (MC) method has been widely implemented in studies of radiation effects on human genetic material. Most of these works have used specific-purpose MC codes to simulate radiation transport in condensed media. PENELOPE is one of the general-purpose MC codes that has been used in many applications related to radiation dosimetry. Based on the fact that PENELOPE can carry out event-by-event coupled electron-photon transport simulations following these particles down to energies of the order of few tens of eV, we have decided to investigate the capacities of this code in the field of nanodosimetry. Single and double strand break probabilities due to the direct impact of gamma rays originated from Co60 and Cs137 isotopes and characteristic x-rays, from Al and C K-shells, have been determined by use of PENELOPE. Indirect damage has not been accounted for in this study. A human genetic material geometrical model has been developed, taking into account five organizational levels. In an article by Friedland et al. [Radiat. Environ. Biophys. 38, 39-47 (1999)], a specific-purpose MC code and a very sophisticated DNA geometrical model were used. We have chosen that work as a reference to compare our results. Single and double strand-break probabilities obtained here underestimate those reported by Friedland and co-workers by 20%-76% and 50%-60%, respectively. However, we obtain RBE values for Cs137, AlK and CK radiations in agreement with those reported in previous works [Radiat. Environ. Biophys. 38, 39-47 (1999)] and [Phys. Med. Biol. 53, 233-244 (2008)]. Some enhancements can be incorporated into the PENELOPE code to improve its results in the nanodosimetry field. PMID:19292002
International Nuclear Information System (INIS)
Thermal-hydraulic (TH) system codes are developed for the evaluation and improvement of the design and safety of nuclear facilities. Since the numerical modeling of the thermal-hydraulic processes is 1D in nature, these programs have only limited capabilities to predict in detail 3D flows and coolant mixing processes. In contrast, computational fluid dynamics (CFD) software tools are used for 3D flow calculations with high spatial resolution. In order to realistically and efficiently simulate the thermal-hydraulic phenomena in a nuclear power plant (NPP), GRS has developed a methodology for the coupling of the TH system code ATHLET with the 3D CFD software ANSYS CFX. Within the European project NURISP validation activities for the 1D-3D code ATHLET - ANSYS CFX based on a Pressurized Thermal Shock (PTS) related experiment are performed. (author)
International Nuclear Information System (INIS)
DYN3D is a nodal diffusion code for 3D steady-state and transient analysis of Light Water Reactor (LWR) cores with hexagonal or square fuel element geometry. In addition to the neutron kinetics, it comprises of a thermal-hydraulics model for flow in parallel coolant channels. Macroscopic cross section data libraries generated with variation of burn-up, reactor poisons concentrations and thermal-hydraulic feedback parameters are linked to the code. Two-group and multi-groups versions of the code are available. Currently, at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), the DYN3D code is being extended and adopted for the application to block-type High Temperature Gas-Cooled Reactors (HTGRs). In this paper, we give an overview of the latest developments of DYN3D concerning block-type HTGR. The simplified P3 (SP3) transport approximation is implemented into the multi-group DYN3D code to take anisotropy of the neutron flux and heterogeneity of the core more precisely into account. The SP3 method previously implemented into DYN3D for square fuel element geometry of LWR is being extended for hexagonal geometry of the graphite blocks, where the hexagons are subdivided into triangular nodes to be able to perform a systematic mesh refinement. One of the main challenges in cross section generation for the HTGR core calculations is the treatment of the so-called “double heterogeneity”. The modified Reactivity-Equivalent Physical Transformation (RPT) approach is applied in order to eliminate the double-heterogeneity of HTGR fuel elements in the deterministic lattice calculations. The main steps of the RPT method are described. The use of the method for the cross section generation of a simplified HTGR core including its verification is presented. A 3D heat conduction module coupled with a channel-type coolant flow model is implemented to take the temperature reactivity feedback to neutronics physically correctly into account. It is shown that there is significant
Development of a 3D-3V PIC code to study PSI processes in tokamak divertor region
International Nuclear Information System (INIS)
A limited overview of the theoretical understanding as well as PIC simulation of edge plasmas in fusion devices is given. The effect of grazing angle on solid surface (divertor) erosion due to ion sputtering in magnetic fusion devices is studied by a 3D-3V PIC-MCC code. For an oblique magnetic field, there exists a different kind of region in front of the solid surface named as Chodura sheath (CS). Important factors like ion energy and impact angle for physical sputtering are highlighted. Because of the presence of the surface itself, the ion distribution in front of the wall is generally not Maxwellian. In spite of this even for an unmagnetized case, presence of sheath can modify the ion distribution, which has been found in different numerical simulation and laboratory experiments. For magnetized plasmas, the distribution can have several peaks at different energies, which brings further complexity in erosion calculation. The dependence of these two parameters on grazing angle is investigated in detail. The code has been written in java and the plots has been generated in VTK based software Paraview developed by Los Alamos National Laboratory. (author)
Biedron, S G; Yu, L H
2000-01-01
One possible design for a fourth-generation light source is the high-gain harmonic generation (HGHG) free-electron laser (FEL). Here, a coherent seed with a wavelength at a subharmonic of the desired output radiation interacts with the electron beam in an energy-modulating section. This energy modulation is then converted into spatial bunching while traversing a dispersive section (a three-dipole chicane). The final step is passage through an undulator tuned to the desired higher harmonic output wavelength. The coherent seed serves to suppress and can be at a much lower subharmonic of the output radiation. Recently, a 3D code that includes multiple frequencies, multiple undulators (both in quantity and/or type), quadrupole magnets, and dipole magnets was developed to easily simulate HGHG. Here, a brief review of the HGHG theory, the code development, the Accelerator Test Facility's (ATF) HGHG FEL experimental parameters, and the parameter analysis from simulations of this specific experiment will be discussed...
International Nuclear Information System (INIS)
One possible design for a fourth-generation light source is the high-gain harmonic generation (HGHG) free-electron laser (FEL). Here, a coherent seed with a wavelength at a subharmonic of the desired output radiation interacts with the electron beam in an energy-modulating section. This energy modulation is then converted into spatial bunching while traversing a dispersive section (a three-dipole chicane). The final step is passage through a radiative section, an undulator tuned to the desired higher harmonic output wavelength. The coherent seed serves to remove noise and can be at a much lower subharmonic of the output radiation, thus eliminating the concerns found in self-amplified spontaneous emission (SASE) and seeded FELs, respectively. Recently, a 3D code that includes multiple frequencies, multiple undulatory (both in quantity and/or type), quadruple magnets, and dipole magnets was developed to easily simulate HGHG. Here, a brief review of the HGHG theory, the code development, the Accelerator Test Facility's (ATF) HGHG FEL experimental parameters, and the parameter analysis from simulations of this specific experiment will be discussed
International Nuclear Information System (INIS)
One possible design for a fourth-generation light source is the high-gain harmonic generation (HGHG) free-electron laser (FEL). Here, a coherent seed with a wavelength at a subharmonic of the desired output radiation interacts with the electron beam in an energy-modulating section. This energy modulation is then converted into spatial bunching while traversing a dispersive section (a three-dipole chicane). The final step is passage through an undulator tuned to the desired higher harmonic output wavelength. The coherent seed serves to suppress and can be at a much lower subharmonic of the output radiation. Recently, a 3D code that includes multiple frequencies, multiple undulators (both in quantity and/or type), quadrupole magnets, and dipole magnets was developed to easily simulate HGHG. Here, a brief review of the HGHG theory, the code development, the Accelerator Test Facility's (ATF) HGHG FEL experimental parameters, and the parameter analysis from simulations of this specific experiment will be discussed
Numerical modeling of the Linac4 negative ion source extraction region by 3D PIC-MCC code ONIX
Mochalskyy, S; Minea, T; Lifschitz, AF; Schmitzer, C; Midttun, O; Steyaert, D
2013-01-01
At CERN, a high performance negative ion (NI) source is required for the 160 MeV H- linear accelerator Linac4. The source is planned to produce 80 mA of H- with an emittance of 0.25 mm mradN-RMS which is technically and scientifically very challenging. The optimization of the NI source requires a deep understanding of the underling physics concerning the production and extraction of the negative ions. The extraction mechanism from the negative ion source is complex involving a magnetic filter in order to cool down electrons’ temperature. The ONIX (Orsay Negative Ion eXtraction) code is used to address this problem. The ONIX is a selfconsistent 3D electrostatic code using Particles-in-Cell Monte Carlo Collisions (PIC-MCC) approach. It was written to handle the complex boundary conditions between plasma, source walls, and beam formation at the extraction hole. Both, the positive extraction potential (25kV) and the magnetic field map are taken from the experimental set-up, in construction at CERN. This contrib...
International Nuclear Information System (INIS)
Runaway of electrons to high energy during plasma disruptions occurs due to large induced toroidal electric fields which tend to maintain the toroidal plasma current, in accord with Lenz law. This has been observed in many tokamaks. Within the closed flux surfaces, the bounce-averaged CQL3D Fokker-Planck code is well suited to obtain the resulting electron distributions, nonthermal contributions to electrical conductivity, and runaway rates. The time-dependent 2D in momentum-space (pparallel and pperpendicular) distributions axe calculated on a radial array of noncircular flux surfaces, including bounce-averaging of the Fokker-Planck equation to account for toroidal trapping effects. In the steady state, the resulting distributions represent a balance between applied toroidal electric field, relativistic Coulomb collisions, and synchrotron radiation. The code can be run in a mode where the electrons are sourced at low velocity and run off the high velocity edge of the computational mesh, giving runaway rates at steady state. At small minor radius, the results closely match previous results reported by Kulsrud et al. It is found that the runaway rate has a strong dependence on inverse aspect ratio e, decreasing by a factor ∼ 5 as e increases from 0.0 to 0.3. The code can also be run with a radial diffusion and pinching term, simulating radial transport with plasma pinching to maintain a given density profile. Results show a transport reduction of runaways in the plasma center, and an enhancement towards the edge due to the electrons from the plasma center. Avalanching of runaways due to a knock-on electron source is being included
Energy Technology Data Exchange (ETDEWEB)
Nishio, Gunji; Watanabe, Kouji; Murazaki, Minoru [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yamazaki, Noboru; Kouno, Kouji
1998-11-01
The CELVA-3D computer code was developed to evaluate thermofluid phenomena and transport behavior of radioactive materials in a cell during hypothetical explosion in the fuel reprocessing plant. The code calculates temperature, pressure, flow velocity in the cell by three-dimensional thermofluid analysis and calculated an ability to confine the radioactive materials by transport analysis taking into consideration the thermofluid in the cell. And the CELVA-3D is separated into CELVA-3D(M) for a deflagration analysis and CELVA-3D(R) for a detonation analysis; the numerical solution of CELVA-3D(M) for the deflagration was applied to SIMPLE and SIMPLEST for a semi-implicit method, and the solution of CELVA-3D(R) for the detonation by ICE for an explicit method. The mathematical models in CELVA-3D were verified by comparison of code calculations with the results of JAERI`s demonstration tests simulating hypothetical explosion in the reprocessing plant. (author)
Determination of dosimetric parameters for 125I seed source using MCNP5 and EGSnrc MC codes
International Nuclear Information System (INIS)
Background: Seed source has become a popular treatment option in the management of various tumors, particularly in the prostate. Purpose: The aim is to develop accurate and reliable dosimetric parameters that could be used to measure the dose delivered to organs at risk. Methods: Dosimetric parameters (dose rate constant, radial dose function and anisotropy function) of model 6711 125I seed source were calculated with MCNP5 and EGSnrc MC codes following AAPM TG43U1 recommendations. Results: The two results were compared with the relative data recommend by AAPM TG43U1, and the data were as follows: dose rate constant with MCNP5 was in agreement with 0.62%, while that with EGSnrc was 2.07%; radial dose function with MCNP5 was within 0.15%-5.12%, while that with EGSnrc was within 0%-2.48%. Conclusion: The results of two MC codes are in accordance with the recommendations. But that with EGSnrc MC code is better. (authors)
Validation of coupled Relap5-3D code in the analysis of RBMK-1500 specific transients
International Nuclear Information System (INIS)
This paper deals with the modelling of RBMK-1500 specific transients taking place at Ignalina NPP. These transients include: measurements of void and fast power reactivity coefficients, change of graphite cooling conditions and reactor power reduction transients. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is very unique and important from the gap between fuel channel and the graphite bricks model validation point of view. The measurement results, obtained during this transient, allowed to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. Reactor power reduction is a regular operation procedure during the entire lifetime of the reactor. In all cases it starts by either a scram or a power reduction signal activation by the reactor control and protection system or by an operator. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviours of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modelling of the neutronic processes taking place in RBMK- 1500 reactor core. And finally, the performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500
Energy Technology Data Exchange (ETDEWEB)
Kansa, E.J.
1989-01-01
The original scope of this task was to simulate the stresses and displacements of a hard rock tunnel experimental design using a suitable three-dimensional finite element code. NIKE3D was selected as a suitable code for performing these primarily approximate linearly elastic 3D analyses, but it required modifications to include initial stress, shear traction boundary condition and excavation options. During the summer of 1988, such capabilities were installed in a special version of NIKE3D. Subsequently, we verified both the LLNL's commonly used version of NIKE3D and our private modified version against the analytic solution for a spherical cavity in an elastic material deforming under a far-field uniaxial stress. We find the results produced by the unmodified and modified versions of NIKE3D to be in good agreement with the analytic solutions, except near the cavity, where the errors in the stress field are large. As can be expected from a code based on a displacement finite element formulation, the displacements are much more accurate than the stresses calculated from the 8-noded brick elements. To reduce these errors to acceptable levels, the grid must be refined further near the cavity wall. The level of grid refinement required to simulate accurately tunneling problems that do not have spatial symmetry in three dimensions using the current NIKE3D code is likely to exceed the memory capacity of the largest CRAY 1 computers at LLNL. 8 refs., 121 figs.
International Nuclear Information System (INIS)
The original scope of this task was to simulate the stresses and displacements of a hard rock tunnel experimental design using a suitable three-dimensional finite element code. NIKE3D was selected as a suitable code for performing these primarily approximate linearly elastic 3D analyses, but it required modifications to include initial stress, shear traction boundary condition and excavation options. During the summer of 1988, such capabilities were installed in a special version of NIKE3D. Subsequently, we verified both the LLNL's commonly used version of NIKE3D and our private modified version against the analytic solution for a spherical cavity in an elastic material deforming under a far-field uniaxial stress. We find the results produced by the unmodified and modified versions of NIKE3D to be in good agreement with the analytic solutions, except near the cavity, where the errors in the stress field are large. As can be expected from a code based on a displacement finite element formulation, the displacements are much more accurate than the stresses calculated from the 8-noded brick elements. To reduce these errors to acceptable levels, the grid must be refined further near the cavity wall. The level of grid refinement required to simulate accurately tunneling problems that do not have spatial symmetry in three dimensions using the current NIKE3D code is likely to exceed the memory capacity of the largest CRAY 1 computers at LLNL. 8 refs., 121 figs
Koncek, O.; Krivonoska, J.
2014-11-01
The MCNP Monte Carlo code was used to simulate the collimating system of the 60Co therapy unit to calculate the primary and scattered photon fluences as well as the electron contamination incident to the isocentric plane as the functions of the irradiation field size. Furthermore, a Monte Carlo simulation for the polyenergetic Pencil Beam Kernels (PBKs) generation was performed using the calculated photon and electron spectra. The PBK was analytically fitted to speed up the dose calculation using the convolution technique in the homogeneous media. The quality of the PBK fit was verified by comparing the calculated and simulated 60Co broad beam profiles and depth dose curves in a homogeneous water medium. The inhomogeneity correction coefficients were derived from the PBK simulation of an inhomogeneous slab phantom consisting of various materials. The inhomogeneity calculation model is based on the changes in the PBK radial displacement and on the change of the forward and backward electron scattering. The inhomogeneity correction is derived from the electron density values gained from a complete 3D CT array and considers different electron densities through which the pencil beam is propagated as well as the electron density values located between the interaction point and the point of dose deposition. Important aspects and details of the algorithm implementation are also described in this study.
International Nuclear Information System (INIS)
In IBRAE 3D CFD modules (CONV code) for safety analysis of the operated Nuclear Power Plants (NPPs) are developed. These modules are based on the developed algorithms with small scheme diffusion, for which the discrete approximations are constructed with use of finite-volume methods and fully staggered grids. For solving of convection problem the regularized nonlinear monotonic operator-splitting scheme is developed. The Richardson iterative method with iterative Fast Fourier Transformation (FFT) solver for Laplace’s operator as preconditioner is applied for solving pressure equation. Such approach for solving of the elliptical equations with variable coefficients gives multiple acceleration in a comparison with a usual method of conjugate gradients. For modeling of 3D turbulent single-phase flows Quasi DNS approach is used. The CONV code is fully parallelized and highly effective at the high performance computers such as “Chebyshev”, “Lomonosov” (Moscow State University). The developed modules were validated on a series of the well known tests in a wide range of Rayleigh numbers from a range 106-1016 and Reynolds numbers from a range 103-105. The software has been applied to the analysis results of test LIVE-L1 (L1 is aimed at investigating the melt pool and crust behaviour during the stages of air circulation at the outer RPV surface with subsequent flooding of the lower head) and joint analyses on transient molten pool thermal hydraulics in the LIVE facility in the framework of ISTC project. Moreover CONV was validated successfully on a series of the experimental tests as: the blind test on simulation of flows in T-junction (OECD/NEA), ERCOFTAC experiment (world database on turbulent flows) natural convection in the closures under extremely high Rayleigh numbers. In all cases the good coincidence of numerical predictions with experimental data was reached, that specifies a possibility of application of the developed approach for a prediction of CFD
International Nuclear Information System (INIS)
PWR steam generators, tubular heat exchangers and condensers, are basic components of nuclear power plants involving two-phase flows in tube bundles. The operation of these components lead to vibration and corrosion inside the tube bundle, and to deposits and thermal shocks on the tube sheet of steam generators. A deep knowledge of the detailed flow patterns on the shell side is necessary to predict, quantify and prevent these risks. Moreover it is also useful to assess the efficiency of new designs, such as the economizer of the N4 nuclear plant steam generator. For these purposes, Electricite de France has developed, since 1986, a general purpose code named THYC (Thermal HYdraulic Code) designed to study three-dimensional single and two-phase flows in tube bundle (pressurized water reactor cores, steam generators, condensers, heat exchangers). Two types of components in the 3D domain are taken into account: fluid and solids (i.e. porous media approach). The THYC model is obtained by space-time averaging of the instantaneous equations (mass, momentum and energy) of each fluid phase over control volumes including fluid and solids. The THYC-EXCHANGER version solves three to five conservation equations of the fluid outside the tubes, plus the energy equation of the fluid inside the tubes. That makes the code able to model all types of heat exchangers, from single phase heat exchangers, to components involving boiling or condensation. First of all, this paper describes the physical model and the numerical method used in THYC-EXCHANGER. Secondly, validation tests (comparison with experiments) and applications are presented. We present successively : (a) the single-phase heat exchanger mock-up VACARM; (b) the PALUEL steam generator with temperature measurement; (c) the steam generator mock-up CLOTAIRE with void fraction and gas velocity measurements. They emphasize the latest developments, the new capabilities and the adaptability of the code to compute the local
International Nuclear Information System (INIS)
Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF's neutronic code COCCINELLE uses the Rowland's formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission's products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on 'low' configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.)
