GASFLOW computer code (physical models and input data)
International Nuclear Information System (INIS)
Muehlbauer, Petr
2007-11-01
The GASFLOW computer code was developed jointly by the Los Alamos National Laboratory, USA, and Forschungszentrum Karlsruhe, Germany. The code is primarily intended for calculations of the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containments and in other facilities. The physical models and the input data are described, and a commented simple calculation is presented
ETFOD: a point model physics code with arbitrary input
International Nuclear Information System (INIS)
Rothe, K.E.; Attenberger, S.E.
1980-06-01
ETFOD is a zero-dimensional code which solves a set of physics equations by minimization. The technique used is different than normally used, in that the input is arbitrary. The user is supplied with a set of variables from which he specifies which variables are input (unchanging). The remaining variables become the output. Presently the code is being used for ETF reactor design studies. The code was written in a manner to allow easy modificaton of equations, variables, and physics calculations. The solution technique is presented along with hints for using the code
Physical model of the nuclear fuel cycle simulation code SITON
International Nuclear Information System (INIS)
Brolly, Á.; Halász, M.; Szieberth, M.; Nagy, L.; Fehér, S.
2017-01-01
Finding answers to main challenges of nuclear energy, like resource utilisation or waste minimisation, calls for transient fuel cycle modelling. This motivation led to the development of SITON v2.0 a dynamic, discrete facilities/discrete materials and also discrete events fuel cycle simulation code. The physical model of the code includes the most important fuel cycle facilities. Facilities can be connected flexibly; their number is not limited. Material transfer between facilities is tracked by taking into account 52 nuclides. Composition of discharged fuel is determined using burnup tables except for the 2400 MW thermal power design of the Gas-Cooled Fast Reactor (GFR2400). For the GFR2400 the FITXS method is used, which fits one-group microscopic cross-sections as polynomial functions of the fuel composition. This method is accurate and fast enough to be used in fuel cycle simulations. Operation of the fuel cycle, i.e. material requests and transfers, is described by discrete events. In advance of the simulation reactors and plants formulate their requests as events; triggered requests are tracked. After that, the events are simulated, i.e. the requests are fulfilled and composition of the material flow between facilities is calculated. To demonstrate capabilities of SITON v2.0, a hypothetical transient fuel cycle is presented in which a 4-unit VVER-440 reactor park was replaced by one GFR2400 that recycled its own spent fuel. It is found that the GFR2400 can be started if the cooling time of its spent fuel is 2 years. However, if the cooling time is 5 years it needs an additional plutonium feed, which can be covered from the spent fuel of a Generation III light water reactor.
INTRA/Mod3.2. Manual and Code Description. Volume I - Physical Modelling
International Nuclear Information System (INIS)
Andersson, Jenny; Edlund, O.; Hermann, J.; Johansson, Lise-Lotte
1999-01-01
The INTRA Manual consists of two volumes. Volume I of the manual is a thorough description of the code INTRA, the Physical modelling of INTRA and the ruling numerical methods and volume II, the User's Manual is an input description. This document, the Physical modelling of INTRA, contains code characteristics, integration methods and applications
INTRA/Mod3.2. Manual and Code Description. Volume I - Physical Modelling
Energy Technology Data Exchange (ETDEWEB)
Andersson, Jenny; Edlund, O; Hermann, J; Johansson, Lise-Lotte
1999-01-01
The INTRA Manual consists of two volumes. Volume I of the manual is a thorough description of the code INTRA, the Physical modelling of INTRA and the ruling numerical methods and volume II, the User`s Manual is an input description. This document, the Physical modelling of INTRA, contains code characteristics, integration methods and applications
Physical model and calculation code for fuel coolant interactions
International Nuclear Information System (INIS)
Goldammer, H.; Kottowski, H.
1976-01-01
A physical model is proposed to describe fuel coolant interactions in shock-tube geometry. According to the experimental results, an interaction model which divides each cycle into three phases is proposed. The first phase is the fuel-coolant-contact, the second one is the ejection and recently of the coolant, and the third phase is the impact and fragmentation. Physical background of these phases are illustrated in the first part of this paper. Mathematical expressions of the model are exposed in the second part. A principal feature of the computational method is the consistent application of the fourier-equation throughout the whole interaction process. The results of some calculations, performed for different conditions are compiled in attached figures. (Aoki, K.)
DEFF Research Database (Denmark)
Fukui, Hironori; Popovski, Petar; Yomo, Hiroyuki
2014-01-01
Physical layer network coding (PLNC) has been proposed to improve throughput of the two-way relay channel, where two nodes communicate with each other, being assisted by a relay node. Most of the works related to PLNC are focused on a simple three-node model and they do not take into account...
Recent progress of an integrated implosion code and modeling of element physics
International Nuclear Information System (INIS)
Nagatomo, H.; Takabe, H.; Mima, K.; Ohnishi, N.; Sunahara, A.; Takeda, T.; Nishihara, K.; Nishiguchu, A.; Sawada, K.
2001-01-01
Physics of the inertial fusion is based on a variety of elements such as compressible hydrodynamics, radiation transport, non-ideal equation of state, non-LTE atomic process, and relativistic laser plasma interaction. In addition, implosion process is not in stationary state and fluid dynamics, energy transport and instabilities should be solved simultaneously. In order to study such complex physics, an integrated implosion code including all physics important in the implosion process should be developed. The details of physics elements should be studied and the resultant numerical modeling should be installed in the integrated code so that the implosion can be simulated with available computer within realistic CPU time. Therefore, this task can be basically separated into two parts. One is to integrate all physics elements into a code, which is strongly related to the development of hydrodynamic equation solver. We have developed 2-D integrated implosion code which solves mass, momentum, electron energy, ion energy, equation of states, laser ray-trace, laser absorption radiation, surface tracing and so on. The reasonable results in simulating Rayleigh-Taylor instability and cylindrical implosion are obtained using this code. The other is code development on each element physics and verification of these codes. We had progress in developing a nonlocal electron transport code and 2 and 3 dimension radiation hydrodynamic code. (author)
Sizing and scaling requirements of a large-scale physical model for code validation
International Nuclear Information System (INIS)
Khaleel, R.; Legore, T.
1990-01-01
Model validation is an important consideration in application of a code for performance assessment and therefore in assessing the long-term behavior of the engineered and natural barriers of a geologic repository. Scaling considerations relevant to porous media flow are reviewed. An analysis approach is presented for determining the sizing requirements of a large-scale, hydrology physical model. The physical model will be used to validate performance assessment codes that evaluate the long-term behavior of the repository isolation system. Numerical simulation results for sizing requirements are presented for a porous medium model in which the media properties are spatially uncorrelated
DEFF Research Database (Denmark)
Fukui, Hironori; Yomo, Hironori; Popovski, Petar
2013-01-01
of interfering nodes and usage of spatial reservation mechanisms. Specifically, we introduce a reserved area in order to protect the nodes involved in two-way relaying from the interference caused by neighboring nodes. We analytically derive the end-to-end rate achieved by PLNC considering the impact......Physical layer network coding (PLNC) has the potential to improve throughput of multi-hop networks. However, most of the works are focused on the simple, three-node model with two-way relaying, not taking into account the fact that there can be other neighboring nodes that can cause....../receive interference. The way to deal with this problem in distributed wireless networks is usage of MAC-layer mechanisms that make a spatial reservation of the shared wireless medium, similar to the well-known RTS/CTS in IEEE 802.11 wireless networks. In this paper, we investigate two-way relaying in presence...
A statistical methodology for quantification of uncertainty in best estimate code physical models
International Nuclear Information System (INIS)
Vinai, Paolo; Macian-Juan, Rafael; Chawla, Rakesh
2007-01-01
A novel uncertainty assessment methodology, based on a statistical non-parametric approach, is presented in this paper. It achieves quantification of code physical model uncertainty by making use of model performance information obtained from studies of appropriate separate-effect tests. Uncertainties are quantified in the form of estimated probability density functions (pdf's), calculated with a newly developed non-parametric estimator. The new estimator objectively predicts the probability distribution of the model's 'error' (its uncertainty) from databases reflecting the model's accuracy on the basis of available experiments. The methodology is completed by applying a novel multi-dimensional clustering technique based on the comparison of model error samples with the Kruskall-Wallis test. This takes into account the fact that a model's uncertainty depends on system conditions, since a best estimate code can give predictions for which the accuracy is affected by the regions of the physical space in which the experiments occur. The final result is an objective, rigorous and accurate manner of assigning uncertainty to coded models, i.e. the input information needed by code uncertainty propagation methodologies used for assessing the accuracy of best estimate codes in nuclear systems analysis. The new methodology has been applied to the quantification of the uncertainty in the RETRAN-3D void model and then used in the analysis of an independent separate-effect experiment. This has clearly demonstrated the basic feasibility of the approach, as well as its advantages in yielding narrower uncertainty bands in quantifying the code's accuracy for void fraction predictions
Physics Based Model for Cryogenic Chilldown and Loading. Part IV: Code Structure
Luchinsky, D. G.; Smelyanskiy, V. N.; Brown, B.
2014-01-01
This is the fourth report in a series of technical reports that describe separated two-phase flow model application to the cryogenic loading operation. In this report we present the structure of the code. The code consists of five major modules: (1) geometry module; (2) solver; (3) material properties; (4) correlations; and finally (5) stability control module. The two key modules - solver and correlations - are further divided into a number of submodules. Most of the physics and knowledge databases related to the properties of cryogenic two-phase flow are included into the cryogenic correlations module. The functional form of those correlations is not well established and is a subject of extensive research. Multiple parametric forms for various correlations are currently available. Some of them are included into correlations module as will be described in details in a separate technical report. Here we describe the overall structure of the code and focus on the details of the solver and stability control modules.
International Nuclear Information System (INIS)
Cooper, R.K.; Jones, M.E.
1989-01-01
The title given this paper is a bit presumptuous, since one can hardly expect to cover the physics incorporated into all the codes already written and currently being written. The authors focus on those codes which have been found to be particularly useful in the analysis and design of linacs. At that the authors will be a bit parochial and discuss primarily those codes used for the design of radio-frequency (rf) linacs, although the discussions of TRANSPORT and MARYLIE have little to do with the time structures of the beams being analyzed. The plan of this paper is first to describe rather simply the concepts of emittance and brightness, then to describe rather briefly each of the codes TRANSPORT, PARMTEQ, TBCI, MARYLIE, and ISIS, indicating what physics is and is not included in each of them. It is expected that the vast majority of what is covered will apply equally well to protons and electrons (and other particles). This material is intended to be tutorial in nature and can in no way be expected to be exhaustive. 31 references, 4 figures
The Physical Models and Statistical Procedures Used in the RACER Monte Carlo Code
International Nuclear Information System (INIS)
Sutton, T.M.; Brown, F.B.; Bischoff, F.G.; MacMillan, D.B.; Ellis, C.L.; Ward, J.T.; Ballinger, C.T.; Kelly, D.J.; Schindler, L.
1999-01-01
capability of performing iterated-source (criticality), multiplied-fixed-source, and fixed-source calculations. MCV uses a highly detailed continuous-energy (as opposed to multigroup) representation of neutron histories and cross section data. The spatial modeling is fully three-dimensional (3-D), and any geometrical region that can be described by quadric surfaces may be represented. The primary results are region-wise reaction rates, neutron production rates, slowing-down-densities, fluxes, leakages, and when appropriate the eigenvalue or multiplication factor. Region-wise nuclidic reaction rates are also computed, which may then be used by other modules in the system to determine time-dependent nuclide inventories so that RACER can perform depletion calculations. Furthermore, derived quantities such as ratios and sums of primary quantities and/or other derived quantities may also be calculated. MCV performs statistical analyses on output quantities, computing estimates of the 95% confidence intervals as well as indicators as to the reliability of these estimates. The remainder of this chapter provides an overview of the MCV algorithm. The following three chapters describe the MCV mathematical, physical, and statistical treatments in more detail. Specifically, Chapter 2 discusses topics related to tracking the histories including: geometry modeling, how histories are moved through the geometry, and variance reduction techniques related to the tracking process. Chapter 3 describes the nuclear data and physical models employed by MCV. Chapter 4 discusses the tallies, statistical analyses, and edits. Chapter 5 provides some guidance as to how to run the code, and Chapter 6 is a list of the code input options
Fission product release from nuclear fuel I. Physical modelling in the ASTEC code
International Nuclear Information System (INIS)
Brillant, G.; Marchetto, C.; Plumecocq, W.
2013-01-01
Highlights: • Physical modeling of FP and SM release in ASTEC is presented. • The release is described as solid state diffusion within fuel for high volatile FP. • The release is described as FP vaporisation for semi volatile FP. • The release is described as fuel vaporisation for low volatile FP. • ASTEC validation is presented in the second paper. - Abstract: This article is the first of a series of two articles dedicated to the mechanisms of fission product release from a degraded core as they are modelled in the ASTEC code. The ASTEC code aims at simulating severe accidents in nuclear reactors from the initiating event up to the radiological consequences on the environment. This code is used for several applications such as nuclear plant safety evaluation including probabilistic studies and emergency preparedness. To cope with the requirements of robustness and low calculation time, the code is based on a semi-empirical approach and only the main limiting phenomena that govern the release from intact rods and from debris beds are considered. For solid fuel, fission products are classified into three groups, depending on their degree of volatility. The kinetics of volatile fission products release depend on the rate-limiting process of solid-state diffusion through fuel grains. For semi-volatile fission products, the release from the open fuel porosities is assumed to be governed by vaporisation and mass transfer processes. The key phenomenon for the release of low volatile fission products is supposed to be fuel volatilisation. A similar approach is used for the release of fission products from a rubble bed. An in-depth validation of the code including both analytical and integral experiments is the subject of the second article
WWER reactor physics code applications
International Nuclear Information System (INIS)
Gado, J.; Kereszturi, A.; Gacs, A.; Telbisz, M.
1994-01-01
The coupled steady-state reactor physics and thermohydraulic code system KARATE has been developed and applied for WWER-1000 and WWER-440 operational calculations. The 3 D coupled kinetic code KIKO3D has been developed and validated for WWER-440 accident analysis applications. The coupled kinetic code SMARTA developed by VTT Helsinki has been applied for WWER-440 accident analysis. The paper gives a summary of the experience in code development and application. (authors). 10 refs., 2 tabs., 5 figs
Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes
International Nuclear Information System (INIS)
Lindley, B.A.; Lillington, J.N.; Kotlyar, D.; Parks, G.T.; Petrovic, B.
2016-01-01
adopts an integral configuration and a fully passive decay heat removal system to provide indefinite cooling capability for a class of accidents. This paper presents the equilibrium cycle core design and reactor physics behaviour of the I"2S-LWR with U_3Si_2 and the advanced steel cladding. The results were obtained using the traditional two-stage approach, in which homogenized macroscopic cross-section sets were generated by WIMS and applied in a full 3D core solution with PANTHER. The results obtained with WIMS/PANTHER were compared against the Monte Carlo Serpent code developed by VTT and previously reported results for the I"2S-LWR. The results were found to be in a good agreement (e.g. <200 pcm in reactivity) among the compared codes, giving confidence that the WIMS/PANTHER reactor physics package can be reliably used in modelling advanced LWR systems. (authors)
ACCELERATION PHYSICS CODE WEB REPOSITORY.
Energy Technology Data Exchange (ETDEWEB)
WEI, J.
2006-06-26
In the framework of the CARE HHH European Network, we have developed a web-based dynamic accelerator-physics code repository. We describe the design, structure and contents of this repository, illustrate its usage, and discuss our future plans, with emphasis on code benchmarking.
Improvements of Physical Models in TRITGO code for Tritium Behavior Analysis in VHTR
International Nuclear Information System (INIS)
Yoo, Jun Soo; Tak, Nam Il; Lim, Hong Sik
2010-01-01
Since tritium is radioactive material with 12.32 year of half-life and is generated by a ternary fission reaction in fuel as well as by neutron absorption reactions of impurities in Very High Temperature gas-cooled Reactor (VHTR) core, accurate prediction of tritium behavior and its concentration in product hydrogen is definitely important in terms of public safety for its construction. In this respect, TRITGO code was developed for estimating the tritium production and distribution in high temperature gas-cooled reactors by General Atomics (GA). However, some models in it are hard-wired to specific reactor type or too simplified, which makes the analysis results less applicable. Thus, major improvements need to be considered for better predictions. In this study, some of model improvements have been suggested and its effect is evaluated based on the analysis work against PMR600 design concept
SERPENT Monte Carlo reactor physics code
International Nuclear Information System (INIS)
Leppaenen, J.
2010-01-01
SERPENT is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, developed at VTT Technical Research Centre of Finland since 2004. The code is specialized in lattice physics applications, but the universe-based geometry description allows transport simulation to be carried out in complicated three-dimensional geometries as well. The suggested applications of SERPENT include generation of homogenized multi-group constants for deterministic reactor simulator calculations, fuel cycle studies involving detailed assembly-level burnup calculations, validation of deterministic lattice transport codes, research reactor applications, educational purposes and demonstration of reactor physics phenomena. The Serpent code has been publicly distributed by the OECD/NEA Data Bank since May 2009 and RSICC in the U. S. since March 2010. The code is being used in some 35 organizations in 20 countries around the world. This paper presents an overview of the methods and capabilities of the Serpent code, with examples in the modelling of WWER-440 reactor physics. (Author)
Modelling of the thermomechanical and physical processes in FR fuel pins using the GERMINAL code
International Nuclear Information System (INIS)
Roche, L.; Pelletier, M.
2000-01-01
In the frame of the R and D on Fast Reactor mixed oxide fuels, CEA/DEC has developed the computer code GERMINAL for studying fuel pin thermal and mechanical behaviour, both during steady-state and incidental conditions, up to high burn-up (25 at%). The first part of this paper is devoted to the description of the main models: fuel evolution (central hole and porosity evolution, Plutonium redistribution, O/M radial profile, transient gas swelling, melting fuel behaviour, minor actinides production), high burn-up models (fission gas, volatile fission products and JOG formation), fuel-cladding heat transfer, fuel-cladding mechanical interaction. The second part gives some examples of calculation results taken from the GERMINAL validation data base (more than 40 experiments from PHENIX, PFR, CABRI reactors), with special emphasis on: local fission gas retention and global release, fuel geometry evolution, radial redistribution of plutonium for high burn-up fuels, solid and annular fuel behaviour during power ramps including fuel melting, helium formation from MA (Am and Np) doped homogeneous fuels. (author)
Wind-US Code Physical Modeling Improvements to Complement Hypersonic Testing and Evaluation
Georgiadis, Nicholas J.; Yoder, Dennis A.; Towne, Charles S.; Engblom, William A.; Bhagwandin, Vishal A.; Power, Greg D.; Lankford, Dennis W.; Nelson, Christopher C.
2009-01-01
This report gives an overview of physical modeling enhancements to the Wind-US flow solver which were made to improve the capabilities for simulation of hypersonic flows and the reliability of computations to complement hypersonic testing. The improvements include advanced turbulence models, a bypass transition model, a conjugate (or closely coupled to vehicle structure) conduction-convection heat transfer capability, and an upgraded high-speed combustion solver. A Mach 5 shock-wave boundary layer interaction problem is used to investigate the benefits of k- s and k-w based explicit algebraic stress turbulence models relative to linear two-equation models. The bypass transition model is validated using data from experiments for incompressible boundary layers and a Mach 7.9 cone flow. The conjugate heat transfer method is validated for a test case involving reacting H2-O2 rocket exhaust over cooled calorimeter panels. A dual-mode scramjet configuration is investigated using both a simplified 1-step kinetics mechanism and an 8-step mechanism. Additionally, variations in the turbulent Prandtl and Schmidt numbers are considered for this scramjet configuration.
Computer codes in particle transport physics
International Nuclear Information System (INIS)
Pesic, M.
2004-01-01
Simulation of transport and interaction of various particles in complex media and wide energy range (from 1 MeV up to 1 TeV) is very complicated problem that requires valid model of a real process in nature and appropriate solving tool - computer code and data library. A brief overview of computer codes based on Monte Carlo techniques for simulation of transport and interaction of hadrons and ions in wide energy range in three dimensional (3D) geometry is shown. Firstly, a short attention is paid to underline the approach to the solution of the problem - process in nature - by selection of the appropriate 3D model and corresponding tools - computer codes and cross sections data libraries. Process of data collection and evaluation from experimental measurements and theoretical approach to establishing reliable libraries of evaluated cross sections data is Ion g, difficult and not straightforward activity. For this reason, world reference data centers and specialized ones are acknowledged, together with the currently available, state of art evaluated nuclear data libraries, as the ENDF/B-VI, JEF, JENDL, CENDL, BROND, etc. Codes for experimental and theoretical data evaluations (e.g., SAMMY and GNASH) together with the codes for data processing (e.g., NJOY, PREPRO and GRUCON) are briefly described. Examples of data evaluation and data processing to generate computer usable data libraries are shown. Among numerous and various computer codes developed in transport physics of particles, the most general ones are described only: MCNPX, FLUKA and SHIELD. A short overview of basic application of these codes, physical models implemented with their limitations, energy ranges of particles and types of interactions, is given. General information about the codes covers also programming language, operation system, calculation speed and the code availability. An example of increasing computation speed of running MCNPX code using a MPI cluster compared to the code sequential option
International Nuclear Information System (INIS)
Armand, Patrick
1995-01-01
The aim of this work consists in the Fluid Mechanics and aerosol Physics coupling. In the first part, the order of magnitude analysis of the particle dynamics is done. This particle is embedded in a non-uniform unsteady flow. Flow approximations around the inclusion are described. Corresponding aerodynamic drag formulae are expressed. Possible situations related to the problem data are extensively listed. In the second part, one studies the turbulent particles transport. Eulerian approach which is particularly well adapted to industrial codes is preferred in comparison with the Lagrangian methods. One chooses the two-fluid formalism in which career gas-particles slip is taken into account. Turbulence modelling gets through a k-epsilon modulated by the inclusions action on the flow. The model is implemented In a finite elements code. Finally, In the third part, one validates the modelling in laminar and turbulent cases. We compare simulations to various experiments (settling battery, inertial impaction in a bend, jets loaded with glass beads particles) which are taken in the literature or done by ourselves at the laboratory. The results are very close. It is a good point when it is thought of the particles transport model and associated software future use. (author) [fr
International Nuclear Information System (INIS)
Cetnar, Jerzy
2014-01-01
The recent development of MCB - Monte Carlo Continuous Energy Burn-up code is directed towards advanced description of modern reactors, including double heterogeneity structures that exist in HTR-s. In this, we exploit the advantages of MCB methodology in integrated approach, where physics, neutronics, burnup, reprocessing, non-stationary process modeling (control rod operation) and refined spatial modeling are carried in a single flow. This approach allows for implementations of advanced statistical options like analysis of error propagation, perturbation in time domain, sensitivity and source convergence analyses. It includes statistical analysis of burnup process, emitted particle collection, thermal-hydraulic coupling, automatic power profile calculations, advanced procedures of burnup step normalization and enhanced post processing capabilities. (author)
International Nuclear Information System (INIS)
Sabchevski, S; Zhelyazkov, I; Benova, E; Atanassov, V; Dankov, P; Thumm, M; Arnold, A; Jin, J; Rzesnicki, T
2006-01-01
Quasi-optical (QO) mode converters are used to transform electromagnetic waves of complex structure and polarization generated in gyrotron cavities into a linearly polarized, Gaussian-like beam suitable for transmission. The efficiency of this conversion as well as the maintenance of low level of diffraction losses are crucial for the implementation of powerful gyrotrons as radiation sources for electron-cyclotron-resonance heating of fusion plasmas. The use of adequate physical models, efficient numerical schemes and up-to-date computer codes may provide the high accuracy necessary for the design and analysis of these devices. In this review, we briefly sketch the most commonly used QO converters, the mathematical base they have been treated on and the basic features of the numerical schemes used. Further on, we discuss the applicability of several commercially available and free software packages, their advantages and drawbacks, for solving QO related problems
Theoretical Atomic Physics code development II: ACE: Another collisional excitation code
International Nuclear Information System (INIS)
Clark, R.E.H.; Abdallah, J. Jr.; Csanak, G.; Mann, J.B.; Cowan, R.D.
1988-12-01
A new computer code for calculating collisional excitation data (collision strengths or cross sections) using a variety of models is described. The code uses data generated by the Cowan Atomic Structure code or CATS for the atomic structure. Collisional data are placed on a random access file and can be displayed in a variety of formats using the Theoretical Atomic Physics Code or TAPS. All of these codes are part of the Theoretical Atomic Physics code development effort at Los Alamos. 15 refs., 10 figs., 1 tab
GALEV evolutionary synthesis models – I. Code, input physics and web
Kotulla, R.; Fritze, U.; Weilbacher, P.; Anders, P.
2009-01-01
GALEV (GALaxy EVolution) evolutionary synthesis models describe the evolution of stellar populations in general, of star clusters as well as of galaxies, both in terms of resolved stellar populations and of integrated light properties over cosmological time-scales of ≥13 Gyr from the onset of star
HOTSPOT Health Physics codes for the PC
Energy Technology Data Exchange (ETDEWEB)
Homann, S.G.
1994-03-01
The HOTSPOT Health Physics codes were created to provide Health Physics personnel with a fast, field-portable calculation tool for evaluating accidents involving radioactive materials. HOTSPOT codes are a first-order approximation of the radiation effects associated with the atmospheric release of radioactive materials. HOTSPOT programs are reasonably accurate for a timely initial assessment. More importantly, HOTSPOT codes produce a consistent output for the same input assumptions and minimize the probability of errors associated with reading a graph incorrectly or scaling a universal nomogram during an emergency. The HOTSPOT codes are designed for short-term (less than 24 hours) release durations. Users requiring radiological release consequences for release scenarios over a longer time period, e.g., annual windrose data, are directed to such long-term models as CAPP88-PC (Parks, 1992). Users requiring more sophisticated modeling capabilities, e.g., complex terrain; multi-location real-time wind field data; etc., are directed to such capabilities as the Department of Energy`s ARAC computer codes (Sullivan, 1993). Four general programs -- Plume, Explosion, Fire, and Resuspension -- calculate a downwind assessment following the release of radioactive material resulting from a continuous or puff release, explosive release, fuel fire, or an area contamination event. Other programs deal with the release of plutonium, uranium, and tritium to expedite an initial assessment of accidents involving nuclear weapons. Additional programs estimate the dose commitment from the inhalation of any one of the radionuclides listed in the database of radionuclides; calibrate a radiation survey instrument for ground-survey measurements; and screen plutonium uptake in the lung (see FIDLER Calibration and LUNG Screening sections).
HOTSPOT Health Physics codes for the PC
International Nuclear Information System (INIS)
Homann, S.G.
1994-03-01
The HOTSPOT Health Physics codes were created to provide Health Physics personnel with a fast, field-portable calculation tool for evaluating accidents involving radioactive materials. HOTSPOT codes are a first-order approximation of the radiation effects associated with the atmospheric release of radioactive materials. HOTSPOT programs are reasonably accurate for a timely initial assessment. More importantly, HOTSPOT codes produce a consistent output for the same input assumptions and minimize the probability of errors associated with reading a graph incorrectly or scaling a universal nomogram during an emergency. The HOTSPOT codes are designed for short-term (less than 24 hours) release durations. Users requiring radiological release consequences for release scenarios over a longer time period, e.g., annual windrose data, are directed to such long-term models as CAPP88-PC (Parks, 1992). Users requiring more sophisticated modeling capabilities, e.g., complex terrain; multi-location real-time wind field data; etc., are directed to such capabilities as the Department of Energy's ARAC computer codes (Sullivan, 1993). Four general programs -- Plume, Explosion, Fire, and Resuspension -- calculate a downwind assessment following the release of radioactive material resulting from a continuous or puff release, explosive release, fuel fire, or an area contamination event. Other programs deal with the release of plutonium, uranium, and tritium to expedite an initial assessment of accidents involving nuclear weapons. Additional programs estimate the dose commitment from the inhalation of any one of the radionuclides listed in the database of radionuclides; calibrate a radiation survey instrument for ground-survey measurements; and screen plutonium uptake in the lung (see FIDLER Calibration and LUNG Screening sections)
Accelerator Physics Code Web Repository
Zimmermann, Frank; Bellodi, G; Benedetto, E; Dorda, U; Giovannozzi, Massimo; Papaphilippou, Y; Pieloni, T; Ruggiero, F; Rumolo, G; Schmidt, F; Todesco, E; Zotter, Bruno W; Payet, J; Bartolini, R; Farvacque, L; Sen, T; Chin, Y H; Ohmi, K; Oide, K; Furman, M; Qiang, J; Sabbi, G L; Seidl, P A; Vay, J L; Friedman, A; Grote, D P; Cousineau, S M; Danilov, V; Holmes, J A; Shishlo, A; Kim, E S; Cai, Y; Pivi, M; Kaltchev, D I; Abell, D T; Katsouleas, Thomas C; Boine-Frankenheim, O; Franchetti, G; Hofmann, I; Machida, S; Wei, J
2006-01-01
In the framework of the CARE HHH European Network, we have developed a web-based dynamic acceleratorphysics code repository. We describe the design, structure and contents of this repository, illustrate its usage, and discuss our future plans, with emphasis on code benchmarking.
International Nuclear Information System (INIS)
Skorek, Tomasz; Crecy, Agnes de
2013-01-01
PREMIUM (Post BEMUSE Reflood Models Input Uncertainty Methods) is an activity launched with the aim to push forward the methods of quantification of physical models uncertainties in thermal-hydraulic codes. It is endorsed by OECD/NEA/CSNI/WGAMA. The benchmark PREMIUM is addressed to all who applies uncertainty evaluation methods based on input uncertainties quantification and propagation. The benchmark is based on a selected case of uncertainty analysis application to the simulation of quench front propagation in an experimental test facility. Application to an experiment enables evaluation and confirmation of the quantified probability distribution functions on the basis of experimental data. The scope of the benchmark comprises a review of the existing methods, selection of potentially important uncertain input parameters, preliminary quantification of the ranges and distributions of the identified parameters, evaluation of the probability density function using experimental results of tests performed on FEBA test facility and confirmation/validation of the performed quantification on the basis of blind calculation of Reflood 2-D PERICLES experiment. (authors)
VOA: a 2-d plasma physics code
International Nuclear Information System (INIS)
Eltgroth, P.G.
1975-12-01
A 2-dimensional relativistic plasma physics code was written and tested. The non-thermal components of the particle distribution functions are represented by expansion into moments in momentum space. These moments are computed directly from numerical equations. Currently three species are included - electrons, ions and ''beam electrons''. The computer code runs on either the 7600 or STAR machines at LLL. Both the physics and the operation of the code are discussed
Yoriyaz, Hélio; Moralles, Maurício; Siqueira, Paulo de Tarso Dalledone; Guimarães, Carla da Costa; Cintra, Felipe Belonsi; dos Santos, Adimir
2009-11-01
Radiopharmaceutical applications in nuclear medicine require a detailed dosimetry estimate of the radiation energy delivered to the human tissues. Over the past years, several publications addressed the problem of internal dose estimate in volumes of several sizes considering photon and electron sources. Most of them used Monte Carlo radiation transport codes. Despite the widespread use of these codes due to the variety of resources and potentials they offered to carry out dose calculations, several aspects like physical models, cross sections, and numerical approximations used in the simulations still remain an object of study. Accurate dose estimate depends on the correct selection of a set of simulation options that should be carefully chosen. This article presents an analysis of several simulation options provided by two of the most used codes worldwide: MCNP and GEANT4. For this purpose, comparisons of absorbed fraction estimates obtained with different physical models, cross sections, and numerical approximations are presented for spheres of several sizes and composed as five different biological tissues. Considerable discrepancies have been found in some cases not only between the different codes but also between different cross sections and algorithms in the same code. Maximum differences found between the two codes are 5.0% and 10%, respectively, for photons and electrons. Even for simple problems as spheres and uniform radiation sources, the set of parameters chosen by any Monte Carlo code significantly affects the final results of a simulation, demonstrating the importance of the correct choice of parameters in the simulation.
International Nuclear Information System (INIS)
Likhanskii, V.; Evdokimov, I.; Khoruzhii, O.; Sorokin, A.; Novikov, V.
2003-01-01
It is appropriate to use the dependences, based on physical models, in the design-analytical codes for improving of reliability of defective fuel rod detection and for determination of defect characteristics by activity measuring in the primary coolant. In the paper the results on development of some physical models and integral mechanistic codes, assigned for prediction of defective fuel rod behaviour are presented. The analysis of mass transfer and mass exchange between fuel rod and coolant showed that the rates of these processes depends on many factors, such as coolant turbulent flow, pressure, effective hydraulic diameter of defect, fuel rod geometric parameters. The models, which describe these dependences, have been created. The models of thermomechanical fuel behaviour, stable gaseous FP release were modified and new computer code RTOP-CA was created thereupon for description of defective fuel rod behaviour and activity release into the primary coolant. The model of fuel oxidation in in-pile conditions, which includes radiolysis and RTOP-LT after validation of physical models are planned to be used for prediction of defective fuel rods behaviour
Data exchange between zero dimensional code and physics platform in the CFETR integrated system code
Energy Technology Data Exchange (ETDEWEB)
Xu, Guoliang [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Shi, Nan [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Zhou, Yifu; Mao, Shifeng [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Jian, Xiang [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronics Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Chen, Jiale [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Liu, Li; Chan, Vincent [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Ye, Minyou, E-mail: yemy@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China)
2016-11-01
Highlights: • The workflow of the zero dimensional code and the multi-dimension physics platform of CFETR integrated system codeis introduced. • The iteration process among the codes in the physics platform. • The data transfer between the zero dimensionalcode and the physical platform, including data iteration and validation, and justification for performance parameters.. - Abstract: The China Fusion Engineering Test Reactor (CFETR) integrated system code contains three parts: a zero dimensional code, a physics platform and an engineering platform. We use the zero dimensional code to identify a set of preliminary physics and engineering parameters for CFETR, which is used as input to initiate multi-dimension studies using the physics and engineering platform for design, verification and validation. Effective data exchange between the zero dimensional code and the physical platform is critical for the optimization of CFETR design. For example, in evaluating the impact of impurity radiation on core performance, an open field line code is used to calculate the impurity transport from the first-wall boundary to the pedestal. The impurity particle in the pedestal are used as boundary conditions in a transport code for calculating impurity transport in the core plasma and the impact of core radiation on core performance. Comparison of the results from the multi-dimensional study to those from the zero dimensional code is used to further refine the controlled radiation model. The data transfer between the zero dimensional code and the physical platform, including data iteration and validation, and justification for performance parameters will be presented in this paper.
International Nuclear Information System (INIS)
Roux, A.
2001-01-01
The diversity of radiological accidents makes difficult the medical prognosis and the therapy choice from only clinical observations. To complete this information, it is important to know the global dose received by the organism and the dose distributions in depth in tissues. The dose estimation can be made by a physical reconstruction of the accident with the help of tools based on experimental techniques or on calculation. The software of the geometry construction (M.G.E.D.), associated to the Monte-Carlo code of photons and neutrons transport (M.O.R.S.E.) replies these constraints. An important result of this work is to determine the principal parameters to know in function of the accident type, as well as the precision level required for these parameters. (N.C.)
Physics options in the plasma code VOA
International Nuclear Information System (INIS)
Eltgroth, P.G.
1976-06-01
A two dimensional relativistic plasma physics code has been modified to accomodate general electromagnetic boundary conditions and various approximations of basic physics. The code can treat internal conductors and insulators, imposed electromagnetic fields, the effects of external circuitry and non-equilibrium starting conditions. Particle dynamics options include a full microscopic treatment, fully relaxed electrons, a low frequency electron approximation and a combination of approximations for specified zones. Electromagnetic options include the full wave treatment, an electrostatic approximation and two varieties of magnetohydrodynamic approximations in specified zones
Studies on DANESS Code Modeling
International Nuclear Information System (INIS)
Jeong, Chang Joon
2009-09-01
The DANESS code modeling study has been performed. DANESS code is widely used in a dynamic fuel cycle analysis. Korea Atomic Energy Research Institute (KAERI) has used the DANESS code for the Korean national nuclear fuel cycle scenario analysis. In this report, the important models such as Energy-demand scenario model, New Reactor Capacity Decision Model, Reactor and Fuel Cycle Facility History Model, and Fuel Cycle Model are investigated. And, some models in the interface module are refined and inserted for Korean nuclear fuel cycle model. Some application studies have also been performed for GNEP cases and for US fast reactor scenarios with various conversion ratios
Operational reactor physics analysis codes (ORPAC)
International Nuclear Information System (INIS)
Kumar, Jainendra; Singh, K.P.; Singh, Kanchhi
2007-07-01
For efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are regularly required. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation of samples requires a prior estimation of the reactivity load due to the sample, the heating rate and the activity developed in it during irradiation. For the safety of personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr.Therefore, a proper shielding and radioactive cooling of the irradiated sample are required to meet the said requirement. Knowledge of xenon load variation with time (Startup-curve) helps in estimating Xenon override time. Monitoring of power in individual fuel channels during reactor operation is essential to know any abnormal power distribution to avoid unsafe situations. Complexities in the estimation of above mentioned reactor parameters and their frequent requirement compel one to use computer codes to avoid possible human errors. For efficient and quick evaluation of parameters related to reactor operations such as xenon load, critical moderator height and nuclear heating and reactivity load of isotope samples/experimental assembly, a computer code ORPAC (Operational Reactor Physics Analysis Codes) has been developed. This code is being used for regular assessment of reactor physics parameters in Dhruva and Cirus. The code ORPAC written in Visual Basic 6.0 environment incorporates several important operational reactor physics aspects on a single platform with graphical user interfaces (GUI) to make it more user-friendly and presentable. (author)
Theoretical atomic physics code development III TAPS: A display code for atomic physics data
International Nuclear Information System (INIS)
Clark, R.E.H.; Abdallah, J. Jr.; Kramer, S.P.
1988-12-01
A large amount of theoretical atomic physics data is becoming available through use of the computer codes CATS and ACE developed at Los Alamos National Laboratory. A new code, TAPS, has been written to access this data, perform averages over terms and configurations, and display information in graphical or text form. 7 refs., 13 figs., 1 tab
Chemistry models in the Victoria code
International Nuclear Information System (INIS)
Grimley, A.J. III
1988-01-01
The VICTORIA Computer code consists of the fission product release and chemistry models for the MELPROG severe accident analysis code. The chemistry models in VICTORIA are used to treat multi-phase interactions in four separate physical regions: fuel grains, gap/open porosity/clad, coolant/aerosols, and structure surfaces. The physical and chemical environment of each region is very different from the others and different models are required for each. The common thread in the modelling is the use of a chemical equilibrium assumption. The validity of this assumption along with a description of the various physical constraints applicable to each region will be discussed. The models that result from the assumptions and constraints will be presented along with samples of calculations in each region
Steam condensation modelling in aerosol codes
International Nuclear Information System (INIS)
Dunbar, I.H.
1986-01-01
The principal subject of this study is the modelling of the condensation of steam into and evaporation of water from aerosol particles. These processes introduce a new type of term into the equation for the development of the aerosol particle size distribution. This new term faces the code developer with three major problems: the physical modelling of the condensation/evaporation process, the discretisation of the new term and the separate accounting for the masses of the water and of the other components. This study has considered four codes which model the condensation of steam into and its evaporation from aerosol particles: AEROSYM-M (UK), AEROSOLS/B1 (France), NAUA (Federal Republic of Germany) and CONTAIN (USA). The modelling in the codes has been addressed under three headings. These are the physical modelling of condensation, the mathematics of the discretisation of the equations, and the methods for modelling the separate behaviour of different chemical components of the aerosol. The codes are least advanced in area of solute effect modelling. At present only AEROSOLS/B1 includes the effect. The effect is greater for more concentrated solutions. Codes without the effect will be more in error (underestimating the total airborne mass) the less condensation they predict. Data are needed on the water vapour pressure above concentrated solutions of the substances of interest (especially CsOH and CsI) if the extent to which aerosols retain water under superheated conditions is to be modelled. 15 refs
Theoretical Atomic Physics code development IV: LINES, A code for computing atomic line spectra
International Nuclear Information System (INIS)
Abdallah, J. Jr.; Clark, R.E.H.
1988-12-01
A new computer program, LINES, has been developed for simulating atomic line emission and absorption spectra using the accurate fine structure energy levels and transition strengths calculated by the (CATS) Cowan Atomic Structure code. Population distributions for the ion stages are obtained in LINES by using the Local Thermodynamic Equilibrium (LTE) model. LINES is also useful for displaying the pertinent atomic data generated by CATS. This report describes the use of LINES. Both CATS and LINES are part of the Theoretical Atomic PhysicS (TAPS) code development effort at Los Alamos. 11 refs., 9 figs., 1 tab
A review on the CIRCE methodology to quantify the uncertainty of the physical models of a code
International Nuclear Information System (INIS)
Jeon, Seong Su; Hong, Soon Joon; Bang, Young Seok
2012-01-01
In the field of nuclear engineering, recent regulatory audit calculations of large break loss of coolant accident (LBLOCA) have been performed with the best estimate code such as MARS, RELAP5 and CATHARE. Since the credible regulatory audit calculation is very important in the evaluation of the safety of the nuclear power plant (NPP), there have been many researches to develop rules and methodologies for the use of best estimate codes. One of the major points is to develop the best estimate plus uncertainty (BEPU) method for uncertainty analysis. As a representative BEPU method, NRC proposes the CSAU (Code scaling, applicability and uncertainty) methodology, which clearly identifies the different steps necessary for an uncertainty analysis. The general idea is 1) to determine all the sources of uncertainty in the code, also called basic uncertainties, 2) quantify them and 3) combine them in order to obtain the final uncertainty for the studied application. Using the uncertainty analysis such as CSAU methodology, an uncertainty band for the code response (calculation result), important from the safety point of view is calculated and the safety margin of the NPP is quantified. An example of such a response is the peak cladding temperature (PCT) for a LBLOCA. However, there is a problem in the uncertainty analysis with the best estimate codes. Generally, it is very difficult to determine the uncertainties due to the empiricism of closure laws (also called correlations or constitutive relationships). So far the only proposed approach is based on the expert judgment. For this case, the uncertainty range of important parameters can be wide and inaccurate so that the confidence level of the BEPU calculation results can be decreased. In order to solve this problem, recently CEA (France) proposes a statistical method of data analysis, called CIRCE. The CIRCE method is intended to quantify the uncertainties of the correlations of a code. It may replace the expert judgment
Cheetah: Starspot modeling code
Walkowicz, Lucianne; Thomas, Michael; Finkestein, Adam
2014-12-01
Cheetah models starspots in photometric data (lightcurves) by calculating the modulation of a light curve due to starspots. The main parameters of the program are the linear and quadratic limb darkening coefficients, stellar inclination, spot locations and sizes, and the intensity ratio of the spots to the stellar photosphere. Cheetah uses uniform spot contrast and the minimum number of spots needed to produce a good fit and ignores bright regions for the sake of simplicity.
High burnup models in computer code fair
Energy Technology Data Exchange (ETDEWEB)
Dutta, B K; Swami Prasad, P; Kushwaha, H S; Mahajan, S C; Kakodar, A [Bhabha Atomic Research Centre, Bombay (India)
1997-08-01
An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ``Light water reactor fuel rod modelling code evaluation`` and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs.
High burnup models in computer code fair
International Nuclear Information System (INIS)
Dutta, B.K.; Swami Prasad, P.; Kushwaha, H.S.; Mahajan, S.C.; Kakodar, A.
1997-01-01
An advanced fuel analysis code FAIR has been developed for analyzing the behavior of fuel rods of water cooled reactors under severe power transients and high burnups. The code is capable of analyzing fuel pins of both collapsible clad, as in PHWR and free standing clad as in LWR. The main emphasis in the development of this code is on evaluating the fuel performance at extended burnups and modelling of the fuel rods for advanced fuel cycles. For this purpose, a number of suitable models have been incorporated in FAIR. For modelling the fission gas release, three different models are implemented, namely Physically based mechanistic model, the standard ANS 5.4 model and the Halden model. Similarly the pellet thermal conductivity can be modelled by the MATPRO equation, the SIMFUEL relation or the Halden equation. The flux distribution across the pellet is modelled by using the model RADAR. For modelling pellet clad interaction (PCMI)/ stress corrosion cracking (SCC) induced failure of sheath, necessary routines are provided in FAIR. The validation of the code FAIR is based on the analysis of fuel rods of EPRI project ''Light water reactor fuel rod modelling code evaluation'' and also the analytical simulation of threshold power ramp criteria of fuel rods of pressurized heavy water reactors. In the present work, a study is carried out by analysing three CRP-FUMEX rods to show the effect of various combinations of fission gas release models and pellet conductivity models, on the fuel analysis parameters. The satisfactory performance of FAIR may be concluded through these case studies. (author). 12 refs, 5 figs
International Nuclear Information System (INIS)
2004-03-01
A working party was organized for two years (2001-2002) on common reactor physics code systems under the Research Committee on Reactor Physics of JAERI. This final report is compilation of activity of the working party on common reactor physics code systems during two years. Objectives of the working party is to clarify basic concept of common reactor physics code systems to improve convenience of reactor physics code systems for reactor physics researchers in Japan on their various field of research and development activities. We have held four meetings during 2 years, investigated status of reactor physics code systems and innovative software technologies, and discussed basic concept of common reactor physics code systems. (author)
Pump Component Model in SPACE Code
International Nuclear Information System (INIS)
Kim, Byoung Jae; Kim, Kyoung Doo
2010-08-01
This technical report describes the pump component model in SPACE code. A literature survey was made on pump models in existing system codes. The models embedded in SPACE code were examined to check the confliction with intellectual proprietary rights. Design specifications, computer coding implementation, and test results are included in this report
The Vulnerability Assessment Code for Physical Protection System
International Nuclear Information System (INIS)
Jang, Sung Soon; Yoo, Ho Sik
2007-01-01
To neutralize the increasing terror threats, nuclear facilities have strong physical protection system (PPS). PPS includes detectors, door locks, fences, regular guard patrols, and a hot line to a nearest military force. To design an efficient PPS and to fully operate it, vulnerability assessment process is required. Evaluating PPS of a nuclear facility is complicate process and, hence, several assessment codes have been developed. The estimation of adversary sequence interruption (EASI) code analyzes vulnerability along a single intrusion path. To evaluate many paths to a valuable asset in an actual facility, the systematic analysis of vulnerability to intrusion (SAVI) code was developed. KAERI improved SAVI and made the Korean analysis of vulnerability to intrusion (KAVI) code. Existing codes (SAVI and KAVI) have limitations in representing the distance of a facility because they use the simplified model of a PPS called adversary sequence diagram. In adversary sequence diagram the position of doors, sensors and fences is described just as the locating area. Thus, the distance between elements is inaccurate and we cannot reflect the range effect of sensors. In this abstract, we suggest accurate and intuitive vulnerability assessment based on raster map modeling of PPS. The raster map of PPS accurately represents the relative position of elements and, thus, the range effect of sensor can be easily incorporable. Most importantly, the raster map is easy to understand
Enhanced Verification Test Suite for Physics Simulation Codes
Energy Technology Data Exchange (ETDEWEB)
Kamm, J R; Brock, J S; Brandon, S T; Cotrell, D L; Johnson, B; Knupp, P; Rider, W; Trucano, T; Weirs, V G
2008-10-10
This document discusses problems with which to augment, in quantity and in quality, the existing tri-laboratory suite of verification problems used by Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and Sandia National Laboratories (SNL). The purpose of verification analysis is demonstrate whether the numerical results of the discretization algorithms in physics and engineering simulation codes provide correct solutions of the corresponding continuum equations. The key points of this document are: (1) Verification deals with mathematical correctness of the numerical algorithms in a code, while validation deals with physical correctness of a simulation in a regime of interest. This document is about verification. (2) The current seven-problem Tri-Laboratory Verification Test Suite, which has been used for approximately five years at the DOE WP laboratories, is limited. (3) Both the methodology for and technology used in verification analysis have evolved and been improved since the original test suite was proposed. (4) The proposed test problems are in three basic areas: (a) Hydrodynamics; (b) Transport processes; and (c) Dynamic strength-of-materials. (5) For several of the proposed problems we provide a 'strong sense verification benchmark', consisting of (i) a clear mathematical statement of the problem with sufficient information to run a computer simulation, (ii) an explanation of how the code result and benchmark solution are to be evaluated, and (iii) a description of the acceptance criterion for simulation code results. (6) It is proposed that the set of verification test problems with which any particular code be evaluated include some of the problems described in this document. Analysis of the proposed verification test problems constitutes part of a necessary--but not sufficient--step that builds confidence in physics and engineering simulation codes. More complicated test cases, including physics models of
Development of multi-physics code systems based on the reactor dynamics code DYN3D
Energy Technology Data Exchange (ETDEWEB)
Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Schuetze, Jochen [ANSYS Germany GmbH, Darmstadt (Germany); Frank, Thomas [ANSYS Germany GmbH, Otterfing (Germany); Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)
2011-07-15
The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)
Development of multi-physics code systems based on the reactor dynamics code DYN3D
International Nuclear Information System (INIS)
Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich; Schuetze, Jochen; Frank, Thomas; Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo
2011-01-01
The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)
User's manual for a process model code
International Nuclear Information System (INIS)
Kern, E.A.; Martinez, D.P.
1981-03-01
The MODEL code has been developed for computer modeling of materials processing facilities associated with the nuclear fuel cycle. However, it can also be used in other modeling applications. This report provides sufficient information for a potential user to apply the code to specific process modeling problems. Several examples that demonstrate most of the capabilities of the code are provided
Development of codes for physical calculations of WWER
International Nuclear Information System (INIS)
Novikov, A.N.
2000-01-01
A package of codes for physical calculations of WWER reactors, used at the RRC 'Kurchatov Institute' is discussed including the purpose of these codes, approximations used, degree of data verification, possibilities of automation of calculations and presentation of results, trends of further development of the codes. (Authors)
Hydrological model in STEALTH 2-D code
International Nuclear Information System (INIS)
Hart, R.; Hofmann, R.
1979-10-01
Porous media fluid flow logic has been added to the two-dimensional version of the STEALTH explicit finite-difference code. It is a first-order hydrological model based upon Darcy's Law. Anisotropic permeability can be prescribed through x and y directional permeabilities. The fluid flow equations are formulated for either two-dimensional translation symmetry or two-dimensional axial symmetry. The addition of the hydrological model to STEALTH is a first step toward analyzing a physical system's response to the coupling of thermal, mechanical, and fluid flow phenomena
The HELIOS-2 lattice physics code
International Nuclear Information System (INIS)
Wemple, C.A.; Gheorghiu, H-N.M.; Stamm'ler, R.J.J.; Villarino, E.A.
2008-01-01
Major advances have been made in the HELIOS code, resulting in the impending release of a new version, HELIOS-2. The new code includes a method of characteristics (MOC) transport solver to supplement the existing collision probabilities (CP) solver. A 177-group, ENDF/B-VII nuclear data library has been developed for inclusion with the new code package. Computational tests have been performed to verify the performance of the MOC solver against the CP solver, and validation testing against computational and measured benchmarks is underway. Results to-date of the verification and validation testing are presented, demonstrating the excellent performance of the new transport solver and nuclear data library. (Author)
Impacts of Model Building Energy Codes
Energy Technology Data Exchange (ETDEWEB)
Athalye, Rahul A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sivaraman, Deepak [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Elliott, Douglas B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Liu, Bing [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bartlett, Rosemarie [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
2016-10-31
The U.S. Department of Energy (DOE) Building Energy Codes Program (BECP) periodically evaluates national and state-level impacts associated with energy codes in residential and commercial buildings. Pacific Northwest National Laboratory (PNNL), funded by DOE, conducted an assessment of the prospective impacts of national model building energy codes from 2010 through 2040. A previous PNNL study evaluated the impact of the Building Energy Codes Program; this study looked more broadly at overall code impacts. This report describes the methodology used for the assessment and presents the impacts in terms of energy savings, consumer cost savings, and reduced CO_{2} emissions at the state level and at aggregated levels. This analysis does not represent all potential savings from energy codes in the U.S. because it excludes several states which have codes which are fundamentally different from the national model energy codes or which do not have state-wide codes. Energy codes follow a three-phase cycle that starts with the development of a new model code, proceeds with the adoption of the new code by states and local jurisdictions, and finishes when buildings comply with the code. The development of new model code editions creates the potential for increased energy savings. After a new model code is adopted, potential savings are realized in the field when new buildings (or additions and alterations) are constructed to comply with the new code. Delayed adoption of a model code and incomplete compliance with the code’s requirements erode potential savings. The contributions of all three phases are crucial to the overall impact of codes, and are considered in this assessment.
Preliminary Coupling of MATRA Code for Multi-physics Analysis
International Nuclear Information System (INIS)
Kim, Seongjin; Choi, Jinyoung; Yang, Yongsik; Kwon, Hyouk; Hwang, Daehyun
2014-01-01
The boundary conditions such as the inlet temperature, mass flux, averaged heat flux, power distributions of the rods, and core geometry is given by constant values or functions of time. These conditions are separately calculated and provided by other codes, such as a neutronics or a system codes, into the MATRA code. In addition, the coupling of several codes in the different physics field is focused and embodied. In this study, multiphysics coupling methods were developed for a subchannel code (MATRA) with neutronics codes (MASTER, DeCART) and a fuel performance code (FRAPCON-3). Preliminary evaluation results for representative sample cases are presented. The MASTER and DeCART codes provide the power distribution of the rods in the core to the MATRA code. In case of the FRAPCON-3 code, the variation of the rod diameter induced by the thermal expansion is yielded and provided. The MATRA code transfers the thermal-hydraulic conditions that each code needs. Moreover, the coupling method with each code is described
Successful vectorization - reactor physics Monte Carlo code
International Nuclear Information System (INIS)
Martin, W.R.
1989-01-01
Most particle transport Monte Carlo codes in use today are based on the ''history-based'' algorithm, wherein one particle history at a time is simulated. Unfortunately, the ''history-based'' approach (present in all Monte Carlo codes until recent years) is inherently scalar and cannot be vectorized. In particular, the history-based algorithm cannot take advantage of vector architectures, which characterize the largest and fastest computers at the current time, vector supercomputers such as the Cray X/MP or IBM 3090/600. However, substantial progress has been made in recent years in developing and implementing a vectorized Monte Carlo algorithm. This algorithm follows portions of many particle histories at the same time and forms the basis for all successful vectorized Monte Carlo codes that are in use today. This paper describes the basic vectorized algorithm along with descriptions of several variations that have been developed by different researchers for specific applications. These applications have been mainly in the areas of neutron transport in nuclear reactor and shielding analysis and photon transport in fusion plasmas. The relative merits of the various approach schemes will be discussed and the present status of known vectorization efforts will be summarized along with available timing results, including results from the successful vectorization of 3-D general geometry, continuous energy Monte Carlo. (orig.)
Fatigue modelling according to the JCSS Probabilistic model code
Vrouwenvelder, A.C.W.M.
2007-01-01
The Joint Committee on Structural Safety is working on a Model Code for full probabilistic design. The code consists out of three major parts: Basis of design, Load Models and Models for Material and Structural Properties. The code is intended as the operational counter part of codes like ISO,
LMFBR models for the ORIGEN2 computer code
International Nuclear Information System (INIS)
Croff, A.G.; McAdoo, J.W.; Bjerke, M.A.
1981-10-01
Reactor physics calculations have led to the development of nine liquid-metal fast breeder reactor (LMFBR) models for the ORIGEN2 computer code. Four of the models are based on the U-Pu fuel cycle, two are based on the Th-U-Pu fuel cycle, and three are based on the Th- 238 U fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST are given
The GNASH preequilibrium-statistical nuclear model code
International Nuclear Information System (INIS)
Arthur, E. D.
1988-01-01
The following report is based on materials presented in a series of lectures at the International Center for Theoretical Physics, Trieste, which were designed to describe the GNASH preequilibrium statistical model code and its use. An overview is provided of the code with emphasis upon code's calculational capabilities and the theoretical models that have been implemented in it. Two sample problems are discussed, the first dealing with neutron reactions on 58 Ni. the second illustrates the fission model capabilities implemented in the code and involves n + 235 U reactions. Finally a description is provided of current theoretical model and code development underway. Examples of calculated results using these new capabilities are also given. 19 refs., 17 figs., 3 tabs
Coding with partially hidden Markov models
DEFF Research Database (Denmark)
Forchhammer, Søren; Rissanen, J.
1995-01-01
Partially hidden Markov models (PHMM) are introduced. They are a variation of the hidden Markov models (HMM) combining the power of explicit conditioning on past observations and the power of using hidden states. (P)HMM may be combined with arithmetic coding for lossless data compression. A general...... 2-part coding scheme for given model order but unknown parameters based on PHMM is presented. A forward-backward reestimation of parameters with a redefined backward variable is given for these models and used for estimating the unknown parameters. Proof of convergence of this reestimation is given....... The PHMM structure and the conditions of the convergence proof allows for application of the PHMM to image coding. Relations between the PHMM and hidden Markov models (HMM) are treated. Results of coding bi-level images with the PHMM coding scheme is given. The results indicate that the PHMM can adapt...
The Los Alamos suite of relativistic atomic physics codes
International Nuclear Information System (INIS)
Fontes, C J; Zhang, H L; Jr, J Abdallah; Clark, R E H; Kilcrease, D P; Colgan, J; Cunningham, R T; Hakel, P; Magee, N H; Sherrill, M E
2015-01-01
The Los Alamos suite of relativistic atomic physics codes is a robust, mature platform that has been used to model highly charged ions in a variety of ways. The suite includes capabilities for calculating data related to fundamental atomic structure, as well as the processes of photoexcitation, electron-impact excitation and ionization, photoionization and autoionization within a consistent framework. These data can be of a basic nature, such as cross sections and collision strengths, which are useful in making predictions that can be compared with experiments to test fundamental theories of highly charged ions, such as quantum electrodynamics. The suite can also be used to generate detailed models of energy levels and rate coefficients, and to apply them in the collisional-radiative modeling of plasmas over a wide range of conditions. Such modeling is useful, for example, in the interpretation of spectra generated by a variety of plasmas. In this work, we provide a brief overview of the capabilities within the Los Alamos relativistic suite along with some examples of its application to the modeling of highly charged ions. (paper)
Resonance interference method in lattice physics code stream
International Nuclear Information System (INIS)
Choi, Sooyoung; Khassenov, Azamat; Lee, Deokjung
2015-01-01
Newly developed resonance interference model is implemented in the lattice physics code STREAM, and the model shows a significant improvement in computing accurate eigenvalues. Equivalence theory is widely used in production calculations to generate the effective multigroup (MG) cross-sections (XS) for commercial reactors. Although a lot of methods have been developed to enhance the accuracy in computing effective XSs, the current resonance treatment methods still do not have a clear resonance interference model. The conventional resonance interference model simply adds the absorption XSs of resonance isotopes to the background XS. However, the conventional models show non-negligible errors in computing effective XSs and eigenvalues. In this paper, a resonance interference factor (RIF) library method is proposed. This method interpolates the RIFs in a pre-generated RIF library and corrects the effective XS, rather than solving the time consuming slowing down calculation. The RIF library method is verified for homogeneous and heterogeneous problems. The verification results using the proposed method show significant improvements of accuracy in treating the interference effect. (author)
Utility subroutine package used by Applied Physics Division export codes
International Nuclear Information System (INIS)
Adams, C.H.; Derstine, K.L.; Henryson, H. II; Hosteny, R.P.; Toppel, B.J.
1983-04-01
This report describes the current state of the utility subroutine package used with codes being developed by the staff of the Applied Physics Division. The package provides a variety of useful functions for BCD input processing, dynamic core-storage allocation and managemnt, binary I/0 and data manipulation. The routines were written to conform to coding standards which facilitate the exchange of programs between different computers
High-fidelity plasma codes for burn physics
Energy Technology Data Exchange (ETDEWEB)
Cooley, James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Graziani, Frank [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marinak, Marty [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Murillo, Michael [Michigan State Univ., East Lansing, MI (United States)
2016-10-19
Accurate predictions of equation of state (EOS), ionic and electronic transport properties are of critical importance for high-energy-density plasma science. Transport coefficients inform radiation-hydrodynamic codes and impact diagnostic interpretation, which in turn impacts our understanding of the development of instabilities, the overall energy balance of burning plasmas, and the efficacy of self-heating from charged-particle stopping. Important processes include thermal and electrical conduction, electron-ion coupling, inter-diffusion, ion viscosity, and charged particle stopping. However, uncertainties in these coefficients are not well established. Fundamental plasma science codes, also called high-fidelity plasma codes, are a relatively recent computational tool that augments both experimental data and theoretical foundations of transport coefficients. This paper addresses the current status of HFPC codes and their future development, and the potential impact they play in improving the predictive capability of the multi-physics hydrodynamic codes used in HED design.
Genetic coding and gene expression - new Quadruplet genetic coding model
Shankar Singh, Rama
2012-07-01
Successful demonstration of human genome project has opened the door not only for developing personalized medicine and cure for genetic diseases, but it may also answer the complex and difficult question of the origin of life. It may lead to making 21st century, a century of Biological Sciences as well. Based on the central dogma of Biology, genetic codons in conjunction with tRNA play a key role in translating the RNA bases forming sequence of amino acids leading to a synthesized protein. This is the most critical step in synthesizing the right protein needed for personalized medicine and curing genetic diseases. So far, only triplet codons involving three bases of RNA, transcribed from DNA bases, have been used. Since this approach has several inconsistencies and limitations, even the promise of personalized medicine has not been realized. The new Quadruplet genetic coding model proposed and developed here involves all four RNA bases which in conjunction with tRNA will synthesize the right protein. The transcription and translation process used will be the same, but the Quadruplet codons will help overcome most of the inconsistencies and limitations of the triplet codes. Details of this new Quadruplet genetic coding model and its subsequent potential applications including relevance to the origin of life will be presented.
The DIT nuclear fuel assembly physics design code
International Nuclear Information System (INIS)
Jonsson, A.
1988-01-01
The DIT code is the Combustion Engineering, Inc. (C-E) nuclear fuel assembly design code. It belongs to a class of codes, all similar in structure and strategy, that may be characterized by the spectrum and spatial calculations being performed in two dimensions and in a single job step for the entire assembly. The forerunner of this class of codes is the United Kingdom Atomic Energy Authority WIMS code, the first version of which was completed 25 yr ago. The structure and strategy of assembly spectrum codes have remained remarkably similar to the original concept thus proving its usefulness. As other organizations, including C-E, have developed their own versions of the concept, many important variations have been added that significantly influence the accuracy and performance of the resulting computational tool. Those features, which are unique to the DIT code and which might be of interest to the community of fuel assembly physics design code users and developers, are described and discussed
Transmutation Fuel Performance Code Thermal Model Verification
Energy Technology Data Exchange (ETDEWEB)
Gregory K. Miller; Pavel G. Medvedev
2007-09-01
FRAPCON fuel performance code is being modified to be able to model performance of the nuclear fuels of interest to the Global Nuclear Energy Partnership (GNEP). The present report documents the effort for verification of the FRAPCON thermal model. It was found that, with minor modifications, FRAPCON thermal model temperature calculation agrees with that of the commercial software ABAQUS (Version 6.4-4). This report outlines the methodology of the verification, code input, and calculation results.
Improvement of MARS code reflood model
International Nuclear Information System (INIS)
Hwang, Moonkyu; Chung, Bub-Dong
2011-01-01
A specifically designed heat transfer model for the reflood process which normally occurs at low flow and low pressure was originally incorporated in the MARS code. The model is essentially identical to that of the RELAP5/MOD3.3 code. The model, however, is known to have under-estimated the peak cladding temperature (PCT) with earlier turn-over. In this study, the original MARS code reflood model is improved. Based on the extensive sensitivity studies for both hydraulic and wall heat transfer models, it is found that the dispersed flow film boiling (DFFB) wall heat transfer is the most influential process determining the PCT, whereas the interfacial drag model most affects the quenching time through the liquid carryover phenomenon. The model proposed by Bajorek and Young is incorporated for the DFFB wall heat transfer. Both space grid and droplet enhancement models are incorporated. Inverted annular film boiling (IAFB) is modeled by using the original PSI model of the code. The flow transition between the DFFB and IABF, is modeled using the TRACE code interpolation. A gas velocity threshold is also added to limit the top-down quenching effect. Assessment calculations are performed for the original and modified MARS codes for the Flecht-Seaset test and RBHT test. Improvements are observed in terms of the PCT and quenching time predictions in the Flecht-Seaset assessment. In case of the RBHT assessment, the improvement over the original MARS code is found marginal. A space grid effect, however, is clearly seen from the modified version of the MARS code. (author)
Applying Physical-Layer Network Coding in Wireless Networks
Directory of Open Access Journals (Sweden)
Liew SoungChang
2010-01-01
Full Text Available A main distinguishing feature of a wireless network compared with a wired network is its broadcast nature, in which the signal transmitted by a node may reach several other nodes, and a node may receive signals from several other nodes, simultaneously. Rather than a blessing, this feature is treated more as an interference-inducing nuisance in most wireless networks today (e.g., IEEE 802.11. This paper shows that the concept of network coding can be applied at the physical layer to turn the broadcast property into a capacity-boosting advantage in wireless ad hoc networks. Specifically, we propose a physical-layer network coding (PNC scheme to coordinate transmissions among nodes. In contrast to "straightforward" network coding which performs coding arithmetic on digital bit streams after they have been received, PNC makes use of the additive nature of simultaneously arriving electromagnetic (EM waves for equivalent coding operation. And in doing so, PNC can potentially achieve 100% and 50% throughput increases compared with traditional transmission and straightforward network coding, respectively, in 1D regular linear networks with multiple random flows. The throughput improvements are even larger in 2D regular networks: 200% and 100%, respectively.
International Nuclear Information System (INIS)
Cheney, J.A.
1981-01-01
The problems of statisfying similarity between a physical model and the prototype in rock wherein fissures and cracks place a role in physical behavior is explored. The need for models of large physical dimensions is explained but also testing of models of the same prototype over a wide range of scales is needed to ascertain the influence of lack of similitude of particular parameters between prototype and model. A large capacity centrifuge would be useful in that respect
Opacity calculations for extreme physical systems: code RACHEL
Drska, Ladislav; Sinor, Milan
1996-08-01
Computer simulations of physical systems under extreme conditions (high density, temperature, etc.) require the availability of extensive sets of atomic data. This paper presents basic information on a self-consistent approach to calculations of radiative opacity, one of the key characteristics of such systems. After a short explanation of general concepts of the atomic physics of extreme systems, the structure of the opacity code RACHEL is discussed and some of its applications are presented.
Modular ORIGEN-S for multi-physics code systems
International Nuclear Information System (INIS)
Yesilyurt, Gokhan; Clarno, Kevin T.; Gauld, Ian C.; Galloway, Jack
2011-01-01
The ORIGEN-S code in the SCALE 6.0 nuclear analysis code suite is a well-validated tool to calculate the time-dependent concentrations of nuclides due to isotopic depletion, decay, and transmutation for many systems in a wide range of time scales. Application areas include nuclear reactor and spent fuel storage analyses, burnup credit evaluations, decay heat calculations, and environmental assessments. Although simple to use within the SCALE 6.0 code system, especially with the ORIGEN-ARP graphical user interface, it is generally complex to use as a component within an externally developed code suite because of its tight coupling within the infrastructure of the larger SCALE 6.0 system. The ORIGEN2 code, which has been widely integrated within other simulation suites, is no longer maintained by Oak Ridge National Laboratory (ORNL), has obsolete data, and has a relatively small validation database. Therefore, a modular version of the SCALE/ORIGEN-S code was developed to simplify its integration with other software packages to allow multi-physics nuclear code systems to easily incorporate the well-validated isotopic depletion, decay, and transmutation capability to perform realistic nuclear reactor and fuel simulations. SCALE/ORIGEN-S was extensively restructured to develop a modular version that allows direct access to the matrix solvers embedded in the code. Problem initialization and the solver were segregated to provide a simple application program interface and fewer input/output operations for the multi-physics nuclear code systems. Furthermore, new interfaces were implemented to access and modify the ORIGEN-S input variables and nuclear cross-section data through external drivers. Three example drivers were implemented, in the C, C++, and Fortran 90 programming languages, to demonstrate the modular use of the new capability. This modular version of SCALE/ORIGEN-S has been embedded within several multi-physics software development projects at ORNL, including
Modular ORIGEN-S for multi-physics code systems
Energy Technology Data Exchange (ETDEWEB)
Yesilyurt, Gokhan; Clarno, Kevin T.; Gauld, Ian C., E-mail: yesilyurtg@ornl.gov, E-mail: clarnokt@ornl.gov, E-mail: gauldi@ornl.gov [Oak Ridge National Laboratory, TN (United States); Galloway, Jack, E-mail: jack@galloways.net [Los Alamos National Laboratory, Los Alamos, NM (United States)
2011-07-01
The ORIGEN-S code in the SCALE 6.0 nuclear analysis code suite is a well-validated tool to calculate the time-dependent concentrations of nuclides due to isotopic depletion, decay, and transmutation for many systems in a wide range of time scales. Application areas include nuclear reactor and spent fuel storage analyses, burnup credit evaluations, decay heat calculations, and environmental assessments. Although simple to use within the SCALE 6.0 code system, especially with the ORIGEN-ARP graphical user interface, it is generally complex to use as a component within an externally developed code suite because of its tight coupling within the infrastructure of the larger SCALE 6.0 system. The ORIGEN2 code, which has been widely integrated within other simulation suites, is no longer maintained by Oak Ridge National Laboratory (ORNL), has obsolete data, and has a relatively small validation database. Therefore, a modular version of the SCALE/ORIGEN-S code was developed to simplify its integration with other software packages to allow multi-physics nuclear code systems to easily incorporate the well-validated isotopic depletion, decay, and transmutation capability to perform realistic nuclear reactor and fuel simulations. SCALE/ORIGEN-S was extensively restructured to develop a modular version that allows direct access to the matrix solvers embedded in the code. Problem initialization and the solver were segregated to provide a simple application program interface and fewer input/output operations for the multi-physics nuclear code systems. Furthermore, new interfaces were implemented to access and modify the ORIGEN-S input variables and nuclear cross-section data through external drivers. Three example drivers were implemented, in the C, C++, and Fortran 90 programming languages, to demonstrate the modular use of the new capability. This modular version of SCALE/ORIGEN-S has been embedded within several multi-physics software development projects at ORNL, including
The Dit nuclear fuel assembly physics design code
International Nuclear Information System (INIS)
Jonsson, A.
1987-01-01
DIT is the Combustion Engineering, Inc. (C-E) nuclear fuel assembly design code. It belongs to a class of codes, all similar in structure and strategy, which may be characterized by the spectrum and spatial calculations being performed in 2D and in a single job step for the entire assembly. The forerunner of this class of codes is the U.K.A.E.A. WIMS code, the first version of which was completed 25 years ago. The structure and strategy of assembly spectrum codes have remained remarkably similar to the original concept thus proving its usefulness. As other organizations, including C-E, have developed their own versions of the concept, many important variations have been added which significantly influence the accuracy and performance of the resulting computational tool. This paper describes and discusses those features which are unique to the DIT code and which might be of interest to the community of fuel assembly physics design code users and developers
Physical-layer network coding in coherent optical OFDM systems.
Guan, Xun; Chan, Chun-Kit
2015-04-20
We present the first experimental demonstration and characterization of the application of optical physical-layer network coding in coherent optical OFDM systems. It combines two optical OFDM frames to share the same link so as to enhance system throughput, while individual OFDM frames can be recovered with digital signal processing at the destined node.
International Nuclear Information System (INIS)
Bartlett, D.V.
1983-06-01
The codes which have been developed for the analysis of electron cyclotron emission measurements in JET are described. Their principal function is to interpret the spectra measured by the diagnostic so as to give the spatial distribution of the electron temperature in the poloidal cross-section. Various systematic effects in the data are corrected using look-up tables generated by an elaborate simulation code. The part of this code responsible for the accurate calculation of single-pass emission and refraction has been written at CNR-Milan and is described in a separate report. The present report is divided into two parts. This first part describes the methods used for the simulation and interpretation of spectra, the physical/mathematical basis of the codes written at CEA-Fontenay and presents some illustrative results
Validation of the VTT's reactor physics code system
International Nuclear Information System (INIS)
Tanskanen, A.
1998-01-01
At VTT Energy several international reactor physics codes and nuclear data libraries are used in a variety of applications. The codes and libraries are under constant development and every now and then new updated versions are released, which are taken in use as soon as they have been validated at VTT Energy. The primary aim of the validation is to ensure that the code works properly, and that it can be used correctly. Moreover, the applicability of the codes and libraries are studied in order to establish their advantages and weak points. The capability of generating program-specific nuclear data for different reactor physics codes starting from the same evaluated data is sometimes of great benefit. VTT Energy has acquired a nuclear data processing system based on the NJOY-94.105 and TRANSX-2.15 processing codes. The validity of the processing system has been demonstrated by generating pointwise (MCNP) and groupwise (ANISN) temperature-dependent cross section sets for the benchmark calculations of the Doppler coefficient of reactivity. At VTT Energy the KENO-VI three-dimensional Monte Carlo code is used in criticality safety analyses. The KENO-VI code and the 44GROUPNDF5 data library have been validated at VTT Energy against the ZR-6 and LR-0 critical experiments. Burnup Credit refers to the reduction in reactivity of burned nuclear fuel due to the change in composition during irradiation. VTT Energy has participated in the calculational VVER-440 burnup credit benchmark in order to validate criticality safety calculation tools. (orig.)
WDEC: A Code for Modeling White Dwarf Structure and Pulsations
Bischoff-Kim, Agnès; Montgomery, Michael H.
2018-05-01
The White Dwarf Evolution Code (WDEC), written in Fortran, makes models of white dwarf stars. It is fast, versatile, and includes the latest physics. The code evolves hot (∼100,000 K) input models down to a chosen effective temperature by relaxing the models to be solutions of the equations of stellar structure. The code can also be used to obtain g-mode oscillation modes for the models. WDEC has a long history going back to the late 1960s. Over the years, it has been updated and re-packaged for modern computer architectures and has specifically been used in computationally intensive asteroseismic fitting. Generations of white dwarf astronomers and dozens of publications have made use of the WDEC, although the last true instrument paper is the original one, published in 1975. This paper discusses the history of the code, necessary to understand why it works the way it does, details the physics and features in the code today, and points the reader to where to find the code and a user guide.
Applications of the ARGUS code in accelerator physics
International Nuclear Information System (INIS)
Petillo, J.J.; Mankofsky, A.; Krueger, W.A.; Kostas, C.; Mondelli, A.A.; Drobot, A.T.
1993-01-01
ARGUS is a three-dimensional, electromagnetic, particle-in-cell (PIC) simulation code that is being distributed to U.S. accelerator laboratories in collaboration between SAIC and the Los Alamos Accelerator Code Group. It uses a modular architecture that allows multiple physics modules to share common utilities for grid and structure input., memory management, disk I/O, and diagnostics, Physics modules are in place for electrostatic and electromagnetic field solutions., frequency-domain (eigenvalue) solutions, time- dependent PIC, and steady-state PIC simulations. All of the modules are implemented with a domain-decomposition architecture that allows large problems to be broken up into pieces that fit in core and that facilitates the adaptation of ARGUS for parallel processing ARGUS operates on either Cray or workstation platforms, and MOTIF-based user interface is available for X-windows terminals. Applications of ARGUS in accelerator physics and design are described in this paper
ALE3D: An Arbitrary Lagrangian-Eulerian Multi-Physics Code
Energy Technology Data Exchange (ETDEWEB)
Noble, Charles R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anderson, Andrew T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Barton, Nathan R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bramwell, Jamie A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Capps, Arlie [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chang, Michael H. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chou, Jin J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dawson, David M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Diana, Emily R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dunn, Timothy A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Faux, Douglas R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fisher, Aaron C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Greene, Patrick T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Heinz, Ines [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kanarska, Yuliya [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Khairallah, Saad A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Liu, Benjamin T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Margraf, Jon D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Nichols, Albert L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Nourgaliev, Robert N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Puso, Michael A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reus, James F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Robinson, Peter B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Shestakov, Alek I. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Solberg, Jerome M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Taller, Daniel [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Tsuji, Paul H. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); White, Christopher A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); White, Jeremy L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
2017-05-23
ALE3D is a multi-physics numerical simulation software tool utilizing arbitrary-Lagrangian- Eulerian (ALE) techniques. The code is written to address both two-dimensional (2D plane and axisymmetric) and three-dimensional (3D) physics and engineering problems using a hybrid finite element and finite volume formulation to model fluid and elastic-plastic response of materials on an unstructured grid. As shown in Figure 1, ALE3D is a single code that integrates many physical phenomena.
Development of M3C code for Monte Carlo reactor physics criticality calculations
International Nuclear Information System (INIS)
Kumar, Anek; Kannan, Umasankari; Krishanani, P.D.