Development of a high order and multi-dimensional nodal code, ACNEC3D, for reactor core analysis
International Nuclear Information System (INIS)
Highlights: ► Outcomes show the ability of ACNEC to model reactor core with high accuracy. ► An adopted iterative solution strategy is implemented in ACNEC. ► Results present the considerable power error of exterior FAs due to high flux gradient. - Abstract: The aim of this work is to develop a coarse mesh code using various orders of average current nodal expansion method to solve the neutron balance equation implementing the proposed adopted iterative solution algorithm for reactor core calculations. Modern nodal methods have the ability to treat diffusion equation with coarse meshes which cause the considerable reduction of CPU memory and computational costs. A major cause of errors in the calculations is the existence of high flux gradient in some nodes. These errors can be decreased by increasing the order of solution and/or decreasing the mesh sizes. A program based on Average Current Nodal Expansion Code, ACNEC3D, has been developed to solve multi group diffusion equation in three dimensional rectangular geometries using zeroth, first and second orders of average current nodal expansion method. Some popular benchmarks have been investigated and results of various orders of solutions and mesh sizes are compared with reference solutions. Results indicate when the order of solution is increased or/and mesh sizes are decreased, the accuracy of solution is enhanced. According to the results, second order solution has adequate accuracy and higher efficiency in calculations with coarse meshes equal to the dimensions of a FA. Moreover, results show the most of major errors are appeared in peripheral FAs because of flux gradient existence between FAs and reflectors. As a result, more error reductions are taken place in these regions which cause the improvement of accuracy
Davis, A. B.; Cahalan, R. F.
2001-05-01
The Intercomparison of 3D Radiation Codes (I3RC) is an on-going initiative involving an international group of over 30 researchers engaged in the numerical modeling of three-dimensional radiative transfer as applied to clouds. Because of their strong variability and extreme opacity, clouds are indeed a major source of uncertainty in the Earth's local radiation budget (at GCM grid scales). Also 3D effects (at satellite pixel scales) invalidate the standard plane-parallel assumption made in the routine of cloud-property remote sensing at NASA and NOAA. Accordingly, the test-cases used in I3RC are based on inputs and outputs which relate to cloud effects in atmospheric heating rates and in real-world remote sensing geometries. The main objectives of I3RC are to (1) enable participants to improve their models, (2) publish results as a community, (3) archive source code, and (4) educate. We will survey the status of I3RC and its plans for the near future with a special emphasis on the mathematical models and computational approaches. We will also describe some of the prime applications of I3RC's efforts in climate models, cloud-resolving models, and remote-sensing observations of clouds, or that of the surface in their presence. In all these application areas, computational efficiency is the main concern and not accuracy. One of I3RC's main goals is to document the performance of as wide a variety as possible of three-dimensional radiative transfer models for a small but representative number of ``cases.'' However, it is dominated by modelers working at the level of linear transport theory (i.e., they solve the radiative transfer equation) and an overwhelming majority of these participants use slow-but-robust Monte Carlo techniques. This means that only a small portion of the efficiency vs. accuracy vs. flexibility domain is currently populated by I3RC participants. To balance this natural clustering the present authors have organized a systematic outreach towards
Sendrier, Nicolas
2014-01-01
The McEliece cryptosystem is the oldest code-based cryptosystem and itis believed to resist to quantum attacks. The system was proposed in1978 by Robert McEliece. It uses irreducible binary Goppa codes, butit can be instantiated with any family of linear codes for which anefficient decoder is available. The security is provably reduced totwo problems: the pseudorandomnes of the family in which the code ischosen and the hardness of decoding in an arbitrary linear codes.In the past few years, m...
基于稀疏编码的多模态3D 检索系统研究%Multi-model 3D retrieval system research via sparse coding
Institute of Scientific and Technical Information of China (English)
黄冰洋; 张承乾
2016-01-01
3D model retrieval is a research focus at home and abroad .In this paper ,we propose a novel 3D object retrieval system via group sparse coding based on multi-model dataset .First ,we extract SIFT feature from a series of 2D model images which recorded from each 3D model .Then the visual topic distribution generated by LDA (latent dirichlet allocation) is selected to represent each 3D model .Finally ,the sparse coding algorithm is utilized to compute the similarity between different 3D models as to solve the retrieval problem .Experimental results demonstrate the effectiveness of the proposed algorithm .%3D模型检索是当前国内外的一个研究热点。针对多模态数据集，提出了一种基于组稀疏编码方法的3D模型检索系统。首先，从表征三维模型的二维视图中提取 SIFT 特征。在此基础上，利用 LDA （latent dirichlet allocation）模型生成3D模型的视图主题分布，并将此分布作为特征以表征三维模型。最后，采用组稀疏编码算法计算不同3D模型间的相似性，从而解决模型的检索问题。实验结果证明了所提出的检索算法的有效性。
Energy Technology Data Exchange (ETDEWEB)
Whirley, R.G.; Engelmann, B.E.
1993-11-01
This report is the User Manual for the 1993 version of DYNA3D, and also serves as a User Guide. DYNA3D is a nonlinear, explicit, finite element code for analyzing the transient dynamic response of three-dimensional solids and structures. The code is fully vectorized and is available on several computer platforms. DYNA3D includes solid, shell, beam, and truss elements to allow maximum flexibility in modeling physical problems. Many material models are available to represent a wide range of material behavior, including elasticity, plasticity, composites, thermal effects, and rate dependence. In addition, DYNA3D has a sophisticated contact interface capability, including frictional sliding and single surface contact. Rigid materials provide added modeling flexibility. A material model driver with interactive graphics display is incorporated into DYNA3D to permit accurate modeling of complex material response based on experimental data. Along with the DYNA3D Example Problem Manual, this document provides the information necessary to apply DYNA3D to solve a wide range of engineering analysis problems.
Solution of the OECD/NEA neutronic SFR benchmark with Serpent-DYN3D and Serpent-PARCS code systems
International Nuclear Information System (INIS)
Highlights: • A large SFR core from the OECD/WPRS SFR benchmark is considered. • 3D full core deterministic neutronic analysis is performed with DYN3D and PARCS. • Homogenized group constants generated by Serpent Monte Carlo code. • DYN3D and PARCS results are verified against full core Monte Carlo solution. - Abstract: In this study, the Serpent Monte Carlo code was used as a tool for preparation of homogenized group constants for the nodal diffusion analysis of a large U-Pu MOX fueled Sodium-cooled Fast Reactor (SFR) core specified in the OECD/WPRS neutronic SFR benchmark. The group constants generated by Serpent were employed by DYN3D and PARCS nodal diffusion codes in 3D full core calculations. The DYN3D and PARCS results were verified against the references full core Serpent Monte Carlo solution. A good agreement between the reference Monte Carlo and nodal diffusion results was observed demonstrating the feasibility of using Serpent as a group constant generator for the deterministic SFR analysis
Tanguy Janin; Renaud Meignen
2010-01-01
In the course of a postulated severe accident in an NPP, Direct Containment Heating (DCH) may occur after an eventual failure of the vessel. DCH is related to dynamical, thermal, and chemical phenomena involved by the eventual fine fragmentation and dispersal of the corium melt out of the vessel pit. It may threaten the integrity of the containment by pressurization of its atmosphere. Several simplified modellings have been proposed in the past but they require a very strong fitting which ren...
Parallel Grand Canonical Monte Carlo (ParaGrandMC) Simulation Code
Yamakov, Vesselin I.
2016-01-01
This report provides an overview of the Parallel Grand Canonical Monte Carlo (ParaGrandMC) simulation code. This is a highly scalable parallel FORTRAN code for simulating the thermodynamic evolution of metal alloy systems at the atomic level, and predicting the thermodynamic state, phase diagram, chemical composition and mechanical properties. The code is designed to simulate multi-component alloy systems, predict solid-state phase transformations such as austenite-martensite transformations, precipitate formation, recrystallization, capillary effects at interfaces, surface absorption, etc., which can aid the design of novel metallic alloys. While the software is mainly tailored for modeling metal alloys, it can also be used for other types of solid-state systems, and to some degree for liquid or gaseous systems, including multiphase systems forming solid-liquid-gas interfaces.
3D Hydrodynamic Simulations with Yguazú-A Code to Model a Jet in a Galaxy Cluster
Haro-Corzo, S. A. R.; Velazquez, P.; Diaz, A.
2009-05-01
We present preliminary results for a galaxy's jet expanding into an intra-cluster medium (ICM). We attempt to model the jet-gas interaction and the evolution of a extragalactic collimated jet placed at center of computational grid, which it is modeled as a cylinder ejecting gas in the z-axis direction with fixed velocity. It has precession motion around z-axis (period of 10^5 sec.) and orbital motion in XY-plane (period of 500 yr.). This jet is embedded in the ICM, which is modeled as surrounding wind in the XZ plane. We carried out 3D hydrodynamical simulations using Yguazú-A code. This simulation do not include radiative losses. In order to compare the numerical results with observations, we generated synthetic X-ray emission images. X-ray observations with high-resolution of rich cluster of galaxies show diffuse emission with filamentary structure (sometimes called as cooling flow or X-ray filament). Radio observations show a jet-like emission of the central region of the cluster. Joining these observations, in this work we explore the possibility that the jet-ambient gas interaction leads to a filamentary morphology in the X-ray domain. We have found that simulation considering orbital motion offers the possibility to explain the diffuse emission observed in the X-ray domain. The circular orbital motion, additional to precession motion, contribute to disperse the shocked gas and the X-ray appearance of the 3D simulation reproduce some important details of Abel 1795 X-ray emission (Rodriguez-Martinez et al. 2006, A&A, 448, 15): A bright bow-shock at north (spot), where interact directly the jet and the ICM and which is observed in the X-ray image. Meanwhile, in the south side there is no bow-shock X-ray emission, but the wake appears as a X-ray source. This wake is part of the diffuse shocked ambient gas region.
Grundmann, Ulrich; Rohde, Ulrich; Mittag, Siegfried; Kliem, Sören
2010-01-01
DYN3D is an best estimate advanced code for the three-dimensional simulation of steady-states and transients in light water reactor cores with quadratic and hexagonal fuel assemblies. Burnup and poison-dynamic calculations can be performed. For the investigation of wide range transients, DYN3D is coupled with system codes as ATHLET and RELAP5. The neutron kinetic model is based on the solution of the three-dimensional two-group neutron diffusion equation by nodal expansion methods. The therm...
International Nuclear Information System (INIS)
This paper compares the predictions of the revised 3D-AIRLOCA computer code to those data available from the Aerospace Systems Test Reactor's (ASTR's) loss-of-coolant-accident (LOCA) tests run in 1964. The theoretical and experimental hot-spot temperature responses compare remarkably well. In the thirteen cases studied, the irradiation powers varied from 0.4 to 8.87 MW; the irradiation times were 300, 1540, 1800, and 104 s. The degrees of agreement between the data and predictions provide an experimental validation of the 3D-AIRLOCA code
Topol, David A.
1999-01-01
TFaNS is the Tone Fan Noise Design/Prediction System developed by Pratt & Whitney under contract to NASA Lewis (presently NASA Glenn). The purpose of this system is to predict tone noise emanating from a fan stage including the effects of reflection and transmission by the rotor and stator and by the duct inlet and nozzle. These effects have been added to an existing annular duct/isolated stator noise prediction capability. TFaNS consists of: The codes that compute the acoustic properties (reflection and transmission coefficients) of the various elements and write them to files. Cup3D: Fan Noise Coupling Code that reads these files, solves the coupling problem, and outputs the desired noise predictions. AWAKEN: CFD/Measured Wake Postprocessor which reformats CFD wake predictions and/or measured wake data so it can be used by the system. This volume of the report provides technical background for TFaNS including the organization of the system and CUP3D technical documentation. This document also provides information for code developers who must write Acoustic Property Files in the CUP3D format. This report is divided into three volumes: Volume I: System Description, CUP3D Technical Documentation, and Manual for Code Developers; Volume II: User's Manual, TFaNS Vers. 1.4; Volume III: Evaluation of System Codes.
A Distinguisher-Based Attack on a Variant of McEliece's Cryptosystem Based on Reed-Solomon Codes
Gauthier, Valérie; Otmani, Ayoub; Tillich, Jean-Pierre
2012-01-01
Baldi et \\textit{al.} proposed a variant of McEliece's cryptosystem. The main idea is to replace its permutation matrix by adding to it a rank 1 matrix. The motivation for this change is twofold: it would allow the use of codes that were shown to be insecure in the original McEliece's cryptosystem, and it would reduce the key size while keeping the same security against generic decoding attacks. The authors suggest to use generalized Reed-Solomon codes instead of Goppa codes. The public code ...
International Nuclear Information System (INIS)
1 - Description of program or function: PARTISN (Parallel, Time-Dependent SN) is the evolutionary successor to CCC-0547/DANTSYS. User input and cross section formats are very similar to that of DANTSYS. The linear Boltzmann transport equation is solved for neutral particles using the deterministic (SN) method. Both the static (fixed source or eigenvalue) and time-dependent forms of the transport equation are solved in forward or adjoint mode. Vacuum, reflective, periodic, white, or inhomogeneous boundary conditions are solved. General anisotropic scattering and inhomogeneous sources are permitted. PARTISN solves the transport equation on orthogonal (single level or block-structured AMR) grids in 1-D (slab, two-angle slab, cylindrical, or spherical), 2-D (X-Y, R-Z, or R-T) and 3-D (X-Y-Z or R-Z-T) geometries. 2 - Methods:PARTISN numerically solves the multigroup form of the neutral-particle Boltzmann transport equation. The discrete-ordinates form of approximation is used for treating the angular variation of the particle distribution. For curvilinear geometries, diamond differencing is used for angular discretization. The spatial discretizations may be either low-order (diamond difference or Adaptive Weighted Diamond Difference (AWDD)) or higher-order (linear discontinuous or exponential discontinuous). Negative fluxes are eliminated by a local set-to-zero-and-correct algorithm for the diamond case (DD/STZ). Time differencing is Crank-Nicholson (diamond), also with a set-to-zero fix-up scheme. Both inner and outer iterations can be accelerated using the diffusion synthetic acceleration method, or transport synthetic acceleration can be used to accelerate the inner iterations. The diffusion solver uses either the conjugate gradient or multigrid method. Chebyshev acceleration of the fission source is used. The angular source terms may be treated either via standard PN expansions or Galerkin scattering. An option is provided for strictly positive scattering sources
International Nuclear Information System (INIS)
The engineering design, performance analysis and safety assessment of Generation IV heavy liquid metal cooled nuclear reactors calls for advanced and qualified numerical tools. These tools need to be qualified before used in decision making process. Computational Fluid Dynamics (CFD) codes provide detailed means for thermal-hydraulics analysis of pool-type nuclear reactors. This paper describes modeling of a forced to natural flow experiment in TALL-3D experimental facility using a commercial CFD code Star-CCM+. TALL-3D facility is 7 meters high LBE loop with two parallel hot legs and a cold leg. One of the hot legs accommodates the 3D test section, a cylindrical pool where the multi-dimensional flow conditions vary between thermal mixing and stratification depending on the mass flow rate and the power of the heater surrounding the pool. The pool outlet temperature which affects the natural convection flow rates in the system is governed by the flow structure in the pool. Therefore, in order to predict the dynamics of the TALL-3D facility it is crucial to resolve the flow inside the 3D test section. Specifically designed measurement instrumentation set-up provides steady state and transient data for calibration and validation of numerical models. The validity of the CFD model is assessed by comparing the computational results to experimental results. (author)
International Nuclear Information System (INIS)
The experiment involving power output drop of one turbine to the house load level was selected for validation of the DYN3D/ATHLET coupled codes. The initial burnup and static calculations were performed with the DYN3DH1.1/M3 reactor dynamics code. The KASSETA library was used for generation of the reactor core neutronic parameters. The transient calculations were performed with the externally coupled codes ATHLET Mod. 1.1 Cycle C and DYN3DH1.1/M3. The VVER-440/213 ATHLET input deck prepared by the GRS for Loviisa-1 conditions was used
International Nuclear Information System (INIS)
The surveillance program of the vessel materials of a BWR reactor requires the determination of the neutron flux in 3D in the core enveloping. To carry out these calculations of the neutron flux, the Regulatory Guide 1.190 of the NRC recommends the use of the following codes: MCNP, TORT and DORT. In the case of using the DORT code, the one which solves the transport equation in discreet coordinates and in two dimensions (xy, rθ, and rz), the regulatory guide in reference, requires to make an approach of the flow in three dimensions by means of the call Synthesis Method. It is denominated like this due to that a flow representation in 3D is achieved 'combining' or 'synthesizing' the calculated flows by DORT in rθ, rz and r. In this work the application of the Synthesis Method it is presented, according to the Regulatory Guide 1.190, to determine the 3D flows in a BWR reactor. To achieve the above mentioned it was implemented the Synthesis Method in a computer program developed in the ININ to which is denominated SYNTHESIS. This program applies the synthesis method, and is 'coupled' with the DORT code to determine by this way the neutronic fluxes in 3D on the enveloping of a BWR reactor. (Author)
THE STUDY OF AN M-ARY MC-CDMA SYSTEM BASED ON CYCLIC SPREADING CODES & PRE-EQUALIZATION
Institute of Scientific and Technical Information of China (English)
Chu Zhenyong; Ying Xiaofan; Yi Kechu; Tian Hongxin
2005-01-01
A novel fast despreading scheme for M-ary Multi-Carrier Code-Division Multiple Access (MC-CDMA) system is proposed based on cyclic spreading codes and pre-equalizer. In the transmitter, the M spreading codes of each user are generated by circularly shifting the prototype spreading code. A feedback pre-equalizer is employed to process the M-ary MCCDMA signal before transmitted. The received signal is multiplied by the Inverse Discrete Fourier Transform (IDFT) result of the mirror image code of the prototype spreading code, and then demodulated by Orthogonal Frequency-Division Multiplexing (OFDM) demodulator. Compared with the conventional M-ary MC-CDMA receiver, the proposed scheme increases bandwidth efficiency, meanwhile, it achieves M-ary despread spectrum and multi-carrier demodulation, which reduces computation complexity remarkably.