2015-06-01
The development of Monte Carlo code (M3C) for reactor design entails use of continuous energy nuclear data and Monte Carlo simulations for each of the neutron interaction processes. BARC has started a concentrated effort for developing a new general geometry continuous energy Monte Carlo code for reactor physics calculation indigenously. The code development required a comprehensive understanding of the basic continuous energy cross section sets. The important features of this code are treatment of heterogeneous lattices by general geometry, use of point cross sections along with unionized energy grid approach, thermal scattering model for low energy treatment, capability of handling the microscopic fuel particles dispersed randomly. The capability of handling the randomly dispersed microscopic fuel particles which is very useful for the modeling of High-Temperature Gas-Cooled reactor fuels which are composed of thousands of microscopic fuel particle (TRISO fuel particle), randomly dispersed in a graphite matrix. The Monte Carlo code for criticality calculation is a pioneering effort and has been used to study several types of lattices including cluster geometries. The code has been verified for its accuracy against more than 60 sample problems covering a wide range from simple (like spherical) to complex geometry (like PHWR lattice). Benchmark results show that the code performs quite well for the criticality calculation of the system. In this report, the current status of the code, features of the code, some of the benchmark results for the testing of the code and input preparation etc. are discussed. (author)
Economic aspects and models for building codes
DEFF Research Database (Denmark)
Bonke, Jens; Pedersen, Dan Ove; Johnsen, Kjeld
It is the purpose of this bulletin to present an economic model for estimating the consequence of new or changed building codes. The object is to allow comparative analysis in order to improve the basis for decisions in this field. The model is applied in a case study.......It is the purpose of this bulletin to present an economic model for estimating the consequence of new or changed building codes. The object is to allow comparative analysis in order to improve the basis for decisions in this field. The model is applied in a case study....
Modeling report of DYMOND code (DUPIC version)
International Nuclear Information System (INIS)
Park, Joo Hwan; Yacout, Abdellatif M.
2003-04-01
The DYMOND code employs the ITHINK dynamic modeling platform to assess the 100-year dynamic evolution scenarios for postulated global nuclear energy parks. Firstly, DYMOND code has been developed by ANL(Argonne National Laboratory) to perform the fuel cycle analysis of LWR once-through and LWR-FBR mixed plant. Since the extensive application of DYMOND code has been requested, the first version of DYMOND has been modified to adapt the DUPIC, MSR and RTF fuel cycle. DYMOND code is composed of three parts; the source language platform, input supply and output. But those platforms are not clearly distinguished. This report described all the equations which were modeled in the modified DYMOND code (which is called as DYMOND-DUPIC version). It divided into five parts;Part A deals model in reactor history which is included amount of the requested fuels and spent fuels. Part B aims to describe model of fuel cycle about fuel flow from the beginning to the end of fuel cycle. Part C is for model in re-processing which is included recovery of burned uranium, plutonium, minor actinide and fission product as well as the amount of spent fuels in storage and disposal. Part D is for model in other fuel cycle which is considered the thorium fuel cycle for MSR and RTF reactor. Part E is for model in economics. This part gives all the information of cost such as uranium mining cost, reactor operating cost, fuel cost etc
Modeling report of DYMOND code (DUPIC version)
Energy Technology Data Exchange (ETDEWEB)
Park, Joo Hwan [KAERI, Taejon (Korea, Republic of); Yacout, Abdellatif M [Argonne National Laboratory, Ilinois (United States)
2003-04-01
The DYMOND code employs the ITHINK dynamic modeling platform to assess the 100-year dynamic evolution scenarios for postulated global nuclear energy parks. Firstly, DYMOND code has been developed by ANL(Argonne National Laboratory) to perform the fuel cycle analysis of LWR once-through and LWR-FBR mixed plant. Since the extensive application of DYMOND code has been requested, the first version of DYMOND has been modified to adapt the DUPIC, MSR and RTF fuel cycle. DYMOND code is composed of three parts; the source language platform, input supply and output. But those platforms are not clearly distinguished. This report described all the equations which were modeled in the modified DYMOND code (which is called as DYMOND-DUPIC version). It divided into five parts;Part A deals model in reactor history which is included amount of the requested fuels and spent fuels. Part B aims to describe model of fuel cycle about fuel flow from the beginning to the end of fuel cycle. Part C is for model in re-processing which is included recovery of burned uranium, plutonium, minor actinide and fission product as well as the amount of spent fuels in storage and disposal. Part D is for model in other fuel cycle which is considered the thorium fuel cycle for MSR and RTF reactor. Part E is for model in economics. This part gives all the information of cost such as uranium mining cost, reactor operating cost, fuel cost etc.
KAMCCO, a reactor physics Monte Carlo neutron transport code
International Nuclear Information System (INIS)
Arnecke, G.; Borgwaldt, H.; Brandl, V.; Lalovic, M.
1976-06-01
KAMCCO is a 3-dimensional reactor Monte Carlo code for fast neutron physics problems. Two options are available for the solution of 1) the inhomogeneous time-dependent neutron transport equation (census time scheme), and 2) the homogeneous static neutron transport equation (generation cycle scheme). The user defines the desired output, e.g. estimates of reaction rates or neutron flux integrated over specified volumes in phase space and time intervals. Such primary quantities can be arbitrarily combined, also ratios of these quantities can be estimated with their errors. The Monte Carlo techniques are mostly analogue (exceptions: Importance sampling for collision processes, ELP/MELP, Russian roulette and splitting). Estimates are obtained from the collision and track length estimators. Elastic scattering takes into account first order anisotropy in the center of mass system. Inelastic scattering is processed via the evaporation model or via the excitation of discrete levels. For the calculation of cross sections, the energy is treated as a continuous variable. They are computed by a) linear interpolation, b) from optionally Doppler broadened single level Breit-Wigner resonances or c) from probability tables (in the region of statistically distributed resonances). (orig.) [de
Models and applications of the UEDGE code
International Nuclear Information System (INIS)
Rensink, M.E.; Knoll, D.A.; Porter, G.D.; Rognlien, T.D.; Smith, G.R.; Wising, F.
1996-09-01
The transport of particles and energy from the core of a tokamak to nearby material surfaces is an important problem for understanding present experiments and for designing reactor-grade devices. A number of fluid transport codes have been developed to model the plasma in the edge and scrape-off layer (SOL) regions. This report will focus on recent model improvements and illustrative results from the UEDGE code. Some geometric and mesh considerations are introduced, followed by a general description of the plasma and neutral fluid models. A few comments on computational issues are given and then two important applications are illustrated concerning benchmarking and the ITER radiative divertor. Finally, we report on some recent work to improve the models in UEDGE by coupling to a Monte Carlo neutrals code and by utilizing an adaptive grid
The nuclear reaction model code MEDICUS
International Nuclear Information System (INIS)
Ibishia, A.I.
2008-01-01
The new computer code MEDICUS has been used to calculate cross sections of nuclear reactions. The code, implemented in MATLAB 6.5, Mathematica 5, and Fortran 95 programming languages, can be run in graphical and command line mode. Graphical User Interface (GUI) has been built that allows the user to perform calculations and to plot results just by mouse clicking. The MS Windows XP and Red Hat Linux platforms are supported. MEDICUS is a modern nuclear reaction code that can compute charged particle-, photon-, and neutron-induced reactions in the energy range from thresholds to about 200 MeV. The calculation of the cross sections of nuclear reactions are done in the framework of the Exact Many-Body Nuclear Cluster Model (EMBNCM), Direct Nuclear Reactions, Pre-equilibrium Reactions, Optical Model, DWBA, and Exciton Model with Cluster Emission. The code can be used also for the calculation of nuclear cluster structure of nuclei. We have calculated nuclear cluster models for some nuclei such as 177 Lu, 90 Y, and 27 Al. It has been found that nucleus 27 Al can be represented through the two different nuclear cluster models: 25 Mg + d and 24 Na + 3 He. Cross sections in function of energy for the reaction 27 Al( 3 He,x) 22 Na, established as a production method of 22 Na, are calculated by the code MEDICUS. Theoretical calculations of cross sections are in good agreement with experimental results. Reaction mechanisms are taken into account. (author)
Porting plasma physics simulation codes to modern computing architectures using the libmrc framework
Germaschewski, Kai; Abbott, Stephen
2015-11-01
Available computing power has continued to grow exponentially even after single-core performance satured in the last decade. The increase has since been driven by more parallelism, both using more cores and having more parallelism in each core, e.g. in GPUs and Intel Xeon Phi. Adapting existing plasma physics codes is challenging, in particular as there is no single programming model that covers current and future architectures. We will introduce the open-source libmrc framework that has been used to modularize and port three plasma physics codes: The extended MHD code MRCv3 with implicit time integration and curvilinear grids; the OpenGGCM global magnetosphere model; and the particle-in-cell code PSC. libmrc consolidates basic functionality needed for simulations based on structured grids (I/O, load balancing, time integrators), and also introduces a parallel object model that makes it possible to maintain multiple implementations of computational kernels, on e.g. conventional processors and GPUs. It handles data layout conversions and enables us to port performance-critical parts of a code to a new architecture step-by-step, while the rest of the code can remain unchanged. We will show examples of the performance gains and some physics applications.
RCS modeling with the TSAR FDTD code
Energy Technology Data Exchange (ETDEWEB)
Pennock, S.T.; Ray, S.L.
1992-03-01
The TSAR electromagnetic modeling system consists of a family of related codes that have been designed to work together to provide users with a practical way to set up, run, and interpret the results from complex 3-D finite-difference time-domain (FDTD) electromagnetic simulations. The software has been in development at the Lawrence Livermore National Laboratory (LLNL) and at other sites since 1987. Active internal use of the codes began in 1988 with limited external distribution and use beginning in 1991. TSAR was originally developed to analyze high-power microwave and EMP coupling problems. However, the general-purpose nature of the tools has enabled us to use the codes to solve a broader class of electromagnetic applications and has motivated the addition of new features. In particular a family of near-to-far field transformation routines have been added to the codes, enabling TSAR to be used for radar-cross section and antenna analysis problems.
MELCOR code modeling for APR1400
Energy Technology Data Exchange (ETDEWEB)
Choi, Young; Park, S. Y.; Kim, D. H.; Ahn, K. I.; Song, Y. M.; Kim, S. D.; Park, J. H
2001-11-01
The severe accident phenomena of nuclear power plant have large uncertainties. For the retention of the containment integrity and improvement of nuclear reactor safety against severe accident, it is essential to understand severe accident phenomena and be able to access the accident progression accurately using computer code. Furthermore, it is important to attain a capability for developing technique and assessment tools for an advanced nuclear reactor design as well as for the severe accident prevention and mitigation. The objective of this report is to establish technical bases for an application of the MELCOR code to the Korean Next Generation Reactor (APR1400) by modeling the plant and analyzing plant steady state. This report shows the data and the input preparation for MELCOR code as well as state-state assessment results using MELCOR code.
Enhanced verification test suite for physics simulation codes
Energy Technology Data Exchange (ETDEWEB)
Kamm, James R.; Brock, Jerry S.; Brandon, Scott T.; Cotrell, David L.; Johnson, Bryan; Knupp, Patrick; Rider, William J.; Trucano, Timothy G.; Weirs, V. Gregory
2008-09-01
This document discusses problems with which to augment, in quantity and in quality, the existing tri-laboratory suite of verification problems used by Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and Sandia National Laboratories (SNL). The purpose of verification analysis is demonstrate whether the numerical results of the discretization algorithms in physics and engineering simulation codes provide correct solutions of the corresponding continuum equations.
ITER Dynamic Tritium Inventory Modeling Code
International Nuclear Information System (INIS)
Cristescu, Ioana-R.; Doerr, L.; Busigin, A.; Murdoch, D.
2005-01-01
A tool for tritium inventory evaluation within each sub-system of the Fuel Cycle of ITER is vital, with respect to both the process of licensing ITER and also for operation. It is very likely that measurements of total tritium inventories may not be possible for all sub-systems, however tritium accounting may be achieved by modeling its hold-up within each sub-system and by validating these models in real-time against the monitored flows and tritium streams between the systems. To get reliable results, an accurate dynamic modeling of the tritium content in each sub-system is necessary. In order to optimize the configuration and operation of the ITER fuel cycle, a dynamic fuel cycle model was developed progressively in the decade up to 2000-2001. As the design for some sub-systems from the fuel cycle (i.e. Vacuum pumping, Neutral Beam Injectors (NBI)) have substantially progressed meanwhile, a new code developed under a different platform to incorporate these modifications has been developed. The new code is taking over the models and algorithms for some subsystems, such as Isotope Separation System (ISS); where simplified models have been previously considered, more detailed have been introduced, as for the Water Detritiation System (WDS). To reflect all these changes, the new code developed inside EU participating team was nominated TRIMO (Tritium Inventory Modeling), to emphasize the use of the code on assessing the tritium inventory within ITER
Identification of physical models
DEFF Research Database (Denmark)
Melgaard, Henrik
1994-01-01
of the model with the available prior knowledge. The methods for identification of physical models have been applied in two different case studies. One case is the identification of thermal dynamics of building components. The work is related to a CEC research project called PASSYS (Passive Solar Components......The problem of identification of physical models is considered within the frame of stochastic differential equations. Methods for estimation of parameters of these continuous time models based on descrete time measurements are discussed. The important algorithms of a computer program for ML or MAP...... design of experiments, which is for instance the design of an input signal that are optimal according to a criterion based on the information provided by the experiment. Also model validation is discussed. An important verification of a physical model is to compare the physical characteristics...
Automatic modeling for the monte carlo transport TRIPOLI code
International Nuclear Information System (INIS)
Zhang Junjun; Zeng Qin; Wu Yican; Wang Guozhong; FDS Team
2010-01-01
TRIPOLI, developed by CEA, France, is Monte Carlo particle transport simulation code. It has been widely applied to nuclear physics, shielding design, evaluation of nuclear safety. However, it is time-consuming and error-prone to manually describe the TRIPOLI input file. This paper implemented bi-directional conversion between CAD model and TRIPOLI model. Its feasibility and efficiency have been demonstrated by several benchmarking examples. (authors)
Hydrogen recycle modeling in transport codes
International Nuclear Information System (INIS)
Howe, H.C.
1979-01-01
The hydrogen recycling models now used in Tokamak transport codes are reviewed and the method by which realistic recycling models are being added is discussed. Present models use arbitrary recycle coefficients and therefore do not model the actual recycling processes at the wall. A model for the hydrogen concentration in the wall serves two purposes: (1) it allows a better understanding of the density behavior in present gas puff, pellet, and neutral beam heating experiments; and (2) it allows one to extrapolate to long pulse devices such as EBT, ISX-C and reactors where the walls are observed or expected to saturate. Several wall models are presently being studied for inclusion in transport codes
International Nuclear Information System (INIS)
Duplex, B.
2011-01-01
The CEA develops and uses scientific software, called physical codes, in various physical disciplines to optimize installation and experimentation costs. During a study, several physical phenomena interact, so a code coupling and some data exchanges between different physical codes are required. Each physical code computes on a particular geometry, usually represented by a mesh composed of thousands to millions of elements. This PhD Thesis focuses on the geometrical modification transfer between specific meshes of each coupled physical code. First, it presents a physical code coupling method where deformations are computed by one of these codes. Next, it discusses the establishment of a model, common to different physical codes, grouping all the shared data. Finally, it covers the deformation transfers between meshes of the same geometry or adjacent geometries. Geometrical modifications are discrete data because they are based on a mesh. In order to permit every code to access deformations and to transfer them, a continuous representation is computed. Two functions are developed, one with a global support, and the other with a local support. Both functions combine a simplification method and a radial basis function network. A whole use case is dedicated to the Jules Horowitz reactor. The effect of differential dilatations on experimental device cooling is studied. (author) [fr
Upgrade and benchmarking of the NIFS physics-engineering-cost code
International Nuclear Information System (INIS)
Dolan, T.J.; Yamazaki, K.
2004-07-01
The NIFS Physics-Engineering-Cost (PEC) code for helical and tokamak fusion reactors is upgraded by adding data from three blanket-shield designs, a new cost section based on the ARIES cost schedule, more recent unit costs, and improved algorithms for various computations. The PEC code is also benchmarked by modeling the ARIES-AT (advanced technology) tokamak and the ARIES-SPPS (stellarator power plant system). The PEC code succeeds in predicting many of the pertinent plasma parameters and reactor component masses within about 10%. There are cost differences greater than 10% for some fusion power core components, which may be attributed to differences of unit costs used by the codes. The COEs estimated by the PEC code differ from the COEs of the ARIES-AT and ARIES-SPPS studies by 5%. (author)
Health physics source document for codes of practice
International Nuclear Information System (INIS)
Pearson, G.W.; Meggitt, G.C.
1989-05-01
Personnel preparing codes of practice often require basic Health Physics information or advice relating to radiological protection problems and this document is written primarily to supply such information. Certain technical terms used in the text are explained in the extensive glossary. Due to the pace of change in the field of radiological protection it is difficult to produce an up-to-date document. This document was compiled during 1988 however, and therefore contains the principle changes brought about by the introduction of the Ionising Radiations Regulations (1985). The paper covers the nature of ionising radiation, its biological effects and the principles of control. It is hoped that the document will provide a useful source of information for both codes of practice and wider areas and stimulate readers to study radiological protection issues in greater depth. (author)
A primer on physical-layer network coding
Liew, Soung Chang; Zhang, Shengli
2015-01-01
The concept of physical-layer network coding (PNC) was proposed in 2006 for application in wireless networks. Since then it has developed into a subfield of communications and networking with a wide following. This book is a primer on PNC. It is the outcome of a set of lecture notes for a course for beginning graduate students at The Chinese University of Hong Kong. The target audience is expected to have some prior background knowledge in communication theory and wireless communications, but not working knowledge at the research level. Indeed, a goal of this book/course is to allow the reader
Channel estimation for physical layer network coding systems
Gao, Feifei; Wang, Gongpu
2014-01-01
This SpringerBrief presents channel estimation strategies for the physical later network coding (PLNC) systems. Along with a review of PLNC architectures, this brief examines new challenges brought by the special structure of bi-directional two-hop transmissions that are different from the traditional point-to-point systems and unidirectional relay systems. The authors discuss the channel estimation strategies over typical fading scenarios, including frequency flat fading, frequency selective fading and time selective fading, as well as future research directions. Chapters explore the performa
The cosmic code quantum physics as the language of nature
Pagels, Heinz R
2012-01-01
""The Cosmic Code can be read by anyone. I heartily recommend it!"" - The New York Times Book Review""A reliable guide for the nonmathematical reader across the highest ridges of physical theory. Pagels is unfailingly lighthearted and confident."" - Scientific American""A sound, clear, vital work that deserves the attention of anyone who takes an interest in the relationship between material reality and the human mind."" - Science 82This is one of the most important books on quantum mechanics ever written for general readers. Heinz Pagels, an eminent physicist and science writer, discusses and
Theoretical atomic physics code development I: CATS: Cowan Atomic Structure Code
International Nuclear Information System (INIS)
Abdallah, J. Jr.; Clark, R.E.H.; Cowan, R.D.
1988-12-01
An adaptation of R.D. Cowan's Atomic Structure program, CATS, has been developed as part of the Theoretical Atomic Physics (TAPS) code development effort at Los Alamos. CATS has been designed to be easy to run and to produce data files that can interface with other programs easily. The CATS produced data files currently include wave functions, energy levels, oscillator strengths, plane-wave-Born electron-ion collision strengths, photoionization cross sections, and a variety of other quantities. This paper describes the use of CATS. 10 refs
DEFF Research Database (Denmark)
Kneubil, Fabiana Botelho
2016-01-01
In this work we show an approach based on models, for an usual subject in an introductory physics course, in order to foster discussions on the nature of physical knowledge. The introduction of elements of the nature of knowledge in physics lessons has been emphasised by many educators and one uses...... the case of metals to show the theoretical and phenomenological dimensions of physics. The discussion is made by means of four questions whose answers cannot be reached neither for theoretical elements nor experimental measurements. Between these two dimensions it is necessary to realise a series...... of reasoning steps to deepen the comprehension of microscopic concepts, such as electrical resistivity, drift velocity and free electrons. When this approach is highlighted, beyond the physical content, aspects of its nature become explicit and may improve the structuring of knowledge for learners...
Development of Advanced Suite of Deterministic Codes for VHTR Physics Analysis
Energy Technology Data Exchange (ETDEWEB)
Kim, Kang Seog; Cho, J. Y.; Lee, K. H. (and others)
2007-07-15
Advanced Suites of deterministic codes for VHTR physics analysis has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. These code suites include the conventional 2-step procedure in which a few group constants are generated by a transport lattice calculation, and the reactor physics analysis is performed by a 3-dimensional diffusion calculation, and a whole core transport code that can model local heterogeneities directly at the core level. Particular modeling issues in physics analysis of the gas-cooled VHTRs were resolved, which include a double heterogeneity of the coated fuel particles, a neutron streaming in the coolant channels, a strong core-reflector interaction, and large spectrum shifts due to changes of the surrounding environment, temperature and burnup. And the geometry handling capability of the DeCART code were extended to deal with the hexagonal fuel elements of the VHTR core. The developed code suites were validated and verified by comparing the computational results with those of the Monte Carlo calculations for the benchmark problems.
Modeling peripheral olfactory coding in Drosophila larvae.
Directory of Open Access Journals (Sweden)
Derek J Hoare
Full Text Available The Drosophila larva possesses just 21 unique and identifiable pairs of olfactory sensory neurons (OSNs, enabling investigation of the contribution of individual OSN classes to the peripheral olfactory code. We combined electrophysiological and computational modeling to explore the nature of the peripheral olfactory code in situ. We recorded firing responses of 19/21 OSNs to a panel of 19 odors. This was achieved by creating larvae expressing just one functioning class of odorant receptor, and hence OSN. Odor response profiles of each OSN class were highly specific and unique. However many OSN-odor pairs yielded variable responses, some of which were statistically indistinguishable from background activity. We used these electrophysiological data, incorporating both responses and spontaneous firing activity, to develop a bayesian decoding model of olfactory processing. The model was able to accurately predict odor identity from raw OSN responses; prediction accuracy ranged from 12%-77% (mean for all odors 45.2% but was always significantly above chance (5.6%. However, there was no correlation between prediction accuracy for a given odor and the strength of responses of wild-type larvae to the same odor in a behavioral assay. We also used the model to predict the ability of the code to discriminate between pairs of odors. Some of these predictions were supported in a behavioral discrimination (masking assay but others were not. We conclude that our model of the peripheral code represents basic features of odor detection and discrimination, yielding insights into the information available to higher processing structures in the brain.
Sodium pool fire model for CONACS code
International Nuclear Information System (INIS)
Yung, S.C.
1982-01-01
The modeling of sodium pool fires constitutes an important ingredient in conducting LMFBR accident analysis. Such modeling capability has recently come under scrutiny at Westinghouse Hanford Company (WHC) within the context of developing CONACS, the Containment Analysis Code System. One of the efforts in the CONACS program is to model various combustion processes anticipated to occur during postulated accident paths. This effort includes the selection or modification of an existing model and development of a new model if it clearly contributes to the program purpose. As part of this effort, a new sodium pool fire model has been developed that is directed at removing some of the deficiencies in the existing models, such as SOFIRE-II and FEUNA
Towards Product Lining Model-Driven Development Code Generators
Roth, Alexander; Rumpe, Bernhard
2015-01-01
A code generator systematically transforms compact models to detailed code. Today, code generation is regarded as an integral part of model-driven development (MDD). Despite its relevance, the development of code generators is an inherently complex task and common methodologies and architectures are lacking. Additionally, reuse and extension of existing code generators only exist on individual parts. A systematic development and reuse based on a code generator product line is still in its inf...
Development of a three dimension multi-physics code for molten salt fast reactor
International Nuclear Information System (INIS)
Cheng Maosong; Dai Zhimin
2014-01-01
Molten Salt Reactor (MSR) was selected as one of the six innovative nuclear reactors by the Generation IV International Forum (GIF). The circulating-fuel in the can-type molten salt fast reactor makes the neutronics and thermo-hydraulics of the reactor strongly coupled and different from that of traditional solid-fuel reactors. In the present paper: a new coupling model is presented that physically describes the inherent relations between the neutron flux, the delayed neutron precursor, the heat transfer and the turbulent flow. Based on the model, integrating nuclear data processing, CAD modeling, structured and unstructured mesh technology, data analysis and visualization application, a three dimension steady state simulation code system (MSR3DS) for the can-type molten salt fast reactor is developed and validated. In order to demonstrate the ability of the code, the three dimension distributions of the velocity, the neutron flux, the delayed neutron precursor and the temperature were obtained for the simplified MOlten Salt Advanced Reactor Transmuter (MOSART) using this code. The results indicate that the MSR3DS code can provide a feasible description of multi-physical coupling phenomena in can-type molten salt fast reactor. Furthermore, the code can well predict the flow effect of fuel salt and the transport effect of the turbulent diffusion. (authors)
Applications of FLUKA Monte Carlo code for nuclear and accelerator physics
Battistoni, Giuseppe; Brugger, Markus; Campanella, Mauro; Carboni, Massimo; Empl, Anton; Fasso, Alberto; Gadioli, Ettore; Cerutti, Francesco; Ferrari, Alfredo; Ferrari, Anna; Lantz, Matthias; Mairani, Andrea; Margiotta, M; Morone, Christina; Muraro, Silvia; Parodi, Katerina; Patera, Vincenzo; Pelliccioni, Maurizio; Pinsky, Lawrence; Ranft, Johannes; Roesler, Stefan; Rollet, Sofia; Sala, Paola R; Santana, Mario; Sarchiapone, Lucia; Sioli, Maximiliano; Smirnov, George; Sommerer, Florian; Theis, Christian; Trovati, Stefania; Villari, R; Vincke, Heinz; Vincke, Helmut; Vlachoudis, Vasilis; Vollaire, Joachim; Zapp, Neil
2011-01-01
FLUKA is a general purpose Monte Carlo code capable of handling all radiation components from thermal energies (for neutrons) or 1keV (for all other particles) to cosmic ray energies and can be applied in many different fields. Presently the code is maintained on Linux. The validity of the physical models implemented in FLUKA has been benchmarked against a variety of experimental data over a wide energy range, from accelerator data to cosmic ray showers in the Earth atmosphere. FLUKA is widely used for studies related both to basic research and to applications in particle accelerators, radiation protection and dosimetry, including the specific issue of radiation damage in space missions, radiobiology (including radiotherapy) and cosmic ray calculations. After a short description of the main features that make FLUKA valuable for these topics, the present paper summarizes some of the recent applications of the FLUKA Monte Carlo code in the nuclear as well high energy physics. In particular it addresses such top...
Rapid installation of numerical models in multiple parent codes
Energy Technology Data Exchange (ETDEWEB)
Brannon, R.M.; Wong, M.K.
1996-10-01
A set of``model interface guidelines``, called MIG, is offered as a means to more rapidly install numerical models (such as stress-strain laws) into any parent code (hydrocode, finite element code, etc.) without having to modify the model subroutines. The model developer (who creates the model package in compliance with the guidelines) specifies the model`s input and storage requirements in a standardized way. For portability, database management (such as saving user inputs and field variables) is handled by the parent code. To date, NUG has proved viable in beta installations of several diverse models in vectorized and parallel codes written in different computer languages. A NUG-compliant model can be installed in different codes without modifying the model`s subroutines. By maintaining one model for many codes, MIG facilitates code-to-code comparisons and reduces duplication of effort potentially reducing the cost of installing and sharing models.
Validation of physics and thermalhydraulic computer codes for advanced Candu reactor applications
International Nuclear Information System (INIS)
Wren, D.J.; Popov, N.; Snell, V.G.
2004-01-01
Atomic Energy of Canada Ltd. (AECL) is developing an Advanced Candu Reactor (ACR) that is an evolutionary advancement of the currently operating Candu 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular Candu concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the Candu 6, most of the key phenomena associated with the safety analyses of ACR are common, and the Candu industry standard tool-set of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design
Digitized forensics: retaining a link between physical and digital crime scene traces using QR-codes
Hildebrandt, Mario; Kiltz, Stefan; Dittmann, Jana
2013-03-01
The digitization of physical traces from crime scenes in forensic investigations in effect creates a digital chain-of-custody and entrains the challenge of creating a link between the two or more representations of the same trace. In order to be forensically sound, especially the two security aspects of integrity and authenticity need to be maintained at all times. Especially the adherence to the authenticity using technical means proves to be a challenge at the boundary between the physical object and its digital representations. In this article we propose a new method of linking physical objects with its digital counterparts using two-dimensional bar codes and additional meta-data accompanying the acquired data for integration in the conventional documentation of collection of items of evidence (bagging and tagging process). Using the exemplary chosen QR-code as particular implementation of a bar code and a model of the forensic process, we also supply a means to integrate our suggested approach into forensically sound proceedings as described by Holder et al.1 We use the example of the digital dactyloscopy as a forensic discipline, where currently progress is being made by digitizing some of the processing steps. We show an exemplary demonstrator of the suggested approach using a smartphone as a mobile device for the verification of the physical trace to extend the chain-of-custody from the physical to the digital domain. Our evaluation of the demonstrator is performed towards the readability and the verification of its contents. We can read the bar code despite its limited size of 42 x 42 mm and rather large amount of embedded data using various devices. Furthermore, the QR-code's error correction features help to recover contents of damaged codes. Subsequently, our appended digital signature allows for detecting malicious manipulations of the embedded data.
MEMOPS: data modelling and automatic code generation.
Fogh, Rasmus H; Boucher, Wayne; Ionides, John M C; Vranken, Wim F; Stevens, Tim J; Laue, Ernest D
2010-03-25
In recent years the amount of biological data has exploded to the point where much useful information can only be extracted by complex computational analyses. Such analyses are greatly facilitated by metadata standards, both in terms of the ability to compare data originating from different sources, and in terms of exchanging data in standard forms, e.g. when running processes on a distributed computing infrastructure. However, standards thrive on stability whereas science tends to constantly move, with new methods being developed and old ones modified. Therefore maintaining both metadata standards, and all the code that is required to make them useful, is a non-trivial problem. Memops is a framework that uses an abstract definition of the metadata (described in UML) to generate internal data structures and subroutine libraries for data access (application programming interfaces--APIs--currently in Python, C and Java) and data storage (in XML files or databases). For the individual project these libraries obviate the need for writing code for input parsing, validity checking or output. Memops also ensures that the code is always internally consistent, massively reducing the need for code reorganisation. Across a scientific domain a Memops-supported data model makes it easier to support complex standards that can capture all the data produced in a scientific area, share them among all programs in a complex software pipeline, and carry them forward to deposition in an archive. The principles behind the Memops generation code will be presented, along with example applications in Nuclear Magnetic Resonance (NMR) spectroscopy and structural biology.
Gyrofluid Modeling of Turbulent, Kinetic Physics
Despain, Kate Marie
2011-12-01
Gyrofluid models to describe plasma turbulence combine the advantages of fluid models, such as lower dimensionality and well-developed intuition, with those of gyrokinetics models, such as finite Larmor radius (FLR) effects. This allows gyrofluid models to be more tractable computationally while still capturing much of the physics related to the FLR of the particles. We present a gyrofluid model derived to capture the behavior of slow solar wind turbulence and describe the computer code developed to implement the model. In addition, we describe the modifications we made to a gyrofluid model and code that simulate plasma turbulence in tokamak geometries. Specifically, we describe a nonlinear phase mixing phenomenon, part of the E x B term, that was previously missing from the model. An inherently FLR effect, it plays an important role in predicting turbulent heat flux and diffusivity levels for the plasma. We demonstrate this importance by comparing results from the updated code to studies done previously by gyrofluid and gyrokinetic codes. We further explain what would be necessary to couple the updated gyrofluid code, gryffin, to a turbulent transport code, thus allowing gryffin to play a role in predicting profiles for fusion devices such as ITER and to explore novel fusion configurations. Such a coupling would require the use of Graphical Processing Units (GPUs) to make the modeling process fast enough to be viable. Consequently, we also describe our experience with GPU computing and demonstrate that we are poised to complete a gryffin port to this innovative architecture.
Verification and Validation of Heat Transfer Model of AGREE Code
Energy Technology Data Exchange (ETDEWEB)
Tak, N. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Seker, V.; Drzewiecki, T. J.; Downar, T. J. [Department of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Michigan (United States); Kelly, J. M. [US Nuclear Regulatory Commission, Washington (United States)
2013-05-15
The AGREE code was originally developed as a multi physics simulation code to perform design and safety analysis of Pebble Bed Reactors (PBR). Currently, additional capability for the analysis of Prismatic Modular Reactor (PMR) core is in progress. Newly implemented fluid model for a PMR core is based on a subchannel approach which has been widely used in the analyses of light water reactor (LWR) cores. A hexagonal fuel (or graphite block) is discretized into triangular prism nodes having effective conductivities. Then, a meso-scale heat transfer model is applied to the unit cell geometry of a prismatic fuel block. Both unit cell geometries of multi-hole and pin-in-hole types of prismatic fuel blocks are considered in AGREE. The main objective of this work is to verify and validate the heat transfer model newly implemented for a PMR core in the AGREE code. The measured data in the HENDEL experiment were used for the validation of the heat transfer model for a pin-in-hole fuel block. However, the HENDEL tests were limited to only steady-state conditions of pin-in-hole fuel blocks. There exist no available experimental data regarding a heat transfer in multi-hole fuel blocks. Therefore, numerical benchmarks using conceptual problems are considered to verify the heat transfer model of AGREE for multi-hole fuel blocks as well as transient conditions. The CORONA and GAMMA+ codes were used to compare the numerical results. In this work, the verification and validation study were performed for the heat transfer model of the AGREE code using the HENDEL experiment and the numerical benchmarks of selected conceptual problems. The results of the present work show that the heat transfer model of AGREE is accurate and reliable for prismatic fuel blocks. Further validation of AGREE is in progress for a whole reactor problem using the HTTR safety test data such as control rod withdrawal tests and loss-of-forced convection tests.
The MESORAD dose assessment model: Computer code
International Nuclear Information System (INIS)
Ramsdell, J.V.; Athey, G.F.; Bander, T.J.; Scherpelz, R.I.
1988-10-01
MESORAD is a dose equivalent model for emergency response applications that is designed to be run on minicomputers. It has been developed by the Pacific Northwest Laboratory for use as part of the Intermediate Dose Assessment System in the US Nuclear Regulatory Commission Operations Center in Washington, DC, and the Emergency Management System in the US Department of Energy Unified Dose Assessment Center in Richland, Washington. This volume describes the MESORAD computer code and contains a listing of the code. The technical basis for MESORAD is described in the first volume of this report (Scherpelz et al. 1986). A third volume of the documentation planned. That volume will contain utility programs and input and output files that can be used to check the implementation of MESORAD. 18 figs., 4 tabs
Tokamak Simulation Code modeling of NSTX
International Nuclear Information System (INIS)
Jardin, S.C.; Kaye, S.; Menard, J.; Kessel, C.; Glasser, A.H.
2000-01-01
The Tokamak Simulation Code [TSC] is widely used for the design of new axisymmetric toroidal experiments. In particular, TSC was used extensively in the design of the National Spherical Torus eXperiment [NSTX]. The authors have now benchmarked TSC with initial NSTX results and find excellent agreement for plasma and vessel currents and magnetic flux loops when the experimental coil currents are used in the simulations. TSC has also been coupled with a ballooning stability code and with DCON to provide stability predictions for NSTX operation. TSC has also been used to model initial CHI experiments where a large poloidal voltage is applied to the NSTX vacuum vessel, causing a force-free current to appear in the plasma. This is a phenomenon that is similar to the plasma halo current that sometimes develops during a plasma disruption
Assessing physical models used in nuclear aerosol transport models
International Nuclear Information System (INIS)
McDonald, B.H.
1987-01-01
Computer codes used to predict the behaviour of aerosols in water-cooled reactor containment buildings after severe accidents contain a variety of physical models. Special models are in place for describing agglomeration processes where small aerosol particles combine to form larger ones. Other models are used to calculate the rates at which aerosol particles are deposited on building structures. Condensation of steam on aerosol particles is currently a very active area in aerosol modelling. In this paper, the physical models incorporated in the current available international codes for all of these processes are reviewed and documented. There is considerable variation in models used in different codes, and some uncertainties exist as to which models are superior. 28 refs
The r-Java 2.0 code: nuclear physics
Kostka, M.; Koning, N.; Shand, Z.; Ouyed, R.; Jaikumar, P.
2014-08-01
Aims: We present r-Java 2.0, a nucleosynthesis code for open use that performs r-process calculations, along with a suite of other analysis tools. Methods: Equipped with a straightforward graphical user interface, r-Java 2.0 is capable of simulating nuclear statistical equilibrium (NSE), calculating r-process abundances for a wide range of input parameters and astrophysical environments, computing the mass fragmentation from neutron-induced fission and studying individual nucleosynthesis processes. Results: In this paper we discuss enhancements to this version of r-Java, especially the ability to solve the full reaction network. The sophisticated fission methodology incorporated in r-Java 2.0 that includes three fission channels (beta-delayed, neutron-induced, and spontaneous fission), along with computation of the mass fragmentation, is compared to the upper limit on mass fission approximation. The effects of including beta-delayed neutron emission on r-process yield is studied. The role of Coulomb interactions in NSE abundances is shown to be significant, supporting previous findings. A comparative analysis was undertaken during the development of r-Java 2.0 whereby we reproduced the results found in the literature from three other r-process codes. This code is capable of simulating the physical environment of the high-entropy wind around a proto-neutron star, the ejecta from a neutron star merger, or the relativistic ejecta from a quark nova. Likewise the users of r-Java 2.0 are given the freedom to define a custom environment. This software provides a platform for comparing proposed r-process sites.
Energy Technology Data Exchange (ETDEWEB)
Joshua J. Cogliati; Abderrafi M. Ougouag
2006-10-01
A comprehensive, high fidelity model for pebble flow has been developed and embodied in the PEBBLES computer code. In this paper, a description of the physical artifacts included in the model is presented and some results from using the computer code for predicting the features of pebble flow and packing in a realistic pebble bed reactor design are shown. The sensitivity of models to various physical parameters is also discussed.
Development status of the lattice physics code in COSINE project
Energy Technology Data Exchange (ETDEWEB)
Chen, Y.; Yu, H.; Li, S.; Liu, Z.; Yan, Y. [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software NEKLS, North Third Ring Road, Beijing 100029 (China)
2013-07-01
LATC is an essential part of COSINE code package, which stands for Core and System Integrated Engine for design and analysis. LATC performs 2D multi-group assembly transport calculation and generates few group constants and the required cross-section data for CORE, the core simulator code. LATC is designed to have the capability of modeling the API 000 series assemblies. The development is a continuously improved process. Currently, LATC uses well-proven technology to achieve the key functions. In the next stage, more advanced methods and modules will be implemented. At present, WIMS and WIMS improved format library could be read in LATC code. For resonance calculation, equivalent relation with rational approximations is utilized. For transport calculation, two options are available. One choice is collision probability method in cell homogenization while discrete coordinate method in assembly homogenization, the other is method of characteristics in assembly homogenization directly. For depletion calculation, an improved linear rate 'constant power' depletion method has been developed. (authors)
Development status of the lattice physics code in COSINE project
International Nuclear Information System (INIS)
Chen, Y.; Yu, H.; Li, S.; Liu, Z.; Yan, Y.
2013-01-01
LATC is an essential part of COSINE code package, which stands for Core and System Integrated Engine for design and analysis. LATC performs 2D multi-group assembly transport calculation and generates few group constants and the required cross-section data for CORE, the core simulator code. LATC is designed to have the capability of modeling the API 000 series assemblies. The development is a continuously improved process. Currently, LATC uses well-proven technology to achieve the key functions. In the next stage, more advanced methods and modules will be implemented. At present, WIMS and WIMS improved format library could be read in LATC code. For resonance calculation, equivalent relation with rational approximations is utilized. For transport calculation, two options are available. One choice is collision probability method in cell homogenization while discrete coordinate method in assembly homogenization, the other is method of characteristics in assembly homogenization directly. For depletion calculation, an improved linear rate 'constant power' depletion method has been developed. (authors)
Energy Technology Data Exchange (ETDEWEB)
Bellantoni, L.
2009-11-01
There are many recent results from searches for fundamental new physics using the TeVatron, the SLAC b-factory and HERA. This talk quickly reviewed searches for pair-produced stop, for gauge-mediated SUSY breaking, for Higgs bosons in the MSSM and NMSSM models, for leptoquarks, and v-hadrons. There is a SUSY model which accommodates the recent astrophysical experimental results that suggest that dark matter annihilation is occurring in the center of our galaxy, and a relevant experimental result. Finally, model-independent searches at D0, CDF, and H1 are discussed.
Recent improvements and new features in the Westinghouse lattice physics codes
International Nuclear Information System (INIS)
Huria, H.C.; Buechel, R.J.
1995-01-01
Westinghouse has been using the ANC three-dimensional, two-energy-group nodal model for nuclear analysis and fuel management calculations for standard pressurized water reactor (PWR) reload design analysis since 1988. The cross sections are obtained from PHOENIX-P, a modified version of the PHOENIX lattice physics code for all square-assembly PWR cores. The PHOENIX-H code was developed for modeling both the VVER-1000 and VVER-440 fuel lattice configurations. The PHOENIX-H code has evolved from PHOENIX-P, the primary difference being in the neutronic solution modules. The PHOENIX-P code determines the assembly flux distribution using integral transport theory-based pin-cell nodal coupling followed by two-dimensional discrete ordinates solution in x-y geometry. The PHOENIX-H code uses the two-dimensional heterogeneous response method. The other infrastructure is identical in both the codes, and they share the same 42-group cross-section library
Hitch code capabilities for modeling AVT chemistry
International Nuclear Information System (INIS)
Leibovitz, J.
1985-01-01
Several types of corrosion have damaged alloy 600 tubing in the secondary side of steam generators. The types of corrosion include wastage, denting, intergranular attack, stress corrosion, erosion-corrosion, etc. The environments which cause attack may originate from leaks of cooling water into the condensate, etc. When the contaminated feedwater is pumped into the generator, the impurities may concentrate first 200 to 400 fold in the bulk water, depending on the blowdown, and then further to saturation and dryness in heated tube support plate crevices. Characterization of local solution chemistries is the first step to predict and correct the type of corrosion that can occur. The pH is of particular importance because it is a major factor governing the rate of corrosion reactions. The pH of a solution at high temperature is not the same as the ambient temperature, since ionic dissociation constants, solubility and solubility products, activity coefficients, etc., all change with temperature. Because the high temperature chemistry of such solutions is not readily characterized experimentally, modeling techniques were developed under EPRI sponsorship to calculate the high temperature chemistry of the relevant solutions. In many cases, the effects of cooling water impurities on steam generator water chemistry with all volatile treatment (AVT), upon concentration by boiling, and in particular the resulting acid or base concentration can be calculated by a simple code, the HITCH code, which is very easy to use. The scope and applicability of the HITCH code are summarized
Top flooding modeling with MAAP4 code
International Nuclear Information System (INIS)
Brunet-Thibault, E.; Marguet, S.
2006-01-01
An engineering top flooding model was developed in MAAP4.04d.4, the severe accident code used in EDF, to simulate the thermal-hydraulic phenomena that should take place if emergency core cooling (ECC) water was injected in hot leg during quenching. In the framework of the ISTC (International Science and Technology Centre), a top flooding test was proposed in the PARAMETER facility (Podolsk, Russia). The MAAP calculation of the PARAMETER top flooding test is presented in this paper. A comparison between top and bottom flooding was made on the bundle test geometry. According to this study, top flooding appears to cool quickly and effectively the upper plenum internals. (author)
Current status of the reactor physics code WIMS and recent developments
International Nuclear Information System (INIS)
Lindley, B.A.; Hosking, J.G.; Smith, P.J.; Powney, D.J.; Tollit, B.S.; Newton, T.D.; Perry, R.; Ware, T.C.; Smith, P.N.
2017-01-01
Highlights: • The current status of the WIMS reactor physics code is presented. • Applications range from 2D lattice calculations up to 3D whole core geometries. • Gamma transport and thermal-hydraulic feedback models added. • Calculations methodologies described for several Gen II, III and IV reactor types. - Abstract: The WIMS modular reactor physics code has been under continuous development for over fifty years. This paper discusses the current status of WIMS and recent developments, in particular developments to the resonance shielding methodology and 3D transport solvers. Traditionally, WIMS is used to perform 2D lattice calculations, typically to generate homogenized reactor physics parameters for a whole core code such as PANTHER. However, with increasing computational resources there has been a growing trend for performing transport calculations on larger problems, up to and including 3D full core models. To this end, a number of the WIMS modules have been parallelised to allow efficient performance for whole core calculations, and WIMS includes a 3D method of characteristics solver with reflective and once-through tracking methods, which can be used to analyse problems of varying size and complexity. A time-dependent flux solver has been incorporated and thermal-hydraulic modelling capability is also being added to allow steady-state and transient coupled calculations to be performed. WIMS has been validated against a range of experimental data and other codes, in particular for water and graphite moderated thermal reactors. Future developments will include improved parallelization, enhancing the thermal-hydraulic feedback models and validating the WIMS/PANTHER code system for BWRs and fast reactors.
Phenomenological optical potentials and optical model computer codes
International Nuclear Information System (INIS)
Prince, A.
1980-01-01
An introduction to the Optical Model is presented. Starting with the purpose and nature of the physical problems to be analyzed, a general formulation and the various phenomenological methods of solution are discussed. This includes the calculation of observables based on assumed potentials such as local and non-local and their forms, e.g. Woods-Saxon, folded model etc. Also discussed are the various calculational methods and model codes employed to describe nuclear reactions in the spherical and deformed regions (e.g. coupled-channel analysis). An examination of the numerical solutions and minimization techniques associated with the various codes, is briefly touched upon. Several computer programs are described for carrying out the calculations. The preparation of input, (formats and options), determination of model parameters and analysis of output are described. The class is given a series of problems to carry out using the available computer. Interpretation and evaluation of the samples includes the effect of varying parameters, and comparison of calculations with the experimental data. Also included is an intercomparison of the results from the various model codes, along with their advantages and limitations. (author)
Improvement of blow down model for LEAP code
International Nuclear Information System (INIS)
Itooka, Satoshi; Fujimata, Kazuhiro
2003-03-01
In Japan Nuclear Cycle Development Institute, the improvement of analysis method for overheating tube rapture was studied for the accident of sodium-water reactions in the steam generator of a fast breeder reactor and the evaluation of heat transfer condition in the tube were carried out based on study of critical heat flux (CHF) and post-CHF heat transfer equation in Light Water Reactors. In this study, the improvement of blow down model for the LEAP code was carried out taking into consideration the above-mentioned evaluation of heat transfer condition. Improvements of the LEAP code were following items. Calculations and verification were performed with the improved LEAP code in order to confirm the code functions. The addition of critical heat flux (CHF) by the formula of Katto and the formula of Tong. The addition of post-CHF heat transfer equation by the formula of Condie-BengstonIV and the formula of Groeneveld 5.9. The physical properties of the water and steam are expanded to the critical conditions of the water. The expansion of the total number of section and the improvement of the input form. The addition of the function to control the valve setting by the PID control model. (author)
24 CFR 200.925c - Model codes.
2010-04-01
... below. (1) Model Building Codes—(i) The BOCA National Building Code, 1993 Edition, The BOCA National..., Administration, for the Building, Plumbing and Mechanical Codes and the references to fire retardant treated wood... number 2 (Chapter 7) of the Building Code, but including the Appendices of the Code. Available from...
Implementation of JAERI's reflood model into TRAC-PF1/MOD1 code
International Nuclear Information System (INIS)
Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio
1993-02-01
Selected physical models of REFLA code, that is a reflood analysis code developed at JAERI, were implemented into the TRAC-PF1/MOD1 code in order to improve the predictive capability of the TRAC-PF1/MOD1 code for the core thermal hydraulic behaviors during the reflood phase in a PWR LOCA. Through comparisons of physical models between both codes, (1) Murao-Iguchi void fraction correlation, (2) the drag coefficient correlation acting to drops, (3) the correlation for wall heat transfer coefficient in the film boiling regime, (4) the quench velocity correlation and (5) heat transfer correlations for the dispersed flow regime were selected from the REFLA code to be implemented into the TRAC-PF1/MOD1 code. A method for the transformation of the void fraction correlation to the equivalent interfacial friction model was developed and the effect of the transformation method on the stability of the solution was discussed. Through assessment calculation using data from CCTF (Cylindrical Core Test Facility) flat power test, it was confirmed that the predictive capability of the TRAC code for the core thermal hydraulic behaviors during the reflood can be improved by the implementation of selected physical models of the REFLA code. Several user guidelines for the modified TRAC code were proposed based on the sensitivity studies on fluid cell number in the hydraulic calculation and on node number and effect of axial heat conduction in the heat conduction calculation of fuel rod. (author)
Containment Modelling with the ASTEC Code
International Nuclear Information System (INIS)
Sadek, Sinisa; Grgic, Davor
2014-01-01
ASTEC is an integral computer code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fur Anlagen-und Reaktorsicherheit (GRS, Germany) to assess the nuclear power plant behaviour during a severe accident (SA). It consists of 13 coupled modules which compute various SA phenomena in primary and secondary circuits of the nuclear power plants (NPP), and in the containment. The ASTEC code was used to model and to simulate NPP behaviour during a postulated station blackout accident in the NPP Krsko, a two-loop pressurized water reactor (PWR) plant. The primary system of the plant was modelled with 110 thermal hydraulic (TH) volumes, 113 junctions and 128 heat structures. The secondary system was modelled with 76 TH volumes, 77 junctions and 87 heat structures. The containment was modelled with 10 TH volumes by taking into account containment representation as a set of distinctive compartments, connected with 23 junctions. A total of 79 heat structures were used to simulate outer containment walls and internal steel and concrete structures. Prior to the transient calculation, a steady state analysis was performed. In order to achieve correct plant initial conditions, the operation of regulation systems was modelled. Parameters which were subjected to regulation were the pressurizer pressure, the pressurizer narrow range level and steam mass flow rates in the steam lines. The accident analysis was focused on containment behaviour, however the complete integral NPP analysis was carried out in order to provide correct boundary conditions for the containment calculation. During the accident, the containment integrity was challenged by release of reactor system coolant through degraded coolant pump seals and, later in the accident following release of the corium out of the reactor pressure vessel, by the molten corium concrete interaction and direct containment heating mechanisms. Impact of those processes on relevant
A graph model for opportunistic network coding
Sorour, Sameh
2015-08-12
© 2015 IEEE. Recent advancements in graph-based analysis and solutions of instantly decodable network coding (IDNC) trigger the interest to extend them to more complicated opportunistic network coding (ONC) scenarios, with limited increase in complexity. In this paper, we design a simple IDNC-like graph model for a specific subclass of ONC, by introducing a more generalized definition of its vertices and the notion of vertex aggregation in order to represent the storage of non-instantly-decodable packets in ONC. Based on this representation, we determine the set of pairwise vertex adjacency conditions that can populate this graph with edges so as to guarantee decodability or aggregation for the vertices of each clique in this graph. We then develop the algorithmic procedures that can be applied on the designed graph model to optimize any performance metric for this ONC subclass. A case study on reducing the completion time shows that the proposed framework improves on the performance of IDNC and gets very close to the optimal performance.
REPFLO model evaluation, physical and numerical consistency
International Nuclear Information System (INIS)
Wilson, R.N.; Holland, D.H.
1978-11-01
This report contains a description of some suggested changes and an evaluation of the REPFLO computer code, which models ground-water flow and nuclear-waste migration in and about a nuclear-waste repository. The discussion contained in the main body of the report is supplemented by a flow chart, presented in the Appendix of this report. The suggested changes are of four kinds: (1) technical changes to make the code compatible with a wider variety of digital computer systems; (2) changes to fill gaps in the computer code, due to missing proprietary subroutines; (3) changes to (a) correct programming errors, (b) correct logical flaws, and (c) remove unnecessary complexity; and (4) changes in the computer code logical structure to make REPFLO a more viable model from the physical point of view
Conservation of concrete structures according to fib Model Code 2010
Matthews, S.; Bigaj-Van Vliet, A.; Ueda, T.
2013-01-01
Conservation of concrete structures forms an essential part of the fib Model Code for Concrete Structures 2010 (fib Model Code 2010). In particular, Chapter 9 of fib Model Code 2010 addresses issues concerning conservation strategies and tactics, conservation management, condition surveys, condition
Gap Conductance model Validation in the TASS/SMR-S code using MARS code
International Nuclear Information System (INIS)
Ahn, Sang Jun; Yang, Soo Hyung; Chung, Young Jong; Lee, Won Jae
2010-01-01
Korea Atomic Energy Research Institute (KAERI) has been developing the TASS/SMR-S (Transient and Setpoint Simulation/Small and Medium Reactor) code, which is a thermal hydraulic code for the safety analysis of the advanced integral reactor. An appropriate work to validate the applicability of the thermal hydraulic models within the code should be demanded. Among the models, the gap conductance model which is describes the thermal gap conductivity between fuel and cladding was validated through the comparison with MARS code. The validation of the gap conductance model was performed by evaluating the variation of the gap temperature and gap width as the changed with the power fraction. In this paper, a brief description of the gap conductance model in the TASS/SMR-S code is presented. In addition, calculated results to validate the gap conductance model are demonstrated by comparing with the results of the MARS code with the test case
A graph model for opportunistic network coding
Sorour, Sameh; Aboutoraby, Neda; Al-Naffouri, Tareq Y.; Alouini, Mohamed-Slim
2015-01-01
© 2015 IEEE. Recent advancements in graph-based analysis and solutions of instantly decodable network coding (IDNC) trigger the interest to extend them to more complicated opportunistic network coding (ONC) scenarios, with limited increase
Code Differentiation for Hydrodynamic Model Optimization
Energy Technology Data Exchange (ETDEWEB)
Henninger, R.J.; Maudlin, P.J.
1999-06-27
Use of a hydrodynamics code for experimental data fitting purposes (an optimization problem) requires information about how a computed result changes when the model parameters change. These so-called sensitivities provide the gradient that determines the search direction for modifying the parameters to find an optimal result. Here, the authors apply code-based automatic differentiation (AD) techniques applied in the forward and adjoint modes to two problems with 12 parameters to obtain these gradients and compare the computational efficiency and accuracy of the various methods. They fit the pressure trace from a one-dimensional flyer-plate experiment and examine the accuracy for a two-dimensional jet-formation problem. For the flyer-plate experiment, the adjoint mode requires similar or less computer time than the forward methods. Additional parameters will not change the adjoint mode run time appreciably, which is a distinct advantage for this method. Obtaining ''accurate'' sensitivities for the j et problem parameters remains problematic.
Generomak: Fusion physics, engineering and costing model
International Nuclear Information System (INIS)
Delene, J.G.; Krakowski, R.A.; Sheffield, J.; Dory, R.A.
1988-06-01
A generic fusion physics, engineering and economics model (Generomak) was developed as a means of performing consistent analysis of the economic viability of alternative magnetic fusion reactors. The original Generomak model developed at Oak Ridge by Sheffield was expanded for the analyses of the Senior Committee on Environmental Safety and Economics of Magnetic Fusion Energy (ESECOM). This report describes the Generomak code as used by ESECOM. The input data used for each of the ten ESECOM fusion plants and the Generomak code output for each case is given. 14 refs., 3 figs., 17 tabs
ER@CEBAF: Modeling code developments
Energy Technology Data Exchange (ETDEWEB)
Meot, F. [Brookhaven National Lab. (BNL), Upton, NY (United States); Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States); Roblin, Y. [Brookhaven National Lab. (BNL), Upton, NY (United States); Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States)
2016-04-13
A proposal for a multiple-pass, high-energy, energy-recovery experiment using CEBAF is under preparation in the frame of a JLab-BNL collaboration. In view of beam dynamics investigations regarding this project, in addition to the existing model in use in Elegant a version of CEBAF is developed in the stepwise ray-tracing code Zgoubi, Beyond the ER experiment, it is also planned to use the latter for the study of polarization transport in the presence of synchrotron radiation, down to Hall D line where a 12 GeV polarized beam can be delivered. This Note briefly reports on the preliminary steps, and preliminary outcomes, based on an Elegant to Zgoubi translation.
Integrated Numerical Experiments (INEX) and the Free-Electron Laser Physical Process Code (FELPPC)
International Nuclear Information System (INIS)
Thode, L.E.; Chan, K.C.D.; Schmitt, M.J.; McKee, J.; Ostic, J.; Elliott, C.J.; McVey, B.D.
1990-01-01
The strong coupling of subsystem elements, such as the accelerator, wiggler, and optics, greatly complicates the understanding and design of a free electron laser (FEL), even at the conceptual level. To address the strong coupling character of the FEL the concept of an Integrated Numerical Experiment (INEX) was proposed. Unique features of the INEX approach are consistency and numerical equivalence of experimental diagnostics. The equivalent numerical diagnostics mitigates the major problem of misinterpretation that often occurs when theoretical and experimental data are compared. The INEX approach has been applied to a large number of accelerator and FEL experiments. Overall, the agreement between INEX and the experiments is very good. Despite the success of INEX, the approach is difficult to apply to trade-off and initial design studies because of the significant manpower and computational requirements. On the other hand, INEX provides a base from which realistic accelerator, wiggler, and optics models can be developed. The Free Electron Laser Physical Process Code (FELPPC) includes models developed from INEX, provides coupling between the subsystem models, and incorporates application models relevant to a specific trade-off or design study. In other words, FELPPC solves the complete physical process model using realistic physics and technology constraints. Because FELPPC provides a detailed design, a good estimate for the FEL mass, cost, and size can be made from a piece-part count of the FEL. FELPPC requires significant accelerator and FEL expertise to operate. The code can calculate complex FEL configurations including multiple accelerator and wiggler combinations
Improved choked flow model for MARS code
International Nuclear Information System (INIS)
Chung, Moon Sun; Lee, Won Jae; Ha, Kwi Seok; Hwang, Moon Kyu
2002-01-01
Choked flow calculation is improved by using a new sound speed criterion for bubbly flow that is derived by the characteristic analysis of hyperbolic two-fluid model. This model was based on the notion of surface tension for the interfacial pressure jump terms in the momentum equations. Real eigenvalues obtained as the closed-form solution of characteristic polynomial represent the sound speed in the bubbly flow regime that agrees well with the existing experimental data. The present sound speed shows more reasonable result in the extreme case than the Nguyens did. The present choked flow criterion derived by the present sound speed is employed in the MARS code and assessed by using the Marviken choked flow tests. The assessment results without any adjustment made by some discharge coefficients demonstrate more accurate predictions of choked flow rate in the bubbly flow regime than those of the earlier choked flow calculations. By calculating the Typical PWR (SBLOCA) problem, we make sure that the present model can reproduce the reasonable transients of integral reactor system
“PROCESS”: A systems code for fusion power plants—Part 1: Physics
Energy Technology Data Exchange (ETDEWEB)
Kovari, M., E-mail: michael.kovari@ccfe.ac.uk; Kemp, R.; Lux, H.; Knight, P.; Morris, J.; Ward, D.J.
2014-12-15
Highlights: • PROCESS is a fusion reactor systems code. • It optimises a figure of merit subject to constraints chosen by the user. • CCFE are working to make the assumptions and equations explicit and public. • The PROCESS homepage is (www.ccfe.ac.uk/powerplants.aspx). - Abstract: PROCESS is a reactor systems code – it assesses the engineering and economic viability of a hypothetical fusion power station using simple models of all parts of a reactor system, from the basic plasma physics to the generation of electricity. It has been used for many years, but details of its operation have not been previously published. This paper describes some of its capabilities. PROCESS is usually used in optimisation mode, in which it finds a set of parameters that maximise (or minimise) a figure of merit chosen by the user, while being consistent with the inputs and the specified constraints. Because the user can apply all the physically relevant constraints, while allowing a large number of parameters to vary, it is in principle only necessary to run the code once to produce a self-consistent, physically plausible reactor model. The scope of PROCESS is very wide and goes well beyond reactor physics, including conversion of heat to electricity, buildings, and costs, but this paper describes only the plasma physics and magnetic field calculations. The capabilities of PROCESS in plasma physics are limited, as its main aim is to combine engineering, physics and economics. A model is described which shows the main plasma features of an inductive ITER scenario. Significant differences between the PROCESS results and the published scenario include the bootstrap current and loop voltage. The PROCESS models for these are being revised. Two new models for DEMO have been obtained. The first, DEMO A, is intended to be “conservative” in that it might be possible to build it using the technology of the near future. For example, since current drive technologies are not yet
PetriCode: A Tool for Template-Based Code Generation from CPN Models
DEFF Research Database (Denmark)
Simonsen, Kent Inge
2014-01-01
Code generation is an important part of model driven methodologies. In this paper, we present PetriCode, a software tool for generating protocol software from a subclass of Coloured Petri Nets (CPNs). The CPN subclass is comprised of hierarchical CPN models describing a protocol system at different...
Mechanistic modelling of gaseous fission product behaviour in UO2 fuel by Rtop code
International Nuclear Information System (INIS)
Kanukova, V.D.; Khoruzhii, O.V.; Kourtchatov, S.Y.; Likhanskii, V.V.; Matveew, L.V.
2002-01-01
The current status of a mechanistic modelling by the RTOP code of the fission product behaviour in polycrystalline UO 2 fuel is described. An outline of the code and implemented physical models is presented. The general approach to code validation is discussed. It is exemplified by the results of validation of the models of fuel oxidation and grain growth. The different models of intragranular and intergranular gas bubble behaviour have been tested and the sensitivity of the code in the framework of these models has been analysed. An analysis of available models of the resolution of grain face bubbles is also presented. The possibilities of the RTOP code are presented through the example of modelling behaviour of WWER fuel over the course of a comparative WWER-PWR experiment performed at Halden and by comparison with Yanagisawa experiments. (author)
Assessment of CANDU physics codes using experimental data - II: CANDU core physics measurements
International Nuclear Information System (INIS)
Roh, Gyu Hong; Jeong, Chang Joon; Choi, Hang Bok
2001-11-01
Benchmark calculations of the advanced CANDU reactor analysis tools (WIMS-AECL, SHETAN and RFSP) and the Monte Carlo code MCNP-4B have been performed using Wolsong Units 2 and 3 Phase-B measurement data. In this study, the benchmark calculations have been done for the criticality, boron worth, reactivity device worth, reactivity coefficient, and flux scan. For the validation of the WIMS-AECL/SHETANRFSP code system, the lattice parameters of the fuel channel were generated by the WIMS-AECL code, and incremental cross sections of reactivity devices and structural material were generated by the SHETAN code. The results have shown that the criticality is under-predicted by -4 mk. The reactivity device worths are generally consistent with the measured data except for the strong absorbers such as shutoff rod and mechanical control absorber. The heat transport system temperature coefficient and flux distributions are in good agreement with the measured data. However, the moderator temperature coefficient has shown a relatively large error, which could be caused by the incremental cross-section generation methodology for the reactivity device. For the MCNP-4B benchmark calculation, cross section libraries were newly generated from ENDF/B-VI release 3 through the NJOY97.114 data processing system and a three-dimensional full core model was developed. The simulation results have shown that the criticality is estimated within 4 mk and the estimated reactivity worth of the control devices are generally consistent with the measurement data, which implies that the MCNP code is valid for CANDU core analysis. In the future, therefore, the MCNP code could be used as a reference tool to benchmark design and analysis codes for the advanced fuels for which experimental data are not available
Françoise Benz
2006-01-01
2005-2006 ACADEMIC TRAINING PROGRAMME LECTURE SERIES 27, 28, 29 June 11:00-12:00 - TH Conference Room, bldg. 4 The use of Monte Carlo radiation transport codes in radiation physics and dosimetry F. Salvat Gavalda,Univ. de Barcelona, A. FERRARI, CERN-AB, M. SILARI, CERN-SC Lecture 1. Transport and interaction of electromagnetic radiation F. Salvat Gavalda,Univ. de Barcelona Interaction models and simulation schemes implemented in modern Monte Carlo codes for the simulation of coupled electron-photon transport will be briefly reviewed. Different schemes for simulating electron transport will be discussed. Condensed algorithms, which rely on multiple-scattering theories, are comparatively fast, but less accurate than mixed algorithms, in which hard interactions (with energy loss or angular deflection larger than certain cut-off values) are simulated individually. The reliability, and limitations, of electron-interaction models and multiple-scattering theories will be analyzed. Benchmark comparisons of simu...
Development of Parallel Code for the Alaska Tsunami Forecast Model
Bahng, B.; Knight, W. R.; Whitmore, P.
2014-12-01
The Alaska Tsunami Forecast Model (ATFM) is a numerical model used to forecast propagation and inundation of tsunamis generated by earthquakes and other means in both the Pacific and Atlantic Oceans. At the U.S. National Tsunami Warning Center (NTWC), the model is mainly used in a pre-computed fashion. That is, results for hundreds of hypothetical events are computed before alerts, and are accessed and calibrated with observations during tsunamis to immediately produce forecasts. ATFM uses the non-linear, depth-averaged, shallow-water equations of motion with multiply nested grids in two-way communications between domains of each parent-child pair as waves get closer to coastal waters. Even with the pre-computation the task becomes non-trivial as sub-grid resolution gets finer. Currently, the finest resolution Digital Elevation Models (DEM) used by ATFM are 1/3 arc-seconds. With a serial code, large or multiple areas of very high resolution can produce run-times that are unrealistic even in a pre-computed approach. One way to increase the model performance is code parallelization used in conjunction with a multi-processor computing environment. NTWC developers have undertaken an ATFM code-parallelization effort to streamline the creation of the pre-computed database of results with the long term aim of tsunami forecasts from source to high resolution shoreline grids in real time. Parallelization will also permit timely regeneration of the forecast model database with new DEMs; and, will make possible future inclusion of new physics such as the non-hydrostatic treatment of tsunami propagation. The purpose of our presentation is to elaborate on the parallelization approach and to show the compute speed increase on various multi-processor systems.
Energy Technology Data Exchange (ETDEWEB)
Chung, Young Jong; Kim, Soo Hyoung; Kim, See Darl (and others)
2008-10-15
The TASS/SMR code has been developed with domestic technologies for the safety analysis of the SMART plant which is an integral type pressurized water reactor. It can be applied to the analysis of design basis accidents including non-LOCA (loss of coolant accident) and LOCA of the SMART plant. The TASS/SMR code can be applied to any plant regardless of the structural characteristics of a reactor since the code solves the same governing equations for both the primary and secondary system. The code has been developed to meet the requirements of the safety analysis code. This report describes the overall structure of the TASS/SMR, input processing, and the processes of a steady state and transient calculations. In addition, basic differential equations, finite difference equations, state relationships, and constitutive models are described in the report. First, the conservation equations, a discretization process for numerical analysis, search method for state relationship are described. Then, a core power model, heat transfer models, physical models for various components, and control and trip models are explained.
40 CFR 194.23 - Models and computer codes.
2010-07-01
... 40 Protection of Environment 24 2010-07-01 2010-07-01 false Models and computer codes. 194.23... General Requirements § 194.23 Models and computer codes. (a) Any compliance application shall include: (1... obtain stable solutions; (iv) Computer models accurately implement the numerical models; i.e., computer...
Influential input parameters for reflood model of MARS code
Energy Technology Data Exchange (ETDEWEB)
Oh, Deog Yeon; Bang, Young Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2012-10-15
Best Estimate (BE) calculation has been more broadly used in nuclear industries and regulations to reduce the significant conservatism for evaluating Loss of Coolant Accident (LOCA). Reflood model has been identified as one of the problems in BE calculation. The objective of the Post BEMUSE Reflood Model Input Uncertainty Methods (PREMIUM) program of OECD/NEA is to make progress the issue of the quantification of the uncertainty of the physical models in system thermal hydraulic codes, by considering an experimental result especially for reflood. It is important to establish a methodology to identify and select the parameters influential to the response of reflood phenomena following Large Break LOCA. For this aspect, a reference calculation and sensitivity analysis to select the dominant influential parameters for FEBA experiment are performed.
Modeling the PUSPATI TRIGA Reactor using MCNP code
International Nuclear Information System (INIS)
Mohamad Hairie Rabir; Mark Dennis Usang; Naim Syauqi Hamzah; Julia Abdul Karim; Mohd Amin Sharifuldin Salleh
2012-01-01
The 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution and depletion study of TRIGA fuel. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core and shielding with literally no physical approximation. (author)
International Nuclear Information System (INIS)
Pijlgroms, B.J.; Oppe, J.; Oudshoorn, H.L.; Slobben, J.
1991-06-01
A brief description is given of the PASC-3 (Petten-AMPX-SCALE) Reactor Physics code system and associated UNIPASC work environment. The PASC-3 code system is used for criticality and reactor calculations and consists of a selection from the Oak Ridge National Laboratory AMPX-SCALE-3 code collection complemented with a number of additional codes and nuclear data bases. The original codes have been adapted to run under the UNIX operating system. The recommended nuclear data base is a complete 219 group cross section library derived from JEF-1 of which some benchmark results are presented. By the addition of the UNIPASC work environment the usage of the code system is greatly simplified. Complex chains of programs can easily be coupled together to form a single job. In addition, the model parameters can be represented by variables instead of literal values which enhances the readability and may improve the integrity of the code inputs. (author). 8 refs.; 6 figs.; 1 tab
Fission Product Transport Models Adopted in REFPAC Code for LOCA Conditions in PWR and WWER NPPS
International Nuclear Information System (INIS)
Strupczewski, A.
2003-01-01
The report presents assumptions and physical models used for calculations of fission product releases from nuclear reactors, their behavior inside the containment and leakages to the environment after large break loss of coolant accident LB LOCA. They are the basis of code REFPAC (RElease of Fission Products under Accident Conditions), designed primarily to represent significant physical processes occurring after LB LOCA. The code describes these processes using three different models. Model 1 corresponds to established US and Russian practice, Model 2 includes all conservative assumptions that are in agreement with the actual state-of-the-art, and Model 3 incorporates formulae and parameter values actually used in EU practice. (author)
A statistical–mechanical view on source coding: physical compression and data compression
International Nuclear Information System (INIS)
Merhav, Neri
2011-01-01
We draw a certain analogy between the classical information-theoretic problem of lossy data compression (source coding) of memoryless information sources and the statistical–mechanical behavior of a certain model of a chain of connected particles (e.g. a polymer) that is subjected to a contracting force. The free energy difference pertaining to such a contraction turns out to be proportional to the rate-distortion function in the analogous data compression model, and the contracting force is proportional to the derivative of this function. Beyond the fact that this analogy may be interesting in its own right, it may provide a physical perspective on the behavior of optimum schemes for lossy data compression (and perhaps also an information-theoretic perspective on certain physical system models). Moreover, it triggers the derivation of lossy compression performance for systems with memory, using analysis tools and insights from statistical mechanics
Energy Technology Data Exchange (ETDEWEB)
Brannon, R.M.; Wong, M.K.
1996-08-01
A set of model interface guidelines, called MIG, is presented as a means by which any compliant numerical material model can be rapidly installed into any parent code without having to modify the model subroutines. Here, {open_quotes}model{close_quotes} usually means a material model such as one that computes stress as a function of strain, though the term may be extended to any numerical operation. {open_quotes}Parent code{close_quotes} means a hydrocode, finite element code, etc. which uses the model and enforces, say, the fundamental laws of motion and thermodynamics. MIG requires the model developer (who creates the model package) to specify model needs in a standardized but flexible way. MIG includes a dictionary of technical terms that allows developers and parent code architects to share a common vocabulary when specifying field variables. For portability, database management is the responsibility of the parent code. Input/output occurs via structured calling arguments. As much model information as possible (such as the lists of required inputs, as well as lists of precharacterized material data and special needs) is supplied by the model developer in an ASCII text file. Every MIG-compliant model also has three required subroutines to check data, to request extra field variables, and to perform model physics. To date, the MIG scheme has proven flexible in beta installations of a simple yield model, plus a more complicated viscodamage yield model, three electromechanical models, and a complicated anisotropic microcrack constitutive model. The MIG yield model has been successfully installed using identical subroutines in three vectorized parent codes and one parallel C++ code, all predicting comparable results. By maintaining one model for many codes, MIG facilitates code-to-code comparisons and reduces duplication of effort, thereby reducing the cost of installing and sharing models in diverse new codes.
The physical closure laws in the CATHARE code
International Nuclear Information System (INIS)
Bestion, D.
1990-01-01
CATHARE is a 2-fluid thermal-hydraulic code capable of simulating thermal and mechanical phenomena occurring in the primary and secondary circuits of PWRs for a wide variety of accidental situations. The description of the flow is essentially 1-dimensional. Closure laws concerning mass, momentum and energy exchanges between phases and between each phase and the walls are required. A set of specifically designed separate effect experiments were performed and analysed. Having regard for some development principles, correlations are established on the basis of experimental data. The mechanical transfer laws are derived first from experiments where thermal non equilibrium is negligible. Using them as a basis for further interpretation of experimental data, interfacial heat transfer laws are then developed. Wall heat transfer correlations then have to be fixed. All these steps are presented with emphasis being placed on the most recent developments. These last investigations concern the direct contact condensation, stratification model, wall friction, droplet break up and the scale effect, geometrical effect and pressure effect on interfacial friction. (orig.)
Noise Residual Learning for Noise Modeling in Distributed Video Coding
DEFF Research Database (Denmark)
Luong, Huynh Van; Forchhammer, Søren
2012-01-01
Distributed video coding (DVC) is a coding paradigm which exploits the source statistics at the decoder side to reduce the complexity at the encoder. The noise model is one of the inherently difficult challenges in DVC. This paper considers Transform Domain Wyner-Ziv (TDWZ) coding and proposes...
International Nuclear Information System (INIS)
Scannapieco, A.J.; Cranfill, C.W.
1978-11-01
There now exists an inertial confinement stability code called DOC, which runs as a postprocessor. DOC (a code that has evolved from a previous code, PANSY) is a spherical harmonic linear stability code that integrates, in time, a set of Lagrangian perturbation equations. Effects due to real equations of state, asymmetric energy deposition, thermal conduction, shock propagation, and a time-dependent zeroth-order state are handled in the code. We present here a detailed derivation of the physical equations that are solved in the code
Energy Technology Data Exchange (ETDEWEB)
Scannapieco, A.J.; Cranfill, C.W.
1978-11-01
There now exists an inertial confinement stability code called DOC, which runs as a postprocessor. DOC (a code that has evolved from a previous code, PANSY) is a spherical harmonic linear stability code that integrates, in time, a set of Lagrangian perturbation equations. Effects due to real equations of state, asymmetric energy deposition, thermal conduction, shock propagation, and a time-dependent zeroth-order state are handled in the code. We present here a detailed derivation of the physical equations that are solved in the code.
Development of Teaching Materials for a Physical Chemistry Experiment Using the QR Code
吉村, 忠与志
2008-01-01
The development of teaching materials with the QR code was attempted in an educational environment using a mobile telephone. The QR code is not sufficiently utilized in education, and the current study is one of the first in the field. The QR code is encrypted. However, the QR code can be deciphered by mobile telephones, thus enabling the expression of text in a small space.Contents of "Physical Chemistry Experiment" which are available on the Internet are briefly summarized and simplified. T...
Fuel analysis code FAIR and its high burnup modelling capabilities
International Nuclear Information System (INIS)
Prasad, P.S.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.
1995-01-01
A computer code FAIR has been developed for analysing performance of water cooled reactor fuel pins. It is capable of analysing high burnup fuels. This code has recently been used for analysing ten high burnup fuel rods irradiated at Halden reactor. In the present paper, the code FAIR and its various high burnup models are described. The performance of code FAIR in analysing high burnup fuels and its other applications are highlighted. (author). 21 refs., 12 figs
The analysis of thermal-hydraulic models in MELCOR code
Energy Technology Data Exchange (ETDEWEB)
Kim, M H; Hur, C; Kim, D K; Cho, H J [POhang Univ., of Science and TECHnology, Pohang (Korea, Republic of)
1996-07-15
The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed.
Calculation codes in radiation protection, radiation physics and dosimetry
International Nuclear Information System (INIS)
2003-01-01
These scientific days had for objective to draw up the situation of calculation codes of radiation transport, of sources estimation, of radiation doses managements and to draw the future perspectives. (N.C.)
A theory manual for multi-physics code coupling in LIME.
Energy Technology Data Exchange (ETDEWEB)
Belcourt, Noel; Bartlett, Roscoe Ainsworth; Pawlowski, Roger Patrick; Schmidt, Rodney Cannon; Hooper, Russell Warren
2011-03-01
The Lightweight Integrating Multi-physics Environment (LIME) is a software package for creating multi-physics simulation codes. Its primary application space is when computer codes are currently available to solve different parts of a multi-physics problem and now need to be coupled with other such codes. In this report we define a common domain language for discussing multi-physics coupling and describe the basic theory associated with multiphysics coupling algorithms that are to be supported in LIME. We provide an assessment of coupling techniques for both steady-state and time dependent coupled systems. Example couplings are also demonstrated.
Tardos fingerprinting codes in the combined digit model
Skoric, B.; Katzenbeisser, S.; Schaathun, H.G.; Celik, M.U.
2009-01-01
We introduce a new attack model for collusion-secure codes, called the combined digit model, which represents signal processing attacks against the underlying watermarking level better than existing models. In this paper, we analyze the performance of two variants of the Tardos code and show that
Once-through CANDU reactor models for the ORIGEN2 computer code
International Nuclear Information System (INIS)
Croff, A.G.; Bjerke, M.A.
1980-11-01
Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % 235 U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given
Interfacial and Wall Transport Models for SPACE-CAP Code
International Nuclear Information System (INIS)
Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul; Choi, Hoon; Ha, Sang Jun
2009-01-01
The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code
Interfacial and Wall Transport Models for SPACE-CAP Code
Energy Technology Data Exchange (ETDEWEB)
Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech., Seoul (Korea, Republic of); Choi, Hoon; Ha, Sang Jun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)
2009-10-15
The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code.