DIF3D: a code to solve one-, two-, and three-dimensional finite-difference diffusion theory problems
International Nuclear Information System (INIS)
The mathematical development and numerical solution of the finite-difference equations are summarized. The report provides a guide for user application and details the programming structure of DIF3D. Guidelines are included for implementing the DIF3D export package on several large scale computers. Optimized iteration methods for the solution of large-scale fast-reactor finite-difference diffusion theory calculations are presented, along with their theoretical basis. The computational and data management considerations that went into their formulation are discussed. The methods utilized include a variant of the Chebyshev acceleration technique applied to the outer fission source iterations and an optimized block successive overrelaxation method for the within-group iterations. A nodal solution option intended for analysis of LMFBR designs in two- and three-dimensional hexagonal geometries is incorporated in the DIF3D package and is documented in a companion report, ANL-83-1
Modeling of tungsten transport in the linear plasma device PSI-2 with the 3D Monte-Carlo code ERO
Marenkov, E.; Eksaeva, A.; Borodin, D.; Kirschner, A.; Laengner, M.; Kurnaev, V.; Kreter, A.; Coenen, J. W.; Rasinski, M.
2015-08-01
The ERO code was modified for modeling of plasma-surface interactions and impurities transport in the PSI-2 installation. Results of experiments on tungsten target irradiation with argon plasma were taken as a benchmark for the new version of the code. Spectroscopy data modeled with the code are in good agreement with experimental ones. Main factors contributing to observed discrepancies are discussed.
International Nuclear Information System (INIS)
TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files
Energy Technology Data Exchange (ETDEWEB)
Cullen, D E
1998-11-22
TART98 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input preparation, running Monte Carlo calculations, and analysis of output results. TART98 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART98 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART98 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART98 and its data files.
A Distinguisher-Based Attack on a Variant of McEliece's Cryptosystem Based on Reed-Solomon Codes
Gauthier, Valérie; Tillich, Jean-Pierre
2012-01-01
Baldi et \\textit{al.} proposed a variant of McEliece's cryptosystem. The main idea is to replace its permutation matrix by adding to it a rank 1 matrix. The motivation for this change is twofold: it would allow the use of codes that were shown to be insecure in the original McEliece's cryptosystem, and it would reduce the key size while keeping the same security against generic decoding attacks. The authors suggest to use generalized Reed-Solomon codes instead of Goppa codes. The public code built with this method is not anymore a generalized Reed-Solomon code. On the other hand, it contains a very large secret generalized Reed-Solomon code. In this paper we present an attack that is built upon a distinguisher which is able to identify elements of this secret code. The distinguisher is constructed by considering the code generated by component-wise products of codewords of the public code (the so-called "square code"). By using square-code dimension considerations, the initial generalized Reed-Solomon code ca...
Energy Technology Data Exchange (ETDEWEB)
Hursin, Mathieu [School of Nuclear Engineering, Purdue University, 400 Central Drive, IN 47907 (United States); Xiao Shanjie [School of Nuclear Engineering, Purdue University, 400 Central Drive, IN 47907 (United States); Jevremovic, Tatjana [School of Nuclear Engineering, Purdue University, 400 Central Drive, IN 47907 (United States)]. E-mail: tatjanaj@purdue.edu
2006-09-15
This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago). The AGENT methodology is based on the unique combination of the three theories: the method of characteristics (MOC) used to simulate the neutron transport in two-dimensional (2D) whole core heterogeneous calculation, the theory of R-functions used as a mathematical tool to describe the true geometry and fuse with the MOC equations, and one-dimensional (1D) higher-order diffusion correction of 2D transport model to account for full 3D heterogeneous whole core representation. The synergism between the radial 2D transport and the 1D axial transport (to take into account the axial neutron interactions and leakage), called the 2D/1D method (used in DeCART and CHAPLET codes), provides a 3D computational solution. The unique synergism between the AGENT geometrical algorithm capable of modeling any current or future reactor core geometry and 3D neutron transport methodology is described in details. The 3D AGENT accuracy and its efficiency are demonstrated showing the eigenvalues, point-wise flux and reaction rate distributions in representative reactor geometries. The AGENT code, comprising this synergism, represents a building block of the computational system, called the virtual reactor. Its main purpose is to perform 'virtual' experiments and demonstrations of various mainly university research reactor experiments.
International Nuclear Information System (INIS)
This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago). The AGENT methodology is based on the unique combination of the three theories: the method of characteristics (MOC) used to simulate the neutron transport in two-dimensional (2D) whole core heterogeneous calculation, the theory of R-functions used as a mathematical tool to describe the true geometry and fuse with the MOC equations, and one-dimensional (1D) higher-order diffusion correction of 2D transport model to account for full 3D heterogeneous whole core representation. The synergism between the radial 2D transport and the 1D axial transport (to take into account the axial neutron interactions and leakage), called the 2D/1D method (used in DeCART and CHAPLET codes), provides a 3D computational solution. The unique synergism between the AGENT geometrical algorithm capable of modeling any current or future reactor core geometry and 3D neutron transport methodology is described in details. The 3D AGENT accuracy and its efficiency are demonstrated showing the eigenvalues, point-wise flux and reaction rate distributions in representative reactor geometries. The AGENT code, comprising this synergism, represents a building block of the computational system, called the virtual reactor. Its main purpose is to perform 'virtual' experiments and demonstrations of various mainly university research reactor experiments
International Nuclear Information System (INIS)
The Main Steam Line Break (MSLB) at End of Life (EOL), Hot Zero Power (HZP) conditions accident was analyzed using a fully time-dependent coupled thermal-hydraulic (T/H) and neutronics method, and compared against conservative Final Safety Analysis Report (FSAR) results, which predict a return-to-power. The development and improvement of coupled neutronics/T/H analysis techniques provide many advantages including the capability to evaluate the impact of modeling assumptions made in previous reactor kinetics and T/H calculations. The coupled STAR kinetics and RETRAN T/H techniques developed here provides a means to evaluate the quasi-static, point kinetics approximation against a fully time-dependent, three-dimensional approach. Using the state-of-the-art 3-D STAR reactor kinetics code with the RETRAN reactor coolant system (RCS) T/H code in a best-estimate approach, it is now possible to evaluate the impact on safety margins imposed by conservative FSAR MSLB assumptions. The method presented shows how the time-dependent 3-D STAR nodal code model was used directly with core inlet conditions determined by RETRAN for a Westinghouse PWR. The STAR/RETRAN results clearly demonstrate that a return-to-power is NOT predicted when a 3-D thermal-hydraulically coupled time-dependent kinetics approach is used. This study shows that: (a) quasi-static and point kinetics methods are not able to describe severe PWR asymmetric transient phenomena adequately; and (b) fully coupled, 3-D time-dependent analysis methods should be used for PWR reactor transients instead. By coupling the RCS response in terms of updated core inlet conditions with 3-D time-dependent core kinetics response, in a tandem manner, the core power and T/H RCS conditions are forced to be self-consistent during the entire event, when non-equilibrium conditions exist. (orig.)
A two-level space-time color-coding method for 3D measurements using structured light
International Nuclear Information System (INIS)
Color-coding methods have significantly improved the measurement efficiency of structured light systems. However, some problems, such as color crosstalk and chromatic aberration, decrease the measurement accuracy of the system. A two-level space-time color-coding method is thus proposed in this paper. The method, which includes a space-code level and a time-code level, is shown to be reliable and efficient. The influence of chromatic aberration is completely mitigated when using this method. Additionally, a self-adaptive windowed Fourier transform is used to eliminate all color crosstalk components. Theoretical analyses and experiments have shown that the proposed coding method solves the problems of color crosstalk and chromatic aberration effectively. Additionally, the method guarantees high measurement accuracy which is very close to the measurement accuracy using monochromatic coded patterns. (paper)
Comparison of ATF and TJ-II stellarator equilibria as computed by the 3-D VMEC and PIES codes
International Nuclear Information System (INIS)
A comparison is made of results from the PIES code, which determines the equilibrium properties of three-dimensional toroidal configurations by direct integration along the magnetic field lines, with those from the VMEC code, which uses an energy minimization in a flux representation to determine the equilibrium configuration, for two devices: the ATF stellarator at Oak Ridge and the TJ-11 heliac which is being built in Madrid. The results obtained from the two codes are in good agreement, providing additional validation for the codes
NEMSQR: A 3-D multi group diffusion theory code based on nodal expansion method for square geometry
International Nuclear Information System (INIS)
Highlights: • A three dimensional diffusion theory code, based on Nodal Expansion Method (NEM), is developed. • The code uses quartic flux expansion with quadratic transverse leakage approximation. • The code is mainly used for calculation of keff, neutron flux and integral kinetics parameters. • The code is tested against several static benchmark problems. • Results corroborate well with the results available in literature. - Abstract: A three dimensional, multigroup, neutron diffusion theory based computer code NEMSQR (Nodal Expansion Method for SQuare geometry) is developed for square geometry in order to perform reactor core calculation. The code is based on Nodal Expansion Method (NEM). In this method, an inhomogeneous matrix equation, which involves spatial moments of nodal flux distribution and surface averaged partial currents across the nodal surfaces, is derived using fourth order polynomial approximation to spatial dependence of nodal flux. Discontinuity factor is incorporated into the code to reduce homogenization error. This code is used for calculation of effective multiplication factor, neutron flux (both direct and adjoint flux), integral kinetics parameters and subcritical count in presence of external neutron source. The numerical studies reported here for several benchmark problems related to light water reactor as well as fast breeder reactor demonstrate the accuracy of the code
Craven, Michael P; Curtis, K. Mervyn
2004-01-01
A complete microcomputer system is described, GesRec3D, which facilitates the data acquisition, segmentation, learning, and recognition of 3-Dimensional arm gestures, with application as a Augmentative and Alternative Communication (AAC) aid for people with motor and speech disability. The gesture data is acquired from a Polhemus electro-magnetic tracker system, with sensors attached to the finger, wrist and elbow of one arm. Coded gestures are linked to user-defined text, to be spoken by a t...
International Nuclear Information System (INIS)
Several years of optimization on the super-scalar architecture has made it more difficult to port the current version of the 3D particle-in-cell code GTC to the CRAY/NEC SX-6 vector architecture. This paper explains the initial work that has been done to port this code to the SX-6 computer and to optimize the most time consuming parts. Early performance results are shown and compared to the same test done on the IBM SP Power 3 and Power 4 machines
基于MC算法的高质量脊柱CT图像三维重建%HIGH-QUALITY 3D RECONSTRUCTION OF SPINE CT IMAGES BASED ON MC ALGORITHM
Institute of Scientific and Technical Information of China (English)
许婉露; 李彬; 田联房
2013-01-01
Reconstructing 3D model of spine from its CT images for providing intuitive preoperative lesion information can effectively assist the high-difficulty spine deformity corrective surgery.As traditional marching cubes (MC) algorithm has the limitations in roughness on reconstruction surface and topological ambiguity,as well as too many fragments in human spine reconstruction,in this paper we propose an improved MC algorithm which is based on edge-preserving local Gaussian filtering and 3D region growing.The algorithm adopts the edge-preserving filtering to eliminate the noises and enhance the edges,and uses the local Gaussian filtering to smooth the pending reconstruction areas for changing original cube types and reducing the number of ambiguous voxels,these effectively solve the problems of roughness on reconstruction surface and topological ambiguity.The dual-threshold segmentation algorithm based on 3D region growing is applied,which can significantly reduce the number of bone fragments reconstruction.Experimental results demonstrate that the 3D spine model reconstructed on this high-quality reconstruction algorithm can serve well the purpose of medical 3D visualisation.%从脊柱CT图像中重建出脊柱的三维模型以提供直观的术前病灶信息,能够有效辅助高难度的脊柱畸形矫正手术.针对传统MC(Marching Cubes)算法存在的重建表面不平滑、结构拓扑歧义的局限以及人体脊柱重构碎片过多的特点,提出一种基于保边局部高斯滤波与三维区域增长的改进型MC算法.该算法采用保边滤波去噪并增强边缘,局部高斯滤波平滑待重建区域以改变原有体素类型,减少二义性体素对数,有效地解决了重建表面不平滑与结构拓扑歧义问题；采用基于三维区域增长的双阈值分割算法,大大减少碎骨重建的数量.实验证明,采用高质量重建算法重建的脊柱三维模型能够满足医学三维可视化的要求.
International Nuclear Information System (INIS)
A supercritical-water-cooled reactor (SCWR) is a high-temperature, high-pressure water cooled reactor that operates above the critical pressure of water. In order to perform efficiently the thermal design of the SCWR, it is important to assess the thermal hydraulics in rod bundles of the core. Experimental conditions of mockup tests, however, may be limited because of technical and financial reasons. Therefore, it is required to establish an analytical design technique that can extrapolate experimental data to various design conditions of the reactor. Japan Atomic Energy Agency (JAEA) has improved the three-dimensional two-fluid model analysis code ACE-3D, which was originally developed for the two-phase flow thermal hydraulics of light water reactors, to handle the thermal hydraulic properties of water in the supercritical region. In the present study, heat transfer experiments of supercritical water flowing in a vertical annular channel around a heater pin, which were performed at JAEA, were analyzed with the improved ACE-3D to assess the prediction performance of the code. As a result, it was implied that the ACE-3D code is applicable to the prediction of wall temperatures of a single rod that simulates the fuel bundle geometry of the SCWR core. (author)
Yoshida, Hiroyuki; Misawa, Takeharu; Takase, Kazuyuki
Two-fluid model can simulate two-phase flow by computational cost less than detailed two-phase flow simulation method such as interface tracking method or particle interaction method. Therefore, two-fluid model is useful for thermal hydraulic analysis in large-scale domain such as a rod bundle. Japan Atomic Energy Agency (JAEA) develops three dimensional two-fluid model analysis code ACE-3D that adopts boundary fitted coordinate system in order to simulate complex shape flow channel. In this paper, boiling two-phase flow analysis in a tight-lattice rod bundle was performed by ACE-3D code. The parallel computation using 126 CPUs was applied to this analysis. In the results, the void fraction, which distributes in outermost region of rod bundle, is lower than that in center region of rod bundle. The tendency of void fraction distribution agreed with the measurement results by neutron radiography qualitatively. To evaluate effects of two-phase flow model used in ACE-3D code, numerical simulation of boiling two-phase in tight-lattice rod bundle with no lift force model was also performed. From the comparison of calculated results, it was concluded that the effects of lift force model were not so large for overall void fraction distribution of tight-lattice rod bundle. However, the lift force model is important for local void fraction distribution of fuel bundles.
International Nuclear Information System (INIS)
Two-fluid model can simulate two-phase flow by computational cost less than detailed two-phase flow simulation method such as interface tracking method or particle interaction method. Therefore, two-fluid model is useful for thermal hydraulic analysis in large-scale domain such as a rod bundle. Japan Atomic Energy Agency (JAEA) develops three dimensional two-fluid model analysis code ACE-3D that adopts boundary fitted coordinate system in order to simulate complex shape flow channel. In this paper, boiling two-phase flow analysis in a tight-lattice rod bundle was performed by ACE-3D code. The parallel computation using 126 CPUs was applied to this analysis. In the results, the void fraction, which distributes in outermost region of rod bundle, is lower than that in center region of rod bundle. The tendency of void fraction distribution agreed with the measurement results by neutron radiography qualitatively. To evaluate effects of two-phase flow model used in ACE-3D code, numerical simulation of boiling two-phase in tight-lattice rod bundle with no lift force model was also performed. From the comparison of calculated results, it was concluded that the effects of lift force model were not so large for overall void fraction distribution of tight-lattice rod bundle. However, the lift force model is important for local void fraction distribution of fuel bundles. (author)
The code IVA 3 for modelling of transient three-phase flows in complicated 3D geometry
International Nuclear Information System (INIS)
IVA3 is a computer code modelling transient, non-equilibrium, multi-phase flows in three-dimensional geometry with arbitrary flow obstacles and volumetric non-flow inclusions. The paper gives a brief description of the code, the verification procedure, and one of the complicated applications, namely, the interaction between molten metal and water in the geometry of a nuclear reactor. (orig.)
A quasi-3D viscous-inviscid interaction code: Q^{3}UIC
DEFF Research Database (Denmark)
Ramos García, Néstor; Sørensen, Jens Nørkær; Shen, Wen Zhong
2014-01-01
inviscid parts. The rotational effects generated by centrifugal and Coriolis forces are introduced in Q3UIC via the streamwise and spanwise integral boundary layer momentum equations. A special inviscid version of the code has been developed to cope with massive separation. To check the ability of the code...
Rohde, Ulrich; Grundmann, Ulrich
2010-01-01
The code DYN3D/M2 is used for investigations of reactivity transients in cores of thermal power reactors with hexagonal fuel elements. The 3-dimensional neutron kinetics model HEXDYN3D of the code is based on a nodal expansion method for solving the two-group neutron diffusion equation. The thermo-hydraulic part FLOCAL consists of a two-phase flow model describing coolant behaviour and a fuel rod model. The fuel elements are simulated by separate coolant channels. Additional, some hot channel...
Use of bar-code technology in MC and A system
International Nuclear Information System (INIS)
Full text: Significant problem during the treatment of nuclear materials is the usage of reliable, rapid, integrate automated systems of nuclear material control and account to reduce the dose loading of personnel. One of the directions to solve the indicated problems is the usage of bar-code technology. Such integrated system should include protection of materials, measuring of materials, and record of materials and drawing up of an inventory list. It is especially important for the enterprises, in which the enriched uranium and other nuclear materials, under IAEA warranties, are utilized. According to US assistance program in the field of MC and A, NSC KIPT has been received indispensable equipment and software, including equipment of nondestructive analysis and automated inventory material accounting system (AIMAS), which was intended for modernizing of nuclear material account system in NSC KIPT. The purpose of operations was estimation of generalized procedures on both MC and A and nondestructive analysis, and updating them so that they might obey the specific conditions of the Enterprise and demands of the Ukraine Regulatory Administration. In NSC KIPT, which is the largest nuclear and physics research center in Ukraine, the measures on enactment of bar-code technology for nuclear materials control and account with the usage of equipment and software of US leading firms (Intermec, Prodigy Max, Tharo Systems, Inc) have been conducting since 1999. During the introduction of this technology, the software on nuclear material control and account (AIMAS data base) has been installed on NSC KIPT computers. The structure of the NSC KIPT's facility has been determined according to demands of the State and IAEA demands. The items of information on the structure of the Facility, and data, which was verified and prepared for input, on nuclear material for each key measuring point of inventory quantity of the material have been set into nuclear material control and
Energy Technology Data Exchange (ETDEWEB)
Kozmenkov, Y.; Grundmann, U.; Kliem, S.; Rohde, U. [Institute of Safety Research, FZR, Dresden (Germany); Bousbia Salahn, A. [Pisa Univ., DIMNP (Italy)
2005-07-01
The modeling of complex transients in Nuclear Power Plants (NPP) remains a challenging topic for Best Estimate (BE) three-dimensional coupled code computational tools. Nowadays, this technique is extensively used since it allows decreasing conservatism in the calculation models by performing more realistic simulations based on a more precise consideration of multidimensional effects under complex transients in NPPs. This paper represents a contribution to the assessment and validation of coupled code technique through the Kozloduy VVER-1000 pump trip test. The coupled RELAP5/3.3-PARCS/2.6 and DYN3D/3-RELAP5/3.3 code systems are used in simulations. The obtained results are assessed against experimental data and also through the code-to-code comparison. The DYN3D/RELAP5 computational model of VVER-1000 has been developed and adjusted for simulations with the parallel running scheme (PVM) of RELAP5/PARCS. Also, the macroscopic cross-section library used in the DYN3D/RELAP5 calculations has been adapted to meet the input requirements of PARCS. Prior to the test simulations, the RELAP5/PARCS model of the plant has been assessed in the stand-alone PARCS and RELAP5 test calculations. A reasonably good agreement between the experimental data and the calculated results is obtained. For the initial state, the observed discrepancies are mainly due to the absence of assembly discontinuity factor (ADF) correction and the evaluation of the Doppler feedback effect. During the transient, the deviations are mainly due to the combined effect of the measurement uncertainty in the control rod axial position and the estimation of the Doppler effect. (authors)
A dynamical systems proof of Kraft-McMillan inequality and its converse for prefix-free codes
Nagaraj, Nithin
2009-03-01
Uniquely decodable codes are central to lossless data compression in both classical and quantum communication systems. The Kraft-McMillan inequality is a basic result in information theory which gives a necessary and sufficient condition for a code to be uniquely decodable and also has a quantum analogue. In this letter, we provide a novel dynamical systems proof of this inequality and its converse for prefix-free codes (no codeword is a prefix of another—the popular Huffman codes are an example). For constrained sources, the problem is still open.
The new deterministic 3-D radiation transport code Multitrans: C5G7 MOX fuel assembly benchmark
International Nuclear Information System (INIS)
The novel deterministic three-dimensional radiation transport code MultiTrans is based on combination of the advanced tree multigrid technique and the simplified P3 (SP3) radiation transport approximation. In the tree multigrid technique, an automatic mesh refinement is performed on material surfaces. The tree multigrid is generated directly from stereo-lithography (STL) files exported by computer-aided design (CAD) systems, thus allowing an easy interface for construction and upgrading of the geometry. The deterministic MultiTrans code allows fast solution of complicated three-dimensional transport problems in detail, offering a new tool for nuclear applications in reactor physics. In order to determine the feasibility of a new code, computational benchmarks need to be carried out. In this work, MultiTrans code is tested for a seven-group three-dimensional MOX fuel assembly transport benchmark without spatial homogenization (NEA C5G7 MOX). (author)
3d and r,z particle simulation of beams for heavy ion fusion: The WARP code
International Nuclear Information System (INIS)
WARP is an electrostatic particle-in-cell (PIC) code that is optimized for studies of space-charge-dominated beams. The authors use the code to understand a number of issues in HIF accelerators and transport systems, including: drift-compression in the presence of misalignments, axial confinement, longitudinal stability, transport around bends, and thermal equilibration processes. In this paper they describe the code architecture and numerical techniques employed to enhance efficiency. They then describe the new simple algorithm for following a beam around a bend, and recent results on bent-beam dynamics and transverse emittance evolution. Finally, they describe the code's most recent feature, a general-lattice capability structured to preserve the efficiency of the particle advance, and present initial results using it
2007-01-01
The results of four gas tracer experiments of atmospheric dispersion on a regional scale are used for the benchmarking of two atmospheric dispersion modeling codes, MINERVE-SPRAY (CEA), and NOSTRADAMUS (IBRAE). The main topic of this comparison is to estimate the Lagrangian code capability to predict the radionuclide atmospheric transfer on a large field, in the case of risk assessment of nuclear power plant for example. For the four experiments, the results of calculations show a rather...
International Nuclear Information System (INIS)
The thermal-hydraulic system code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is developed at Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors. The aim of the code development is to cover the whole spectrum of design basis and beyond design basis accidents (without core degradation) for PWRs and BWRs. In order to extend the simulation capabilities of the 1D system code ATHLET, different approaches are applied at GRS to enable multidimensional thermal-hydraulic representation of relevant primary circuit geometries. One of the current major strategies at the technical safety organization is the coupling of ATHLET with the commercial 3D Computational Fluid Dynamics (CFD) software package ANSYS CFX. This code is a general purpose CFD software program that combines an advanced solver with powerful pre- and post-processing capabilities. It is an efficient tool for simulating the behavior of systems involving fluid flow, heat transfer, and other related physical processes. In the frame of the German CFD Network on Nuclear Reactor Safety, GRS and ANSYS Germany developed a general computer interface for the coupling of both codes. This paper focuses on the methodology and the challenges related to the coupling process. A great number of simulations including test cases with closed loop configurations have been carried out to evaluate and improve the performance of the coupled code system. Selected results of the 1D-3D thermal-hydraulic calculations are presented and analyzed. Preliminary comparative calculations with CFX-ATHLET and ATHLET stand alone showed very good agreement. Nevertheless, an extensive validation of the developed coupled code is planned. Finally, the optimization potential of the coupling methodology is discussed. (author)
Energy Technology Data Exchange (ETDEWEB)
Salah, Anis Bousbia [Facolta di Ingegneria, DIMNP, Universita di Pisa (Italy)]. E-mail: b.salah@ing.unipi.it; Kliem, Soeren [Forschungszentrum Rossendorf (FZR) (Germany); Rohde, Ulrich [Forschungszentrum Rossendorf (FZR) (Germany); D' Auria, Francesco [Facolta di Ingegneria, DIMNP, Universita di Pisa (Italy); Petruzzi, Alessandro [Facolta di Ingegneria, DIMNP, Universita di Pisa (Italy)
2006-06-15
The modeling of complex transients in nuclear power plants (NPP) remains a challenging topic for best estimate three-dimensional coupled code computational tools. This technique is, nowadays, extensively used since it allows decreasing conservatism in the calculation models and performs more realistic simulation and more precise consideration of multidimensional effects under complex transients in NPPs. Therefore, large international activities are in progress aiming to assess the capabilities of coupled codes and the new frontiers for the nuclear technology that could be opened by this technique. In the current paper, a contribution to the assessment and validation of coupled code technique through the Kozloduy VVER100 pump trip test is performed. For this purpose, the coupled RELAP5/3.3-PARCS/2.6 code is used. The code results were assessed against experimental data. Deviations between code predictions and measurements are mainly due to the used models for evaluating and modeling of the Doppler feedback effect. Further investigations through the use of two 'antagonist' uncertainty GRS and the CIAU methods, were considered in order to evaluate and quantify the origin of the observed discrepancies. It was revealed on one hand that relative error quantification discrepancies exist between the two approaches, and further enhancements for both methods are needed.
Analysis of the beam halo in negative ion sources by using 3D3V PIC code
Energy Technology Data Exchange (ETDEWEB)
Miyamoto, K., E-mail: kmiyamot@naruto-u.ac.jp [Naruto University of Education, 748 Nakashima, Takashima, Naruto-cho, Naruto-shi, Tokushima 772-8502 (Japan); Nishioka, S.; Goto, I.; Hatayama, A. [Faculty of Science and Technology, Keio University, 3-14-1 Hiyoshi, Kohoku-ku, Yokohama 223-8522 (Japan); Hanada, M.; Kojima, A.; Hiratsuka, J. [Japan Atomic Energy Agency, 801-1 Mukouyama, Naka 319-0913 (Japan)
2016-02-15
The physical mechanism of the formation of the negative ion beam halo and the heat loads of the multi-stage acceleration grids are investigated with the 3D PIC (particle in cell) simulation. The following physical mechanism of the beam halo formation is verified: The beam core and the halo consist of the negative ions extracted from the center and the periphery of the meniscus, respectively. This difference of negative ion extraction location results in a geometrical aberration. Furthermore, it is shown that the heat loads on the first acceleration grid and the second acceleration grid are quantitatively improved compared with those for the 2D PIC simulation result.
International Nuclear Information System (INIS)
A small scale experimental facility was designed to study the thermal hydraulic phenomena in the Reactor Cavity Cooling System (RCCS). The facility was scaled down from the full scale RCCS system by applying scaling laws. A set of RELAP5-3D simulations were performed to confirm the scaling calculations, and to refine and optimize the facility's configuration, instrumentation selection, and layout. Computational Fluid Dynamics (CFD) calculations using StarCCM+ were performed in order to study the flow patterns and two-phase water behavior in selected locations of the facility where expected complex flow structure occurs. (author)
Analysis of the beam halo in negative ion sources by using 3D3V PIC code
Miyamoto, K.; Nishioka, S.; Goto, I.; Hatayama, A.; Hanada, M.; Kojima, A.; Hiratsuka, J.
2016-02-01
The physical mechanism of the formation of the negative ion beam halo and the heat loads of the multi-stage acceleration grids are investigated with the 3D PIC (particle in cell) simulation. The following physical mechanism of the beam halo formation is verified: The beam core and the halo consist of the negative ions extracted from the center and the periphery of the meniscus, respectively. This difference of negative ion extraction location results in a geometrical aberration. Furthermore, it is shown that the heat loads on the first acceleration grid and the second acceleration grid are quantitatively improved compared with those for the 2D PIC simulation result.
Analysis of the beam halo in negative ion sources by using 3D3V PIC code
International Nuclear Information System (INIS)
The physical mechanism of the formation of the negative ion beam halo and the heat loads of the multi-stage acceleration grids are investigated with the 3D PIC (particle in cell) simulation. The following physical mechanism of the beam halo formation is verified: The beam core and the halo consist of the negative ions extracted from the center and the periphery of the meniscus, respectively. This difference of negative ion extraction location results in a geometrical aberration. Furthermore, it is shown that the heat loads on the first acceleration grid and the second acceleration grid are quantitatively improved compared with those for the 2D PIC simulation result
Ervik, Åsmund; Müller, Bernhard
2014-01-01
To leverage the last two decades' transition in High-Performance Computing (HPC) towards clusters of compute nodes bound together with fast interconnects, a modern scalable CFD code must be able to efficiently distribute work amongst several nodes using the Message Passing Interface (MPI). MPI can enable very large simulations running on very large clusters, but it is necessary that the bulk of the CFD code be written with MPI in mind, an obstacle to parallelizing an existing serial code. In this work we present the results of extending an existing two-phase 3D Navier-Stokes solver, which was completely serial, to a parallel execution model using MPI. The 3D Navier-Stokes equations for two immiscible incompressible fluids are solved by the continuum surface force method, while the location of the interface is determined by the level-set method. We employ the Portable Extensible Toolkit for Scientific Computing (PETSc) for domain decomposition (DD) in a framework where only a fraction of the code needs to be a...
Chitra, S.; Kumaratharan, N.
2015-01-01
Multi-carrier code division multiple access (MC-CDMA) technique is one of the strong candidates for next generation wireless mobile communication systems. Multi-carrier systems are very much sensitive to carrier frequency offset (CFO) results in intercarrier interference (ICI). To mitigate ICI without any spectral loss, a second order duobinary coded phase rotated conjugate cancellation algorithm is proposed in this paper. In the conventional phase rotated conjugate cancellation (PRCC) techni...
International Nuclear Information System (INIS)
One of the most important tasks in today's nuclear power plant safety analysis is a simulation of physical processes at nuclear facilities which accounts for 3-dimensional effects in the core and downcomer of reactor. System coupled thermo-hydraulic/neutron-kinetic code RELAP5-3D, which is a modeling tool provided to University of Kyiv by US DOE in a frame of International Nuclear Safety Program, allows simulation of variable in time spatial distribution of neutron flux in a core and also includes special components for 3D modeling of thermo-hydraulics. A model of Rivne NPP Unit 1 with WWER-440/V-213 type reactor has been developed for RELAP5-3D code. A scenario of 'Main steam line break' design basis accident has been calculated using this model. Such a problem can be characterized by intensive overcooling of a primary coolant in affected loop and, taking into account partial mixing of coolant from different primary loops, a non-uniform cooling of reactor core. Obtained results have been compared with the results obtained by model, which has been used at Design Based Accidents analysis, performed at specified unit.(author)
34/45-Mbps 3D HDTV digital coding scheme using modified motion compensation with disparity vectors
Naito, Sei; Matsumoto, Shuichi
1998-12-01
This paper describes a digital compression coding scheme for transmitting three dimensional stereo HDTV signals with full resolution at bit-rates around 30 to 40 Mbps to be adapted for PDH networks of the CCITT 3rd digital hierarchy, 34 Mbps and 45 Mbps, SDH networks of 52 Mbps and ATM networks. In order to achieve a satisfactory quality for stereo HDTV pictures, three advanced key technologies are introduced into the MPEG-2 Multi-View Profile, i.e., a modified motion compensation using disparity vectors estimated between the left and right pictures, an adaptive rate control using a common buffer memory for left and right pictures encoding, and a discriminatory bit allocation which results in the improvement of left pictures quality without any degradation of right pictures. From the results of coding experiment conducted to evaluate the coding picture achieved by this coding scheme, it is confirmed that our coding scheme gives satisfactory picture quality even at 34 Mbps including audio and FEC data.
Lively, Michael
2010-01-01
Professional Papervision3D describes how Papervision3D works and how real world applications are built, with a clear look at essential topics such as building websites and games, creating virtual tours, and Adobe's Flash 10. Readers learn important techniques through hands-on applications, and build on those skills as the book progresses. The companion website contains all code examples, video step-by-step explanations, and a collada repository.
Energy Technology Data Exchange (ETDEWEB)
Shapiro, A.B.
1983-08-01
The computer code FACET calculates the radiation geometric view factor (alternatively called shape factor, angle factor, or configuration factor) between surfaces for axisymmetric, two-dimensional planar and three-dimensional geometries with interposed third surface obstructions. FACET was developed to calculate view factors for input to finite-element heat-transfer analysis codes. The first section of this report is a brief review of previous radiation-view-factor computer codes. The second section presents the defining integral equation for the geometric view factor between two surfaces and the assumptions made in its derivation. Also in this section are the numerical algorithms used to integrate this equation for the various geometries. The third section presents the algorithms used to detect self-shadowing and third-surface shadowing between the two surfaces for which a view factor is being calculated. The fourth section provides a user's input guide followed by several example problems.
Jouiad, Mustapha
2012-01-01
An unprecedented investigation consisting of the association of X-Ray tomography and Scanning Electron Microscopy combined with Focus Ion Beam (SEM-FIB) is conducted to perform a 3D reconstruction imaging. These techniques are applied to study the non-isothermal creep behavior of close (111) oriented samples of MC2 nickel base superalloys single crystal. The issue here is to develop a strategy to come out with the 3D rafting of γ\\' particles and its interaction whether with dislocation structures or/and with the preexisting voids. This characterization is uncommonly performed away from the conventional studied orientation [001] in order to feed the viscoplastic modeling leading to its improvement by taking into account the crystal anisotropy. The creep tests were performed at two different conditions: classical isothermal tests at 1050°C under 140 MPa and a non isothermal creep test consisting of one overheating at 1200°C and 30 seconds dwell time during the isothermal creep life. The X-Ray tomography shows a great deformation heterogeneity that is pronounced for the non-isothermal tested samples. This deformation localization seems to be linked to the preexisting voids. Nevertheless, for both tested samples, the voids coalescence is the precursor of the observed damage leading to failure. SEM-FIB investigation by means of slice and view technique gives 3D views of the rafted γ\\' particles and shows that γ corridors evolution seems to be the main creep rate controlling parameter. © 2012 Trans Tech Publications, Switzerland.
International Nuclear Information System (INIS)
In the context of the FP7 European THINS Project, complex thermal-hydraulic phenomena relevant for the Generation IV of nuclear reactors are investigated. KTH (Sweden) built the TALL-3D facility to investigate the transition from forced to natural circulation of the Lead-Bismuth Eutectic (LBE) in a pool connected to a 3-leg primary circuit with two heaters and a heat exchanger. The simulation of such 3D phenomena is a challenging task. GRS (Germany) developed the coupling between the Computational Fluid Dynamics (CFD) code ANSYS CFX and the System Analysis code ATHLET. Such coupled codes combine the advantages of CFD, which allow a fine resolution of 3D phenomena, and of System Analysis codes, which are fast running. TUM (Germany) is responsible for the Uncertainty and Sensitivity Analysis of the coupled ATHLET-CFX model in the THINS Project. The influence of modeling uncertainty on simulation results needs to be assessed to characterize and to improve the model and, eventually, to assess its performance against experimental data. TUM has developed a computational framework capable of propagating model input uncertainty through coupled codes. This framework can also be used to apply different approaches for the assessment of the influence of the uncertain input parameters on the model output (Sensitivity Analysis). The work reported in this paper focuses on three methods for the assessment of the sensitivity of the results to the modeling uncertainty. The first method (Morris) allows for the computation of the Elementary Effects resulting from the input parameters. This method is widely used to perform Screening Analysis. The second method (Spearman's rank correlation) relies on regression-based non-parametric measures. This method is suitable if the relation between the input and the output variables is at least monotonic, with the advantage of a low computational cost. The last method (Sobol') computes so-called total effect indices which account for
Directory of Open Access Journals (Sweden)
Vítek Oldřich
2016-06-01
Full Text Available The paper deals with CCV knowledge transfer from reference data (either experiments or 3-D CFD data into system simulation SW tools (based on 0-D/1-D CFD. It was verified that CCV phenomenon can be modeled by means of combustion model perturbations. The proposed methodology consists of two major steps. First, individual cycle data have to be matched with the 0-D/1-D model, i.e., combustion model parameters are varied to achieve the best possible match of in-cylinder pressure traces. Second, the combustion model parameters (obtained in previous step are statistically evaluated to obtain PDFs and cross-correlations. Then such information is imposed to the 0-D/1-D tool to mimic pressure traces CCV. Good correspondence with the reference data is achieved only if both PDFs and cross-correlations are imposed simultaneously.
Dynamic 3D shape of the plantar surface of the foot using coded structured light: a technical report
Thabet, Ali Kassem
2014-01-23
Background The foot provides a crucial contribution to the balance and stability of the musculoskeletal system, and accurate foot measurements are important in applications such as designing custom insoles/footwear. With better understanding of the dynamic behavior of the foot, dynamic foot reconstruction techniques are surfacing as useful ways to properly measure the shape of the foot. This paper presents a novel design and implementation of a structured-light prototype system providing dense three dimensional (3D) measurements of the foot in motion. The input to the system is a video sequence of a foot during a single step; the output is a 3D reconstruction of the plantar surface of the foot for each frame of the input. Methods Engineering and clinical tests were carried out to test the accuracy and repeatability of the system. Accuracy experiments involved imaging a planar surface from different orientations and elevations and measuring the fitting errors of the data to a plane. Repeatability experiments were done using reconstructions from 27 different subjects, where for each one both right and left feet were reconstructed in static and dynamic conditions over two different days. Results The static accuracy of the system was found to be 0.3 mm with planar test objects. In tests with real feet, the system proved repeatable, with reconstruction differences between trials one week apart averaging 2.4 mm (static case) and 2.8 mm (dynamic case). Conclusion The results obtained in the experiments show positive accuracy and repeatability results when compared to current literature. The design also shows to be superior to the systems available in the literature in several factors. Further studies need to be done to quantify the reliability of the system in clinical environments.