MARS code manual volume I: code structure, system models, and solution methods
International Nuclear Information System (INIS)
Chung, Bub Dong; Kim, Kyung Doo; Bae, Sung Won; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu; Yoon, Churl
2010-02-01
Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This theory manual provides a complete list of overall information of code structure and major function of MARS including code architecture, hydrodynamic model, heat structure, trip / control system and point reactor kinetics model. Therefore, this report would be very useful for the code users. The overall structure of the manual is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible
HTR core physics and transient analyses by the Panthermix code system
Energy Technology Data Exchange (ETDEWEB)
Haas, J.B.M. de; Kuijper, J.C.; Oppe, J. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)
2005-07-01
At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes.
HTR core physics and transient analyses by the Panthermix code system
International Nuclear Information System (INIS)
Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.
2005-01-01
At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes
Computation of Thermodynamic Equilibria Pertinent to Nuclear Materials in Multi-Physics Codes
Piro, Markus Hans Alexander
Nuclear energy plays a vital role in supporting electrical needs and fulfilling commitments to reduce greenhouse gas emissions. Research is a continuing necessity to improve the predictive capabilities of fuel behaviour in order to reduce costs and to meet increasingly stringent safety requirements by the regulator. Moreover, a renewed interest in nuclear energy has given rise to a "nuclear renaissance" and the necessity to design the next generation of reactors. In support of this goal, significant research efforts have been dedicated to the advancement of numerical modelling and computational tools in simulating various physical and chemical phenomena associated with nuclear fuel behaviour. This undertaking in effect is collecting the experience and observations of a past generation of nuclear engineers and scientists in a meaningful way for future design purposes. There is an increasing desire to integrate thermodynamic computations directly into multi-physics nuclear fuel performance and safety codes. A new equilibrium thermodynamic solver is being developed with this matter as a primary objective. This solver is intended to provide thermodynamic material properties and boundary conditions for continuum transport calculations. There are several concerns with the use of existing commercial thermodynamic codes: computational performance; limited capabilities in handling large multi-component systems of interest to the nuclear industry; convenient incorporation into other codes with quality assurance considerations; and, licensing entanglements associated with code distribution. The development of this software in this research is aimed at addressing all of these concerns. The approach taken in this work exploits fundamental principles of equilibrium thermodynamics to simplify the numerical optimization equations. In brief, the chemical potentials of all species and phases in the system are constrained by estimates of the chemical potentials of the system
Improving the quality of clinical coding: a comprehensive audit model
Directory of Open Access Journals (Sweden)
Hamid Moghaddasi
2014-04-01
Full Text Available Introduction: The review of medical records with the aim of assessing the quality of codes has long been conducted in different countries. Auditing medical coding, as an instructive approach, could help to review the quality of codes objectively using defined attributes, and this in turn would lead to improvement of the quality of codes. Method: The current study aimed to present a model for auditing the quality of clinical codes. The audit model was formed after reviewing other audit models, considering their strengths and weaknesses. A clear definition was presented for each quality attribute and more detailed criteria were then set for assessing the quality of codes. Results: The audit tool (based on the quality attributes included legibility, relevancy, completeness, accuracy, definition and timeliness; led to development of an audit model for assessing the quality of medical coding. Delphi technique was then used to reassure the validity of the model. Conclusion: The inclusive audit model designed could provide a reliable and valid basis for assessing the quality of codes considering more quality attributes and their clear definition. The inter-observer check suggested in the method of auditing is of particular importance to reassure the reliability of coding.
Physical models for high burnup fuel
International Nuclear Information System (INIS)
Kanyukova, V.; Khoruzhii, O.; Likhanskii, V.; Solodovnikov, G.; Sorokin, A.
2003-01-01
In this paper some models of processes in high burnup fuel developed in Src of Russia Troitsk Institute for Innovation and Fusion Research are presented. The emphasis is on the description of the degradation of the fuel heat conductivity, radial profiles of the burnup and the plutonium accumulation, restructuring of the pellet rim, mechanical pellet-cladding interaction. The results demonstrate the possibility of rather accurate description of the behaviour of the fuel of high burnup on the base of simplified models in frame of the fuel performance code if the models are physically ground. The development of such models requires the performance of the detailed physical analysis to serve as a test for a correct choice of allowable simplifications. This approach was applied in the SRC of Russia TRINITI to develop a set of models for the WWER fuel resulting in high reliability of predictions in simulation of the high burnup fuel
Repairing business process models as retrieved from source code
Fernández-Ropero, M.; Reijers, H.A.; Pérez-Castillo, R.; Piattini, M.; Nurcan, S.; Proper, H.A.; Soffer, P.; Krogstie, J.; Schmidt, R.; Halpin, T.; Bider, I.
2013-01-01
The static analysis of source code has become a feasible solution to obtain underlying business process models from existing information systems. Due to the fact that not all information can be automatically derived from source code (e.g., consider manual activities), such business process models
2005-01-01
Tech-X Corporation releases simulation code for solving complex problems in plasma physics : VORPAL code provides a robust environment for simulating plasma processes in high-energy physics, IC fabrications and material processing applications
Content Coding of Psychotherapy Transcripts Using Labeled Topic Models.
Gaut, Garren; Steyvers, Mark; Imel, Zac E; Atkins, David C; Smyth, Padhraic
2017-03-01
Psychotherapy represents a broad class of medical interventions received by millions of patients each year. Unlike most medical treatments, its primary mechanisms are linguistic; i.e., the treatment relies directly on a conversation between a patient and provider. However, the evaluation of patient-provider conversation suffers from critical shortcomings, including intensive labor requirements, coder error, nonstandardized coding systems, and inability to scale up to larger data sets. To overcome these shortcomings, psychotherapy analysis needs a reliable and scalable method for summarizing the content of treatment encounters. We used a publicly available psychotherapy corpus from Alexander Street press comprising a large collection of transcripts of patient-provider conversations to compare coding performance for two machine learning methods. We used the labeled latent Dirichlet allocation (L-LDA) model to learn associations between text and codes, to predict codes in psychotherapy sessions, and to localize specific passages of within-session text representative of a session code. We compared the L-LDA model to a baseline lasso regression model using predictive accuracy and model generalizability (measured by calculating the area under the curve (AUC) from the receiver operating characteristic curve). The L-LDA model outperforms the lasso logistic regression model at predicting session-level codes with average AUC scores of 0.79, and 0.70, respectively. For fine-grained level coding, L-LDA and logistic regression are able to identify specific talk-turns representative of symptom codes. However, model performance for talk-turn identification is not yet as reliable as human coders. We conclude that the L-LDA model has the potential to be an objective, scalable method for accurate automated coding of psychotherapy sessions that perform better than comparable discriminative methods at session-level coding and can also predict fine-grained codes.
Experimental benchmark of non-local-thermodynamic-equilibrium plasma atomic physics codes
International Nuclear Information System (INIS)
Nagels-Silvert, V.
2004-09-01
The main purpose of this thesis is to get experimental data for the testing and validation of atomic physics codes dealing with non-local-thermodynamical-equilibrium plasmas. The first part is dedicated to the spectroscopic study of xenon and krypton plasmas that have been produced by a nanosecond laser pulse interacting with a gas jet. A Thomson scattering diagnostic has allowed us to measure independently plasma parameters such as electron temperature, electron density and the average ionisation state. We have obtained time integrated spectra in the range between 5 and 10 angstroms. We have identified about one hundred xenon rays between 8.6 and 9.6 angstroms via the use of the Relac code. We have discovered unknown rays for the krypton between 5.2 and 7.5 angstroms. In a second experiment we have extended the wavelength range to the X UV domain. The Averroes/Transpec code has been tested in the ranges from 9 to 15 angstroms and from 10 to 130 angstroms, the first range has been well reproduced while the second range requires a more complex data analysis. The second part is dedicated to the spectroscopic study of aluminium, selenium and samarium plasmas in femtosecond operating rate. We have designed an interferometry diagnostic in the frequency domain that has allowed us to measure the expanding speed of the target's backside. Via the use of an adequate isothermal model this parameter has led us to know the plasma electron temperature. Spectra and emission times of various rays from the aluminium and selenium plasmas have been computed satisfactorily with the Averroes/Transpec code coupled with Film and Multif hydrodynamical codes. (A.C.)
GENII-LIN: a Multipurpose Health Physics Code Built on GENII-1.485
Directory of Open Access Journals (Sweden)
Marco Sumini
2006-10-01
Full Text Available The aim of the GENII-LIN project was to develop an open source multipurpose health physics code running on Linux platform, for calculating radiation dose and risk from radionuclides released to the environment. The general features of the GENII-LIN system include [1] capabilities for calculating radiation dose both for acute and chronic releases, with options for annual dose, committed dose and accumulated dose [2] capabilities for evaluating exposure pathways including direct exposure via water (swimming, boating, fishing, soil (buried and surface sources and air (semi-infinite cloud and finite cloud model, inhalation pathways and ingestion pathways. The release scenarios considered are: - acute release to air, from ground level or elevated sources, or to water; - chronic release to air, from ground level or elevated sources, or to water; - initial contamination of soil or surfaces. Keywords: radiation protection, Linux, health physics, risk analysis.
WWER radial reflector modeling by diffusion codes
International Nuclear Information System (INIS)
Petkov, P. T.; Mittag, S.
2005-01-01
The two commonly used approaches to describe the WWER radial reflectors in diffusion codes, by albedo on the core-reflector boundary and by a ring of diffusive assembly size nodes, are discussed. The advantages and disadvantages of the first approach are presented first, then the Koebke's equivalence theory is outlined and its implementation for the WWER radial reflectors is discussed. Results for the WWER-1000 reactor are presented. Then the boundary conditions on the outer reflector boundary are discussed. The possibility to divide the library into fuel assembly and reflector parts and to generate each library by a separate code package is discussed. Finally, the homogenization errors for rodded assemblies are presented and discussed (Author)
COMPBRN III: a computer code for modeling compartment fires
International Nuclear Information System (INIS)
Ho, V.; Siu, N.; Apostolakis, G.; Flanagan, G.F.
1986-07-01
The computer code COMPBRN III deterministically models the behavior of compartment fires. This code is an improvement of the original COMPBRN codes. It employs a different air entrainment model and numerical scheme to estimate properties of the ceiling hot gas layer model. Moreover, COMPBRN III incorporates a number of improvements in shape factor calculations and error checking, which distinguish it from the COMPBRN II code. This report presents the ceiling hot gas layer model employed by COMPBRN III as well as several other modifications. Information necessary to run COMPBRN III, including descriptions of required input and resulting output, are also presented. Simulation of experiments and a sample problem are included to demonstrate the usage of the code. 37 figs., 46 refs
DEFF Research Database (Denmark)
Taasti, Vicki Trier; Knudsen, Helge; Holzscheiter, Michael
2015-01-01
The Monte Carlo particle transport code SHIELD-HIT12A is designed to simulate therapeutic beams for cancer radiotherapy with fast ions. SHIELD-HIT12A allows creation of antiproton beam kernels for the treatment planning system TRiP98, but first it must be benchmarked against experimental data. An...
International Nuclear Information System (INIS)
Darby, J.L.
1986-01-01
The Adversary Sequence Diagram (ASD) concept was developed by Sandia National Laboratories (SNL) to examine physical security system effectiveness. Sandia also developed a mainframe computer code, PANL, to analyze the ASD. The authors have developed a microcomputer code, SEAPATH, which also analyzes ASD's. The Authors are supporting SNL in software development of the SAVI code; SAVI utilizes the SEAPATH algorithm to identify and quantify paths
Plasma simulation studies using multilevel physics models
International Nuclear Information System (INIS)
Park, W.; Belova, E.V.; Fu, G.Y.; Tang, X.Z.; Strauss, H.R.; Sugiyama, L.E.
1999-01-01
The question of how to proceed toward ever more realistic plasma simulation studies using ever increasing computing power is addressed. The answer presented here is the M3D (Multilevel 3D) project, which has developed a code package with a hierarchy of physics levels that resolve increasingly complete subsets of phase-spaces and are thus increasingly more realistic. The rationale for the multilevel physics models is given. Each physics level is described and examples of its application are given. The existing physics levels are fluid models (3D configuration space), namely magnetohydrodynamic (MHD) and two-fluids; and hybrid models, namely gyrokinetic-energetic-particle/MHD (5D energetic particle phase-space), gyrokinetic-particle-ion/fluid-electron (5D ion phase-space), and full-kinetic-particle-ion/fluid-electron level (6D ion phase-space). Resolving electron phase-space (5D or 6D) remains a future project. Phase-space-fluid models are not used in favor of δf particle models. A practical and accurate nonlinear fluid closure for noncollisional plasmas seems not likely in the near future. copyright 1999 American Institute of Physics
Standard interface files and procedures for reactor physics codes. Version IV
International Nuclear Information System (INIS)
O'Dell, R.D.
1977-09-01
Standards, procedures, and recommendations of the Committee on Computer Code Coordination for promoting the exchange of reactor physics codes are updated to Version IV status. Standards and procedures covering general programming, program structure, standard interface files, and file management and handling subroutines are included
Impurity seeding in ASDEX upgrade tokamak modeled by COREDIV code
Energy Technology Data Exchange (ETDEWEB)
Galazka, K.; Ivanova-Stanik, I.; Czarnecka, A.; Zagoerski, R. [Institute of Plasma Physics and Laser Microfusion, Warsaw (Poland); Bernert, M.; Kallenbach, A. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); Collaboration: ASDEX Upgrade Team
2016-08-15
The self-consistent COREDIV code is used to simulate discharges in a tokamak plasma, especially the influence of impurities during nitrogen and argon seeding on the key plasma parameters. The calculations are performed with and without taking into account the W prompt redeposition in the divertor area and are compared to the experimental results acquired on ASDEX Upgrade tokamak (shots 29254 and 29257). For both impurities the modeling shows a better agreement with the experiment in the case without prompt redeposition. It is attributed to higher average tungsten concentration, which on the other hand seriously exceeds the experimental value. By turning the prompt redeposition process on, the W concentration is lowered, what, in turn, results in underestimation of the radiative power losses. By analyzing the influence of the transport coefficients on the radiative power loss and average W concentration it is concluded that the way to compromise the opposing tendencies is to include the edge-localized mode flushing mechanism into the code, which dominates the experimental particle and energy balance. Also performing the calculations with both anomalous and neoclassical diffusion transport mechanisms included is suggested. (copyright 2016 The Authors. Contributions to Plasma Physics published by Wiley-VCH Verlag GmbH and Co. KGaA Weinheim. This)
RELAP5/MOD3 code manual: Code structure, system models, and solution methods. Volume 1
International Nuclear Information System (INIS)
1995-08-01
The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling, approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. RELAP5/MOD3 code documentation is divided into seven volumes: Volume I provides modeling theory and associated numerical schemes
CORESAFE: A Formal Approach against Code Replacement Attacks on Cyber Physical Systems
2018-04-19
AFRL-AFOSR-JP-TR-2018-0035 CORESAFE:A Formal Approach against Code Replacement Attacks on Cyber Physical Systems Sandeep Shukla INDIAN INSTITUTE OF...Formal Approach against Code Replacement Attacks on Cyber Physical Systems 5a. CONTRACT NUMBER 5b. GRANT NUMBER FA2386-16-1-4099 5c. PROGRAM ELEMENT...SUPPLEMENTARY NOTES 14. ABSTRACT Industrial Control Systems (ICS) used in manufacturing, power generators and other critical infrastructure monitoring and
Light water reactor fuel analysis code FEMAXI-7; model and structure
International Nuclear Information System (INIS)
Suzuki, Motoe; Udagawa, Yutaka; Saitou, Hiroaki
2011-03-01
A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report describes in detail the design, basic theory and structure, models and numerical method, and improvements and extensions. (author)
Light water reactor fuel analysis code FEMAXI-7. Model and structure
International Nuclear Information System (INIS)
Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa; Saitou, Hiroaki
2013-07-01
A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report describes in detail the design, basic theory and structure, models and numerical method of FEMAXI-7, and its improvements and extensions. (author)
Radiation transport phenomena and modeling. Part A: Codes; Part B: Applications with examples
International Nuclear Information System (INIS)
Lorence, L.J. Jr.; Beutler, D.E.
1997-09-01
This report contains the notes from the second session of the 1997 IEEE Nuclear and Space Radiation Effects Conference Short Course on Applying Computer Simulation Tools to Radiation Effects Problems. Part A discusses the physical phenomena modeled in radiation transport codes and various types of algorithmic implementations. Part B gives examples of how these codes can be used to design experiments whose results can be easily analyzed and describes how to calculate quantities of interest for electronic devices
The MCUCN simulation code for ultracold neutron physics
Zsigmond, G.
2018-02-01
Ultracold neutrons (UCN) have very low kinetic energies 0-300 neV, thereby can be stored in specific material or magnetic confinements for many hundreds of seconds. This makes them a very useful tool in probing fundamental symmetries of nature (for instance charge-parity violation by neutron electric dipole moment experiments) and contributing important parameters for the Big Bang nucleosynthesis (neutron lifetime measurements). Improved precision experiments are in construction at new and planned UCN sources around the world. MC simulations play an important role in the optimization of such systems with a large number of parameters, but also in the estimation of systematic effects, in benchmarking of analysis codes, or as part of the analysis. The MCUCN code written at PSI has been extensively used for the optimization of the UCN source optics and in the optimization and analysis of (test) experiments within the nEDM project based at PSI. In this paper we present the main features of MCUCN and interesting benchmark and application examples.
Calculation codes in radioprotection, radio-physics and dosimetry
International Nuclear Information System (INIS)
Jan, S.; Laedermann, J.P.; Bochud, F.; Ferragut, A.; Bordy, J.M.; Parisi, L.L.; Abou-Khalil, R.; Longeot, M.; Kitsos, S.; Groetz, J.E.; Villagrasa, C.; Daures, J.; Martin, E.; Henriet, J.; Tsilanizara, A.; Farah, J.; Uyttenhove, W.; Perrot, Y.; De Carlan, L.; Vivier, A.; Kodeli, I.; Sayah, R.; Hadid, L.; Courageot, E.; Fritsch, P.; Davesne, E.; Michel, X.
2010-01-01
This document gathers the slides of the available presentations given during these conference days. Twenty seven presentations are assembled in the document and deal with: 1 - GATE: calculation code for medical imaging, radiotherapy and dosimetry (S. Jan); 2 - estimation of conversion factors for the measurement of the ambient dose equivalent rate by in-situ spectroscopy (J.P. Laedermann); 3 - geometry specific calibration factors for nuclear medicine activity meters (F. Bochud); 4 - Monte Carlo simulation of a rare gases measurement system - calculation and validation, ASGA/VGM system (A. Ferragut); 5 - design of a realistic radiation field for the calibration of the dosemeters used in interventional radiology/cardiology (medical personnel dosimetry) (J.M. Bordy); 6 - determination of the position and height of the KALINA facility chimney at CEA Cadarache (L.L. Parisi); 7 - MERCURAD TM - 3D simulation software for dose rates calculation (R. Abou-Khalil); 8 - PANTHERE - 3D software for gamma dose rates simulation of complex nuclear facilities (M. Longeot); 9 - radioprotection, from the design to the exploitation of radioactive materials transportation containers (S. Kitsos); 10 - post-simulation processing of MCNPX responses in neutron spectroscopy (J.E. Groetz); 11 - last developments of the Geant4 Monte Carlo code for trace amounts simulation in liquid water at the molecular scale (C. Villagrasa); 12 - Calculation of H p (3)/K air conversion coefficients using PENELOPE Monte-Carlo code and comparison with MCNP calculation results (J. Daures); 13 - artificial neural networks, a new alternative to Monte Carlo calculations for radiotherapy (E. Martin); 14 - use of case-based reasoning for the reconstruction and handling of voxelized fantoms (J. Henriet); 15 - resolution of the radioactive decay inverse problem for dose calculation in radioprotection (A. Tsilanizara); 16 - use of NURBS-type fantoms for the study of the morphological factors influencing the pulmonary
Fuel behavior modeling using the MARS computer code
International Nuclear Information System (INIS)
Faya, S.C.S.; Faya, A.J.G.
1983-01-01
The fuel behaviour modeling code MARS against experimental data, was evaluated. Two cases were selected: an early comercial PWR rod (Maine Yankee rod) and an experimental rod from the Canadian BWR program (Canadian rod). The MARS predictions are compared with experimental data and predictions made by other fuel modeling codes. Improvements are suggested for some fuel behaviour models. Mars results are satisfactory based on the data available. (Author) [pt
MIDAS/PK code development using point kinetics model
International Nuclear Information System (INIS)
Song, Y. M.; Park, S. H.
1999-01-01
In this study, a MIDAS/PK code has been developed for analyzing the ATWS (Anticipated Transients Without Scram) which can be one of severe accident initiating events. The MIDAS is an integrated computer code based on the MELCOR code to develop a severe accident risk reduction strategy by Korea Atomic Energy Research Institute. In the mean time, the Chexal-Layman correlation in the current MELCOR, which was developed under a BWR condition, is appeared to be inappropriate for a PWR. So as to provide ATWS analysis capability to the MIDAS code, a point kinetics module, PKINETIC, has first been developed as a stand-alone code whose reference model was selected from the current accident analysis codes. In the next step, the MIDAS/PK code has been developed via coupling PKINETIC with the MIDAS code by inter-connecting several thermal hydraulic parameters between the two codes. Since the major concern in the ATWS analysis is the primary peak pressure during the early few minutes into the accident, the peak pressure from the PKINETIC module and the MIDAS/PK are compared with the RETRAN calculations showing a good agreement between them. The MIDAS/PK code is considered to be valuable for analyzing the plant response during ATWS deterministically, especially for the early domestic Westinghouse plants which rely on the operator procedure instead of an AMSAC (ATWS Mitigating System Actuation Circuitry) against ATWS. This capability of ATWS analysis is also important from the view point of accident management and mitigation
Modeling Vortex Generators in a Navier-Stokes Code
Dudek, Julianne C.
2011-01-01
A source-term model that simulates the effects of vortex generators was implemented into the Wind-US Navier-Stokes code. The source term added to the Navier-Stokes equations simulates the lift force that would result from a vane-type vortex generator in the flowfield. The implementation is user-friendly, requiring the user to specify only three quantities for each desired vortex generator: the range of grid points over which the force is to be applied and the planform area and angle of incidence of the physical vane. The model behavior was evaluated for subsonic flow in a rectangular duct with a single vane vortex generator, subsonic flow in an S-duct with 22 corotating vortex generators, and supersonic flow in a rectangular duct with a counter-rotating vortex-generator pair. The model was also used to successfully simulate microramps in supersonic flow by treating each microramp as a pair of vanes with opposite angles of incidence. The validation results indicate that the source-term vortex-generator model provides a useful tool for screening vortex-generator configurations and gives comparable results to solutions computed using gridded vanes.
Coupling a Basin Modeling and a Seismic Code using MOAB
Yan, Mi; Jordan, Kirk; Kaushik, Dinesh; Perrone, Michael; Sachdeva, Vipin; Tautges, Timothy J.; Magerlein, John
2012-01-01
We report on a demonstration of loose multiphysics coupling between a basin modeling code and a seismic code running on a large parallel machine. Multiphysics coupling, which is one critical capability for a high performance computing (HPC) framework, was implemented using the MOAB open-source mesh and field database. MOAB provides for code coupling by storing mesh data and input and output field data for the coupled analysis codes and interpolating the field values between different meshes used by the coupled codes. We found it straightforward to use MOAB to couple the PBSM basin modeling code and the FWI3D seismic code on an IBM Blue Gene/P system. We describe how the coupling was implemented and present benchmarking results for up to 8 racks of Blue Gene/P with 8192 nodes and MPI processes. The coupling code is fast compared to the analysis codes and it scales well up to at least 8192 nodes, indicating that a mesh and field database is an efficient way to implement loose multiphysics coupling for large parallel machines.
Coupling a Basin Modeling and a Seismic Code using MOAB
Yan, Mi
2012-06-02
We report on a demonstration of loose multiphysics coupling between a basin modeling code and a seismic code running on a large parallel machine. Multiphysics coupling, which is one critical capability for a high performance computing (HPC) framework, was implemented using the MOAB open-source mesh and field database. MOAB provides for code coupling by storing mesh data and input and output field data for the coupled analysis codes and interpolating the field values between different meshes used by the coupled codes. We found it straightforward to use MOAB to couple the PBSM basin modeling code and the FWI3D seismic code on an IBM Blue Gene/P system. We describe how the coupling was implemented and present benchmarking results for up to 8 racks of Blue Gene/P with 8192 nodes and MPI processes. The coupling code is fast compared to the analysis codes and it scales well up to at least 8192 nodes, indicating that a mesh and field database is an efficient way to implement loose multiphysics coupling for large parallel machines.
Status of computer codes available in AEOI for reactor physics analysis
International Nuclear Information System (INIS)
Karbassiafshar, M.
1986-01-01
Many of the nuclear computer codes available in Atomic Energy Organization of Iran AEOI can be used for physics analysis of an operating reactor or design purposes. Grasp of the various methods involved and practical experience with these codes would be the starting point for interesting design studies or analysis of operating conditions of presently existing and future reactors. A review of the objectives and flowchart of commonly practiced procedures in reactor physics analysis of LWRs and related computer codes was made, extrapolating to the nationally and internationally available resources. Finally, effective utilization of the existing facilities is discussed and called upon
EM modeling for GPIR using 3D FDTD modeling codes
Energy Technology Data Exchange (ETDEWEB)
Nelson, S.D.
1994-10-01
An analysis of the one-, two-, and three-dimensional electrical characteristics of structural cement and concrete is presented. This work connects experimental efforts in characterizing cement and concrete in the frequency and time domains with the Finite Difference Time Domain (FDTD) modeling efforts of these substances. These efforts include Electromagnetic (EM) modeling of simple lossless homogeneous materials with aggregate and targets and the modeling dispersive and lossy materials with aggregate and complex target geometries for Ground Penetrating Imaging Radar (GPIR). Two- and three-dimensional FDTD codes (developed at LLNL) where used for the modeling efforts. Purpose of the experimental and modeling efforts is to gain knowledge about the electrical properties of concrete typically used in the construction industry for bridges and other load bearing structures. The goal is to optimize the performance of a high-sample-rate impulse radar and data acquisition system and to design an antenna system to match the characteristics of this material. Results show agreement to within 2 dB of the amplitudes of the experimental and modeled data while the frequency peaks correlate to within 10% the differences being due to the unknown exact nature of the aggregate placement.
Subgroup A: nuclear model codes report to the Sixteenth Meeting of the WPEC
International Nuclear Information System (INIS)
Talou, P.; Chadwick, M.B.; Dietrich, F.S.; Herman, M.; Kawano, T.; Konig, A.; Oblozinsky, P.
2004-01-01
The Subgroup A activities focus on the development of nuclear reaction models and codes, used in evaluation work for nuclear reactions from the unresolved energy region up to the pion threshold production limit, and for target nuclides from the low teens and heavier. Much of the efforts are devoted by each participant to the continuing development of their own Institution codes. Progresses in this arena are reported in detail for each code in the present document. EMPIRE-II is of public access. The release of the TALYS code has been announced for the ND2004 Conference in Santa Fe, NM, October 2004. McGNASH is still under development and is not expected to be released in the very near future. In addition, Subgroup A members have demonstrated a growing interest in working on common modeling and codes capabilities, which would significantly reduce the amount of duplicate work, help manage efficiently the growing lines of existing codes, and render codes inter-comparison much easier. A recent and important activity of the Subgroup A has therefore been to develop the framework and the first bricks of the ModLib library, which is constituted of mostly independent pieces of codes written in Fortran 90 (and above) to be used in existing and future nuclear reaction codes. Significant progresses in the development of ModLib have been made during the past year. Several physics modules have been added to the library, and a few more have been planned in detail for the coming year.
Cavitation Modeling in Euler and Navier-Stokes Codes
Deshpande, Manish; Feng, Jinzhang; Merkle, Charles L.
1993-01-01
Many previous researchers have modeled sheet cavitation by means of a constant pressure solution in the cavity region coupled with a velocity potential formulation for the outer flow. The present paper discusses the issues involved in extending these cavitation models to Euler or Navier-Stokes codes. The approach taken is to start from a velocity potential model to ensure our results are compatible with those of previous researchers and available experimental data, and then to implement this model in both Euler and Navier-Stokes codes. The model is then augmented in the Navier-Stokes code by the inclusion of the energy equation which allows the effect of subcooling in the vicinity of the cavity interface to be modeled to take into account the experimentally observed reduction in cavity pressures that occurs in cryogenic fluids such as liquid hydrogen. Although our goal is to assess the practicality of implementing these cavitation models in existing three-dimensional, turbomachinery codes, the emphasis in the present paper will center on two-dimensional computations, most specifically isolated airfoils and cascades. Comparisons between velocity potential, Euler and Navier-Stokes implementations indicate they all produce consistent predictions. Comparisons with experimental results also indicate that the predictions are qualitatively correct and give a reasonable first estimate of sheet cavitation effects in both cryogenic and non-cryogenic fluids. The impact on CPU time and the code modifications required suggests that these models are appropriate for incorporation in current generation turbomachinery codes.
Physical Modeling Modular Boxes: PHOXES
DEFF Research Database (Denmark)
Gelineck, Steven; Serafin, Stefania
2010-01-01
This paper presents the development of a set of musical instruments, which are based on known physical modeling sound synthesis techniques. The instruments are modular, meaning that they can be combined in various ways. This makes it possible to experiment with physical interaction and sonic...
RELAP5/MOD3 code coupling model
International Nuclear Information System (INIS)
Martin, R.P.; Johnsen, G.W.
1994-01-01
A new capability has been incorporated into RELAP5/MOD3 that enables the coupling of RELAP5/MOD3 to other computer codes. The new capability has been designed to support analysis of the new advanced reactor concepts. Its user features rely solely on new RELAP5 open-quotes styledclose quotes input and the Parallel Virtual Machine (PVM) software, which facilitates process management and distributed communication of multiprocess problems. RELAP5/MOD3 manages the input processing, communication instruction, process synchronization, and its own send and receive data processing. The flexible capability requires that an explicit coupling be established, which updates boundary conditions at discrete time intervals. Two test cases are presented that demonstrate the functionality, applicability, and issues involving use of this capability
Plasma simulation studies using multilevel physics models
International Nuclear Information System (INIS)
Park, W.; Belova, E.V.; Fu, G.Y.
2000-01-01
The question of how to proceed toward ever more realistic plasma simulation studies using ever increasing computing power is addressed. The answer presented here is the M3D (Multilevel 3D) project, which has developed a code package with a hierarchy of physics levels that resolve increasingly complete subsets of phase-spaces and are thus increasingly more realistic. The rationale for the multilevel physics models is given. Each physics level is described and examples of its application are given. The existing physics levels are fluid models (3D configuration space), namely magnetohydrodynamic (MHD) and two-fluids; and hybrid models, namely gyrokinetic-energetic-particle/MHD (5D energetic particle phase-space), gyrokinetic-particle-ion/fluid-electron (5D ion phase-space), and full-kinetic-particle-ion/fluid-electron level (6D ion phase-space). Resolving electron phase-space (5D or 6D) remains a future project. Phase-space-fluid models are not used in favor of delta f particle models. A practical and accurate nonlinear fluid closure for noncollisional plasmas seems not likely in the near future
"SMART": A Compact and Handy FORTRAN Code for the Physics of Stellar Atmospheres
Sapar, A.; Poolamäe, R.
2003-01-01
A new computer code SMART (Spectra from Model Atmospheres by Radiative Transfer) for computing the stellar spectra, forming in plane-parallel atmospheres, has been compiled by us and A. Aret. To guarantee wide compatibility of the code with shell environment, we chose FORTRAN-77 as programming language and tried to confine ourselves to common part of its numerous versions both in WINDOWS and LINUX. SMART can be used for studies of several processes in stellar atmospheres. The current version of the programme is undergoing rapid changes due to our goal to elaborate a simple, handy and compact code. Instead of linearisation (being a mathematical method of recurrent approximations) we propose to use the physical evolutionary changes or in other words relaxation of quantum state populations rates from LTE to NLTE has been studied using small number of NLTE states. This computational scheme is essentially simpler and more compact than the linearisation. This relaxation scheme enables using instead of the Λ-iteration procedure a physically changing emissivity (or the source function) which incorporates in itself changing Menzel coefficients for NLTE quantum state populations. However, the light scattering on free electrons is in the terms of Feynman graphs a real second-order quantum process and cannot be reduced to consequent processes of absorption and emission as in the case of radiative transfer in spectral lines. With duly chosen input parameters the code SMART enables computing radiative acceleration to the matter of stellar atmosphere in turbulence clumps. This also enables to connect the model atmosphere in more detail with the problem of the stellar wind triggering. Another problem, which has been incorporated into the computer code SMART, is diffusion of chemical elements and their isotopes in the atmospheres of chemically peculiar (CP) stars due to usual radiative acceleration and the essential additional acceleration generated by the light-induced drift. As
ABAREX -- A neutron spherical optical-statistical-model code -- A user`s manual
Energy Technology Data Exchange (ETDEWEB)
Smith, A.B. [ed.; Lawson, R.D.
1998-06-01
The contemporary version of the neutron spherical optical-statistical-model code ABAREX is summarized with the objective of providing detailed operational guidance for the user. The physical concepts involved are very briefly outlined. The code is described in some detail and a number of explicit examples are given. With this document one should very quickly become fluent with the use of ABAREX. While the code has operated on a number of computing systems, this version is specifically tailored for the VAX/VMS work station and/or the IBM-compatible personal computer.
Case studies in Gaussian process modelling of computer codes
International Nuclear Information System (INIS)
Kennedy, Marc C.; Anderson, Clive W.; Conti, Stefano; O'Hagan, Anthony
2006-01-01
In this paper we present a number of recent applications in which an emulator of a computer code is created using a Gaussian process model. Tools are then applied to the emulator to perform sensitivity analysis and uncertainty analysis. Sensitivity analysis is used both as an aid to model improvement and as a guide to how much the output uncertainty might be reduced by learning about specific inputs. Uncertainty analysis allows us to reflect output uncertainty due to unknown input parameters, when the finished code is used for prediction. The computer codes themselves are currently being developed within the UK Centre for Terrestrial Carbon Dynamics
Altarelli, Guido
1999-01-01
Introduction structure of gauge theories. The QEDand QCD examples. Chiral theories. The electroweak theory. Spontaneous symmetry breaking. The Higgs mechanism Gauge boson and fermion masses. Yukawa coupling. Charges current couplings. The Cabibo-Kobayashi-Maskawa matrix and CP violation. Neutral current couplings. The Glasow-Iliopoulos-Maiani mechanism. Gauge boson and Higgs coupling. Radiative corrections and loops. Cancellation of the chiral anomaly. Limits on the Higgs comparaison. Problems of the Standard Model. Outlook.
Relativistic modeling capabilities in PERSEUS extended MHD simulation code for HED plasmas
Energy Technology Data Exchange (ETDEWEB)
Hamlin, Nathaniel D., E-mail: nh322@cornell.edu [438 Rhodes Hall, Cornell University, Ithaca, NY, 14853 (United States); Seyler, Charles E., E-mail: ces7@cornell.edu [Cornell University, Ithaca, NY, 14853 (United States)
2014-12-15
We discuss the incorporation of relativistic modeling capabilities into the PERSEUS extended MHD simulation code for high-energy-density (HED) plasmas, and present the latest hybrid X-pinch simulation results. The use of fully relativistic equations enables the model to remain self-consistent in simulations of such relativistic phenomena as X-pinches and laser-plasma interactions. By suitable formulation of the relativistic generalized Ohm’s law as an evolution equation, we have reduced the recovery of primitive variables, a major technical challenge in relativistic codes, to a straightforward algebraic computation. Our code recovers expected results in the non-relativistic limit, and reveals new physics in the modeling of electron beam acceleration following an X-pinch. Through the use of a relaxation scheme, relativistic PERSEUS is able to handle nine orders of magnitude in density variation, making it the first fluid code, to our knowledge, that can simulate relativistic HED plasmas.
Quo vadis code optimization in high energy physics
International Nuclear Information System (INIS)
Jarp, S.
1994-01-01
Although performance tuning and optimization can be considered less critical than in the past, there are still many High Energy Physics (HEP) applications and application domains that can profit from such an undertaking. In CERN's CORE (Centrally Operated RISC Environment) where all major RISC vendors are present, this implies an understanding of the various computer architectures, instruction sets and performance analysis tools from each of these vendors. This paper discusses some initial observations after having evaluated the situation and makes some recommendations for further progress
International Nuclear Information System (INIS)
Peccei, R.D.
1986-01-01
Possible small extensions of the standard model are considered, which are motivated by the strong CP problem and by the baryon asymmetry of the Universe. Phenomenological arguments are given which suggest that imposing a PQ symmetry to solve the strong CP problem is only tenable if the scale of the PQ breakdown is much above M W . Furthermore, an attempt is made to connect the scale of the PQ breakdown to that of the breakdown of lepton number. It is argued that in these theories the same intermediate scale may be responsible for the baryon number of the Universe, provided the Kuzmin Rubakov Shaposhnikov (B+L) erasing mechanism is operative. (orig.)
ATHENA code manual. Volume 1. Code structure, system models, and solution methods
International Nuclear Information System (INIS)
Carlson, K.E.; Roth, P.A.; Ransom, V.H.
1986-09-01
The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code has been developed to perform transient simulation of the thermal hydraulic systems which may be found in fusion reactors, space reactors, and other advanced systems. A generic modeling approach is utilized which permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of a complete facility. Several working fluids are available to be used in one or more interacting loops. Different loops may have different fluids with thermal connections between loops. The modeling theory and associated numerical schemes are documented in Volume I in order to acquaint the user with the modeling base and thus aid effective use of the code. The second volume contains detailed instructions for input data preparation
Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF
Blyth, Taylor S.
The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physics-based Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR) cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermal-hydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code.
Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF
Energy Technology Data Exchange (ETDEWEB)
Blyth, Taylor S. [Pennsylvania State Univ., University Park, PA (United States); Avramova, Maria [North Carolina State Univ., Raleigh, NC (United States)
2017-04-01
The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physics- based Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR) cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermal- hydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code.
Research on V and V strategy of reactor physics code of COSINE
International Nuclear Information System (INIS)
Liu Zhanquan; Chen Yixue; Yang Chao; Dang Halei
2013-01-01
Verification and validation (V and V) is very important for the software quality assurance. Reasonable and efficient V and V strategy can achieve twice the result with half the effort. Core and system integrated engine for design and analysis (COSINE) software package contains three reactor physics codes, the lattice code (LATC), the core simulator (CORE) and the kinetics code (KIND), which is called the reactor physics subsystem. The V and V strategy for the physics subsystem was researched based on the foundation of scientific software's V and V method. The module based verification method and the function based validation method were proposed, composing the physical subsystem V and V strategy of COSINE software package. (authors)
Dilution physics modeling: Dissolution/precipitation chemistry
International Nuclear Information System (INIS)
Onishi, Y.; Reid, H.C.; Trent, D.S.