International Nuclear Information System (INIS)
The M.E.R.C.U.R.E. calculation code (version 6.3) simulate the photons transport from 15 keV to 10 MeV in three dimensional geometries between volume sources and calculation points. It is based in the integration of attenuation punctual nuclei in straight line with accumulation factors. The accumulation factors take into account the following physical phenomena: photoelectric effect, coherent diffusion, incoherent diffusion, pairs production, radiation secondary sources coming from Bremsstrahlung and fluorescence. The code determines the accumulation factor of a succession of several screens with an innovative iterative method. M.E.R.C.U.R.E. -6.3 integers the punctual nuclei by a Monte Carlo method for which it automatically determines the importance distributions. The results of this code are compared with these ones of the Sn T.W.O.D.A.N.T. code in two one-dimensional configurations. One includes five screens composed of four different materials and the other one three screens. In the configuration with three screens, the second screen is of an infinitesimal thickness. (N.C.)
Benchmark analysis of high temperature engineering test reactor core using McCARD code
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A benchmark calculation has been performed for a startup core physics test of Japan's High Temperature Engineering Test Reactor (HTTR). The calculation is carried out by the McCARD code, which adopts the Monte Carlo method. The cross section library is ENDF-B/VII.0. The fuel cell is modeled by the reactivity-equivalent physical transform (RPT) method. Effective multiplication factors with different numbers of fuel columns have been analyzed. The calculation shows that the HTTR becomes critical with 19 fuel columns with an excess reactivity of 0.84% Δk/k. The discrepancies between the measurements and Monte Carlo calculations are 2.2 and 1.4 % Δk/k for 24 and 30 columns, respectively. The reasons for the discrepancy are thought to be the current version of cross section library and the impurity in the graphite which is represented by the boron concentration. In the future, the depletion results will be proposed for further benchmark calculations. (authors)
Harvey, R. W. (Bob); Petrov, Yu. V.; Jaeger, E. F.; Berry, L. A.; Bonoli, P. T.; Bader, A.
2015-11-01
A time-dependent simulation of C-Mod pulsed ICRF power is made calculating minority hydrogen ion distribution functions with the CQL3D-Hybrid-FOW finite-orbit-width Fokker-Planck code. ICRF fields are calculated with the AORSA full wave code, and RF diffusion coefficients are obtained from these fields using the DC Lorentz gyro-orbit code. Prior results with a zero-banana-width simulation using the CQL3D/AORSA/DC time-cycles showed a pronounced enhancement of the H distribution in the perpendicular velocity direction compared to results obtained from Stix's quasilinear theory, in general agreement with experiment. The present study compares the new FOW results, including relevant gyro-radius effects, to determine the importance of these effects on the the NPA synthetic diagnostic time-dependence. The new NPA results give increased agreement with experiment, particularly in the ramp-down time after the ICRF pulse. Funded, through subcontract with Massachusetts Institute of Technology, by USDOE sponsored SciDAC Center for Simulation of Wave-Plasma Interactions.
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Improved TORT solutions to the 3D transport codes suite of benchmarks exercise are presented in this study. Preliminary TORT solutions to this benchmark indicate that the majority of benchmark quantities for most benchmark cases are computed with good accuracy, and that accuracy improves with model refinement. However, TORT fails to compute accurate results for some benchmark cases with aspect ratios drastically different from 1, possibly due to ray effects. In this work, we employ the standard approach of splitting the solution to the transport equation into an uncollided flux and a fully collided flux via the code sequence GRTUNCL3D and TORT to mitigate ray effects. The results of this code sequence presented in this paper show that the accuracy of most benchmark cases improved substantially. Furthermore, the iterative convergence problems reported for the preliminary TORT solutions have been resolved by bringing the computational cells' aspect ratio closer to unity and, more importantly, by using 64-bit arithmetic precision in the calculation sequence. Results of this study are also reported
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The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.)
Initial Self-Consistent 3D Electron-Cloud Simulations of the LHC Beam with the Code WARP+POSINST
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We present initial results for the self-consistent beam-cloud dynamics simulations for a sample LHC beam, using a newly developed set of modeling capability based on a merge [1] of the three-dimensional parallel Particle-In-Cell (PIC) accelerator code WARP [2] and the electron-cloud code POSINST [3]. Although the storage ring model we use as a test bed to contain the beam is much simpler and shorter than the LHC, its lattice elements are realistically modeled, as is the beam and the electron cloud dynamics. The simulated mechanisms for generation and absorption of the electrons at the walls are based on previously validated models available in POSINST [3, 4
Misselt, K A; Clayton, G C; Wolff, M J
2000-01-01
In this paper and a companion paper we present the DIRTY model, a Monte Carlo radiative transfer code, self-consistently including dust heating and emission, and accounting for the effects of the transient heating of small grains. The code is completely general; the density structure of the dust, the number and type of heating sources, and their geometric configurations can be specified arbitrarily within the model space. Source photons are tracked through the scattering and absorbing medium using Monte Carlo techniques and the effects of multiple scattering are included. The dust scattering, absorbing, and emitting properties are calculated from realistic dust models derived by fitting observed extinction curves in Local Group galaxies including the Magellanic Clouds and the Milky Way. The dust temperature and the emitted dust spectrum are calculated self consistently from the absorbed energy including the effects of temperature fluctuations in small grains. Dust self-absorption is also accounted for, allowi...
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The reasons for large discrepancies between the computed and measured values of the efficiency of control rods observed during start-up experiments on the Russian pressurized water type VVER reactors are discussed. The numerical simulation of the measurements including the prediction of the ex-core detector signals was used to resolve the discrepancies. The time and space dependent neutron flux in the core during these measurements have been calculated by the KIKO3D nodal kinetic code. For calculating the ionization chamber signals the Green function technique has been applied. The Green functions of ionization chambers have been evaluated via solving the neutron transport equation in the reflector regions with the MCNP Monte Carlo code. The detector signals have been calculated and compared with measured ones using the inverse point kinetics transformation. Large number of asymmetric rod drop measurements (with one rod stuck) and some differential rod worth measurements from the Zero Power Physics Tests were provided by the Paks NPP for validation. The experiments cover different fuels (without and with enrichment zoning) and loading patterns. The intermediate range ionization chambers have been used during the scram measurements. The newly developed method provides fairly sufficient match of measured and calculated results. The time behavior of the detector readings observed in the measurements are described by the code in a consistent manner. As a further application the uncertainty of scram rod worth of the KARATE-440 code system was determined by static calculations and subsequent simulation of rod drop with the KIKO3D code. The calculated results were compared to measurements carried out by the Paks NPP. The uncertainty of scram rod worth is established by statistical analysis.
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All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finland. For the first exercise, the plant simulation with point-kinetic neutronics, the thermal-hydraulics code SMABRE was used. The second exercise was calculated with the three-dimensional reactor dynamics code TRAB-3D, and the third exercise with the combination TRAB-3D/SMABRE. VTT has over ten years' experience of coupling neutronic and thermal-hydraulic codes, but this benchmark was the first time these two codes, both developed at VTT, were coupled together. The coupled code system is fast and efficient; the total computation time of the 100-s transient in the third exercise was 16 min on a modern UNIX workstation. The results of all the exercises are similar to those of the other participants. In order to demonstrate the effect of secondary circuit modeling on the results, three different cases were calculated. In case 1 there is no phase separation in the steam lines and no flow reversal in the aspirator. In case 2 the flow reversal in the aspirator is allowed, but there is no phase separation in the steam lines. Finally, in case 3 the drift-flux model is used for the phase separation in the steam lines, but the aspirator flow reversal is not allowed. With these two modeling variations, it is possible to cover a remarkably broad range of results. The maximum power level reached after the reactor trip varies from 534 to 904 MW, the range of the time of the power maximum being close to 30 s. Compared to the total calculated transient time of 100 s, the effect of the secondary side modeling is extremely important
P1 adaptation of TRIPOLI-4 code for the use of 3D realistic core multigroup cross section generation
International Nuclear Information System (INIS)
In this paper, we discuss some improvements we recently implemented in the Monte-Carlo code TRIPOLI-4 associated with the homogenization and collapsing of subassemblies cross sections. The improvement offered us another approach to get critical multigroup cross sections with Monte-Carlo method. The new calculation method in TRIPOLI-4 tries to ensure the neutronic balances, the multiplicative factors and the critical flux spectra for some realistic geometries. We make it by at first improving the treatment of the energy transfer probability, the neutron excess weight and the neutron fission spectrum. This step is necessary for infinite geometries. The second step which will be enlarged in this paper is aimed at better dealing with the multigroup anisotropy distribution law for finite geometries. Usually, Monte-Carlo homogenized multi-group cross sections are validated within a core calculation by a deterministic code. Here, the validation of multigroup constants will also be carried out by Monte-Carlo core calculation code. Different subassemblies are tested with the new collapsing method, especially for the fast neutron reactors subassemblies. (authors)
Directory of Open Access Journals (Sweden)
François Van Dorpe
2007-01-01
Full Text Available The results of four gas tracer experiments of atmospheric dispersion on a regional scale are used for the benchmarking of two atmospheric dispersion modeling codes, MINERVE-SPRAY (CEA, and NOSTRADAMUS (IBRAE. The main topic of this comparison is to estimate the Lagrangian code capability to predict the radionuclide atmospheric transfer on a large field, in the case of risk assessment of nuclear power plant for example. For the four experiments, the results of calculations show a rather good agreement between the two codes, and the order of magnitude of the concentrations measured on the soil is predicted. Simulation is best for sampling points located ten kilometers from the source, while we note a divergence for more distant points results (difference in concentrations by a factor 2 to 5. This divergence may be explained by the fact that, for these four experiments, only one weather station (near the point source was used on a field of 10 000 km2, generating the simulation of a uniform wind field throughout the calculation domain.
International Nuclear Information System (INIS)
The results of four gas tracer experiments of atmospheric dispersion on a regional scale are used for the benchmarking of two atmospheric dispersion modeling codes, MINERVE-SPRAY (CEA), and NOSTRADAMUS (IBRAE). The main topic of this comparison is to estimate the Lagrangian code capability to predict the radionuclide atmospheric transfer on a large field, in the case of risk assessment of nuclear power plant for example. For the four experiments, the results of calculations show a rather good agreement between the two codes, and the order of magnitude of the concentrations measured on the soil is predicted. Simulation is best for sampling points located ten kilometers from the source, while we note a divergence for more distant points results (difference in concentrations by a factor 2 to 5). This divergence may be explained by the fact that, for these four experiments, only one weather station (near the point source) was used on a field of 10000 km2, generating the simulation of a uniform wind field throughout the calculation domain.
Darazi, R.; Gouze, A.; Macq, B.
2009-01-01
Reproducing a natural and real scene as we see in the real world everyday is becoming more and more popular. Stereoscopic and multi-view techniques are used for this end. However due to the fact that more information are displayed requires supporting technologies such as digital compression to ensure the storage and transmission of the sequences. In this paper, a new scheme for stereo image coding is proposed. The original left and right images are jointly coded. The main idea is to optimally exploit the existing correlation between the two images. This is done by the design of an efficient transform that reduces the existing redundancy in the stereo image pair. This approach was inspired by Lifting Scheme (LS). The novelty in our work is that the prediction step is been replaced by an hybrid step that consists in disparity compensation followed by luminance correction and an optimized prediction step. The proposed scheme can be used for lossless and for lossy coding. Experimental results show improvement in terms of performance and complexity compared to recently proposed methods.
MOSRA-LIGHT, High Speed 3-D X-Y-Z Nodal Diffusion Code for Vector Computers
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1 - Description of program or function: MOSRA-Light is a three-dimensional neutron diffusion calculation code for X-Y-Z geometry. 2 - Methods: It is based on the 4. order polynomial nodal expansion method (NEM). As the 4. order NEM is not sensitive to mesh sizes, accurate calculation is possible by the use of coarse meshes of about 20 cm. The drastic decrease of number of unknowns is a 3-dimensional problem results in very fast computation. Furthermore, it employs newly developed computation algorithm 'boundary separated checkerboard sweep method' appropriate to vector computers. This method is very efficient because the speedup factor by vectorization increases, as a scale of problem becomes lager. Speed-up factor compared to the scalar calculation is from 20 to 40 in the case of PWR core calculation. Considering the both effects by the vectorization and the coarse mesh method, total speedup factor is more than 1000 as compared with conventional scalar code with the finite difference method. The general theory of NEM, the fast computation algorithm, benchmark calculation results and detailed information for usage of this code including input data instruction and sample input data is described in the documentation
Cai, Li; Pénéliau, Yannick; Diop, Cheikh M.; Malvagi, Fausto
2014-06-01
In this paper, we discuss some improvements we recently implemented in the Monte-Carlo code TRIPOLI-4® associated with the homogenization and collapsing of subassemblies cross sections. The improvement offered us another approach to get critical multigroup cross sections with Monte-Carlo method. The new calculation method in TRIPOLI-4® tries to ensure the neutronic balances, the multiplicative factors and the critical flux spectra for some realistic geometries. We make it by at first improving the treatment of the energy transfer probability, the neutron excess weight and the neutron fission spectrum. This step is necessary for infinite geometries. The second step which will be enlarged in this paper is aimed at better dealing with the multigroup anisotropy distribution law for finite geometries. Usually, Monte-Carlo homogenized multi-group cross sections are validated within a core calculation by a deterministic code. Here, the validation of multigroup constants will also be carried out by Monte-Carlo core calculation code. Different subassemblies are tested with the new collapsing method, especially for the fast neutron reactors subassemblies.