1995-09-01
This report documents progress made to date on integrating dilution/precipitation chemistry and new physical models into the TEMPEST thermal-hydraulics computer code. Implementation of dissolution/precipitation chemistry models is necessary for predicting nonhomogeneous, time-dependent, physical/chemical behavior of tank wastes with and without a variety of possible engineered remediation and mitigation activities. Such behavior includes chemical reactions, gas retention, solids resuspension, solids dissolution and generation, solids settling/rising, and convective motion of physical and chemical species. Thus this model development is important from the standpoint of predicting the consequences of various engineered activities, such as mitigation by dilution, retrieval, or pretreatment, that can affect safe operations. The integration of a dissolution/precipitation chemistry module allows the various phase species concentrations to enter into the physical calculations that affect the TEMPEST hydrodynamic flow calculations. The yield strength model of non-Newtonian sludge correlates yield to a power function of solids concentration. Likewise, shear stress is concentration-dependent, and the dissolution/precipitation chemistry calculations develop the species concentration evolution that produces fluid flow resistance changes. Dilution of waste with pure water, molar concentrations of sodium hydroxide, and other chemical streams can be analyzed for the reactive species changes and hydrodynamic flow characteristics
ADVANCED ELECTRIC AND MAGNETIC MATERIAL MODELS FOR FDTD ELECTROMAGNETIC CODES
Energy Technology Data Exchange (ETDEWEB)
Poole, B R; Nelson, S D; Langdon, S
2005-05-05
The modeling of dielectric and magnetic materials in the time domain is required for pulse power applications, pulsed induction accelerators, and advanced transmission lines. For example, most induction accelerator modules require the use of magnetic materials to provide adequate Volt-sec during the acceleration pulse. These models require hysteresis and saturation to simulate the saturation wavefront in a multipulse environment. In high voltage transmission line applications such as shock or soliton lines the dielectric is operating in a highly nonlinear regime, which require nonlinear models. Simple 1-D models are developed for fast parameterization of transmission line structures. In the case of nonlinear dielectrics, a simple analytic model describing the permittivity in terms of electric field is used in a 3-D finite difference time domain code (FDTD). In the case of magnetic materials, both rate independent and rate dependent Hodgdon magnetic material models have been implemented into 3-D FDTD codes and 1-D codes.
ADVANCED ELECTRIC AND MAGNETIC MATERIAL MODELS FOR FDTD ELECTROMAGNETIC CODES
International Nuclear Information System (INIS)
Poole, B R; Nelson, S D; Langdon, S
2005-01-01
The modeling of dielectric and magnetic materials in the time domain is required for pulse power applications, pulsed induction accelerators, and advanced transmission lines. For example, most induction accelerator modules require the use of magnetic materials to provide adequate Volt-sec during the acceleration pulse. These models require hysteresis and saturation to simulate the saturation wavefront in a multipulse environment. In high voltage transmission line applications such as shock or soliton lines the dielectric is operating in a highly nonlinear regime, which require nonlinear models. Simple 1-D models are developed for fast parameterization of transmission line structures. In the case of nonlinear dielectrics, a simple analytic model describing the permittivity in terms of electric field is used in a 3-D finite difference time domain code (FDTD). In the case of magnetic materials, both rate independent and rate dependent Hodgdon magnetic material models have been implemented into 3-D FDTD codes and 1-D codes
The physics of compensating calorimetry and the new CALOR89 code system
International Nuclear Information System (INIS)
Gabriel, T.A.; Brau, J.E.; Bishop, B.L.
1989-03-01
Much of the understanding of the physics of calorimetry has come from the use of excellent radiation transport codes. A new understanding of compensating calorimetry was introduced four years ago following detailed studies with a new CALOR system. Now, the CALOR system has again been revised to reflect a better comprehension of high energy nuclear collisions by incorporating a modified high energy fragmentation model from FLUKA87. This revision will allow for the accurate analysis of calorimeters at energies of 100's of GeV. Presented in this paper is a discussion of compensating calorimetry, the new CALOR system, the revisions to HETC, and recently generated calorimeter related data on modes of energy deposition and secondary neutron production (E < 50 MeV) in infinite iron and uranium blocks. 38 refs., 5 figs., 5 tabs
Code Generation for Protocols from CPN models Annotated with Pragmatics
DEFF Research Database (Denmark)
Simonsen, Kent Inge; Kristensen, Lars Michael; Kindler, Ekkart
software implementation satisfies the properties verified for the model. Coloured Petri Nets (CPNs) have been widely used to model and verify protocol software, but limited work exists on using CPN models of protocol software as a basis for automated code generation. In this report, we present an approach...... modelling languages, MDE further has the advantage that models are amenable to model checking which allows key behavioural properties of the software design to be verified. The combination of formally verified models and automated code generation contributes to a high degree of assurance that the resulting...... for generating protocol software from a restricted class of CPN models. The class of CPN models considered aims at being descriptive in that the models are intended to be helpful in understanding and conveying the operation of the protocol. At the same time, a descriptive model is close to a verifiable version...
Fusion safety codes International modeling with MELCOR and ATHENA- INTRA
Marshall, T; Topilski, L; Merrill, B
2002-01-01
For a number of years, the world fusion safety community has been involved in benchmarking their safety analyses codes against experiment data to support regulatory approval of a next step fusion device. This paper discusses the benchmarking of two prominent fusion safety thermal-hydraulic computer codes. The MELCOR code was developed in the US for fission severe accident safety analyses and has been modified for fusion safety analyses. The ATHENA code is a multifluid version of the US-developed RELAP5 code that is also widely used for fusion safety analyses. The ENEA Fusion Division uses ATHENA in conjunction with the INTRA code for its safety analyses. The INTRA code was developed in Germany and predicts containment building pressures, temperatures and fluid flow. ENEA employs the French-developed ISAS system to couple ATHENA and INTRA. This paper provides a brief introduction of the MELCOR and ATHENA-INTRA codes and presents their modeling results for the following breaches of a water cooling line into the...
Physical Processes and Applications of the Monte Carlo Radiative Energy Deposition (MRED) Code
Reed, Robert A.; Weller, Robert A.; Mendenhall, Marcus H.; Fleetwood, Daniel M.; Warren, Kevin M.; Sierawski, Brian D.; King, Michael P.; Schrimpf, Ronald D.; Auden, Elizabeth C.
2015-08-01
MRED is a Python-language scriptable computer application that simulates radiation transport. It is the computational engine for the on-line tool CRÈME-MC. MRED is based on c++ code from Geant4 with additional Fortran components to simulate electron transport and nuclear reactions with high precision. We provide a detailed description of the structure of MRED and the implementation of the simulation of physical processes used to simulate radiation effects in electronic devices and circuits. Extensive discussion and references are provided that illustrate the validation of models used to implement specific simulations of relevant physical processes. Several applications of MRED are summarized that demonstrate its ability to predict and describe basic physical phenomena associated with irradiation of electronic circuits and devices. These include effects from single particle radiation (including both direct ionization and indirect ionization effects), dose enhancement effects, and displacement damage effects. MRED simulations have also helped to identify new single event upset mechanisms not previously observed by experiment, but since confirmed, including upsets due to muons and energetic electrons.
CMCpy: Genetic Code-Message Coevolution Models in Python
Becich, Peter J.; Stark, Brian P.; Bhat, Harish S.; Ardell, David H.
2013-01-01
Code-message coevolution (CMC) models represent coevolution of a genetic code and a population of protein-coding genes (“messages”). Formally, CMC models are sets of quasispecies coupled together for fitness through a shared genetic code. Although CMC models display plausible explanations for the origin of multiple genetic code traits by natural selection, useful modern implementations of CMC models are not currently available. To meet this need we present CMCpy, an object-oriented Python API and command-line executable front-end that can reproduce all published results of CMC models. CMCpy implements multiple solvers for leading eigenpairs of quasispecies models. We also present novel analytical results that extend and generalize applications of perturbation theory to quasispecies models and pioneer the application of a homotopy method for quasispecies with non-unique maximally fit genotypes. Our results therefore facilitate the computational and analytical study of a variety of evolutionary systems. CMCpy is free open-source software available from http://pypi.python.org/pypi/CMCpy/. PMID:23532367
Improvement of a combustion model in MELCOR code
International Nuclear Information System (INIS)
Ogino, Masao; Hashimoto, Takashi
1999-01-01
NUPEC has been improving a hydrogen combustion model in MELCOR code for severe accident analysis. In the proposed combustion model, the flame velocity in a node was predicted using five different flame front shapes of fireball, prism, bubble, spherical jet, and plane jet. For validation of the proposed model, the results of the Battelle multi-compartment hydrogen combustion test were used. The selected test cases for the study were Hx-6, 13, 14, 20 and Ix-2 which had two, three or four compartments under homogeneous hydrogen concentration of 5 to 10 vol%. The proposed model could predict well the combustion behavior in multi-compartment containment geometry on the whole. MELCOR code, incorporating the present combustion model, can simulate combustion behavior during severe accident with acceptable computing time and some degree of accuracy. The applicability study of the improved MELCOR code to the actual reactor plants will be further continued. (author)
International Nuclear Information System (INIS)
1988-03-01
HYDROCOIN is an international study for examining ground-water flow modeling strategies and their influence on safety assessments of geologic repositories for nuclear waste. This report summarizes only the combined NRC project temas' simulation efforts on the computer code bench-marking problems. The codes used to simulate thesee seven problems were SWIFT II, FEMWATER, UNSAT2M USGS-3D, AND TOUGH. In general, linear problems involving scalars such as hydraulic head were accurately simulated by both finite-difference and finite-element solution algorithms. Both types of codes produced accurate results even for complex geometrics such as intersecting fractures. Difficulties were encountered in solving problems that invovled nonlinear effects such as density-driven flow and unsaturated flow. In order to fully evaluate the accuracy of these codes, post-processing of results using paricle tracking algorithms and calculating fluxes were examined. This proved very valuable by uncovering disagreements among code results even through the hydraulic-head solutions had been in agreement. 9 refs., 111 figs., 6 tabs
JPEG2000 COMPRESSION CODING USING HUMAN VISUAL SYSTEM MODEL
Institute of Scientific and Technical Information of China (English)
Xiao Jiang; Wu Chengke
2005-01-01
In order to apply the Human Visual System (HVS) model to JPEG2000 standard,several implementation alternatives are discussed and a new scheme of visual optimization isintroduced with modifying the slope of rate-distortion. The novelty is that the method of visual weighting is not lifting the coefficients in wavelet domain, but is complemented by code stream organization. It remains all the features of Embedded Block Coding with Optimized Truncation (EBCOT) such as resolution progressive, good robust for error bit spread and compatibility of lossless compression. Well performed than other methods, it keeps the shortest standard codestream and decompression time and owns the ability of VIsual Progressive (VIP) coding.
The ELOCA fuel modelling code: past, present and future
International Nuclear Information System (INIS)
Williams, A.F.
2005-01-01
ELOCA is the Industry Standard Toolset (IST) computer code for modelling CANDU fuel under the transient coolant conditions typical of an accident scenario. Since its original inception in the early 1970's, the code has undergone continual development and improvement. The code now embodies much of the knowledge and experience of fuel behaviour gained by the Canadian nuclear industry over this period. ELOCA has proven to be a valuable tool for the safety analyst, and continues to be used extensively to support the licensing cases of CANDU reactors. This paper provides a brief and much simplified view of this development history, its current status, and plans for future development. (author)
Hamming Code Based Watermarking Scheme for 3D Model Verification
Directory of Open Access Journals (Sweden)
Jen-Tse Wang
2014-01-01
Full Text Available Due to the explosive growth of the Internet and maturing of 3D hardware techniques, protecting 3D objects becomes a more and more important issue. In this paper, a public hamming code based fragile watermarking technique is proposed for 3D objects verification. An adaptive watermark is generated from each cover model by using the hamming code technique. A simple least significant bit (LSB substitution technique is employed for watermark embedding. In the extraction stage, the hamming code based watermark can be verified by using the hamming code checking without embedding any verification information. Experimental results shows that 100% vertices of the cover model can be watermarked, extracted, and verified. It also shows that the proposed method can improve security and achieve low distortion of stego object.
Modeling Guidelines for Code Generation in the Railway Signaling Context
Ferrari, Alessio; Bacherini, Stefano; Fantechi, Alessandro; Zingoni, Niccolo
2009-01-01
Modeling guidelines constitute one of the fundamental cornerstones for Model Based Development. Their relevance is essential when dealing with code generation in the safety-critical domain. This article presents the experience of a railway signaling systems manufacturer on this issue. Introduction of Model-Based Development (MBD) and code generation in the industrial safety-critical sector created a crucial paradigm shift in the development process of dependable systems. While traditional software development focuses on the code, with MBD practices the focus shifts to model abstractions. The change has fundamental implications for safety-critical systems, which still need to guarantee a high degree of confidence also at code level. Usage of the Simulink/Stateflow platform for modeling, which is a de facto standard in control software development, does not ensure by itself production of high-quality dependable code. This issue has been addressed by companies through the definition of modeling rules imposing restrictions on the usage of design tools components, in order to enable production of qualified code. The MAAB Control Algorithm Modeling Guidelines (MathWorks Automotive Advisory Board)[3] is a well established set of publicly available rules for modeling with Simulink/Stateflow. This set of recommendations has been developed by a group of OEMs and suppliers of the automotive sector with the objective of enforcing and easing the usage of the MathWorks tools within the automotive industry. The guidelines have been published in 2001 and afterwords revisited in 2007 in order to integrate some additional rules developed by the Japanese division of MAAB [5]. The scope of the current edition of the guidelines ranges from model maintainability and readability to code generation issues. The rules are conceived as a reference baseline and therefore they need to be tailored to comply with the characteristics of each industrial context. Customization of these
MARS CODE MANUAL VOLUME V: Models and Correlations
International Nuclear Information System (INIS)
Chung, Bub Dong; Bae, Sung Won; Lee, Seung Wook; Yoon, Churl; Hwang, Moon Kyu; Kim, Kyung Doo; Jeong, Jae Jun
2010-02-01
Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This models and correlations manual provides a complete list of detailed information of the thermal-hydraulic models used in MARS, so that this report would be very useful for the code users. The overall structure of the manual is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible
The drift flux model in the ASSERT subchannel code
International Nuclear Information System (INIS)
Carver, M.B.; Judd, R.A.; Kiteley, J.C.; Tahir, A.
1987-01-01
The ASSERT subchannel code has been developed specifically to model flow and phase distributions within CANDU fuel bundles. ASSERT uses a drift-flux model that permits the phases to have unequal velocities, and can thus model phase separation tendencies that may occur in horizontal flow. The basic principles of ASSERT are outlined, and computed results are compared against data from various experiments for validation purposes. The paper concludes with an example of the use of the code to predict critical heat flux in CANDU geometries
Fuel rod modelling during transients: The TOUTATIS code
International Nuclear Information System (INIS)
Bentejac, F.; Bourreau, S.; Brochard, J.; Hourdequin, N.; Lansiart, S.
2001-01-01
The TOUTATIS code is devoted to the PCI local phenomena simulation, in correlation with the METEOR code for the global behaviour of the fuel rod. More specifically, the TOUTATIS objective is to evaluate the mechanical constraints on the cladding during a power transient thus predicting its behaviour in term of stress corrosion cracking. Based upon the finite element computation code CASTEM 2000, TOUTATIS is a set of modules written in a macro language. The aim of this paper is to present both code modules: The axisymmetric bi-dimensional module, modeling a unique block pellet; The tri dimensional module modeling a radially fragmented pellet. Having shown the boundary conditions and the algorithms used, the application will be illustrated by: A short presentation of the bidimensional axisymmetric modeling performances as well as its limits; The enhancement due to the three dimensional modeling will be displayed by sensitivity studies to the geometry, in this case the pellet height/diameter ratio. Finally, we will show the easiness of the development inherent to the CASTEM 2000 system by depicting the process of a modeling enhancement by adding the possibility of an axial (horizontal) fissuration of the pellet. As conclusion, the future improvements planned for the code are depicted. (author)
Physical model of Nernst element
International Nuclear Information System (INIS)
Nakamura, Hiroaki; Ikeda, Kazuaki; Yamaguchi, Satarou
1998-08-01
Generation of electric power by the Nernst effect is a new application of a semiconductor. A key point of this proposal is to find materials with a high thermomagnetic figure-of-merit, which are called Nernst elements. In order to find candidates of the Nernst element, a physical model to describe its transport phenomena is needed. As the first model, we began with a parabolic two-band model in classical statistics. According to this model, we selected InSb as candidates of the Nernst element and measured their transport coefficients in magnetic fields up to 4 Tesla within a temperature region from 270 K to 330 K. In this region, we calculated transport coefficients numerically by our physical model. For InSb, experimental data are coincident with theoretical values in strong magnetic field. (author)
Study of no-man's land physics in the total-f gyrokinetic code XGC1
Ku, Seung Hoe; Chang, C. S.; Lang, J.
2014-10-01
While the ``transport shortfall'' in the ``no-man's land'' has been observed often in delta-f codes, it has not yet been observed in the global total-f gyrokinetic particle code XGC1. Since understanding the interaction between the edge and core transport appears to be a critical element in the prediction for ITER performance, understanding the no-man's land issue is an important physics research topic. Simulation results using the Holland case will be presented and the physics causing the shortfall phenomenon will be discussed. Nonlinear nonlocal interaction of turbulence, secondary flows, and transport appears to be the key.
Statistical physics inspired energy-efficient coded-modulation for optical communications.
Djordjevic, Ivan B; Xu, Lei; Wang, Ting
2012-04-15
Because Shannon's entropy can be obtained by Stirling's approximation of thermodynamics entropy, the statistical physics energy minimization methods are directly applicable to the signal constellation design. We demonstrate that statistical physics inspired energy-efficient (EE) signal constellation designs, in combination with large-girth low-density parity-check (LDPC) codes, significantly outperform conventional LDPC-coded polarization-division multiplexed quadrature amplitude modulation schemes. We also describe an EE signal constellation design algorithm. Finally, we propose the discrete-time implementation of D-dimensional transceiver and corresponding EE polarization-division multiplexed system. © 2012 Optical Society of America
International Nuclear Information System (INIS)
Rastogi, B.P.
1989-01-01
This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)
Apar-T: code, validation, and physical interpretation of particle-in-cell results
Melzani, Mickaël; Winisdoerffer, Christophe; Walder, Rolf; Folini, Doris; Favre, Jean M.; Krastanov, Stefan; Messmer, Peter
2013-10-01
We present the parallel particle-in-cell (PIC) code Apar-T and, more importantly, address the fundamental question of the relations between the PIC model, the Vlasov-Maxwell theory, and real plasmas. First, we present four validation tests: spectra from simulations of thermal plasmas, linear growth rates of the relativistic tearing instability and of the filamentation instability, and nonlinear filamentation merging phase. For the filamentation instability we show that the effective growth rates measured on the total energy can differ by more than 50% from the linear cold predictions and from the fastest modes of the simulation. We link these discrepancies to the superparticle number per cell and to the level of field fluctuations. Second, we detail a new method for initial loading of Maxwell-Jüttner particle distributions with relativistic bulk velocity and relativistic temperature, and explain why the traditional method with individual particle boosting fails. The formulation of the relativistic Harris equilibrium is generalized to arbitrary temperature and mass ratios. Both are required for the tearing instability setup. Third, we turn to the key point of this paper and scrutinize the question of what description of (weakly coupled) physical plasmas is obtained by PIC models. These models rely on two building blocks: coarse-graining, i.e., grouping of the order of p ~ 1010 real particles into a single computer superparticle, and field storage on a grid with its subsequent finite superparticle size. We introduce the notion of coarse-graining dependent quantities, i.e., quantities depending on p. They derive from the PIC plasma parameter ΛPIC, which we show to behave as ΛPIC ∝ 1/p. We explore two important implications. One is that PIC collision- and fluctuation-induced thermalization times are expected to scale with the number of superparticles per grid cell, and thus to be a factor p ~ 1010 smaller than in real plasmas, a fact that we confirm with
Physical activity and influenza-coded outpatient visits, a population-based cohort study.
Directory of Open Access Journals (Sweden)
Eric Siu
Full Text Available Although the benefits of physical activity in preventing chronic medical conditions are well established, its impacts on infectious diseases, and seasonal influenza in particular, are less clearly defined. We examined the association between physical activity and influenza-coded outpatient visits, as a proxy for influenza infection.We conducted a cohort study of Ontario respondents to Statistics Canada's population health surveys over 12 influenza seasons. We assessed physical activity levels through survey responses, and influenza-coded physician office and emergency department visits through physician billing claims. We used logistic regression to estimate the risk of influenza-coded outpatient visits during influenza seasons. The cohort comprised 114,364 survey respondents who contributed 357,466 person-influenza seasons of observation. Compared to inactive individuals, moderately active (OR 0.83; 95% CI 0.74-0.94 and active (OR 0.87; 95% CI 0.77-0.98 individuals were less likely to experience an influenza-coded visit. Stratifying by age, the protective effect of physical activity remained significant for individuals <65 years (active OR 0.86; 95% CI 0.75-0.98, moderately active: OR 0.85; 95% CI 0.74-0.97 but not for individuals ≥ 65 years. The main limitations of this study were the use of influenza-coded outpatient visits rather than laboratory-confirmed influenza as the outcome measure, the reliance on self-report for assessing physical activity and various covariates, and the observational study design.Moderate to high amounts of physical activity may be associated with reduced risk of influenza for individuals <65 years. Future research should use laboratory-confirmed influenza outcomes to confirm the association between physical activity and influenza.
Instream Physical Habitat Modelling Types
DEFF Research Database (Denmark)
Conallin, John; Boegh, Eva; Krogsgaard, Jørgen
2010-01-01
The introduction of the EU Water Framework Directive (WFD) is providing member state water resource managers with significant challenges in relation to meeting the deadline for 'Good Ecological Status' by 2015. Overall, instream physical habitat modelling approaches have advantages and disadvanta......The introduction of the EU Water Framework Directive (WFD) is providing member state water resource managers with significant challenges in relation to meeting the deadline for 'Good Ecological Status' by 2015. Overall, instream physical habitat modelling approaches have advantages...... suit their situations. This paper analyses the potential of different methods available for water managers to assess hydrological and geomorphological impacts on the habitats of stream biota, as requested by the WFD. The review considers both conventional and new advanced research-based instream...... physical habitat models. In parametric and non-parametric regression models, model assumptions are often not satisfied and the models are difficult to transfer to other regions. Research-based methods such as the artificial neural networks and individual-based modelling have promising potential as water...
Improving system modeling accuracy with Monte Carlo codes
International Nuclear Information System (INIS)
Johnson, A.S.
1996-01-01
The use of computer codes based on Monte Carlo methods to perform criticality calculations has become common-place. Although results frequently published in the literature report calculated k eff values to four decimal places, people who use the codes in their everyday work say that they only believe the first two decimal places of any result. The lack of confidence in the computed k eff values may be due to the tendency of the reported standard deviation to underestimate errors associated with the Monte Carlo process. The standard deviation as reported by the codes is the standard deviation of the mean of the k eff values for individual generations in the computer simulation, not the standard deviation of the computed k eff value compared with the physical system. A more subtle problem with the standard deviation of the mean as reported by the codes is that all the k eff values from the separate generations are not statistically independent since the k eff of a given generation is a function of k eff of the previous generation, which is ultimately based on the starting source. To produce a standard deviation that is more representative of the physical system, statistically independent values of k eff are needed
A predictive transport modeling code for ICRF-heated tokamaks
International Nuclear Information System (INIS)
Phillips, C.K.; Hwang, D.Q.
1992-02-01
In this report, a detailed description of the physic included in the WHIST/RAZE package as well as a few illustrative examples of the capabilities of the package will be presented. An in depth analysis of ICRF heating experiments using WHIST/RAZE will be discussed in a forthcoming report. A general overview of philosophy behind the structure of the WHIST/RAZE package, a summary of the features of the WHIST code, and a description of the interface to the RAZE subroutines are presented in section 2 of this report. Details of the physics contained in the RAZE code are examined in section 3. Sample results from the package follow in section 4, with concluding remarks and a discussion of possible improvements to the package discussed in section 5
International Nuclear Information System (INIS)
Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng
2011-01-01
This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are
Energy Technology Data Exchange (ETDEWEB)
Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng
2011-03-01
This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are
International Nuclear Information System (INIS)
Yokoyama, Kenji; Uto, Nariaki; Kasahara, Naoto; Ishikawa, Makoto
2003-04-01
In the fast reactor development, numerical simulation using analytical codes plays an important role for complementing theory and experiment. It is necessary that the engineering models and analysis methods can be flexibly changed, because the phenomena to be investigated become more complicated due to the diversity of the needs for research. And, there are large problems in combining physical properties and engineering models in many different fields. Aiming to the realization of the next generation code system which can solve those problems, the authors adopted three methods, (1) Multi-language (SoftWIRE.NET, Visual Basic.NET and Fortran) (2) Fortran 90 and (3) Python to make a prototype of the next generation code system. As this result, the followings were confirmed. (1) It is possible to reuse a function of the existing codes written in Fortran as an object of the next generation code system by using Visual Basic.NET. (2) The maintainability of the existing code written by Fortran 77 can be improved by using the new features of Fortran 90. (3) The toolbox-type code system can be built by using Python. (author)
Modelling of thermalhydraulics and reactor physics in simulators
International Nuclear Information System (INIS)
Miettinen, J.
1994-01-01
The evolution of thermalhydraulic analysis methods for analysis and simulator purposes has brought closer the thermohydraulic models in both application areas. In large analysis codes like RELAP5, TRAC, CATHARE and ATHLET the accuracy for calculating complicated phenomena has been emphasized, but in spite of large development efforts many generic problems remain unsolved. For simulator purposes fast running codes have been developed and these include only limited assessment efforts. But these codes have more simulator friendly features than large codes, like portability and modular code structure. In this respect the simulator experiences with SMABRE code are discussed. Both large analysis codes and special simulator codes have their advances in simulator applications. The evolution of reactor physical calculation methods in simulator applications has started from simple point kinetic models. For analysis purposes accurate 1-D and 3-D codes have been developed being capable for fast and complicated transients. For simulator purposes capability for simulation of instruments has been emphasized, but the dynamic simulation capability has been less significant. The approaches for 3-dimensionality in simulators requires still quite much development, before the analysis accuracy is reached. (orig.) (8 refs., 2 figs., 2 tabs.)
Accelerator physics and modeling: Proceedings
International Nuclear Information System (INIS)
Parsa, Z.
1991-01-01
This report contains papers on the following topics: Physics of high brightness beams; radio frequency beam conditioner for fast-wave free-electron generators of coherent radiation; wake-field and space-charge effects on high brightness beams. Calculations and measured results for BNL-ATF; non-linear orbit theory and accelerator design; general problems of modeling for accelerators; development and application of dispersive soft ferrite models for time-domain simulation; and bunch lengthening in the SLC damping rings
Wave Generation in Physical Models
DEFF Research Database (Denmark)
Andersen, Thomas Lykke; Frigaard, Peter
The present book describes the most important aspects of wave generation techniques in physical models. Moreover, the book serves as technical documentation for the wave generation software AwaSys 6, cf. Aalborg University (2012). In addition to the two main authors also Tue Hald and Michael...
Development of the next generation code system as an engineering modeling language (1)
International Nuclear Information System (INIS)
Yokoyama, Kenji; Uto, Nariaki; Kasahara, Naoto; Nagura, Fuminori; Ishikawa, Makoto; Ohira, Masanori; Kato, Masayuki
2002-11-01
In the fast reactor development, numerical simulation using analytical codes plays an important role for complementing theory and experiment. It is necessary that the engineering models and analysis methods can be flexibly changed, because the phenamine to be investigated become more complicated due to the diversity of the needs for research. And, there are large problems in combining physical properties and engineering models in many different fields. In this study, the goal is to develop a flexible and general-purposive analysis system, in which the physical properties and engineering models are represented as a programming language or a diagrams that are easily understandable for humans and executable for computers. The authors named this concept the Engineering Modeling Language (EML). This report describes the result of the investigation for latest computer technologies and software development techniques which seem to be usable for a realization of the analysis code system for nuclear engineering as an EML. (author)
Energy Technology Data Exchange (ETDEWEB)
Nagels-Silvert, V
2004-09-15
The main purpose of this thesis is to get experimental data for the testing and validation of atomic physics codes dealing with non-local-thermodynamical-equilibrium plasmas. The first part is dedicated to the spectroscopic study of xenon and krypton plasmas that have been produced by a nanosecond laser pulse interacting with a gas jet. A Thomson scattering diagnostic has allowed us to measure independently plasma parameters such as electron temperature, electron density and the average ionisation state. We have obtained time integrated spectra in the range between 5 and 10 angstroms. We have identified about one hundred xenon rays between 8.6 and 9.6 angstroms via the use of the Relac code. We have discovered unknown rays for the krypton between 5.2 and 7.5 angstroms. In a second experiment we have extended the wavelength range to the X UV domain. The Averroes/Transpec code has been tested in the ranges from 9 to 15 angstroms and from 10 to 130 angstroms, the first range has been well reproduced while the second range requires a more complex data analysis. The second part is dedicated to the spectroscopic study of aluminium, selenium and samarium plasmas in femtosecond operating rate. We have designed an interferometry diagnostic in the frequency domain that has allowed us to measure the expanding speed of the target's backside. Via the use of an adequate isothermal model this parameter has led us to know the plasma electron temperature. Spectra and emission times of various rays from the aluminium and selenium plasmas have been computed satisfactorily with the Averroes/Transpec code coupled with Film and Multif hydrodynamical codes. (A.C.)
Energy Technology Data Exchange (ETDEWEB)
Nagels-Silvert, V
2004-09-15
The main purpose of this thesis is to get experimental data for the testing and validation of atomic physics codes dealing with non-local-thermodynamical-equilibrium plasmas. The first part is dedicated to the spectroscopic study of xenon and krypton plasmas that have been produced by a nanosecond laser pulse interacting with a gas jet. A Thomson scattering diagnostic has allowed us to measure independently plasma parameters such as electron temperature, electron density and the average ionisation state. We have obtained time integrated spectra in the range between 5 and 10 angstroms. We have identified about one hundred xenon rays between 8.6 and 9.6 angstroms via the use of the Relac code. We have discovered unknown rays for the krypton between 5.2 and 7.5 angstroms. In a second experiment we have extended the wavelength range to the X UV domain. The Averroes/Transpec code has been tested in the ranges from 9 to 15 angstroms and from 10 to 130 angstroms, the first range has been well reproduced while the second range requires a more complex data analysis. The second part is dedicated to the spectroscopic study of aluminium, selenium and samarium plasmas in femtosecond operating rate. We have designed an interferometry diagnostic in the frequency domain that has allowed us to measure the expanding speed of the target's backside. Via the use of an adequate isothermal model this parameter has led us to know the plasma electron temperature. Spectra and emission times of various rays from the aluminium and selenium plasmas have been computed satisfactorily with the Averroes/Transpec code coupled with Film and Multif hydrodynamical codes. (A.C.)
Development of the physical model
International Nuclear Information System (INIS)
Liu Zunqi; Morsy, Samir
2001-01-01
Full text: The Physical Model was developed during Program 93+2 as a technical tool to aid enhanced information analysis and now is an integrated part of the Department's on-going State evaluation process. This paper will describe the concept of the Physical Model, including its objectives, overall structure and the development of indicators with designated strengths, followed by a brief description of using the Physical Model in implementing the enhanced information analysis. The work plan for expansion and update of the Physical Model is also presented at the end of the paper. The development of the Physical Model is an attempt to identify, describe and characterize every known process for carrying out each step necessary for the acquisition of weapons-usable material, i.e., all plausible acquisition paths for highly enriched uranium (HEU) and separated plutonium (Pu). The overall structure of the Physical Model has a multilevel arrangement. It includes at the top level all the main steps (technologies) that may be involved in the nuclear fuel cycle from the source material production up to the acquisition of weapons-usable material, and then beyond the civilian fuel cycle to the development of nuclear explosive devices (weaponization). Each step is logically interconnected with the preceding and/or succeeding steps by nuclear material flows. It contains at its lower levels every known process that is associated with the fuel cycle activities presented at the top level. For example, uranium enrichment is broken down into three branches at the second level, i.e., enrichment of UF 6 , UCl 4 and U-metal respectively; and then further broken down at the third level into nine processes: gaseous diffusion, gas centrifuge, aerodynamic, electromagnetic, molecular laser (MLIS), atomic vapor laser (AVLIS), chemical exchange, ion exchange and plasma. Narratives are presented at each level, beginning with a general process description then proceeding with detailed
Universal Regularizers For Robust Sparse Coding and Modeling
Ramirez, Ignacio; Sapiro, Guillermo
2010-01-01
Sparse data models, where data is assumed to be well represented as a linear combination of a few elements from a dictionary, have gained considerable attention in recent years, and their use has led to state-of-the-art results in many signal and image processing tasks. It is now well understood that the choice of the sparsity regularization term is critical in the success of such models. Based on a codelength minimization interpretation of sparse coding, and using tools from universal coding...
Model-Driven Engineering of Machine Executable Code
Eichberg, Michael; Monperrus, Martin; Kloppenburg, Sven; Mezini, Mira
Implementing static analyses of machine-level executable code is labor intensive and complex. We show how to leverage model-driven engineering to facilitate the design and implementation of programs doing static analyses. Further, we report on important lessons learned on the benefits and drawbacks while using the following technologies: using the Scala programming language as target of code generation, using XML-Schema to express a metamodel, and using XSLT to implement (a) transformations and (b) a lint like tool. Finally, we report on the use of Prolog for writing model transformations.
An improved thermal model for the computer code NAIAD
International Nuclear Information System (INIS)
Rainbow, M.T.
1982-12-01
An improved thermal model, based on the concept of heat slabs, has been incorporated as an option into the thermal hydraulic computer code NAIAD. The heat slabs are one-dimensional thermal conduction models with temperature independent thermal properties which may be internal and/or external to the fluid. Thermal energy may be added to or removed from the fluid via heat slabs and passed across the external boundary of external heat slabs at a rate which is a linear function of the external surface temperatures. The code input for the new option has been restructured to simplify data preparation. A full description of current input requirements is presented
Anthropomorphic Coding of Speech and Audio: A Model Inversion Approach
Directory of Open Access Journals (Sweden)
W. Bastiaan Kleijn
2005-06-01
Full Text Available Auditory modeling is a well-established methodology that provides insight into human perception and that facilitates the extraction of signal features that are most relevant to the listener. The aim of this paper is to provide a tutorial on perceptual speech and audio coding using an invertible auditory model. In this approach, the audio signal is converted into an auditory representation using an invertible auditory model. The auditory representation is quantized and coded. Upon decoding, it is then transformed back into the acoustic domain. This transformation converts a complex distortion criterion into a simple one, thus facilitating quantization with low complexity. We briefly review past work on auditory models and describe in more detail the components of our invertible model and its inversion procedure, that is, the method to reconstruct the signal from the output of the auditory model. We summarize attempts to use the auditory representation for low-bit-rate coding. Our approach also allows the exploitation of the inherent redundancy of the human auditory system for the purpose of multiple description (joint source-channel coding.
Data model description for the DESCARTES and CIDER codes
International Nuclear Information System (INIS)
Miley, T.B.; Ouderkirk, S.J.; Nichols, W.E.; Eslinger, P.W.
1993-01-01
The primary objective of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate the radiation dose that individuals could have received as a result of emissions since 1944 from the US Department of Energy's (DOE) Hanford Site near Richland, Washington. One of the major objectives of the HEDR Project is to develop several computer codes to model the airborne releases. transport and envirorunental accumulation of radionuclides resulting from Hanford operations from 1944 through 1972. In July 1992, the HEDR Project Manager determined that the computer codes being developed (DESCARTES, calculation of environmental accumulation from airborne releases, and CIDER, dose calculations from environmental accumulation) were not sufficient to create accurate models. A team of HEDR staff members developed a plan to assure that computer codes would meet HEDR Project goals. The plan consists of five tasks: (1) code requirements definition. (2) scoping studies, (3) design specifications, (4) benchmarking, and (5) data modeling. This report defines the data requirements for the DESCARTES and CIDER codes
COCOA code for creating mock observations of star cluster models
Askar, Abbas; Giersz, Mirek; Pych, Wojciech; Dalessandro, Emanuele
2018-04-01
We introduce and present results from the COCOA (Cluster simulatiOn Comparison with ObservAtions) code that has been developed to create idealized mock photometric observations using results from numerical simulations of star cluster evolution. COCOA is able to present the output of realistic numerical simulations of star clusters carried out using Monte Carlo or N-body codes in a way that is useful for direct comparison with photometric observations. In this paper, we describe the COCOA code and demonstrate its different applications by utilizing globular cluster (GC) models simulated with the MOCCA (MOnte Carlo Cluster simulAtor) code. COCOA is used to synthetically observe these different GC models with optical telescopes, perform point spread function photometry, and subsequently produce observed colour-magnitude diagrams. We also use COCOA to compare the results from synthetic observations of a cluster model that has the same age and metallicity as the Galactic GC NGC 2808 with observations of the same cluster carried out with a 2.2 m optical telescope. We find that COCOA can effectively simulate realistic observations and recover photometric data. COCOA has numerous scientific applications that maybe be helpful for both theoreticians and observers that work on star clusters. Plans for further improving and developing the code are also discussed in this paper.
Code-code comparisons of DIVIMP's 'onion-skin model' and the EDGE2D fluid code
International Nuclear Information System (INIS)
Stangeby, P.C.; Elder, J.D.; Horton, L.D.; Simonini, R.; Taroni, A.; Matthews, O.F.; Monk, R.D.
1997-01-01
In onion-skin modelling, O-SM, of the edge plasma, the cross-field power and particle flows are treated very simply e.g. as spatially uniform. The validity of O-S modelling requires demonstration that such approximations can still result in reasonable solutions for the edge plasma. This is demonstrated here by comparison of O-SM with full 2D fluid edge solutions generated by the EDGE2D code. The target boundary conditions for the O-SM are taken from the EDGE2D output and the complete O-SM solutions are then compared with the EDGE2D ones. Agreement is generally within 20% for n e , T e , T i and parallel particle flux density Γ for the medium and high recycling JET cases examined and somewhat less good for a strongly detached CMOD example. (orig.)
Advanced Electric and Magnetic Material Models for FDTD Electromagnetic Codes
Poole, Brian R; Nelson, Scott D
2005-01-01
The modeling of dielectric and magnetic materials in the time domain is required for pulse power applications, pulsed induction accelerators, and advanced transmission lines. For example, most induction accelerator modules require the use of magnetic materials to provide adequate Volt-sec during the acceleration pulse. These models require hysteresis and saturation to simulate the saturation wavefront in a multipulse environment. In high voltage transmission line applications such as shock or soliton lines the dielectric is operating in a highly nonlinear regime, which requires nonlinear models. Simple 1-D models are developed for fast parameterization of transmission line structures. In the case of nonlinear dielectrics, a simple analytic model describing the permittivity in terms of electric field is used in a 3-D finite difference time domain code (FDTD). In the case of magnetic materials, both rate independent and rate dependent Hodgdon magnetic material models have been implemented into 3-D FDTD codes an...
Radiation transport phenomena and modeling - part A: Codes
International Nuclear Information System (INIS)
Lorence, L.J.
1997-01-01
The need to understand how particle radiation (high-energy photons and electrons) from a variety of sources affects materials and electronics has motivated the development of sophisticated computer codes that describe how radiation with energies from 1.0 keV to 100.0 GeV propagates through matter. Predicting radiation transport is the necessary first step in predicting radiation effects. The radiation transport codes that are described here are general-purpose codes capable of analyzing a variety of radiation environments including those produced by nuclear weapons (x-rays, gamma rays, and neutrons), by sources in space (electrons and ions) and by accelerators (x-rays, gamma rays, and electrons). Applications of these codes include the study of radiation effects on electronics, nuclear medicine (imaging and cancer treatment), and industrial processes (food disinfestation, waste sterilization, manufacturing.) The primary focus will be on coupled electron-photon transport codes, with some brief discussion of proton transport. These codes model a radiation cascade in which electrons produce photons and vice versa. This coupling between particles of different types is important for radiation effects. For instance, in an x-ray environment, electrons are produced that drive the response in electronics. In an electron environment, dose due to bremsstrahlung photons can be significant once the source electrons have been stopped
Development of improved methods for the LWR lattice physics code EPRI-CELL
International Nuclear Information System (INIS)
Williams, M.L.; Wright, R.Q.; Barhen, J.
1982-07-01
A number of improvements have been made by ORNL to the lattice physics code EPRI-CELL (E-C) which is widely used by utilities for analysis of power reactors. The code modifications were made mainly in the thermal and epithermal routines and resulted in improved reactor physics approximations and more efficient running times. The improvements in the thermal flux calculation included implementation of a group-dependent rebalance procedure to accelerate the iterative process and a more rigorous calculation of interval-to-interval collision probabilities. The epithermal resonance shielding methods used in the code have been extensively studied to determine its major approximations and to examine the sensitivity of computed results to these approximations. The study has resulted in several improvements in the original methodology
Three-dimensional modeling with finite element codes
Energy Technology Data Exchange (ETDEWEB)
Druce, R.L.
1986-01-17
This paper describes work done to model magnetostatic field problems in three dimensions. Finite element codes, available at LLNL, and pre- and post-processors were used in the solution of the mathematical model, the output from which agreed well with the experimentally obtained data. The geometry used in this work was a cylinder with ports in the periphery and no current sources in the space modeled. 6 refs., 8 figs.
Code Development for Control Design Applications: Phase I: Structural Modeling
International Nuclear Information System (INIS)
Bir, G. S.; Robinson, M.
1998-01-01
The design of integrated controls for a complex system like a wind turbine relies on a system model in an explicit format, e.g., state-space format. Current wind turbine codes focus on turbine simulation and not on system characterization, which is desired for controls design as well as applications like operating turbine model analysis, optimal design, and aeroelastic stability analysis. This paper reviews structural modeling that comprises three major steps: formation of component equations, assembly into system equations, and linearization
Light water reactor fuel analysis code FEMAXI-7. Model and structure [Revised edition
International Nuclear Information System (INIS)
Suzuki, Motoe; Udagawa, Yutaka; Amaya, Masaki; Saitou, Hiroaki
2014-03-01
A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in both normal conditions and anticipated transient conditions. This code is an advanced version which has been produced by incorporating the former version FEMAXI-6 with numerous functional improvements and extensions. In FEMAXI-7, many new models have been added and parameters have been clearly arranged. Also, to facilitate effective maintenance and accessibility of the code, modularization of subroutines and functions have been attained, and quality comment descriptions of variables or physical quantities have been incorporated in the source code. With these advancements, the FEMAXI-7 code has been upgraded to a versatile analytical tool for high burnup fuel behavior analyses. This report is the revised edition of the first one which describes in detail the design, basic theory and structure, models and numerical method, and improvements and extensions. The first edition, JAEA-Data/Code 2010-035, was published in 2010. The first edition was extended by orderly addition and disposition of explanations of models and organized as the revised edition after three years interval. (author)
Model-implementation fidelity in cyber physical system design
Fabre, Christian
2017-01-01
This book puts in focus various techniques for checking modeling fidelity of Cyber Physical Systems (CPS), with respect to the physical world they represent. The authors' present modeling and analysis techniques representing different communities, from very different angles, discuss their possible interactions, and discuss the commonalities and differences between their practices. Coverage includes model driven development, resource-driven development, statistical analysis, proofs of simulator implementation, compiler construction, power/temperature modeling of digital devices, high-level performance analysis, and code/device certification. Several industrial contexts are covered, including modeling of computing and communication, proof architectures models and statistical based validation techniques. Addresses CPS design problems such as cross-application interference, parsimonious modeling, and trustful code production Describes solutions, such as simulation for extra-functional properties, extension of cod...
Physical models of cell motility
2016-01-01
This book surveys the most recent advances in physics-inspired cell movement models. This synergetic, cross-disciplinary effort to increase the fidelity of computational algorithms will lead to a better understanding of the complex biomechanics of cell movement, and stimulate progress in research on related active matter systems, from suspensions of bacteria and synthetic swimmers to cell tissues and cytoskeleton.Cell motility and collective motion are among the most important themes in biology and statistical physics of out-of-equilibrium systems, and crucial for morphogenesis, wound healing, and immune response in eukaryotic organisms. It is also relevant for the development of effective treatment strategies for diseases such as cancer, and for the design of bioactive surfaces for cell sorting and manipulation. Substrate-based cell motility is, however, a very complex process as regulatory pathways and physical force generation mechanisms are intertwined. To understand the interplay between adhesion, force ...
Revised uranium--plutonium cycle PWR and BWR models for the ORIGEN computer code
International Nuclear Information System (INIS)
Croff, A.G.; Bjerke, M.A.; Morrison, G.W.; Petrie, L.M.
1978-09-01
Reactor physics calculations and literature searches have been conducted, leading to the creation of revised enriched-uranium and enriched-uranium/mixed-oxide-fueled PWR and BWR reactor models for the ORIGEN computer code. These ORIGEN reactor models are based on cross sections that have been taken directly from the reactor physics codes and eliminate the need to make adjustments in uncorrected cross sections in order to obtain correct depletion results. Revised values of the ORIGEN flux parameters THERM, RES, and FAST were calculated along with new parameters related to the activation of fuel-assembly structural materials not located in the active fuel zone. Recommended fuel and structural material masses and compositions are presented. A summary of the new ORIGEN reactor models is given
Performance Theories for Sentence Coding: Some Quantitative Models
Aaronson, Doris; And Others
1977-01-01
This study deals with the patterns of word-by-word reading times over a sentence when the subject must code the linguistic information sufficiently for immediate verbatim recall. A class of quantitative models is considered that would account for reading times at phrase breaks. (Author/RM)
Modeling of PHWR fuel elements using FUDA code
International Nuclear Information System (INIS)
Tripathi, Rahul Mani; Soni, Rakesh; Prasad, P.N.; Pandarinathan, P.R.
2008-01-01
The computer code FUDA (Fuel Design Analysis) is used for modeling PHWR fuel bundle operation history and carry out fuel element thermo-mechanical analysis. The radial temperature profile across fuel and sheath, fission gas release, internal gas pressure, sheath stress and strains during the life of fuel bundle are estimated
28 CFR 36.608 - Guidance concerning model codes.
2010-07-01
... Section 36.608 Judicial Administration DEPARTMENT OF JUSTICE NONDISCRIMINATION ON THE BASIS OF DISABILITY BY PUBLIC ACCOMMODATIONS AND IN COMMERCIAL FACILITIES Certification of State Laws or Local Building... private entity responsible for developing a model code, the Assistant Attorney General may review the...
Code Shift: Grid Specifications and Dynamic Wind Turbine Models
DEFF Research Database (Denmark)
Ackermann, Thomas; Ellis, Abraham; Fortmann, Jens
2013-01-01
Grid codes (GCs) and dynamic wind turbine (WT) models are key tools to allow increasing renewable energy penetration without challenging security of supply. In this article, the state of the art and the further development of both tools are discussed, focusing on the European and North American e...
EMPIRE-II statistical model code for nuclear reaction calculations
Energy Technology Data Exchange (ETDEWEB)
Herman, M [International Atomic Energy Agency, Vienna (Austria)
2001-12-15
EMPIRE II is a nuclear reaction code, comprising various nuclear models, and designed for calculations in the broad range of energies and incident particles. A projectile can be any nucleon or Heavy Ion. The energy range starts just above the resonance region, in the case of neutron projectile, and extends up to few hundreds of MeV for Heavy Ion induced reactions. The code accounts for the major nuclear reaction mechanisms, such as optical model (SCATB), Multistep Direct (ORION + TRISTAN), NVWY Multistep Compound, and the full featured Hauser-Feshbach model. Heavy Ion fusion cross section can be calculated within the simplified coupled channels approach (CCFUS). A comprehensive library of input parameters covers nuclear masses, optical model parameters, ground state deformations, discrete levels and decay schemes, level densities, fission barriers (BARFIT), moments of inertia (MOMFIT), and {gamma}-ray strength functions. Effects of the dynamic deformation of a fast rotating nucleus can be taken into account in the calculations. The results can be converted into the ENDF-VI format using the accompanying code EMPEND. The package contains the full EXFOR library of experimental data. Relevant EXFOR entries are automatically retrieved during the calculations. Plots comparing experimental results with the calculated ones can be produced using X4TOC4 and PLOTC4 codes linked to the rest of the system through bash-shell (UNIX) scripts. The graphic user interface written in Tcl/Tk is provided. (author)
Physical model of reactor pulse
International Nuclear Information System (INIS)
Petrovic, A.; Ravnik, M.
2004-01-01
Pulse experiments have been performed at J. Stefan Institute TRIGA reactor since 1991. In total, more than 130 pulses have been performed. Extensive experimental information on the pulse physical characteristics has been accumulated. Fuchs-Hansen adiabatic model has been used for predicting and analysing the pulse parameters. The model is based on point kinetics equation, neglecting the delayed neutrons and assuming constant inserted reactivity in form of step function. Deficiencies of the Fuchs-Hansen model and systematic experimental errors have been observed and analysed. Recently, the pulse model was improved by including the delayed neutrons and time dependence of inserted reactivity. The results explain the observed non-linearity of the pulse energy for high pulses due to finite time of pulse rod withdrawal and the contribution of the delayed neutrons after the prompt part of the pulse. The results of the improved model are in good agreement with experimental results. (author)
Preliminary analyses for HTTR`s start-up physics tests by Monte Carlo code MVP
Energy Technology Data Exchange (ETDEWEB)
Nojiri, Naoki [Science and Technology Agency, Tokyo (Japan); Nakano, Masaaki; Ando, Hiroei; Fujimoto, Nozomu; Takeuchi, Mitsuo; Fujisaki, Shingo; Yamashita, Kiyonobu
1998-08-01
Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)
Preliminary analyses for HTTR's start-up physics tests by Monte Carlo code MVP
International Nuclear Information System (INIS)
Nojiri, Naoki; Nakano, Masaaki; Ando, Hiroei; Fujimoto, Nozomu; Takeuchi, Mitsuo; Fujisaki, Shingo; Yamashita, Kiyonobu
1998-08-01
Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)
Recent improvements of the TNG statistical model code
International Nuclear Information System (INIS)
Shibata, K.; Fu, C.Y.
1986-08-01
The applicability of the nuclear model code TNG to cross-section evaluations has been extended. The new TNG is capable of using variable bins for outgoing particle energies. Moreover, three additional quantities can now be calculated: capture gamma-ray spectrum, the precompound mode of the (n,γ) reaction, and fission cross section. In this report, the new features of the code are described together with some sample calculations and a brief explanation of the input data. 15 refs., 6 figs., 2 tabs
Modeling RERTR experimental fuel plates using the PLATE code
International Nuclear Information System (INIS)
Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Snelgrove, J.L.; Brazener, R.A.
2003-01-01
Modeling results using the PLATE dispersion fuel performance code are presented for the U-Mo/Al experimental fuel plates from the RERTR-1, -2, -3 and -5 irradiation tests. Agreement of the calculations with experimental data obtained in post-irradiation examinations of these fuels, where available, is shown to be good. Use of the code to perform a series of parametric evaluations highlights the sensitivity of U-Mo dispersion fuel performance to fabrication variables, especially fuel particle shape and size distributions. (author)
Generation of initial geometries for the simulation of the physical system in the DualPHYsics code
International Nuclear Information System (INIS)
Segura Q, E.
2013-01-01
In the diverse research areas of the Instituto Nacional de Investigaciones Nucleares (ININ) are different activities related to science and technology, one of great interest is the study and treatment of the collection and storage of radioactive waste. Therefore at ININ the draft on the simulation of the pollutants diffusion in the soil through a porous medium (third stage) has this problem inherent aspects, hence a need for such a situation is to generate the initial geometry of the physical system For the realization of the simulation method is implemented smoothed particle hydrodynamics (SPH). This method runs in DualSPHysics code, which has great versatility and ability to simulate phenomena of any physical system where hydrodynamic aspects combine. In order to simulate a physical system DualSPHysics code, you need to preset the initial geometry of the system of interest, then this is included in the input file of the code. The simulation sets the initial geometry through regular geometric bodies positioned at different points in space. This was done through a programming language (Fortran, C + +, Java, etc..). This methodology will provide the basis to simulate more complex geometries future positions and form. (Author)
Thermohydraulic modeling of nuclear thermal rockets: The KLAXON code
International Nuclear Information System (INIS)
Hall, M.L.; Rider, W.J.; Cappiello, M.W.
1992-01-01
The hydrogen flow from the storage tanks, through the reactor core, and out the nozzle of a Nuclear Thermal Rocket is an integral design consideration. To provide an analysis and design tool for this phenomenon, the KLAXON code is being developed. A shock-capturing numerical methodology is used to model the gas flow (the Harten, Lax, and van Leer method, as implemented by Einfeldt). Preliminary results of modeling the flow through the reactor core and nozzle are given in this paper
Automatic modeling for the Monte Carlo transport code Geant4
International Nuclear Information System (INIS)
Nie Fanzhi; Hu Liqin; Wang Guozhong; Wang Dianxi; Wu Yican; Wang Dong; Long Pengcheng; FDS Team
2015-01-01
Geant4 is a widely used Monte Carlo transport simulation package. Its geometry models could be described in Geometry Description Markup Language (GDML), but it is time-consuming and error-prone to describe the geometry models manually. This study implemented the conversion between computer-aided design (CAD) geometry models and GDML models. This method has been Studied based on Multi-Physics Coupling Analysis Modeling Program (MCAM). The tests, including FDS-Ⅱ model, demonstrated its accuracy and feasibility. (authors)
Steam generator and circulator model for the HELAP code
International Nuclear Information System (INIS)
Ludewig, H.
1975-07-01
An outline is presented of the work carried out in the 1974 fiscal year on the GCFBR safety research project consisting of the development of improved steam generator and circulator (steam turbine driven helium compressor) models which will eventually be inserted in the HELAP (1) code. Furthermore, a code was developed which will be used to generate steady state input for the primary and secondary sides of the steam generator. The following conclusions and suggestions for further work are made: (1) The steam-generator and circulator model are consistent with the volume and junction layout used in HELAP, (2) with minor changes these models, when incorporated in HELAP, could be used to simulate a direct cycle plant, (3) an explicit control valve model is still to be developed and would be very desirable to control the flow to the turbine during a transient (initially this flow will be controlled by using the existing check valve model); (4) the friction factor in the laminar flow region is computed inaccurately, this might cause significant errors in loss-of-flow accidents; and (5) it is felt that HELAP will still use a large amount of computer time and will thus be limited to design basis accidents without scram or loss of flow transients with and without scram. Finally it may also be used as a test bed for the development of prototype component models which would be incorporated in a more sophisticated system code, developed specifically for GCFBR's
Background-Modeling-Based Adaptive Prediction for Surveillance Video Coding.
Zhang, Xianguo; Huang, Tiejun; Tian, Yonghong; Gao, Wen
2014-02-01
The exponential growth of surveillance videos presents an unprecedented challenge for high-efficiency surveillance video coding technology. Compared with the existing coding standards that were basically developed for generic videos, surveillance video coding should be designed to make the best use of the special characteristics of surveillance videos (e.g., relative static background). To do so, this paper first conducts two analyses on how to improve the background and foreground prediction efficiencies in surveillance video coding. Following the analysis results, we propose a background-modeling-based adaptive prediction (BMAP) method. In this method, all blocks to be encoded are firstly classified into three categories. Then, according to the category of each block, two novel inter predictions are selectively utilized, namely, the background reference prediction (BRP) that uses the background modeled from the original input frames as the long-term reference and the background difference prediction (BDP) that predicts the current data in the background difference domain. For background blocks, the BRP can effectively improve the prediction efficiency using the higher quality background as the reference; whereas for foreground-background-hybrid blocks, the BDP can provide a better reference after subtracting its background pixels. Experimental results show that the BMAP can achieve at least twice the compression ratio on surveillance videos as AVC (MPEG-4 Advanced Video Coding) high profile, yet with a slightly additional encoding complexity. Moreover, for the foreground coding performance, which is crucial to the subjective quality of moving objects in surveillance videos, BMAP also obtains remarkable gains over several state-of-the-art methods.
Comparison Study on Low Energy Physics Model of GEANT4
International Nuclear Information System (INIS)
Park, So Hyun; Jung, Won Gyun; Suh, Tae Suk
2010-01-01
The Geant4 simulation toolkit provides improved or renewed physics model according to the version. The latest Geant4.9.3 which has been recoded by developers applies inserted Livermore data and renewed physics model to the low energy electromagnetic physics model. And also, Geant4.9.3 improved the physics factors by modified code. In this study, the stopping power and CSDA(Continuously Slowing Down Approximation) range data of electron or particles were acquired in various material and then, these data were compared with NIST(National Institute of Standards and Technology) data. Through comparison between data of Geant4 simulation and NIST, the improvement of physics model on low energy electromagnetic of Geant4.9.3 was evaluated by comparing the Geant4.9.2
Evaluating nuclear physics inputs in core-collapse supernova models
Lentz, E.; Hix, W. R.; Baird, M. L.; Messer, O. E. B.; Mezzacappa, A.
Core-collapse supernova models depend on the details of the nuclear and weak interaction physics inputs just as they depend on the details of the macroscopic physics (transport, hydrodynamics, etc.), numerical methods, and progenitors. We present preliminary results from our ongoing comparison studies of nuclear and weak interaction physics inputs to core collapse supernova models using the spherically-symmetric, general relativistic, neutrino radiation hydrodynamics code Agile-Boltztran. We focus on comparisons of the effects of the nuclear EoS and the effects of improving the opacities, particularly neutrino--nucleon interactions.
An improved steam generator model for the SASSYS code
International Nuclear Information System (INIS)
Pizzica, P.A.
1989-01-01
A new steam generator model has been developed for the SASSYS computer code, which analyzes accident conditions in a liquid-metal-cooled fast reactor. It has been incorporated into the new SASSYS balance-of-plant model, but it can also function as a stand-alone model. The model provides a full solution of the steady-state condition before the transient calculation begins for given sodium and water flow rates, inlet and outlet sodium temperatures, and inlet enthalpy and region lengths on the water side
Joint ICTP-IAEA advanced workshop on model codes for spallation reactions
International Nuclear Information System (INIS)
Filges, D.; Leray, S.; Yariv, Y.; Mengoni, A.; Stanculescu, A.; Mank, G.
2008-08-01
The International Atomic Energy Agency (IAEA) and the Abdus Salam International Centre for Theoretical Physics (ICTP) organised an expert meeting at the ICTP from 4 to 8 February 2008 to discuss model codes for spallation reactions. These nuclear reactions play an important role in a wide domain of applications ranging from neutron sources for condensed matter and material studies, transmutation of nuclear waste and rare isotope production to astrophysics, simulation of detector set-ups in nuclear and particle physics experiments, and radiation protection near accelerators or in space. The simulation tools developed for these domains use nuclear model codes to compute the production yields and characteristics of all the particles and nuclei generated in these reactions. These codes are generally Monte-Carlo implementations of Intra-Nuclear Cascade (INC) or Quantum Molecular Dynamics (QMD) models, followed by de-excitation (principally evaporation/fission) models. Experts have discussed in depth the physics contained within the different models in order to understand their strengths and weaknesses. Such codes need to be validated against experimental data in order to determine their accuracy and reliability with respect to all forms of application. Agreement was reached during the course of the workshop to organise an international benchmark of the different models developed by different groups around the world. The specifications of the benchmark, including the set of selected experimental data to be compared to the models, were also defined during the workshop. The benchmark will be organised under the auspices of the IAEA in 2008, and the first results will be discussed at the next Accelerator Applications Conference (AccApp'09) to be held in Vienna in May 2009. (author)
Dual coding: a cognitive model for psychoanalytic research.
Bucci, W
1985-01-01
Four theories of mental representation derived from current experimental work in cognitive psychology have been discussed in relation to psychoanalytic theory. These are: verbal mediation theory, in which language determines or mediates thought; perceptual dominance theory, in which imagistic structures are dominant; common code or propositional models, in which all information, perceptual or linguistic, is represented in an abstract, amodal code; and dual coding, in which nonverbal and verbal information are each encoded, in symbolic form, in separate systems specialized for such representation, and connected by a complex system of referential relations. The weight of current empirical evidence supports the dual code theory. However, psychoanalysis has implicitly accepted a mixed model-perceptual dominance theory applying to unconscious representation, and verbal mediation characterizing mature conscious waking thought. The characterization of psychoanalysis, by Schafer, Spence, and others, as a domain in which reality is constructed rather than discovered, reflects the application of this incomplete mixed model. The representations of experience in the patient's mind are seen as without structure of their own, needing to be organized by words, thus vulnerable to distortion or dissolution by the language of the analyst or the patient himself. In these terms, hypothesis testing becomes a meaningless pursuit; the propositions of the theory are no longer falsifiable; the analyst is always more or less "right." This paper suggests that the integrated dual code formulation provides a more coherent theoretical framework for psychoanalysis than the mixed model, with important implications for theory and technique. In terms of dual coding, the problem is not that the nonverbal representations are vulnerable to distortion by words, but that the words that pass back and forth between analyst and patient will not affect the nonverbal schemata at all. Using the dual code
Fuel Management Study for a CANDU reactor Using New Physics Codes Suite
International Nuclear Information System (INIS)
Kim, Won Young; Kim, Bong Ghi; Park, Joo Hwan
2008-01-01
A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. The primary reactivity control in a CANDU reactor is the on-power refueling on a daily basis and an additional reactivity control is provided through an individual reactivity device movement, which includes 21 adjusters, 6 liquid zone controllers, 4 mechanical control absorbers and 2 shutdown systems. The refueling in CANDU is carried out on power and this makes the in-core fuel management different from that in a reactor refueled during shutdowns. The objective of a fuel management is to determine a fuel loading and fuel replacement procedure which will result in a minimum total unit energy cost in a safe and reliable operation. In this article, the in-core fuel management for the CANDU reactor was studied by using the new physics code suite of WIMS-IST/DRAGON-IST/RFSP-IST with the model of Wolsong-1 NPP
Self-shielding models of MICROX-2 code: Review and updates
International Nuclear Information System (INIS)
Hou, J.; Choi, H.; Ivanov, K.N.
2014-01-01
Highlights: • The MICROX-2 code has been improved to expand its application to advanced reactors. • New fine-group cross section libraries based on ENDF/B-VII have been generated. • Resonance self-shielding and spatial self-shielding models have been improved. • The improvements were assessed by a series of benchmark calculations against MCNPX. - Abstract: The MICROX-2 is a transport theory code that solves for the neutron slowing-down and thermalization equations of a two-region lattice cell. The MICROX-2 code has been updated to expand its application to advanced reactor concepts and fuel cycle simulations, including generation of new fine-group cross section libraries based on ENDF/B-VII. In continuation of previous work, the MICROX-2 methods are reviewed and updated in this study, focusing on its resonance self-shielding and spatial self-shielding models for neutron spectrum calculations. The improvement of self-shielding method was assessed by a series of benchmark calculations against the Monte Carlo code, using homogeneous and heterogeneous pin cell models. The results have shown that the implementation of the updated self-shielding models is correct and the accuracy of physics calculation is improved. Compared to the existing models, the updates reduced the prediction error of the infinite multiplication factor by ∼0.1% and ∼0.2% for the homogeneous and heterogeneous pin cell models, respectively, considered in this study
QR-codes as a tool to increase physical activity level among school children during class hours
DEFF Research Database (Denmark)
Christensen, Jeanette Reffstrup; Kristensen, Allan; Bredahl, Thomas Viskum Gjelstrup
the students physical activity level during class hours. Methods: A before-after study was used to examine 12 students physical activity level, measured with pedometers for six lessons. Three lessons of traditional teaching and three lessons, where QR-codes were used to make orienteering in school area...... as old fashioned. The students also felt positive about being physically active in teaching. Discussion and conclusion: QR-codes as a tool for teaching are usable for making students more physically active in teaching. The students were exited for using QR-codes and they experienced a good motivation......QR-codes as a tool to increase physical activity level among school children during class hours Introduction: Danish students are no longer fulfilling recommendations for everyday physical activity. Since August 2014, Danish students in public schools are therefore required to be physically active...
GEMSFITS: Code package for optimization of geochemical model parameters and inverse modeling
International Nuclear Information System (INIS)
Miron, George D.; Kulik, Dmitrii A.; Dmytrieva, Svitlana V.; Wagner, Thomas
2015-01-01
Highlights: • Tool for generating consistent parameters against various types of experiments. • Handles a large number of experimental data and parameters (is parallelized). • Has a graphical interface and can perform statistical analysis on the parameters. • Tested on fitting the standard state Gibbs free energies of aqueous Al species. • Example on fitting interaction parameters of mixing models and thermobarometry. - Abstract: GEMSFITS is a new code package for fitting internally consistent input parameters of GEM (Gibbs Energy Minimization) geochemical–thermodynamic models against various types of experimental or geochemical data, and for performing inverse modeling tasks. It consists of the gemsfit2 (parameter optimizer) and gfshell2 (graphical user interface) programs both accessing a NoSQL database, all developed with flexibility, generality, efficiency, and user friendliness in mind. The parameter optimizer gemsfit2 includes the GEMS3K chemical speciation solver ( (http://gems.web.psi.ch/GEMS3K)), which features a comprehensive suite of non-ideal activity- and equation-of-state models of solution phases (aqueous electrolyte, gas and fluid mixtures, solid solutions, (ad)sorption. The gemsfit2 code uses the robust open-source NLopt library for parameter fitting, which provides a selection between several nonlinear optimization algorithms (global, local, gradient-based), and supports large-scale parallelization. The gemsfit2 code can also perform comprehensive statistical analysis of the fitted parameters (basic statistics, sensitivity, Monte Carlo confidence intervals), thus supporting the user with powerful tools for evaluating the quality of the fits and the physical significance of the model parameters. The gfshell2 code provides menu-driven setup of optimization options (data selection, properties to fit and their constraints, measured properties to compare with computed counterparts, and statistics). The practical utility, efficiency, and
The 2010 fib Model Code for Structural Concrete: A new approach to structural engineering
Walraven, J.C.; Bigaj-Van Vliet, A.
2011-01-01
The fib Model Code is a recommendation for the design of reinforced and prestressed concrete which is intended to be a guiding document for future codes. Model Codes have been published before, in 1978 and 1990. The draft for fib Model Code 2010 was published in May 2010. The most important new
Plutonium explosive dispersal modeling using the MACCS2 computer code
International Nuclear Information System (INIS)
Steele, C.M.; Wald, T.L.; Chanin, D.I.
1998-01-01
The purpose of this paper is to derive the necessary parameters to be used to establish a defensible methodology to perform explosive dispersal modeling of respirable plutonium using Gaussian methods. A particular code, MACCS2, has been chosen for this modeling effort due to its application of sophisticated meteorological statistical sampling in accordance with the philosophy of Nuclear Regulatory Commission (NRC) Regulatory Guide 1.145, ''Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants''. A second advantage supporting the selection of the MACCS2 code for modeling purposes is that meteorological data sets are readily available at most Department of Energy (DOE) and NRC sites. This particular MACCS2 modeling effort focuses on the calculation of respirable doses and not ground deposition. Once the necessary parameters for the MACCS2 modeling are developed and presented, the model is benchmarked against empirical test data from the Double Tracks shot of project Roller Coaster (Shreve 1965) and applied to a hypothetical plutonium explosive dispersal scenario. Further modeling with the MACCS2 code is performed to determine a defensible method of treating the effects of building structure interaction on the respirable fraction distribution as a function of height. These results are related to the Clean Slate 2 and Clean Slate 3 bunkered shots of Project Roller Coaster. Lastly a method is presented to determine the peak 99.5% sector doses on an irregular site boundary in the manner specified in NRC Regulatory Guide 1.145 (1983). Parametric analyses are performed on the major analytic assumptions in the MACCS2 model to define the potential errors that are possible in using this methodology
Plutonium explosive dispersal modeling using the MACCS2 computer code
Energy Technology Data Exchange (ETDEWEB)
Steele, C.M.; Wald, T.L.; Chanin, D.I.
1998-11-01
The purpose of this paper is to derive the necessary parameters to be used to establish a defensible methodology to perform explosive dispersal modeling of respirable plutonium using Gaussian methods. A particular code, MACCS2, has been chosen for this modeling effort due to its application of sophisticated meteorological statistical sampling in accordance with the philosophy of Nuclear Regulatory Commission (NRC) Regulatory Guide 1.145, ``Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants``. A second advantage supporting the selection of the MACCS2 code for modeling purposes is that meteorological data sets are readily available at most Department of Energy (DOE) and NRC sites. This particular MACCS2 modeling effort focuses on the calculation of respirable doses and not ground deposition. Once the necessary parameters for the MACCS2 modeling are developed and presented, the model is benchmarked against empirical test data from the Double Tracks shot of project Roller Coaster (Shreve 1965) and applied to a hypothetical plutonium explosive dispersal scenario. Further modeling with the MACCS2 code is performed to determine a defensible method of treating the effects of building structure interaction on the respirable fraction distribution as a function of height. These results are related to the Clean Slate 2 and Clean Slate 3 bunkered shots of Project Roller Coaster. Lastly a method is presented to determine the peak 99.5% sector doses on an irregular site boundary in the manner specified in NRC Regulatory Guide 1.145 (1983). Parametric analyses are performed on the major analytic assumptions in the MACCS2 model to define the potential errors that are possible in using this methodology.
Cabin Environment Physics Risk Model
Mattenberger, Christopher J.; Mathias, Donovan Leigh
2014-01-01
This paper presents a Cabin Environment Physics Risk (CEPR) model that predicts the time for an initial failure of Environmental Control and Life Support System (ECLSS) functionality to propagate into a hazardous environment and trigger a loss-of-crew (LOC) event. This physics-of failure model allows a probabilistic risk assessment of a crewed spacecraft to account for the cabin environment, which can serve as a buffer to protect the crew during an abort from orbit and ultimately enable a safe return. The results of the CEPR model replace the assumption that failure of the crew critical ECLSS functionality causes LOC instantly, and provide a more accurate representation of the spacecraft's risk posture. The instant-LOC assumption is shown to be excessively conservative and, moreover, can impact the relative risk drivers identified for the spacecraft. This, in turn, could lead the design team to allocate mass for equipment to reduce overly conservative risk estimates in a suboptimal configuration, which inherently increases the overall risk to the crew. For example, available mass could be poorly used to add redundant ECLSS components that have a negligible benefit but appear to make the vehicle safer due to poor assumptions about the propagation time of ECLSS failures.
Evaluation of the HTR-10 Reactor as a Benchmark for Physics Code QA
International Nuclear Information System (INIS)
William K. Terry; Soon Sam Kim; Leland M. Montierth; Joshua J. Cogliati; Abderrafi M. Ougouag
2006-01-01
The HTR-10 is a small (10 MWt) pebble-bed research reactor intended to develop pebble-bed reactor (PBR) technology in China. It will be used to test and develop fuel, verify PBR safety features, demonstrate combined electricity production and co-generation of heat, and provide experience in PBR design, operation, and construction. As the only currently operating PBR in the world, the HTR-10 can provide data of great interest to everyone involved in PBR technology. In particular, if it yields data of sufficient quality, it can be used as a benchmark for assessing the accuracy of computer codes proposed for use in PBR analysis. This paper summarizes the evaluation for the International Reactor Physics Experiment Evaluation Project (IRPhEP) of data obtained in measurements of the HTR-10's initial criticality experiment for use as benchmarks for reactor physics codes
International Nuclear Information System (INIS)
Luneville, L.; Chiron, M.; Toubon, H.; Dogny, S.; Huver, M.; Berger, L.
2001-01-01
The research performed in common these last 3 years by the French Atomic Commission CEA, COGEMA and Eurisys Mesures had for main subject the realization of a complete tool of modelization for the largest range of realistic cases, the Pascalys modelization software. The main purpose of the modelization was to calculate the global measurement efficiency, which delivers the most accurate relationship between the photons emitted by the nuclear source in volume, punctual or deposited form and the germanium hyper pure detector, which detects and analyzes the received photons. It has been stated since long time that experimental global measurement efficiency becomes more and more difficult to address especially for complex scene as we can find in decommissioning and dismantling or in case of high activities for which the use of high activity reference sources become difficult to use for both health physics point of view and regulations. The choice of a calculation code is fundamental if accurate modelization is searched. MCNP represents the reference code but its use is long time calculation consuming and then not practicable in line on the field. Direct line-of-sight point kernel code as the French Atomic Commission 3-D analysis Mercure code can represent the practicable compromise between the most accurate MCNP reference code and the realistic performances needed in modelization. The comparison between the results of Pascalys-Mercure and MCNP code taking in account the last improvements of Mercure in the low energy range where the most important errors can occur, is presented in this paper, Mercure code being supported in line by the recent Pascalys 3-D modelization scene software. The incidence of the intrinsic efficiency of the Germanium detector is also approached for the total efficiency of measurement. (authors)
The WARP Code: Modeling High Intensity Ion Beams
International Nuclear Information System (INIS)
Grote, David P.; Friedman, Alex; Vay, Jean-Luc; Haber, Irving
2005-01-01
The Warp code, developed for heavy-ion driven inertial fusion energy studies, is used to model high intensity ion (and electron) beams. Significant capability has been incorporated in Warp, allowing nearly all sections of an accelerator to be modeled, beginning with the source. Warp has as its core an explicit, three-dimensional, particle-in-cell model. Alongside this is a rich set of tools for describing the applied fields of the accelerator lattice, and embedded conducting surfaces (which are captured at sub-grid resolution). Also incorporated are models with reduced dimensionality: an axisymmetric model and a transverse ''slice'' model. The code takes advantage of modern programming techniques, including object orientation, parallelism, and scripting (via Python). It is at the forefront in the use of the computational technique of adaptive mesh refinement, which has been particularly successful in the area of diode and injector modeling, both steady-state and time-dependent. In the presentation, some of the major aspects of Warp will be overviewed, especially those that could be useful in modeling ECR sources. Warp has been benchmarked against both theory and experiment. Recent results will be presented showing good agreement of Warp with experimental results from the STS500 injector test stand
The WARP Code: Modeling High Intensity Ion Beams
International Nuclear Information System (INIS)
Grote, D P; Friedman, A; Vay, J L; Haber, I
2004-01-01
The Warp code, developed for heavy-ion driven inertial fusion energy studies, is used to model high intensity ion (and electron) beams. Significant capability has been incorporated in Warp, allowing nearly all sections of an accelerator to be modeled, beginning with the source. Warp has as its core an explicit, three-dimensional, particle-in-cell model. Alongside this is a rich set of tools for describing the applied fields of the accelerator lattice, and embedded conducting surfaces (which are captured at sub-grid resolution). Also incorporated are models with reduced dimensionality: an axisymmetric model and a transverse ''slice'' model. The code takes advantage of modern programming techniques, including object orientation, parallelism, and scripting (via Python). It is at the forefront in the use of the computational technique of adaptive mesh refinement, which has been particularly successful in the area of diode and injector modeling, both steady-state and time-dependent. In the presentation, some of the major aspects of Warp will be overviewed, especially those that could be useful in modeling ECR sources. Warp has been benchmarked against both theory and experiment. Recent results will be presented showing good agreement of Warp with experimental results from the STS500 injector test stand. Additional information can be found on the web page http://hif.lbl.gov/theory/WARP( ) summary.html
DEFF Research Database (Denmark)
Johansen, Peter Meincke
1996-01-01
New uniform closed-form expressions for physical theory of diffraction equivalent edge currents are derived for truncated incremental wedge strips. In contrast to previously reported expressions, the new expressions are well-behaved for all directions of incidence and observation and take a finite...... value for zero strip length. Consequently, the new equivalent edge currents are, to the knowledge of the author, the first that are well-suited for implementation in general computer codes...
Physical-layer Network Coding in Two-Way Heterogeneous Cellular Networks with Power Imbalance
Thampi, Ajay K; Liew, Soung Chang; Armour, Simon M D; Fan, Zhong; You, Lizhao; Kaleshi, Dritan
2016-01-01
The growing demand for high-speed data, quality of service (QoS) assurance and energy efficiency has triggered the evolution of 4G LTE-A networks to 5G and beyond. Interference is still a major performance bottleneck. This paper studies the application of physical-layer network coding (PNC), a technique that exploits interference, in heterogeneous cellular networks. In particular, we propose a rate-maximising relay selection algorithm for a single cell with multiple relays assuming the decode...
Geochemical modelling of groundwater evolution using chemical equilibrium codes
International Nuclear Information System (INIS)
Pitkaenen, P.; Pirhonen, V.
1991-01-01
Geochemical equilibrium codes are a modern tool in studying interaction between groundwater and solid phases. The most common used programs and application subjects are shortly presented in this article. The main emphasis is laid on the approach method of using calculated results in evaluating groundwater evolution in hydrogeological system. At present in geochemical equilibrium modelling also kinetic as well as hydrologic constrains along a flow path are taken into consideration
Optical model calculations with the code ECIS95
Energy Technology Data Exchange (ETDEWEB)
Carlson, B V [Departamento de Fisica, Instituto Tecnologico da Aeronautica, Centro Tecnico Aeroespacial (Brazil)
2001-12-15
The basic features of elastic and inelastic scattering within the framework of the spherical and deformed nuclear optical models are discussed. The calculation of cross sections, angular distributions and other scattering quantities using J. Raynal's code ECIS95 is described. The use of the ECIS method (Equations Couplees en Iterations Sequentielles) in coupled-channels and distorted-wave Born approximation calculations is also reviewed. (author)
Excellence in Physics Education Award: Modeling Theory for Physics Instruction
Hestenes, David
2014-03-01
All humans create mental models to plan and guide their interactions with the physical world. Science has greatly refined and extended this ability by creating and validating formal scientific models of physical things and processes. Research in physics education has found that mental models created from everyday experience are largely incompatible with scientific models. This suggests that the fundamental problem in learning and understanding science is coordinating mental models with scientific models. Modeling Theory has drawn on resources of cognitive science to work out extensive implications of this suggestion and guide development of an approach to science pedagogy and curriculum design called Modeling Instruction. Modeling Instruction has been widely applied to high school physics and, more recently, to chemistry and biology, with noteworthy results.