Use of bar-code technology in MC and A system
International Nuclear Information System (INIS)
Full text: Significant problem during the treatment with nuclear materials is the usage of reliable, rapid, integrant automated systems of nuclear material control and account. Thus the dose loading of attending technical personnel is essentially reduced. One of the directions of the solution of the indicated problems is the usage of bar-code technology. Such integrated system should include protection of materials, measuring of materials, and record of materials and drawing up of an inventory list. Especially it is important for the enterprises, on which the enriched uranium and other nuclear materials, which are under IAEA warranties, are utilized. According to US assistance program in the field of MC and A, NSC KIPT has been received indispensable equipment and software, including equipment of nondestructive analysis and automated inventory material accounting system (AIMAS), which was intended for modernizing of nuclear material account system in NSC KIPT. The purpose of operations was estimation of generalized procedures on both MC and A and nondestructive analysis, and updating them so that they might obey the specific conditions of the Enterprise and demands of the Ukraine Regulatory Administration. In NSC KIPT, which is the largest nuclear and physics research center in Ukraine, the measures on enactment of bar-code technology for nuclear materials control and account with the usage of equipment and software of US leading firms (Intermec, Prodigy Max, Tharo Systems, Inc) have been conducting since 1999. During the introduction of this technology, it has been installed the software on nuclear material control and account (AIMAS data base), which was intended for this activities, on NSC KIPT computers. The structure of the NSC KIPT's facility has been determined according to demands of the State and IAEA demands. The key measuring points of inventory quantity has been determined in nuclear material balance zone and the concrete computers, on which is kept
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In the deterministic safety analysis, codes are required in order to provide evaluations of potential nuclear power plant accidents. In the fields of the core transient behaviour, the computer codes have achieved a high degree of realistic modelling. Nevertheless, some further tools for the investigations of the wide range of physical phenomena in the whole plant transient, such as modeling the ex-core detector signals and the malfunctioning of the emergency control system are unavoidable, too. The main objective of this work is to show the status of the code package based on the best estimate coupled code system Athlet/Kiko-3D and to present a safety analysis. The programs and methods used in KFKI-AEKI for safety analysis of VVER-440 NPP are presented. The accident analysis methodology for a boron dilution scenario, in which an inactive coolant loop is started, is shown. The cooling and the strong dilution increase the reactivity resulting in increasing power level especially in the affected sector. Due to the use of time dependent signals of the ex-core detectors the SCRAM (emergency shutdown) is delayed. Investigating the DNBR (burnout ratio) value by the TRABCO code, no dangerous hot spot was found
International Nuclear Information System (INIS)
MC21 is a continuous-energy Monte Carlo radiation transport code for the calculation of the steady-state spatial distributions of reaction rates in three-dimensional models. The code supports neutron and photon transport in fixed source problems, as well as iterated-fission-source (eigenvalue) neutron transport problems. MC21 has been designed and optimized to support large-scale problems in reactor physics, shielding, and criticality analysis applications. The code also supports many in-line reactor feedback effects, including depletion, thermal feedback, xenon feedback, eigenvalue search, and neutron and photon heating. MC21 uses continuous-energy neutron/nucleus interaction physics over the range from 10-5 eV to 20 MeV. The code treats all common neutron scattering mechanisms, including fast-range elastic and non-elastic scattering, and thermal- and epithermal-range scattering from molecules and crystalline materials. For photon transport, MC21 uses continuous-energy interaction physics over the energy range from 1 keV to 100 GeV. The code treats all common photon interaction mechanisms, including Compton scattering, pair production, and photoelectric interactions. All of the nuclear data required by MC21 is provided by the NDEX system of codes, which extracts and processes data from EPDL-, ENDF-, and ACE-formatted source files. For geometry representation, MC21 employs a flexible constructive solid geometry system that allows users to create spatial cells from first- and second-order surfaces. The system also allows models to be built up as hierarchical collections of previously defined spatial cells, with interior detail provided by grids and template overlays. Results are collected by a generalized tally capability which allows users to edit integral flux and reaction rate information. Results can be collected over the entire problem or within specific regions of interest through the use of phase filters that control which particles are allowed to score each
Application of multi-thread computing and domain decomposition to the 3-D neutronics Fem code Cronos
International Nuclear Information System (INIS)
The purpose of this paper is to present the parallelization of the flux solver and the isotopic depletion module of the code, either using Message Passing Interface (MPI) or OpenMP. Thread parallelism using OpenMP was used to parallelize the mixed dual FEM (finite element method) flux solver MINOS. Investigations regarding the opportunity of mixing parallelism paradigms will be discussed. The isotopic depletion module was parallelized using domain decomposition and MPI. An attempt at using OpenMP was unsuccessful and will be explained. This paper is organized as follows: the first section recalls the different types of parallelism. The mixed dual flux solver and its parallelization are then presented. In the third section, we describe the isotopic depletion solver and its parallelization; and finally conclude with some future perspectives. Parallel applications are mandatory for fine mesh 3-dimensional transport and simplified transport multigroup calculations. The MINOS solver of the FEM neutronics code CRONOS2 was parallelized using the directive based standard OpenMP. An efficiency of 80% (resp. 60%) was achieved with 2 (resp. 4) threads. Parallelization of the isotopic depletion solver was obtained using domain decomposition principles and MPI. Efficiencies greater than 90% were reached. These parallel implementations were tested on a shared memory symmetric multiprocessor (SMP) cluster machine. The OpenMP implementation in the solver MINOS is only the first step towards fully using the SMPs cluster potential with a mixed mode parallelism. Mixed mode parallelism can be achieved by combining message passing interface between clusters with OpenMP implicit parallelism within a cluster
International Nuclear Information System (INIS)
To evaluate thermal hydraulic characteristics of a fuel assembly of supercritical water-cooled fast reactor (Super Fast Reactor), a simplified fuel assembly was analyzed with a three-dimensional two-fluid model analysis code ACE-3D which has been enhanced by Japan Atomic Energy Agency. In the ACE-3D code, the two-phase flow turbulent model based on the k-ε model were adopted. The analytical geometry simulates a 19-rod fuel assembly, which is a simplified geometry of the 271-rod fuel assembly and includes all three kinds of different subchannel types; (1): adjoining to the channel box, (2): next to type (1), and (3): located inside types (1) and (2). In this calculation, one-twelfth model is adopted as the computational domain taking advantage of symmetry. As the boundary conditions, mass velocity, inlet enthalpy and power per rod are to be the same as the steady state condition of the Super Fast Reactor. Cross-sectional local power distribution in the fuel assembly is set to be flat. Rod surface temperatures take peak values near the top of the rods. Maximum clad surface temperature (MCST) is observed at the position facing to the narrowest gap on the center rod near the outlet and the value is 902 K (629 deg. C). It was confirmed that the predicted MCST satisfies a thermal design criteria to ensure fuel and cladding integrity: the MCST should be less than 650 deg. C. (author)
International Nuclear Information System (INIS)
When dealing with nuclear reactor calculation schemes, the need for 3D transport-based reference solutions is essential for validation and optimization purposes. As SN transport method may be considered promising with respect to comprehensive parallel computations, a 3D full PWR core benchmark was proposed to challenge the capabilities of the PENTRAN parallel SN code utilizing an IBM-BG/P computer. After a brief description of the benchmark, a parallel performance analysis is carried out, and shows that the parallelizable (Amdahl) fraction of PENTRAN is comprised between 0.994 ≤ f ≤ 0.996 for a number of BG/P nodes ranging from 17 to 1156. The associated speedup reaches a value greater than 200 with 1156 nodes. Using a best estimate model, PENTRAN results are then compared to Monte Carlo results rendered using the MCNP5 code. Good consistency is observed between the two methods (SN and Monte Carlo), with discrepancies less than 65 pcm for the keff, and less than 2.5% for the flux at the pincell level. (author)
International Nuclear Information System (INIS)
A new specific purpose Monte Carlo code called McENL for modeling the time response of epithermal neutron lifetime tools is described. The code was developed so that the Monte Carlo neophyte can easily use it. A minimum amount of input preparation is required and specified fixed values of the parameters used to control the code operation can be used. The weight windows technique, employing splitting and Russian Roulette, is used with an automated importance function based on the solution of an adjoint diffusion model to improve the code efficiency. Complete composition and density correlated sampling is also included in the code and can be used to study the effect on tool response of small variations in the formation, borehole, or logging tool composition and density. An illustration of the latter application is given here for the density of a thermal neutron filter. McENL was benchmarked against test-pit data for the Mobil pulsed neutron porosity (PNP) tool and found to be very accurate. Results of the experimental validation and details of code performance are presented
International Nuclear Information System (INIS)
Thermal hydraulic system codes are being successfully used in the last decades for the analyses of the behavior of nuclear power plants (NPPs) under off-normal or accidental conditions to evaluate and improve the design, operation and safety of these installations. These programs use simplifications in the mathematical models describing the simulated systems and provide mean values for relevant physical parameters. CFD codes are capable to predict three-dimensional fluid flow behavior in complex geometries and can provide detailed distributions of the physical parameters in space and time. Unfortunately, CFD simulations require very high computation time so that a full CFD representation of the primary circuit of a NPP is currently not feasible. In order to overcome the deficiencies of CFD and system codes, a direct coupling of these simulation tools is pursued. The aim of the current development of the coupled code ATHLET - ANSYS CFX is focused on the extension of the physical models for the application to innovative reactor concepts. Furthermore, first validation activities on the TALL-3D facility, operated with lead bismuth eutectic are already in progress, and described in this paper. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Papukchiev, A.; Lerchl, G. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany)
2013-07-01
Thermal hydraulic system codes are being successfully used in the last decades for the analyses of the behavior of nuclear power plants (NPPs) under off-normal or accidental conditions to evaluate and improve the design, operation and safety of these installations. These programs use simplifications in the mathematical models describing the simulated systems and provide mean values for relevant physical parameters. CFD codes are capable to predict three-dimensional fluid flow behavior in complex geometries and can provide detailed distributions of the physical parameters in space and time. Unfortunately, CFD simulations require very high computation time so that a full CFD representation of the primary circuit of a NPP is currently not feasible. In order to overcome the deficiencies of CFD and system codes, a direct coupling of these simulation tools is pursued. The aim of the current development of the coupled code ATHLET - ANSYS CFX is focused on the extension of the physical models for the application to innovative reactor concepts. Furthermore, first validation activities on the TALL-3D facility, operated with lead bismuth eutectic are already in progress, and described in this paper. (orig.)
Cegla, H. M.; Lovis, C.; Bourrier, V.; Beeck, B.; Watson, C. A.; Pepe, F.
2016-04-01
When a planet transits its host star, it blocks regions of the stellar surface from view; this causes a distortion of the spectral lines and a change in the line-of-sight (LOS) velocities, known as the Rossiter-McLaughlin (RM) effect. Since the LOS velocities depend, in part, on the stellar rotation, the RM waveform is sensitive to the star-planet alignment (which provides information on the system's dynamical history). We present a new RM modelling technique that directly measures the spatially-resolved stellar spectrum behind the planet. This is done by scaling the continuum flux of the (HARPS) spectra by the transit light curve, and then subtracting the in- from the out-of-transit spectra to isolate the starlight behind the planet. This technique does not assume any shape for the intrinsic local profiles. In it, we also allow for differential stellar rotation and centre-to-limb variations in the convective blueshift. We apply this technique to HD 189733 and compare to 3D magnetohydrodynamic (MHD) simulations. We reject rigid body rotation with high confidence (>99% probability), which allows us to determine the occulted stellar latitudes and measure the stellar inclination. In turn, we determine both the sky-projected (λ ≈ -0.4 ± 0.2°) and true 3D obliquity (ψ ≈ 7+12-4°). We also find good agreement with the MHD simulations, with no significant centre-to-limb variations detectable in the local profiles. Hence, this technique provides a new powerful tool that can probe stellar photospheres, differential rotation, determine 3D obliquities, and remove sky-projection biases in planet migration theories. This technique can be implemented with existing instrumentation, but will become even more powerful with the next generation of high-precision radial velocity spectrographs.
2014-01-01
In this paper we estimate the performance of 2*4 MIMO-MC-CDMA system using convolution code in MATLAB which highly reduces BER. MC-CDMA (Multi Carrier Code Division for Multiple Access) is a multiuser and multiple access system which is formed by the combination of OFDM and CDMA and convolution encoding scheme is used in encoder of CDMA as FEC (Forward Error Correction) code to reduce BER (Bit Error Rate). MC-CDMA system is a multicarrier system in which single wideband frequency selective ca...
Directory of Open Access Journals (Sweden)
Antonella Lombardi Costa
2008-01-01
Full Text Available Boiling water reactor (BWR instabilities may occur when, starting from a stable operating condition, changes in system parameters bring the reactor towards an unstable region. In order to design more stable and safer core configurations, experimental and theoretical studies about BWR stability have been performed to characterise the phenomenon and to predict the conditions for its occurrence. In this work, contributions to the study of BWR instability phenomena are presented. The RELAP5/MOD3.3 thermal-hydraulic (TH system code and the PARCS-2.4 3D neutron kinetic (NK code were coupled to simulate BWR transients. Different algorithms were used to calculate the decay ratio (DR and the natural frequency (NF from the power oscillation predicted by the transient calculations as two typical parameters used to provide a quantitative description of instabilities. The validation of the code model set up for the Peach Bottom Unit 2 BWR plant is performed against low-flow stability tests (LFSTs. The four series of LFST have been performed during the first quarter of 1977 at the end of cycle 2 in Pennsylvania. The tests were intended to measure the reactor core stability margins at the limiting conditions used in design and safety analyses.
International Nuclear Information System (INIS)
For safety analyses to support conversion of MNSR reactors from HEU fuel to LEU fuel, a RELAP5-3D model was set up to simulate the entire MNSR system. This model includes the core, the beryllium reflectors, the water in the tank and the water in the surrounding pool. The MCNP code was used to obtain the power distributions in the core and to obtain reactivity feedback coefficients for the transient analyses. The RELAP5-3D model was validated by comparing measured and calculated data for the NIRR-1 reactor in Nigeria. Comparisons include normal operation at constant power and a 3.77 mk rod withdrawal transient. Excellent agreement was obtained for core coolant inlet and outlet temperatures for operation at constant power, and for power level, coolant inlet temperature, and coolant outlet temperature for the rod withdrawal transient. In addition to the negative reactivity feedbacks from increasing core moderator and fuel temperatures, it was necessary to calculate and include positive reactivity feedback from temperature changes in the radial beryllium reflector and changes in the temperature and density of the water in the tank above the core and at the side of the core. The validated RELAP5-3D model was then used to analyze 3.77 mk rod withdrawal transients for LEU cores with two UO2 fuel pin designs. The impact of cracking of oxide LEU fuel is discussed. In addition, steady-state power operation at elevated power levels was evaluated to determine steady-state safety margins for onset of nucleate boiling and for onset of significant voiding. (author)
International Nuclear Information System (INIS)
This work contains description of the physical and mathematical basis on which the IVA3 computer code relies. After describing the state of the art of the 3D modeling for transient multiphase flows, the model assumptions and the modeling technique used in IVA3 are described. Starting with the principles of conservation of mass, momentum, and energy, the non averaged conservation equations are derived for each of the velocity fields which consist of different isothermal components. Thereafter averaging is applied and the working form of the system of 21 partial differential equations is derived. Special attention is paid to the strict consistence of the modeling technique used in IVA3 with the second principle of thermodynamics. The entropy concept used is derived starting with the unaveraged conservation equations and subsequent averaging. The source terms of the entropy production are carefully defined and the final form of the averaged entropy equation is given ready for direct practical applications. The idea of strong analytical thermodynamic coupling between pressure field and changes of the other thermodynamic properties, which is used for the first time in 3D multi fluid modeling, is presented in detail. After obtaining the working form of the conservation equations, the discretization procedure and the reduction to algebraic problems is presented. The mathematical solution method together with some information about the architecture of IVA3 including the local momentum decoupling and accuracy control is presented too. (orig./GL)
Dzhalandinov, A.; Tsofin, V.; Kochkin, V.; Panferov, P.; Timofeev, A.; Reshetnikov, A.; Makhotin, D.; Erak, D.; Voloschenko, A.
2016-02-01
Usually the synthesis of two-dimensional and one-dimensional discrete ordinate calculations is used to evaluate neutron fluence on VVER-1000 reactor pressure vessel (RPV) for prognosis of radiation embrittlement. But there are some cases when this approach is not applicable. For example the latest projects of VVER-1000 have upgraded surveillance program. Containers with surveillance specimens are located on the inner surface of RPV with fast neutron flux maximum. Therefore, the synthesis approach is not suitable enough for calculation of local disturbance of neutron field in RPV inner surface behind the surveillance specimens because of their complicated and heterogeneous structure. In some cases the VVER-1000 core loading consists of fuel assemblies with different fuel height and the applicability of synthesis approach is also ambiguous for these fuel cycles. Also, the synthesis approach is not enough correct for the neutron fluence estimation at the RPV area above core top. Because of these reasons only the 3D neutron transport codes seem to be satisfactory for calculation of neutron fluence on the VVER-1000 RPV. The direct 3D calculations are also recommended by modern regulations.
Lampson, Alan I.; Plummer, David N.; Erkkila, John H.; Crowell, Peter G.; Helms, Charles A.
1998-05-01
This paper describes a series of analyses using the 3-d MINT Navier-Stokes and OCELOT wave optics codes to calculate beam quality in a COIL laser cavity. To make this analysis tractable, the problem was broken into two contributions to the medium quality; that associated with microscale disturbances primarily from the transverse iodine injectors, and that associated with the macroscale including boundary layers and shock-like effects. Results for both microscale and macroscale medium quality are presented for the baseline layer operating point in terms of single pass wavefront error. These results show that the microscale optical path difference effects are 1D in nature and of low spatial order. The COIL medium quality is shown to be dominated by macroscale effects; primarily pressure waves generated from flow/boundary layer interactions on the cavity shrouds.
International Nuclear Information System (INIS)
Full text: We show how the outer magnetic surfaces can be broken up in a spherical tokamak, by breaking the axisymmetry using an inner tilted coil. The configuration chosen for this work is that of the MEDUSA small spherical tokamak, a small glass chamber device, which allows the introduction of such a coil. The simulation is carried out with the 3D-MAPTOR code developed by the authors. Given an initial condition for the magnetic field, it is integrated from the plasma current profile and the external currents, such as the toroidal and the vertical field. Poincare maps along the toroidal angle and the image of the field, as seen from above can be plotted. The latter allows the identification of parameters for which the ripple effect is significant. (author)
International Nuclear Information System (INIS)
Recently several projects have been initiated in Hungary aiming at the introduction of new fuel type, increased maximum allowed power and economic fuel cycle. The planned upgraded power and parallel application of new fuel type require the renewal of the relevant chapter of the Final Safety Analysis Report (FSAR). One of the main tools used for analyzing transient scenarios initiating by reactivity and power distribution anomalies was the ATHLET/KIKO3D coupled neutron kinetic / thermal-hydraulic code. This paper gives an overview of two analyses, which was prepared in the frame of the revision of Paks FSAR, namely the ''withdrawal of one control rod'' and ''initial phase of main steam line break'' events. (author)
High-Fidelity RF Gun Simulations with the Parallel 3D Finite Element Particle-In-Cell Code Pic3P
Energy Technology Data Exchange (ETDEWEB)
Candel, A; Kabel, A.; Lee, L.; Li, Z.; Limborg, C.; Ng, C.; Schussman, G.; Ko, K.; /SLAC
2009-06-19
SLAC's Advanced Computations Department (ACD) has developed the first parallel Finite Element 3D Particle-In-Cell (PIC) code, Pic3P, for simulations of RF guns and other space-charge dominated beam-cavity interactions. Pic3P solves the complete set of Maxwell-Lorentz equations and thus includes space charge, retardation and wakefield effects from first principles. Pic3P uses higher-order Finite Elementmethods on unstructured conformal meshes. A novel scheme for causal adaptive refinement and dynamic load balancing enable unprecedented simulation accuracy, aiding the design and operation of the next generation of accelerator facilities. Application to the Linac Coherent Light Source (LCLS) RF gun is presented.
Chitra, S; Kumaratharan, N
2015-01-01
Multi-carrier code division multiple access (MC-CDMA) technique is one of the strong candidates for next generation wireless mobile communication systems. Multi-carrier systems are very much sensitive to carrier frequency offset (CFO) results in intercarrier interference (ICI). To mitigate ICI without any spectral loss, a second order duobinary coded phase rotated conjugate cancellation algorithm is proposed in this paper. In the conventional phase rotated conjugate cancellation (PRCC) technique, one path carries the MC-CDMA signal with a phase spin of ϕ and the other path carries the conjugate of the first path signal with -ϕ phase spin. This artificial phase rotation allows the transmitter to tune the transmitted signals so that the ICI effects could be mutually cancelled at the receiver. Although the PRCC technique reduces the spectral efficiency, the limitation can be overcome by the joint second order duobinary coding scheme with PRCC technique. In the proposed method, the correlative coding between the binary symbols modulated on adjacent subcarriers is used to reduce the ICI without any spectral loss. Simulation results show that the proposed PRCC method provides better carrier to interference ratio (CIR) and bit error rate (BER) performances compared to the conventional conjugate cancellation (CC) technique. PMID:25790029
Picot-Colbeaux, Géraldine; Devau, Nicolas; Thiéry, Dominique; Pettenati, Marie; Surdyk, Nicolas; Parmentier, Marc; Amraoui, Nadia; Crastes de Paulet, François; André, Laurent
2016-04-01
Chalk aquifer is the main water resource for domestic water supply in many parts in northern France. In same basin, groundwater is frequently affected by quality problems concerning nitrates. Often close to or above the drinking water standards, nitrate concentration in groundwater is mainly due to historical agriculture practices, combined with leakage and aquifer recharge through the vadose zone. The complexity of processes occurring into such an environment leads to take into account a lot of knowledge on agronomy, geochemistry and hydrogeology in order to understand, model and predict the spatiotemporal evolution of nitrate content and provide a decision support tool for the water producers and stakeholders. To succeed in this challenge, conceptual and numerical models representing accurately the Chalk aquifer specificity need to be developed. A multidisciplinary approach is developed to simulate storage and transport from the ground surface until groundwater. This involves a new agronomic module "NITRATE" (NItrogen TRansfer for Arable soil to groundwaTEr), a soil-crop model allowing to calculate nitrogen mass balance in arable soil, and the "PHREEQC" numerical code for geochemical calculations, both coupled with the 3D transient groundwater numerical code "MARTHE". Otherwise, new development achieved on MARTHE code allows the use of dual porosity and permeability calculations needed in the fissured Chalk aquifer context. This method concerning the integration of existing multi-disciplinary tools is a real challenge to reduce the number of parameters by selecting the relevant equations and simplifying the equations without altering the signal. The robustness and the validity of these numerical developments are tested step by step with several simulations constrained by climate forcing, land use and nitrogen inputs over several decades. In the first time, simulations are performed in a 1D vertical unsaturated soil column for representing experimental nitrates
Energy Technology Data Exchange (ETDEWEB)
Pescarini, M.; Orsi, R.; Martinelli, T. [ENEA, Ente per le Nuove Tecnologie, l' Energia e l' Ambiente, Centro Ricerche Ezio Clementel Bologna (Italy)
2003-07-01
In many practical radiation transport applications today the cost for solving refined, large size and complex multi-dimensional problems is not so much computing but is linked to the cumbersome effort required by an expert to prepare a detailed geometrical model, verify and validate that it is correct and represents, to a specified tolerance, the real design or facility. This situation is, in particular, relevant and frequent in reactor core criticality and shielding calculations, with three-dimensional (3D) general purpose radiation transport codes, requiring a very large number of meshes and high performance computers. The need for developing tools that make easier the task to the physicist or engineer, by reducing the time required, by facilitating through effective graphical display the verification of correctness and, finally, that help the interpretation of the results obtained, has clearly emerged. The paper shows the results of efforts in this field through detailed simulations of a complex shielding benchmark experiment. In the context of the activities proposed by the OECD/NEA Nuclear Science Committee (NSC) Task Force on Computing Radiation Dose and Modelling of Radiation-Induced Degradation of Reactor Components (TFRDD), the ENEA-Bologna Nuclear Data Centre contributed with an analysis of the VENUS-3 low-flux neutron shielding benchmark experiment (SCK/CEN-Mol, Belgium). One of the targets of the work was to test the BOT3P system, originally developed at the Nuclear Data Centre in ENEA-Bologna and actually released to OECD/NEA Data Bank for free distribution. BOT3P, ancillary system of the DORT (2D) and TORT (3D) SN codes, permits a flexible automatic generation of spatial mesh grids in Cartesian or cylindrical geometry, through combinatorial geometry algorithms, following a simplified user-friendly approach. This system demonstrated its validity also in core criticality analyses, as for example the Lewis MOX fuel benchmark, permitting to easily
International Nuclear Information System (INIS)
In many practical radiation transport applications today the cost for solving refined, large size and complex multi-dimensional problems is not so much computing but is linked to the cumbersome effort required by an expert to prepare a detailed geometrical model, verify and validate that it is correct and represents, to a specified tolerance, the real design or facility. This situation is, in particular, relevant and frequent in reactor core criticality and shielding calculations, with three-dimensional (3D) general purpose radiation transport codes, requiring a very large number of meshes and high performance computers. The need for developing tools that make easier the task to the physicist or engineer, by reducing the time required, by facilitating through effective graphical display the verification of correctness and, finally, that help the interpretation of the results obtained, has clearly emerged. The paper shows the results of efforts in this field through detailed simulations of a complex shielding benchmark experiment. In the context of the activities proposed by the OECD/NEA Nuclear Science Committee (NSC) Task Force on Computing Radiation Dose and Modelling of Radiation-Induced Degradation of Reactor Components (TFRDD), the ENEA-Bologna Nuclear Data Centre contributed with an analysis of the VENUS-3 low-flux neutron shielding benchmark experiment (SCK/CEN-Mol, Belgium). One of the targets of the work was to test the BOT3P system, originally developed at the Nuclear Data Centre in ENEA-Bologna and actually released to OECD/NEA Data Bank for free distribution. BOT3P, ancillary system of the DORT (2D) and TORT (3D) SN codes, permits a flexible automatic generation of spatial mesh grids in Cartesian or cylindrical geometry, through combinatorial geometry algorithms, following a simplified user-friendly approach. This system demonstrated its validity also in core criticality analyses, as for example the Lewis MOX fuel benchmark, permitting to easily
Cegla, H M; Bourrier, V; Beeck, B; Watson, C A; Pepe, F
2016-01-01
When a planet transits its host star, it blocks regions of the stellar surface from view; this causes a distortion of the spectral lines and a change in the line-of-sight (LOS) velocities, known as the Rossiter-McLaughlin (RM) effect. Since the LOS velocities depend, in part, on the stellar rotation, the RM waveform is sensitive to the star-planet alignment (which provides information on the system's dynamical history). We present a new RM modelling technique that directly measures the spatially-resolved stellar spectrum behind the planet. This is done by scaling the continuum flux of the (HARPS) spectra by the transit light curve, and then subtracting the in- from the out-of-transit spectra to isolate the starlight behind the planet. This technique does not assume any shape for the intrinsic local profiles. In it, we also allow for differential stellar rotation and centre-to-limb variations in the convective blueshift. We apply this technique to HD189733 and compare to 3D magnetohydrodynamic (MHD) simulation...
Energy Technology Data Exchange (ETDEWEB)
Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.
1998-03-01
The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [Deutsch] Das Reaktorkernmodell DYN3D mit 3D Neutronenkinetik wurde an den Thermohydraulik-Systemcode ATHLET angekoppelt. Im vorliegenden Bericht werden Arbeiten zur Qualifizierung des gekoppelten Codekomplexes zu einem validierten Hilfsmittel fuer Stoerfallablaufanalysen zu Reaktoren des russischen Typs WWER dargestellt. Diese umfassten im einzelnen: - Beitraege zur Validierung der Einzelcodes ATHLET und DYN3D anhand der Nachrechnung von Experimenten zum
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Libraries of two-group neutron-diffusion parameters for a Siemens-KWU-Konvoi Pressurized Water Reactor have been generated at Forschungszentrum Rossendorf and TUeV Bau und Betrieb GmbH by using the codes HELIOS and CASMO, respectively. The libraries have been coupled to the reactor-dynamics code DYN3D. For a generic PWR core containing MOX fuel elements, DYN3D macro-burnup calculations and the calculation of different operation states have been carried out. The results will be used for the investigation of possible accident scenarios. Reactivity coefficients calculated by DYN3D are needed for accident analyses by the 1-D thermal-hydraulic code ATHLET. Using the cross section data, more detailed analyses can be carried out by applying the coupled-code system DYN3D-ATHLET, considering 3D neutron kinetics. The comparison of the results calculated by DYN3D with two different diffusion-parameter libraries can give an idea of how uncertainties in diffusion data influence the accuracy of reactor simulation. (orig.)
MT3D was first developed by Chunmiao Zheng in 1990 at S.S. Papadopulos & Associates, Inc. with partial support from the U.S. Environmental Protection Agency (USEPA). Starting in 1990, MT3D was released as a pubic domain code from the USEPA. Commercial versions with enhanced capab...
International Nuclear Information System (INIS)
1 - Description of program or function: BOT3P was originally conceived as a set of standard FORTRAN 77 language programs in order to give the users of the DORT and TORT deterministic transport codes some useful diagnostic tools to prepare and check their input data files. Later versions extended the possibility to produce the geometrical, material distribution and fixed neutron source data to other deterministic transport codes such as TWODANT/THREEDANT of the DANTSYS system, PARTISN and, potentially, to any transport code through BOT3P binary output files that can be easily interfaced (see, for example, the Russian two-dimensional (2D) and three-dimensional (3D) discrete ordinates neutron, photon and charged particle transport codes KASKAD-S-2.5 and KATRIN-2.0). As from Version 5.1 BOT3P contained important additions specifically addressed to radiation transport analysis for medical applications. BOT3P-5.2 contains new graphics capabilities. Some of them enable users to select space sub-domains of the total mesh grid in order to improve the zoom simulation of the geometry, both in 2D cuts and in 3D. Moreover the new BOT3P module (PDTM) may improve the interface of BOT3P geometrical models to transport analysis codes. The following programs are included in the BOT3P software package: GGDM, DDM, GGTM, DTM2, DTM3, RVARSCL, COMPARE, MKSRC, CATSM, DTET, and PDTM. The main features of these different programs are described. 2 - Methods: GGDM and GGTM work similarly from the logical point of view. Since the 3D case is more general, the following description refers to GGTM. All the co-ordinate values that characterise the geometrical scheme at the basis of the 3D transport code geometrical and material model are read, sorted and all stored if different from the neighbouring ones more than an input tolerance established by the user. These co-ordinates are always present in the fine-mesh boundary arrays independently of the mesh grid refinement options, because they
International Nuclear Information System (INIS)
We present the TORT solutions to the 3D transport codes' suite of benchmarks exercise. An overview of benchmark configurations is provided, followed by a description of the TORT computational model we developed to solve the cases comprising the benchmark suite. In the numerical experiments reported in this paper, we chose to refine the spatial and angular discretizations simultaneously, from the coarsest model (40 x 40 x 40, 200 angles) to the finest model (160 x 160 x 160, 800 angles). The MCNP reference solution is used for evaluating the effect of model-refinement on the accuracy of the TORT solutions. The presented results show that the majority of benchmark quantities are computed with good accuracy by TORT, and that the accuracy improves with model refinement. However, this deliberately severe test has exposed some deficiencies in both deterministic and stochastic solution approaches. Specifically, TORT fails to converge the inner iterations in some benchmark configurations while MCNP produces zero tallies, or drastically poor statistics for some benchmark quantities. We conjecture that TORT's failure to converge is driven by ray effects in configurations with low scattering ratio and/or highly skewed computational cells, i.e. aspect ratio far from unity. The failure of MCNP occurs in quantities tallied over a very small area or volume in physical space, or quantities tallied many (∼25) mean free paths away from the source. Hence automated, robust, and reliable variance reduction techniques are essential for obtaining high quality reference values of the benchmark quantities. Preliminary results of the benchmark exercise indicate that the occasionally poor performance of TORT is shared with other deterministic codes. Armed with this information, method developers can now direct their attention to regions in parameter space where such failures occur and design alternative solution approaches for such instances
DEFF Research Database (Denmark)
Lavstsen, Thomas; Salanti, Ali; Jensen, Anja T R;
2003-01-01
-domain type dominant of groups B and C. Two sequences belonging to the var1 and var2 subfamilies formed independent groups. A rif subgroup transcribed towards the centromere was found neighbouring var genes of group A such that the rif and var 5' regions merged. This organization appeared to be unique...... organization of the 3D7 PfEMP1 repertoire was investigated on the basis of the complete genome sequence. METHODS: Using two tree-building methods we analysed the coding and non-coding sequences of 3D7 var and rif genes as well as var genes of other parasite strains. RESULTS: var genes can be sub-grouped into...... three major groups (group A, B and C) and two intermediate groups B/A and B/C representing transitions between the three major groups. The best defined var group, group A, comprises telomeric genes transcribed towards the telomere encoding PfEMP1s with complex domain structures different from the 4...
3-D neutron transport benchmarks
International Nuclear Information System (INIS)
A set of 3-D neutron transport benchmark problems proposed by the Osaka University to NEACRP in 1988 has been calculated by many participants and the corresponding results are summarized in this report. The results of Keff, control rod worth and region-averaged fluxes for the four proposed core models, calculated by using various 3-D transport codes are compared and discussed. The calculational methods used were: Monte Carlo, Discrete Ordinates (Sn), Spherical Harmonics (Pn), Nodal Transport and others. The solutions of the four core models are quite useful as benchmarks for checking the validity of 3-D neutron transport codes
Audigane, Pascal; Chiaberge, Christophe; Mathurin, Frédéric; Lions, Julie; Picot-Colbeaux, Géraldine
2011-04-01
This paper is addressed to the TOUGH2 user community. It presents a new tool for handling simulations run with the TOUGH2 code with specific application to CO 2 geological storage. This tool is composed of separate FORTRAN subroutines (or modules) that can be run independently, using input and output files in ASCII format for TOUGH2. These modules have been developed specifically for modeling of carbon dioxide geological storage and their use with TOUGH2 and the Equation of State module ECO2N, dedicated to CO 2-water-salt mixture systems, with TOUGHREACT, which is an adaptation of TOUGH2 with ECO2N and geochemical fluid-rock interactions, and with TOUGH2 and the EOS7C module dedicated to CO 2-CH 4 gas mixture is described. The objective is to save time for the pre-processing, execution and visualization of complex geometry for geological system representation. The workflow is rapid and user-friendly and future implementation to other TOUGH2 EOS modules for other contexts (e.g. nuclear waste disposal, geothermal production) is straightforward. Three examples are shown for validation: (i) leakage of CO 2 up through an abandoned well; (ii) 3D reactive transport modeling of CO 2 in a sandy aquifer formation in the Sleipner gas Field, (North Sea, Norway); and (iii) an estimation of enhanced gas recovery technology using CO 2 as the injected and stored gas to produce methane in the K12B Gas Field (North Sea, Denmark).
International Nuclear Information System (INIS)
Cavitation is one of the most demanding physical phenomena influencing the performance of hydraulic machines. It is therefore important to predict correctly its inception and development, in order to quantify the performance drop it induces, and also to characterize the resulting flow instabilities. The aim of this work is to develop an unsteady 3D algorithm for the numerical simulation of cavitation in an industrial CFD solver 'Code Saturne'. It is based on a fractional step method and preserves the minimum/maximum principle of the void fraction. An implicit solver, based on a transport equation of the void fraction coupled with the Navier-Stokes equations is proposed. A specific numerical treatment of the cavitation source terms provides physical values of the void fraction (between 0 and 1) without including any artificial numerical limitation. The influence of RANS turbulence models on the simulation of cavitation on 2D geometries (Venturi and Hydrofoil) is then studied. It confirms the capability of the two-equation eddy viscosity models, k-epsilon and k-omega-SST, with the modification proposed by Reboud et al. (1998) to reproduce the main features of the unsteady sheet cavity behavior. The second order model RSM-SSG, based on the Reynolds stress transport, appears able to reproduce the highly unsteady flow behavior without including any arbitrary modification. The three-dimensional effects involved in the instability mechanisms are also analyzed. This work allows us to achieve a numerical tool, validated on complex configurations of cavitating flows, to improve the understanding of the physical mechanisms that control the three-dimensional unsteady effects involved in the mechanisms of instability. (author)
Energy Technology Data Exchange (ETDEWEB)
Tuan, Hoang Sy Minh; Sun, Gwang Min [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2011-05-15
The HANARO (KAERI) research reactor has been developed a neutron guide system for cold neutron (CN) research facilities since July, 2003. The neutron guide system plays an important role in transporting cold neutrons from the CN source to the neutron facilities as CN-NDP, CN-PGAA, SANS, etc. The CN activation station is being installed in the HANARO cold-neutron research project. The CN-NDP and CN-PGAA were selected as two facilities using at this station. At the end position of CG1 and CG2B beam guides, the CN-NDP and CN-PGAA will be installed in the CN guide hall. In order to predict the neutron flux and intensity values at the CG1 and CG2B beam guides, the simulation results of neutron flux at the CG1 and CG2B beam guides are presented by using several Monte Carlo (MC) neutron ray-tracing simulation codes. The intercomparison of neutron flux values between McStas, VITESS and RESTRAX are performed for getting fairly correct results at two neutron beam guides
International Nuclear Information System (INIS)
Document available in extended abstract form only. Highly alkaline environments induced by cement based materials are likely to deteriorate the physical and/or chemical properties of the bentonite buffer materials in radioactive waste repositories. Predicting long-term alteration of concrete/clay systems requires physico-chemical models and a number of input parameters. In our previous studies, alkaline alteration of bentonite and diffusive mass transport in micropores in cement based materials were empirically investigated and modeled. In order to provide reliability to the long-term prediction of bentonite buffer performance under disposal conditions, it is necessary to develop and verify reactive transport codes for concrete/clay systems. In this study, a PHREEQC-based, reactive transport analysis code (MC-CEMENT ver.2) was developed and was verified by comparing results of the calculations with in situ observations. The finite differential method (FDM) is applied in MC-CEMENT ver.2 for discretization. PHREEQC ver. 2.18 and HST3D are in charge of chemical reaction analysis and mass transport analysis, respectively. The sequential non-iterative approach (SNIA) is used to couple mass transport and chemical reactions. A model of diffusive mass transport in micropores in cement based materials (4) was incorporated to take low diffusivity in cement based materials into account. The code sequentially calculates the groundwater flow, heat transfer, mass transport, geochemical reactions, changes in properties such as porosity and effective montmorillonite dry density, changes in parameters on mass transport such as hydraulic conductivity and diffusivity. Space-dependent mass transport parameters such as diffusivity, hydraulic conductivity and porosity and their changes with time as a result of dissolution and formation of minerals can be calculated by the code. An in situ engineered analogue of a cement/clay interface which has undergone 15 years of interaction at IRSN
Performance Estimation of 2*3 MIMO-MC-CDMA using Convolution Code
2014-01-01
In this paper we estimate the performance of 2by3 MIMOMCCDMA system using convolution code in MATLAB which highly reduces BER by increasing the efficiency of system. MIMO and MCCDMA system combination is used to reduce bit error rate and also for forming a new system called MCCDMA which is multi user and multiple access schemes used to increase the performance of the system. MCCDMA system is a narrowband flat fading in nature which converts frequency selective to numerous narrowband flat fadi...
International Nuclear Information System (INIS)
IRSN is developing the DRACCAR computational software within the scope of its safety analyses on pressurised water reactors (PWR). This software is used to study loss-of-coolant accidents in the reactor core (LOCA) or in a spent fuel storage tank, for example. During such an accident, the coolant vaporises and the fuel rods dry out, which leads to an increase of their temperature, a swelling and fuel cladding failure. This swelling is responsible for major blockage in port of the core and can jeopardize the possibility of core cooling by means of back-up systems. The 3D multi-rod software is designed to model a fuel assembly so as to assess rod cooling and the blockage rate caused by deformed rods, by taking into account mechanical and thermal interactions between rods. The software can provide a consistent interpretation of the entire experimental database for a 'single-rod' configuration or a 'rod-bundle' configuration with either real or simulator fuel, transpose these results onto a reactor scale to determine what kind of research still needs to be conducted and finally, carry out safety studies. The models developed for this software cover: Heat transfers by conduction, convection and radiation. Oxidation of Zircaloy elements (cladding, guide tubes, inner shroud layer..) as well as hydriding process which can change mechanical properties. Thermomechanical behavior of fuel cladding (deformation and failure), including bowing phenomenon. Thermohydraulics on the scale of an assembly (to couple with an appropriate software), including a reflooding model. Fuel relocation and release of fission gases. A first version (DRACCAR V1) was delivered in March 2008 and is being validated on the basis of available experimental data (EDGAR, PHEBUS LOCA, PERICLES, REBEKA, HALDEN, etc.). A second version will be released in 2012 for which a coupling, in particular in the frame of the European NURISP project, is planned to an advanced sub-channel thermal-hydraulics code CATHARE
Lucas, Laurent; Loscos, Céline
2013-01-01
While 3D vision has existed for many years, the use of 3D cameras and video-based modeling by the film industry has induced an explosion of interest for 3D acquisition technology, 3D content and 3D displays. As such, 3D video has become one of the new technology trends of this century.The chapters in this book cover a large spectrum of areas connected to 3D video, which are presented both theoretically and technologically, while taking into account both physiological and perceptual aspects. Stepping away from traditional 3D vision, the authors, all currently involved in these areas, provide th
Beane, Andy
2012-01-01
The essential fundamentals of 3D animation for aspiring 3D artists 3D is everywhere--video games, movie and television special effects, mobile devices, etc. Many aspiring artists and animators have grown up with 3D and computers, and naturally gravitate to this field as their area of interest. Bringing a blend of studio and classroom experience to offer you thorough coverage of the 3D animation industry, this must-have book shows you what it takes to create compelling and realistic 3D imagery. Serves as the first step to understanding the language of 3D and computer graphics (CG)Covers 3D anim
3D Computations and Experiments
Energy Technology Data Exchange (ETDEWEB)
Couch, R; Faux, D; Goto, D; Nikkel, D
2004-04-05
This project consists of two activities. Task A, Simulations and Measurements, combines all the material model development and associated numerical work with the materials-oriented experimental activities. The goal of this effort is to provide an improved understanding of dynamic material properties and to provide accurate numerical representations of those properties for use in analysis codes. Task B, ALE3D Development, involves general development activities in the ALE3D code with the focus of improving simulation capabilities for problems of mutual interest to DoD and DOE. Emphasis is on problems involving multi-phase flow, blast loading of structures and system safety/vulnerability studies.
Sliding Adjustment for 3D Video Representation
Directory of Open Access Journals (Sweden)
Galpin Franck
2002-01-01
Full Text Available This paper deals with video coding of static scenes viewed by a moving camera. We propose an automatic way to encode such video sequences using several 3D models. Contrary to prior art in model-based coding where 3D models have to be known, the 3D models are automatically computed from the original video sequence. We show that several independent 3D models provide the same functionalities as one single 3D model, and avoid some drawbacks of the previous approaches. To achieve this goal we propose a novel algorithm of sliding adjustment, which ensures consistency of successive 3D models. The paper presents a method to automatically extract the set of 3D models and associate camera positions. The obtained representation can be used for reconstructing the original sequence, or virtual ones. It also enables 3D functionalities such as synthetic object insertion, lightning modification, or stereoscopic visualization. Results on real video sequences are presented.
Analysis of select BEAVRS PWR benchmark cycle 1 results using MC21 and OpenMC
International Nuclear Information System (INIS)
MC21 and OpenMC Monte Carlo results have been compared with hot zero power measurements from an operating pressurized water reactor (PWR), as specified in the recently released BEAVRS (Benchmark for Evaluation And Validation of Reactor Simulations) flux map data. Included in this comparison are axially-integrated full core detector measurements and axial detector profiles. In addition, MC21 was used in the 3-D analysis of cycle 1 using integrated reactor feedback capabilities. A quarter core model was chosen for this study because of difficulties converging a full core model. This depletion encompassed quarter assembly temperature feedback with more than 200 fission products in more than 379,000 regions. An approximate power history was utilized to model the actual ragged power history present in cycle 1. Detailed analysis of in-core detector data was performed with MC21 at three different exposure levels. These also included comparisons with SIMULATE-3 in order to further validate the MC21 code. The MC21 results agreed well with experiment and were comparable to SIMULATE-3 indicating that these initial analysis methods in MC21 provide a highly accurate solution to the analysis of a 3-D PWR. (author)
International Nuclear Information System (INIS)
TRAB-3D is a reactor dynamics code with three-dimensional neutronics coupled to core and circuit thermal-hydraulics. The code, entirely developed at VTT, can be used in transient and accident analyses of boiling (BWR) and pressurized water (PWR) reactors with rectangular fuel bundle geometry. The validation history of TRAB-3D includes calculation of international benchmark exercises, as well as comparisons with measured data from real plant transients. The most recent validation case is a load rejection test performed at the Olkiluoto 1 nuclear power plant in 1998 in connection with the power uprating project. The fact that there is local power measurement data available from this test makes it a suitable case for three-dimensional core model validation. The agreement between the results of the TRAB-3D calculation and the measurements is very good. (orig.)
International Nuclear Information System (INIS)
The investigation of the action of ionising radiation on biological structures requires a detailed analysis of the various stages underlying damage induction and evolution. In order to take into account the stochastic aspects characterising the process of interest ab initio models and MC simulation codes are required, which start from the physical track structure and follow its time evolution, taking into account the various levels of organisation of the biological targets (DNA, chromosomes etc.). Representative examples of the activities in this area of the Universities of Milan and Pavia will be presented, focusing on the development of models aimed: a) to better understand the action mechanisms of ionising radiation, in the framework of the EC project Low Dose Risk Models coordinated by the GSF Institute of munich; b) to study the induction of chromosome aberrations and their possible use as biomarkers, mainly in the framework of the INFN experiment DOSBI, developed in collaboration with the University of Naples; c) to provide basic data for applicative tools developed for hadron therapy and space radiation protection, in the framework of the INFN projects ATER.FIBI and FLUKA and the ASI (Italian Space Agency) project Influence of the shielding in the space radiation biological effectiveness
Kuepper, A H W; Kroupa, P; Baumgardt, H
2011-01-01
By analysing models of the young massive cluster R136 in 30 Doradus, set-up using the herewith introduced and publicly made available code McLuster, we investigate and compare different methods for detecting and quantifying mass segregation and substructure in non-seeing limited N-body data. For this purpose we generate star cluster models with different degrees of mass segregation and fractal substructure and analyse them. We quantify mass segregation by measuring, from the projected 2d model data, the mass function slope in radial annuli, by looking for colour gradients in radial colour profiles, by measuring Allison's Lambda parameter, and by determining the local stellar surface density around each star. We find that these methods for quantifying mass segregation often produce ambiguous results. Most reliable for detecting mass segregation is the mass function slope method, whereas the colour gradient method is the least practical in an R136-like configuration. The other two methods are more sensitive to ...
Harvey, R. W.; Petrov, Yu.; Jaeger, E. F.; Berry, L. A.; Bonoli, P. T.; Bader, A.
2015-12-01
A time-dependent simulation of C-Mod pulsed TCRF power is made obtaining minority hydrogen ion distributions with the CQL3D-Hybrid-FOW finite-orbit-width Fokker-Planck code. Cyclotron-resonant TCRF fields are calculated with the AORSA full wave code. The RF diffusion coefficients used in CQL3D are obtained with the DC Lorentz gyro-orbit code for perturbed particle trajectories in the combined equilibrium and TCRF electromagnetic fields. Prior results with a zero-banana-width simulation using the CQL3D/AORSA/DC time-cycles showed a pronounced enhancement of the H distribution in the perpendicular velocity direction compared to results obtained from Stix's quasilinear theory, and this substantially increased the rampup rate of the observed vertically-viewed neutral particle analyzer (NPA) flux, in general agreement with experiment. However, ramp down of the NPA flux after the pulse, remained long compared to the experiment. The present study compares the new FOW results, including relevant gyro-radius effects, to determine the importance of these new effects on the the NPA time-dependence.
Directory of Open Access Journals (Sweden)
D. Pletinckx
2012-09-01
Full Text Available The current 3D hype creates a lot of interest in 3D. People go to 3D movies, but are we ready to use 3D in our homes, in our offices, in our communication? Are we ready to deliver real 3D to a general public and use interactive 3D in a meaningful way to enjoy, learn, communicate? The CARARE project is realising this for the moment in the domain of monuments and archaeology, so that real 3D of archaeological sites and European monuments will be available to the general public by 2012. There are several aspects to this endeavour. First of all is the technical aspect of flawlessly delivering 3D content over all platforms and operating systems, without installing software. We have currently a working solution in PDF, but HTML5 will probably be the future. Secondly, there is still little knowledge on how to create 3D learning objects, 3D tourist information or 3D scholarly communication. We are still in a prototype phase when it comes to integrate 3D objects in physical or virtual museums. Nevertheless, Europeana has a tremendous potential as a multi-facetted virtual museum. Finally, 3D has a large potential to act as a hub of information, linking to related 2D imagery, texts, video, sound. We describe how to create such rich, explorable 3D objects that can be used intuitively by the generic Europeana user and what metadata is needed to support the semantic linking.
International Nuclear Information System (INIS)
This book explains modeling of solid works 3D and application of 3D CAD/CAM. The contents of this book are outline of modeling such as CAD and 2D and 3D, solid works composition, method of sketch, writing measurement fixing, selecting projection, choosing condition of restriction, practice of sketch, making parts, reforming parts, modeling 3D, revising 3D modeling, using pattern function, modeling necessaries, assembling, floor plan, 3D modeling method, practice floor plans for industrial engineer data aided manufacturing, processing of CAD/CAM interface.
International Nuclear Information System (INIS)
The results of calculation of the drop of one turbine to house load level are presented. Static calculations were performed by using the DYN3DH1.1/M3 reactor dynamics code, whereas the externally coupled codes ATHLET Mod. 1.1 Cycle C and DYN3DH1.1/M3 were employed during the calculations of the transients. The results are compared with observed values. (P.A.)
Directory of Open Access Journals (Sweden)
Felician ALECU
2010-01-01
Full Text Available Many professionals and 3D artists consider Blender as being the best open source solution for 3D computer graphics. The main features are related to modeling, rendering, shading, imaging, compositing, animation, physics and particles and realtime 3D/game creation.
3d-3d correspondence revisited
Chung, Hee-Joong; Dimofte, Tudor; Gukov, Sergei; Sułkowski, Piotr
2016-04-01
In fivebrane compactifications on 3-manifolds, we point out the importance of all flat connections in the proper definition of the effective 3d {N}=2 theory. The Lagrangians of some theories with the desired properties can be constructed with the help of homological knot invariants that categorify colored Jones polynomials. Higgsing the full 3d theories constructed this way recovers theories found previously by Dimofte-Gaiotto-Gukov. We also consider the cutting and gluing of 3-manifolds along smooth boundaries and the role played by all flat connections in this operation.
Brdnik, Lovro
2015-01-01
Diplomsko delo analizira trenutno stanje 3D tiskalnikov na trgu. Prikazan je razvoj in principi delovanja 3D tiskalnikov. Predstavljeni so tipi 3D tiskalnikov, njihove prednosti in slabosti. Podrobneje je predstavljena zgradba in delovanje koračnih motorjev. Opravljene so meritve koračnih motorjev. Opisana je programska oprema za rokovanje s 3D tiskalniki in komponente, ki jih potrebujemo za izdelavo. Diploma se oklepa vprašanja, ali je izdelava 3D tiskalnika bolj ekonomična kot pa naložba v ...
Jingjing Xu; Wei Yang; Linyuan Zhang; Ruisong Han; Xiaotao Shao
2015-01-01
In this paper, a wireless sensor network (WSN) technology adapted to underground channel conditions is developed, which has important theoretical and practical value for safety monitoring in underground coal mines. According to the characteristics that the space, time and frequency resources of underground tunnel are open, it is proposed to constitute wireless sensor nodes based on multicarrier code division multiple access (MC-CDMA) to make full use of these resources. To improve the wireles...
International Nuclear Information System (INIS)
Initial burnup and static calculations were performed by using the DYN3DH1.1/M3 reactor dynamic code. The KASSETA library was used to generate the reactor core neutron parameters. Transient calculations were carried out by means of the ATHLET Mod.1.1 Cycle C and DYN3DH1.1/M3 externally coupled codes. The VVER 440/213 ATHLET input deck prepared by the GRS for the conditions of the Loviisa-1 reactor unit was employed as well
a Fast Method for Measuring the Similarity Between 3d Model and 3d Point Cloud
Zhang, Zongliang; Li, Jonathan; Li, Xin; Lin, Yangbin; Zhang, Shanxin; Wang, Cheng
2016-06-01
This paper proposes a fast method for measuring the partial Similarity between 3D Model and 3D point Cloud (SimMC). It is crucial to measure SimMC for many point cloud-related applications such as 3D object retrieval and inverse procedural modelling. In our proposed method, the surface area of model and the Distance from Model to point Cloud (DistMC) are exploited as measurements to calculate SimMC. Here, DistMC is defined as the weighted distance of the distances between points sampled from model and point cloud. Similarly, Distance from point Cloud to Model (DistCM) is defined as the average distance of the distances between points in point cloud and model. In order to reduce huge computational burdens brought by calculation of DistCM in some traditional methods, we define SimMC as the ratio of weighted surface area of model to DistMC. Compared to those traditional SimMC measuring methods that are only able to measure global similarity, our method is capable of measuring partial similarity by employing distance-weighted strategy. Moreover, our method is able to be faster than other partial similarity assessment methods. We demonstrate the superiority of our method both on synthetic data and laser scanning data.
Xu, Jingjing; Yang, Wei; Zhang, Linyuan; Han, Ruisong; Shao, Xiaotao
2015-01-01
In this paper, a wireless sensor network (WSN) technology adapted to underground channel conditions is developed, which has important theoretical and practical value for safety monitoring in underground coal mines. According to the characteristics that the space, time and frequency resources of underground tunnel are open, it is proposed to constitute wireless sensor nodes based on multicarrier code division multiple access (MC-CDMA) to make full use of these resources. To improve the wireless transmission performance of source sensor nodes, it is also proposed to utilize cooperative sensors with good channel conditions from the sink node to assist source sensors with poor channel conditions. Moreover, the total power of the source sensor and its cooperative sensors is allocated on the basis of their channel conditions to increase the energy efficiency of the WSN. To solve the problem that multiple access interference (MAI) arises when multiple source sensors transmit monitoring information simultaneously, a kind of multi-sensor detection (MSD) algorithm with particle swarm optimization (PSO), namely D-PSO, is proposed for the time-frequency coded cooperative MC-CDMA WSN. Simulation results show that the average bit error rate (BER) performance of the proposed WSN in an underground coal mine is improved significantly by using wireless sensor nodes based on MC-CDMA, adopting time-frequency coded cooperative transmission and D-PSO algorithm with particle swarm optimization. PMID:26343660
Directory of Open Access Journals (Sweden)
Jingjing Xu
2015-08-01
Full Text Available In this paper, a wireless sensor network (WSN technology adapted to underground channel conditions is developed, which has important theoretical and practical value for safety monitoring in underground coal mines. According to the characteristics that the space, time and frequency resources of underground tunnel are open, it is proposed to constitute wireless sensor nodes based on multicarrier code division multiple access (MC-CDMA to make full use of these resources. To improve the wireless transmission performance of source sensor nodes, it is also proposed to utilize cooperative sensors with good channel conditions from the sink node to assist source sensors with poor channel conditions. Moreover, the total power of the source sensor and its cooperative sensors is allocated on the basis of their channel conditions to increase the energy efficiency of the WSN. To solve the problem that multiple access interference (MAI arises when multiple source sensors transmit monitoring information simultaneously, a kind of multi-sensor detection (MSD algorithm with particle swarm optimization (PSO, namely D-PSO, is proposed for the time-frequency coded cooperative MC-CDMA WSN. Simulation results show that the average bit error rate (BER performance of the proposed WSN in an underground coal mine is improved significantly by using wireless sensor nodes based on MC-CDMA, adopting time-frequency coded cooperative transmission and D-PSO algorithm with particle swarm optimization.
Xu, Jingjing; Yang, Wei; Zhang, Linyuan; Han, Ruisong; Shao, Xiaotao
2015-01-01
In this paper, a wireless sensor network (WSN) technology adapted to underground channel conditions is developed, which has important theoretical and practical value for safety monitoring in underground coal mines. According to the characteristics that the space, time and frequency resources of underground tunnel are open, it is proposed to constitute wireless sensor nodes based on multicarrier code division multiple access (MC-CDMA) to make full use of these resources. To improve the wireless transmission performance of source sensor nodes, it is also proposed to utilize cooperative sensors with good channel conditions from the sink node to assist source sensors with poor channel conditions. Moreover, the total power of the source sensor and its cooperative sensors is allocated on the basis of their channel conditions to increase the energy efficiency of the WSN. To solve the problem that multiple access interference (MAI) arises when multiple source sensors transmit monitoring information simultaneously, a kind of multi-sensor detection (MSD) algorithm with particle swarm optimization (PSO), namely D-PSO, is proposed for the time-frequency coded cooperative MC-CDMA WSN. Simulation results show that the average bit error rate (BER) performance of the proposed WSN in an underground coal mine is improved significantly by using wireless sensor nodes based on MC-CDMA, adopting time-frequency coded cooperative transmission and D-PSO algorithm with particle swarm optimization. PMID:26343660
International Nuclear Information System (INIS)
The objective of the first phase of the research of CNAT and the UPM project is the construction of several three-dimensional models detailed GOTHIC 8.0 code of containment of a buildings plant type PWR-W and KWU, corresponding to the Central Nuclear de Almaraz (CNA) and Trillo (CNT) respectively. (Author)
DYNA3D2000*, Explicit 3-D Hydrodynamic FEM Program
International Nuclear Information System (INIS)
1 - Description of program or function: DYNA3D2000 is a nonlinear explicit finite element code for analyzing 3-D structures and solid continuum. The code is vectorized and available on several computer platforms. The element library includes continuum, shell, beam, truss and spring/damper elements to allow maximum flexibility in modeling physical problems. Many materials are available to represent a wide range of material behavior, including elasticity, plasticity, composites, thermal effects and rate dependence. In addition, DYNA3D has a sophisticated contact interface capability, including frictional sliding, single surface contact and automatic contact generation. 2 - Method of solution: Discretization of a continuous model transforms partial differential equations into algebraic equations. A numerical solution is then obtained by solving these algebraic equations through a direct time marching scheme. 3 - Restrictions on the complexity of the problem: Recent software improvements have eliminated most of the user identified limitations with dynamic memory allocation and a very large format description that has pushed potential problem sizes beyond the reach of most users. The dominant restrictions remain in code execution speed and robustness, which the developers constantly strive to improve
Meulien Ohlmann, Odile
2013-02-01
Today the industry offers a chain of 3D products. Learning to "read" and to "create in 3D" becomes an issue of education of primary importance. 25 years professional experience in France, the United States and Germany, Odile Meulien set up a personal method of initiation to 3D creation that entails the spatial/temporal experience of the holographic visual. She will present some different tools and techniques used for this learning, their advantages and disadvantages, programs and issues of educational policies, constraints and expectations related to the development of new techniques for 3D imaging. Although the creation of display holograms is very much reduced compared to the creation of the 90ies, the holographic concept is spreading in all scientific, social, and artistic activities of our present time. She will also raise many questions: What means 3D? Is it communication? Is it perception? How the seeing and none seeing is interferes? What else has to be taken in consideration to communicate in 3D? How to handle the non visible relations of moving objects with subjects? Does this transform our model of exchange with others? What kind of interaction this has with our everyday life? Then come more practical questions: How to learn creating 3D visualization, to learn 3D grammar, 3D language, 3D thinking? What for? At what level? In which matter? for whom?
International Nuclear Information System (INIS)
A formula is derived for predicting multiprocessing efficiency on Cray supercomputers equipped with the Cray Time-Sharing System (CTSS). The model is applicable to an intensive time-sharing environment. The actual efficiency estimate depends on three factors: the code size, task length, and job mix. The implementation of multitasking in a three-dimensional plasma magnetohydrodynamics (MHD) code, TEMCO, is discussed. TEMCO solves the primitive one-fluid compressible MHD equations and includes resistive and Hall effects in Ohm's law. Virtually all segments of the main time-integration loop are multitasked. The multiprocessing efficiency model is applied to TEMCO. Excellent agreement is obtained between the actual multiprocessing efficiency and the theoretical prediction