A Multivariate Model of Physics Problem Solving
Taasoobshirazi, Gita; Farley, John
2013-01-01
A model of expertise in physics problem solving was tested on undergraduate science, physics, and engineering majors enrolled in an introductory-level physics course. Structural equation modeling was used to test hypothesized relationships among variables linked to expertise in physics problem solving including motivation, metacognitive planning,…
International codes and model intercomparison for intermediate energy activation yields
International Nuclear Information System (INIS)
Rolf, M.; Nagel, P.
1997-01-01
The motivation for this intercomparison came from data needs of accelerator-based waste transmutation, energy amplification and medical therapy. The aim of this exercise is to determine the degree of reliability of current nuclear reaction models and codes when calculating activation yields in the intermediate energy range up to 5000 MeV. Emphasis has been placed for a wide range of target elements ( O, Al, Fe, Co, Zr and Au). This work is mainly based on calculation of (P,xPyN) integral cross section for incident proton. A qualitative description of some of the nuclear models and code options employed is made. The systematics of graphical presentation of the results allows a quick quantitative measure of agreement or deviation. This code intercomparison highlights the fact that modeling calculations of energy activation yields may at best have uncertainties of a factor of two. The causes of such discrepancies are multi-factorial. Problems are encountered which are connected with the calculation of nuclear masses, binding energies, Q-values, shell effects, medium energy fission and Fermi break-up. (A.C.)
Film grain noise modeling in advanced video coding
Oh, Byung Tae; Kuo, C.-C. Jay; Sun, Shijun; Lei, Shawmin
2007-01-01
A new technique for film grain noise extraction, modeling and synthesis is proposed and applied to the coding of high definition video in this work. The film grain noise is viewed as a part of artistic presentation by people in the movie industry. On one hand, since the film grain noise can boost the natural appearance of pictures in high definition video, it should be preserved in high-fidelity video processing systems. On the other hand, video coding with film grain noise is expensive. It is desirable to extract film grain noise from the input video as a pre-processing step at the encoder and re-synthesize the film grain noise and add it back to the decoded video as a post-processing step at the decoder. Under this framework, the coding gain of the denoised video is higher while the quality of the final reconstructed video can still be well preserved. Following this idea, we present a method to remove film grain noise from image/video without distorting its original content. Besides, we describe a parametric model containing a small set of parameters to represent the extracted film grain noise. The proposed model generates the film grain noise that is close to the real one in terms of power spectral density and cross-channel spectral correlation. Experimental results are shown to demonstrate the efficiency of the proposed scheme.
Models and structures: mathematical physics
International Nuclear Information System (INIS)
2003-01-01
This document gathers research activities along 5 main directions. 1) Quantum chaos and dynamical systems. Recent results concern the extension of the exact WKB method that has led to a host of new results on the spectrum and wave functions. Progress have also been made in the description of the wave functions of chaotic quantum systems. Renormalization has been applied to the analysis of dynamical systems. 2) Combinatorial statistical physics. We see the emergence of new techniques applied to various such combinatorial problems, from random walks to random lattices. 3) Integrability: from structures to applications. Techniques of conformal field theory and integrable model systems have been developed. Progress is still made in particular for open systems with boundary conditions, in connection to strings and branes physics. Noticeable links between integrability and exact WKB quantization to 2-dimensional disordered systems have been highlighted. New correlations of eigenvalues and better connections to integrability have been formulated for random matrices. 4) Gravities and string theories. We have developed aspects of 2-dimensional string theory with a particular emphasis on its connection to matrix models as well as non-perturbative properties of M-theory. We have also followed an alternative path known as loop quantum gravity. 5) Quantum field theory. The results obtained lately concern its foundations, in flat or curved spaces, but also applications to second-order phase transitions in statistical systems
Overview of the Graphical User Interface for the GERM Code (GCR Event-Based Risk Model
Kim, Myung-Hee; Cucinotta, Francis A.
2010-01-01
The descriptions of biophysical events from heavy ions are of interest in radiobiology, cancer therapy, and space exploration. The biophysical description of the passage of heavy ions in tissue and shielding materials is best described by a stochastic approach that includes both ion track structure and nuclear interactions. A new computer model called the GCR Event-based Risk Model (GERM) code was developed for the description of biophysical events from heavy ion beams at the NASA Space Radiation Laboratory (NSRL). The GERM code calculates basic physical and biophysical quantities of high-energy protons and heavy ions that have been studied at NSRL for the purpose of simulating space radiobiological effects. For mono-energetic beams, the code evaluates the linear-energy transfer (LET), range (R), and absorption in tissue equivalent material for a given Charge (Z), Mass Number (A) and kinetic energy (E) of an ion. In addition, a set of biophysical properties are evaluated such as the Poisson distribution of ion or delta-ray hits for a specified cellular area, cell survival curves, and mutation and tumor probabilities. The GERM code also calculates the radiation transport of the beam line for either a fixed number of user-specified depths or at multiple positions along the Bragg curve of the particle. The contributions from primary ion and nuclear secondaries are evaluated. The GERM code accounts for the major nuclear interaction processes of importance for describing heavy ion beams, including nuclear fragmentation, elastic scattering, and knockout-cascade processes by using the quantum multiple scattering fragmentation (QMSFRG) model. The QMSFRG model has been shown to be in excellent agreement with available experimental data for nuclear fragmentation cross sections, and has been used by the GERM code for application to thick target experiments. The GERM code provides scientists participating in NSRL experiments with the data needed for the interpretation of their
EBT time-dependent point model code: description and user's guide
International Nuclear Information System (INIS)
Roberts, J.F.; Uckan, N.A.
1977-07-01
A D-T time-dependent point model has been developed to assess the energy balance in an EBT reactor plasma. Flexibility is retained in the model to permit more recent data to be incorporated as they become available from the theoretical and experimental studies. This report includes the physics models involved, the program logic, and a description of the variables and routines used. All the files necessary for execution are listed, and the code, including a post-execution plotting routine, is discussed
Direct containment heating models in the CONTAIN code
International Nuclear Information System (INIS)
Washington, K.E.; Williams, D.C.
1995-08-01
The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment of direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale
Direct containment heating models in the CONTAIN code
Energy Technology Data Exchange (ETDEWEB)
Washington, K.E.; Williams, D.C.
1995-08-01
The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment of direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale.
Channel modeling, signal processing and coding for perpendicular magnetic recording
Wu, Zheng
With the increasing areal density in magnetic recording systems, perpendicular recording has replaced longitudinal recording to overcome the superparamagnetic limit. Studies on perpendicular recording channels including aspects of channel modeling, signal processing and coding techniques are presented in this dissertation. To optimize a high density perpendicular magnetic recording system, one needs to know the tradeoffs between various components of the system including the read/write transducers, the magnetic medium, and the read channel. We extend the work by Chaichanavong on the parameter optimization for systems via design curves. Different signal processing and coding techniques are studied. Information-theoretic tools are utilized to determine the acceptable region for the channel parameters when optimal detection and linear coding techniques are used. Our results show that a considerable gain can be achieved by the optimal detection and coding techniques. The read-write process in perpendicular magnetic recording channels includes a number of nonlinear effects. Nonlinear transition shift (NLTS) is one of them. The signal distortion induced by NLTS can be reduced by write precompensation during data recording. We numerically evaluate the effect of NLTS on the read-back signal and examine the effectiveness of several write precompensation schemes in combating NLTS in a channel characterized by both transition jitter noise and additive white Gaussian electronics noise. We also present an analytical method to estimate the bit-error-rate and use it to help determine the optimal write precompensation values in multi-level precompensation schemes. We propose a mean-adjusted pattern-dependent noise predictive (PDNP) detection algorithm for use on the channel with NLTS. We show that this detector can offer significant improvements in bit-error-rate (BER) compared to conventional Viterbi and PDNP detectors. Moreover, the system performance can be further improved by
Modelling of LOCA Tests with the BISON Fuel Performance Code
Energy Technology Data Exchange (ETDEWEB)
Williamson, Richard L [Idaho National Laboratory; Pastore, Giovanni [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Spencer, Benjamin Whiting [Idaho National Laboratory; Hales, Jason Dean [Idaho National Laboratory
2016-05-01
BISON is a modern finite-element based, multidimensional nuclear fuel performance code that is under development at Idaho National Laboratory (USA). Recent advances of BISON include the extension of the code to the analysis of LWR fuel rod behaviour during loss-of-coolant accidents (LOCAs). In this work, BISON models for the phenomena relevant to LWR cladding behaviour during LOCAs are described, followed by presentation of code results for the simulation of LOCA tests. Analysed experiments include separate effects tests of cladding ballooning and burst, as well as the Halden IFA-650.2 fuel rod test. Two-dimensional modelling of the experiments is performed, and calculations are compared to available experimental data. Comparisons include cladding burst pressure and temperature in separate effects tests, as well as the evolution of fuel rod inner pressure during ballooning and time to cladding burst. Furthermore, BISON three-dimensional simulations of separate effects tests are performed, which demonstrate the capability to reproduce the effect of azimuthal temperature variations in the cladding. The work has been carried out in the frame of the collaboration between Idaho National Laboratory and Halden Reactor Project, and the IAEA Coordinated Research Project FUMAC.
The top-down reflooding model in the Cathare code
International Nuclear Information System (INIS)
Bartak, J.; Bestion, D.; Haapalehto, T.
1993-01-01
A top-down reflooding model was developed for the French best-estimate thermalhydraulic code CATHARE. The paper presents the current state of development of this model. Based on a literature survey and on compatibility considerations with respect to the existing CATHARE bottom reflooding package, a falling film top-down reflooding model was developed and implemented into CATHARE version 1.3E. Following a brief review of previous work, the paper describes the most important features of the model. The model was validated with the WINFRITH single tube top-down reflooding experiment and with the REWET - II simultaneous bottom and top-down reflooding experiment in rod bundle geometry. The results demonstrate the ability of the new package to describe the falling film rewetting phenomena and the main parametric trends both in a simple analytical experimental setup and in a much more complex rod bundle reflooding experiment. (authors). 9 figs., 28 refs
Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code
International Nuclear Information System (INIS)
Leppaenen, J.
2007-01-01
High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels. (authors)
Toward a Probabilistic Automata Model of Some Aspects of Code-Switching.
Dearholt, D. W.; Valdes-Fallis, G.
1978-01-01
The purpose of the model is to select either Spanish or English as the language to be used; its goals at this stage of development include modeling code-switching for lexical need, apparently random code-switching, dependency of code-switching upon sociolinguistic context, and code-switching within syntactic constraints. (EJS)
Subotin, Michael; Davis, Anthony R
2016-09-01
Natural language processing methods for medical auto-coding, or automatic generation of medical billing codes from electronic health records, generally assign each code independently of the others. They may thus assign codes for closely related procedures or diagnoses to the same document, even when they do not tend to occur together in practice, simply because the right choice can be difficult to infer from the clinical narrative. We propose a method that injects awareness of the propensities for code co-occurrence into this process. First, a model is trained to estimate the conditional probability that one code is assigned by a human coder, given than another code is known to have been assigned to the same document. Then, at runtime, an iterative algorithm is used to apply this model to the output of an existing statistical auto-coder to modify the confidence scores of the codes. We tested this method in combination with a primary auto-coder for International Statistical Classification of Diseases-10 procedure codes, achieving a 12% relative improvement in F-score over the primary auto-coder baseline. The proposed method can be used, with appropriate features, in combination with any auto-coder that generates codes with different levels of confidence. The promising results obtained for International Statistical Classification of Diseases-10 procedure codes suggest that the proposed method may have wider applications in auto-coding. © The Author 2016. Published by Oxford University Press on behalf of the American Medical Informatics Association. All rights reserved. For Permissions, please email: journals.permissions@oup.com.
Benchmarking of computer codes and approaches for modeling exposure scenarios
International Nuclear Information System (INIS)
Seitz, R.R.; Rittmann, P.D.; Wood, M.I.; Cook, J.R.
1994-08-01
The US Department of Energy Headquarters established a performance assessment task team (PATT) to integrate the activities of DOE sites that are preparing performance assessments for the disposal of newly generated low-level waste. The PATT chartered a subteam with the task of comparing computer codes and exposure scenarios used for dose calculations in performance assessments. This report documents the efforts of the subteam. Computer codes considered in the comparison include GENII, PATHRAE-EPA, MICROSHIELD, and ISOSHLD. Calculations were also conducted using spreadsheets to provide a comparison at the most fundamental level. Calculations and modeling approaches are compared for unit radionuclide concentrations in water and soil for the ingestion, inhalation, and external dose pathways. Over 30 tables comparing inputs and results are provided
Modeling RIA scenarios with the FRAPTRAN and SCANAIR codes
International Nuclear Information System (INIS)
Sagrado Garcia, I. C.; Vallejo, I.; Herranz, L. E.
2013-01-01
The need of defining new RIA safety criteria has pointed out the importance of performing a rigorous assessment of the transient codes capabilities. The present work is a comparative exercise devoted to identify the origin of the key deviations found between the predictions of FRAPTRAN-1.4 and SCANAIR-7.1. To do so, the calculations submitted by CIEMAT to the OECD/NEA RIA benchmark have been exploited. This work shows that deviations in clad temperatures mainly come from the treatment of the oxide layer. The systematically higher deformations calculated by FRAPTRAN-1.4 in early failed tests are caused by the different gap closure estimation. Besides, the dissimilarities observed in the FGR predictions are inherent to the different modeling strategies adopted in each code.
Modeling RIA scenarios with the FRAPTRAN and SCANAIR codes
Energy Technology Data Exchange (ETDEWEB)
Sagrado Garcia, I. C.; Vallejo, I.; Herranz, L. E.
2013-07-01
The need of defining new RIA safety criteria has pointed out the importance of performing a rigorous assessment of the transient codes capabilities. The present work is a comparative exercise devoted to identify the origin of the key deviations found between the predictions of FRAPTRAN-1.4 and SCANAIR-7.1. To do so, the calculations submitted by CIEMAT to the OECD/NEA RIA benchmark have been exploited. This work shows that deviations in clad temperatures mainly come from the treatment of the oxide layer. The systematically higher deformations calculated by FRAPTRAN-1.4 in early failed tests are caused by the different gap closure estimation. Besides, the dissimilarities observed in the FGR predictions are inherent to the different modeling strategies adopted in each code.
Simplified model for radioactive contaminant transport: the TRANSS code
International Nuclear Information System (INIS)
Simmons, C.S.; Kincaid, C.T.; Reisenauer, A.E.
1986-09-01
A simplified ground-water transport model called TRANSS was devised to estimate the rate of migration of a decaying radionuclide that is subject to sorption governed by a linear isotherm. Transport is modeled as a contaminant mass transmitted along a collection of streamlines constituting a streamtube, which connects a source release zone with an environmental arrival zone. The probability-weighted contaminant arrival distribution along each streamline is represented by an analytical solution of the one-dimensional advection-dispersion equation with constant velocity and dispersion coefficient. The appropriate effective constant velocity for each streamline is based on the exact travel time required to traverse a streamline with a known length. An assumption used in the model to facilitate the mathematical simplification is that transverse dispersion within a streamtube is negligible. Release of contaminant from a source is described in terms of a fraction-remaining curve provided as input information. However, an option included in the code is the calculation of a fraction-remaining curve based on four specialized release models: (1) constant release rate, (2) solubility-controlled release, (3) adsorption-controlled release, and (4) diffusion-controlled release from beneath an infiltration barrier. To apply the code, a user supplies only a certain minimal number of parameters: a probability-weighted list of travel times for streamlines, a local-scale dispersion coefficient, a sorption distribution coefficient, total initial radionuclide inventory, radioactive half-life, a release model choice, and size dimensions of the source. The code is intended to provide scoping estimates of contaminant transport and does not predict the evolution of a concentration distribution in a ground-water flow field. Moreover, the required travel times along streamlines must be obtained from a prior ground-water flow simulation
A critical flow model for the Cathena thermalhydraulic code
International Nuclear Information System (INIS)
Popov, N.K.; Hanna, B.N.
1990-01-01
The calculation of critical flow rate, e.g., of choked flow through a break, is required for simulating a loss of coolant transient in a reactor or reactor-like experimental facility. A model was developed to calculate the flow rate through the break for given geometrical parameters near the break and fluid parameters upstream of the break for ordinary water, as well as heavy water, with or without non- condensible gases. This model has been incorporated in the CATHENA, one-dimensional, two-fluid thermalhydraulic code. In the CATHENA code a standard staggered-mesh, finite-difference representation is used to solve the thermalhydraulic equations. This model compares the fluid mixture velocity, calculated using the CATHENA momentum equations, with a critical velocity. When the mixture velocity is smaller than the critical velocity, the flow is assumed to be subcritical, and the model remains passive. When the fluid mixture velocity is higher than the critical velocity, the model sets the fluid mixture velocity equal to the critical velocity. In this paper the critical velocity at a link (momentum cell) is first estimated separately for single-phase liquid, two- phase, or single-phase gas flow condition at the upstream node (mass/energy cell). In all three regimes non-condensible gas can be present in the flow. For single-phase liquid flow, the critical velocity is estimated using a Bernoulli- type of equation, the pressure at the link is estimated by the pressure undershoot method
Implementation of 3D models in the Monte Carlo code MCNP
International Nuclear Information System (INIS)
Lopes, Vivaldo; Millian, Felix M.; Guevara, Maria Victoria M.; Garcia, Fermin; Sena, Isaac; Menezes, Hugo
2009-01-01
On the area of numerical dosimetry Applied to medical physics, the scientific community focuses on the elaboration of new hybrids models based on 3D models. But different steps of the process of simulation with 3D models needed improvement and optimization in order to expedite the calculations and accuracy using this methodology. This project was developed with the aim of optimize the process of introduction of 3D models within the simulation code of radiation transport by Monte Carlo (MCNP). The fast implementation of these models on the simulation code allows the estimation of the dose deposited on the patient organs on a more personalized way, increasing the accuracy with this on the estimates and reducing the risks to health, caused by ionizing radiations. The introduction o these models within the MCNP was made through a input file, that was constructed through a sequence of images, bi-dimensional in the 3D model, generated using the program '3DSMAX', imported by the program 'TOMO M C' and thus, introduced as INPUT FILE of the MCNP code. (author)
High Energy Transport Code HETC
International Nuclear Information System (INIS)
Gabriel, T.A.
1985-09-01
The physics contained in the High Energy Transport Code (HETC), in particular the collision models, are discussed. An application using HETC as part of the CALOR code system is also given. 19 refs., 5 figs., 3 tabs
AX-GADGET: a new code for cosmological simulations of Fuzzy Dark Matter and Axion models
Nori, Matteo; Baldi, Marco
2018-05-01
We present a new module of the parallel N-Body code P-GADGET3 for cosmological simulations of light bosonic non-thermal dark matter, often referred as Fuzzy Dark Matter (FDM). The dynamics of the FDM features a highly non-linear Quantum Potential (QP) that suppresses the growth of structures at small scales. Most of the previous attempts of FDM simulations either evolved suppressed initial conditions, completely neglecting the dynamical effects of QP throughout cosmic evolution, or resorted to numerically challenging full-wave solvers. The code provides an interesting alternative, following the FDM evolution without impairing the overall performance. This is done by computing the QP acceleration through the Smoothed Particle Hydrodynamics (SPH) routines, with improved schemes to ensure precise and stable derivatives. As an extension of the P-GADGET3 code, it inherits all the additional physics modules implemented up to date, opening a wide range of possibilities to constrain FDM models and explore its degeneracies with other physical phenomena. Simulations are compared with analytical predictions and results of other codes, validating the QP as a crucial player in structure formation at small scales.
Modeling of fission product release in integral codes
International Nuclear Information System (INIS)
Obaidurrahman, K.; Raman, Rupak K.; Gaikwad, Avinash J.
2014-01-01
The Great Tohoku earthquake and tsunami that stroke the Fukushima-Daiichi nuclear power station in March 11, 2011 has intensified the needs of detailed nuclear safety research and with this objective all streams associated with severe accident phenomenology are being revisited thoroughly. The present paper would cover an overview of state of art FP release models being used, the important phenomenon considered in semi-mechanistic models and knowledge gaps in present FP release modeling. Capability of FP release module, ELSA of ASTEC integral code in appropriate prediction of FP release under several diversified core degraded conditions will also be demonstrated. Use of semi-mechanistic fission product release models at AERB in source-term estimation shall be briefed. (author)
Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes
International Nuclear Information System (INIS)
Holowach, M.J.; Hochreiter, L.E.; Cheung, F.B.
2002-01-01
A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation. (authors)
International Nuclear Information System (INIS)
Lindemuth, I.R.
1979-01-01
This report describes ANIMAL, a two-dimensional Eulerian magnetohydrodynamic computer code. ANIMAL's physical model also appears. Formulated are temporal and spatial finite-difference equations in a manner that facilitates implementation of the algorithm. Outlined are the functions of the algorithm's FORTRAN subroutines and variables
DEFF Research Database (Denmark)
Elesin, Y; Gerya, T; Artemieva, Irina
2010-01-01
We present a new 2D finite difference code, Samovar, for high-resolution numerical modeling of complex geodynamic processes. Examples are collision of lithospheric plates (including mountain building and subduction) and lithosphere extension (including formation of sedimentary basins, regions...... of extended crust, and rift zones). The code models deformation of the lithosphere with viscoelastoplastic rheology, including erosion/sedimentation processes and formation of shear zones in areas of high stresses. It also models steady-state and transient conductive and advective thermal processes including...... partial melting and magma transport in the lithosphere. The thermal and mechanical parts of the code are tested for a series of physical problems with analytical solutions. We apply the code to geodynamic modeling by examining numerically the processes of lithosphere extension and basin formation...
Recent developments of JAEA's Monte Carlo Code MVP for reactor physics applications
International Nuclear Information System (INIS)
Nagaya, Y.; Okumura, K.; Mori, T.
2013-01-01
MVP is a general-purpose continuous-energy Monte Carlo code for neutron and photon transport calculations that has been developed since the late 1980's at Japan Atomic Energy Agency (JAEA, formerly JAERI). The MVP code is designed for nuclear reactor applications such as reactor core design/analysis, criticality safety and reactor shielding. This paper describes the MVP code and present its latest developments. Among the new capabilities of MVP we find: -) the perturbation method has been implemented for the change in k(eff); -) the eigenvalue calculations can be performed with an explicit treatment of delayed neutrons in which their fission spectra are taken into account; -) the capability of tallying the scattering matrix (group-to-group scattering cross sections); -) the implementation of an exact model for resonance elastic scattering; and -) a Monte Carlo perturbation technique is used to calculate reactor kinetics parameters
Dataset of coded handwriting features for use in statistical modelling
Directory of Open Access Journals (Sweden)
Anna Agius
2018-02-01
Full Text Available The data presented here is related to the article titled, “Using handwriting to infer a writer's country of origin for forensic intelligence purposes” (Agius et al., 2017 [1]. This article reports original writer, spatial and construction characteristic data for thirty-seven English Australian11 In this study, English writers were Australians whom had learnt to write in New South Wales (NSW. writers and thirty-seven Vietnamese writers. All of these characteristics were coded and recorded in Microsoft Excel 2013 (version 15.31. The construction characteristics coded were only extracted from seven characters, which were: ‘g’, ‘h’, ‘th’, ‘M’, ‘0’, ‘7’ and ‘9’. The coded format of the writer, spatial and construction characteristics is made available in this Data in Brief in order to allow others to perform statistical analyses and modelling to investigate whether there is a relationship between the handwriting features and the nationality of the writer, and whether the two nationalities can be differentiated. Furthermore, to employ mathematical techniques that are capable of characterising the extracted features from each participant.
Auditory information coding by modeled cochlear nucleus neurons.
Wang, Huan; Isik, Michael; Borst, Alexander; Hemmert, Werner
2011-06-01
In this paper we use information theory to quantify the information in the output spike trains of modeled cochlear nucleus globular bushy cells (GBCs). GBCs are part of the sound localization pathway. They are known for their precise temporal processing, and they code amplitude modulations with high fidelity. Here we investigated the information transmission for a natural sound, a recorded vowel. We conclude that the maximum information transmission rate for a single neuron was close to 1,050 bits/s, which corresponds to a value of approximately 5.8 bits per spike. For quasi-periodic signals like voiced speech, the transmitted information saturated as word duration increased. In general, approximately 80% of the available information from the spike trains was transmitted within about 20 ms. Transmitted information for speech signals concentrated around formant frequency regions. The efficiency of neural coding was above 60% up to the highest temporal resolution we investigated (20 μs). The increase in transmitted information to that precision indicates that these neurons are able to code information with extremely high fidelity, which is required for sound localization. On the other hand, only 20% of the information was captured when the temporal resolution was reduced to 4 ms. As the temporal resolution of most speech recognition systems is limited to less than 10 ms, this massive information loss might be one of the reasons which are responsible for the lack of noise robustness of these systems.
MMA, A Computer Code for Multi-Model Analysis
Poeter, Eileen P.; Hill, Mary C.
2007-01-01
This report documents the Multi-Model Analysis (MMA) computer code. MMA can be used to evaluate results from alternative models of a single system using the same set of observations for all models. As long as the observations, the observation weighting, and system being represented are the same, the models can differ in nearly any way imaginable. For example, they may include different processes, different simulation software, different temporal definitions (for example, steady-state and transient models could be considered), and so on. The multiple models need to be calibrated by nonlinear regression. Calibration of the individual models needs to be completed before application of MMA. MMA can be used to rank models and calculate posterior model probabilities. These can be used to (1) determine the relative importance of the characteristics embodied in the alternative models, (2) calculate model-averaged parameter estimates and predictions, and (3) quantify the uncertainty of parameter estimates and predictions in a way that integrates the variations represented by the alternative models. There is a lack of consensus on what model analysis methods are best, so MMA provides four default methods. Two are based on Kullback-Leibler information, and use the AIC (Akaike Information Criterion) or AICc (second-order-bias-corrected AIC) model discrimination criteria. The other two default methods are the BIC (Bayesian Information Criterion) and the KIC (Kashyap Information Criterion) model discrimination criteria. Use of the KIC criterion is equivalent to using the maximum-likelihood Bayesian model averaging (MLBMA) method. AIC, AICc, and BIC can be derived from Frequentist or Bayesian arguments. The default methods based on Kullback-Leibler information have a number of theoretical advantages, including that they tend to favor more complicated models as more data become available than do the other methods, which makes sense in many situations. Many applications of MMA will
Modeling of severe accident sequences with the new modules CESAR and DIVA of ASTEC system code
International Nuclear Information System (INIS)
Pignet, Sophie; Guillard, Gaetan; Barre, Francois; Repetto, Georges
2003-01-01
Systems of computer codes, so-called 'integral' codes, are being developed to simulate the scenario of a hypothetical severe accident in a light water reactor, from the initial event until the possible radiological release of fission products out of the containment. They couple the predominant physical phenomena that occur in the different reactor zones and simulate the actuation of safety systems by procedures and by operators. In order to allow to study a great number of scenarios, a compromise must be found between precision of results and calculation time: one day of accident time should take less than one day of real time to simulate on a PC computer. This search of compromise is a real challenge for such integral codes. The development of the ASTEC integral code was initiated jointly by IRSN and GRS as an international reference code. The latest version 1.0 of ASTEC, including the new modules CESAR and DIVA which model the behaviour of the reactor cooling system and the core degradation, is presented here. Validation of the modules and one plant application are described
Assessment CANDU physics codes using experimental data - part 1: criticality measurement
International Nuclear Information System (INIS)
Roh, Gyu Hong; Choi, Hang Bok; Jeong, Chang Joon
2001-08-01
In order to assess the applicability of MCNP-4B code to the heavy water moderated, light water cooled and pressure-tube type reactor, the MCNP-4B physics calculations has been carried out for the Deuterium Critical Assembly (DCA), and the results were compared with those of the experimental data. In this study, the key safety parameters like as the multiplication factor, void coefficient, local power peaking factor and bundle power distribution in the scattered core are simulated. In order to use the cross section data consistently for the fuels to be analyzed in the future, new MCNP libraries have been generated from ENDF/B-VI release 3. Generally, the MCNP-4B calculation results show a good agreement with experimental data of DCA core. After benchmarking MCNP-4B against available experimental data, it will be used as the reference tool to benchmark design and analysis codes for the advanced CANDU fuels
Verification of MVP-II and SRAC2006 code to the core physics vera benchmark problem
International Nuclear Information System (INIS)
Jati Susilo
2014-01-01
In this research, verification calculation for VERA core physics benchmark on the Zero Power Physical Test (ZPPT) of the nuclear reactor Watts Bar 1. The reactor is a 1000 MWe class of PWR designed by. Westinghouse, arranged from 193 unit of 17 x 17 fuel assembly consisting 3 type enrichment of UO2 that are 2.1wt%, 2.619wt% and 3.1wt%. Core power factor distribution and k-eff calculation has been done for the first cycle operation of the core at beginning of cycle (BOC) and hot zero power (HZP). In this calculation, MVP-II and CITATION module of SRAC2006 computer code has been used with ENDF/B-VII.0. cross section data library. Calculation result showed that differences value of k-eff for the core at controlled and uncontrolled condition between reference with MVP-II (-0,07% and -0,014%) and SRAC2006 (0,92% and 0,99%) are very small or below 1%. Differences value of radial power peaking factor at controlled and uncontrolled of the core between reference value with MVP-II are 0,38% and 1,53%, even though with SRAC2006 are 1,13% and -2,45%. It can be said that the calculation result by both computer code showing suitability with reference value. In order to determinate of criticality of the core, the calculation result using MVP-II code is more conservative compare with SRAC2006 code. (author)
An improved steam generator model for the SASSYS code
International Nuclear Information System (INIS)
Pizzica, P.A.
1989-01-01
A new steam generator model has been developed for the SASSYS computer code, which analyzes accident conditions in a liquid metal cooled fast reactor. It has been incorporated into the new SASSYS balance-of-plant model but it can also function on a stand-alone basis. The steam generator can be used in a once-through mode, or a variant of the model can be used as a separate evaporator and a superheater with recirculation loop. The new model provides for an exact steady-state solution as well as the transient calculation. There was a need for a faster and more flexible model than the old steam generator model. The new model provides for more detail with its multi-mode treatment as opposed to the previous model's one node per region approach. Numerical instability problems which were the result of cell-centered spatial differencing, fully explicit time differencing, and the moving boundary treatment of the boiling crisis point in the boiling region have been reduced. This leads to an increase in speed as larger time steps can now be taken. The new model is an improvement in many respects. 2 refs., 3 figs
Auxiliary plasma heating and fueling models for use in particle simulation codes
International Nuclear Information System (INIS)
Procassini, R.J.; Cohen, B.I.
1989-01-01
Computational models of a radiofrequency (RF) heating system and neutral-beam injector are presented. These physics packages, when incorporated into a particle simulation code allow one to simulate the auxiliary heating and fueling of fusion plasmas. The RF-heating package is based upon a quasilinear diffusion equation which describes the slow evolution of the heated particle distribution. The neutral-beam injector package models the charge exchange and impact ionization processes which transfer energy and particles from the beam to the background plasma. Particle simulations of an RF-heated and a neutral-beam-heated simple-mirror plasma are presented. 8 refs., 5 figs
SPIDERMAN: an open-source code to model phase curves and secondary eclipses
Louden, Tom; Kreidberg, Laura
2018-03-01
We present SPIDERMAN (Secondary eclipse and Phase curve Integrator for 2D tempERature MAppiNg), a fast code for calculating exoplanet phase curves and secondary eclipses with arbitrary surface brightness distributions in two dimensions. Using a geometrical algorithm, the code solves exactly the area of sections of the disc of the planet that are occulted by the star. The code is written in C with a user-friendly Python interface, and is optimised to run quickly, with no loss in numerical precision. Approximately 1000 models can be generated per second in typical use, making Markov Chain Monte Carlo analyses practicable. The modular nature of the code allows easy comparison of the effect of multiple different brightness distributions for the dataset. As a test case we apply the code to archival data on the phase curve of WASP-43b using a physically motivated analytical model for the two dimensional brightness map. The model provides a good fit to the data; however, it overpredicts the temperature of the nightside. We speculate that this could be due to the presence of clouds on the nightside of the planet, or additional reflected light from the dayside. When testing a simple cloud model we find that the best fitting model has a geometric albedo of 0.32 ± 0.02 and does not require a hot nightside. We also test for variation of the map parameters as a function of wavelength and find no statistically significant correlations. SPIDERMAN is available for download at https://github.com/tomlouden/spiderman.
SPIDERMAN: an open-source code to model phase curves and secondary eclipses
Louden, Tom; Kreidberg, Laura
2018-06-01
We present SPIDERMAN (Secondary eclipse and Phase curve Integrator for 2D tempERature MAppiNg), a fast code for calculating exoplanet phase curves and secondary eclipses with arbitrary surface brightness distributions in two dimensions. Using a geometrical algorithm, the code solves exactly the area of sections of the disc of the planet that are occulted by the star. The code is written in C with a user-friendly Python interface, and is optimized to run quickly, with no loss in numerical precision. Approximately 1000 models can be generated per second in typical use, making Markov Chain Monte Carlo analyses practicable. The modular nature of the code allows easy comparison of the effect of multiple different brightness distributions for the data set. As a test case, we apply the code to archival data on the phase curve of WASP-43b using a physically motivated analytical model for the two-dimensional brightness map. The model provides a good fit to the data; however, it overpredicts the temperature of the nightside. We speculate that this could be due to the presence of clouds on the nightside of the planet, or additional reflected light from the dayside. When testing a simple cloud model, we find that the best-fitting model has a geometric albedo of 0.32 ± 0.02 and does not require a hot nightside. We also test for variation of the map parameters as a function of wavelength and find no statistically significant correlations. SPIDERMAN is available for download at https://github.com/tomlouden/spiderman.
Modeling of the CTEx subcritical unit using MCNPX code
International Nuclear Information System (INIS)
Santos, Avelino; Silva, Ademir X. da; Rebello, Wilson F.; Cunha, Victor L. Lassance
2011-01-01
The present work aims at simulating the subcritical unit of Army Technology Center (CTEx) namely ARGUS pile (subcritical uranium-graphite arrangement) by using the computational code MCNPX. Once such modeling is finished, it could be used in k-effective calculations for systems using natural uranium as fuel, for instance. ARGUS is a subcritical assembly which uses reactor-grade graphite as moderator of fission neutrons and metallic uranium fuel rods with aluminum cladding. The pile is driven by an Am-Be spontaneous neutron source. In order to achieve a higher value for k eff , a higher concentration of U235 can be proposed, provided it safely remains below one. (author)
Status of emergency spray modelling in the integral code ASTEC
International Nuclear Information System (INIS)
Plumecocq, W.; Passalacqua, R.
2001-01-01
Containment spray systems are emergency systems that would be used in very low probability events which may lead to severe accidents in Light Water Reactors. In most cases, the primary function of the spray would be to remove heat and condense steam in order to reduce pressure and temperature in the containment building. Spray would also wash out fission products (aerosols and gaseous species) from the containment atmosphere. The efficiency of the spray system in the containment depressurization as well as in the removal of aerosols, during a severe accident, depends on the evolution of the spray droplet size distribution with the height in the containment, due to kinetic and thermal relaxation, gravitational agglomeration and mass transfer with the gas. A model has been developed taking into account all of these phenomena. This model has been implemented in the ASTEC code with a validation of the droplets relaxation against the CARAIDAS experiment (IPSN). Applications of this modelling to a PWR 900, during a severe accident, with special emphasis on the effect of spray on containment hydrogen distribution have been performed in multi-compartment configuration with the ASTEC V0.3 code. (author)
49 CFR 41.120 - Acceptable model codes.
2010-10-01
... 1991 International Conference of Building Officials (ICBO) Uniform Building Code, published by the... Supplement to the Building Officials and Code Administrators International (BOCA) National Building Code, published by the Building Officials and Code Administrators, 4051 West Flossmoor Rd., Country Club Hills...
NaI(Tl) detectors modeling in MCNP-X and Gate/Geant4 codes
Energy Technology Data Exchange (ETDEWEB)
Affonso, Renato Raoni Werneck; Silva, Ademir Xavier da, E-mail: raoniwa@yahoo.com.br, E-mail: ademir@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Salgado, Cesar Marques, E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2017-07-01
NaI (Tl) detectors are widely used in gamma-ray densitometry, but their modeling in Monte Carlo codes, such as MCNP-X and Gate/Geant4, needs a lot of work and does not yield comparable results with experimental arrangements, possibly due to non-simulated physical phenomena, such as light transport within the scintillator. Therefore, it is necessary a methodology that positively impacts the results of the simulations while maintaining the real dimensions of the detectors and other objects to allow validating a modeling that matches up with the experimental arrangement. Thus, the objective of this paper is to present the studies conducted with the MCNPX and Gate/Geant4 codes, in which the comparisons of their results were satisfactory, showing that both can be used for the same purposes. (author)
Model comparisons of the reactive burn model SURF in three ASC codes
Energy Technology Data Exchange (ETDEWEB)
Whitley, Von Howard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stalsberg, Krista Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reichelt, Benjamin Lee [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Shipley, Sarah Jayne [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2018-01-12
A study of the SURF reactive burn model was performed in FLAG, PAGOSA and XRAGE. In this study, three different shock-to-detonation transition experiments were modeled in each code. All three codes produced similar model results for all the experiments modeled and at all resolutions. Buildup-to-detonation time, particle velocities and resolution dependence of the models was notably similar between the codes. Given the current PBX 9502 equations of state and SURF calibrations, each code is equally capable of predicting the correct detonation time and distance when impacted by a 1D impactor at pressures ranging from 10-16 GPa, as long as the resolution of the mesh is not too coarse.
Energy Technology Data Exchange (ETDEWEB)
NONE
2003-07-01
These scientific days had for objective to draw up the situation of calculation codes of radiation transport, of sources estimation, of radiation doses managements and to draw the future perspectives. (N.C.)
Effects of physics change in Monte Carlo code on electron pencil beam dose distributions
International Nuclear Information System (INIS)
Toutaoui, Abdelkader; Khelassi-Toutaoui, Nadia; Brahimi, Zakia; Chami, Ahmed Chafik
2012-01-01
Pencil beam algorithms used in computerized electron beam dose planning are usually described using the small angle multiple scattering theory. Alternatively, the pencil beams can be generated by Monte Carlo simulation of electron transport. In a previous work, the 4th version of the Electron Gamma Shower (EGS) Monte Carlo code was used to obtain dose distributions from monoenergetic electron pencil beam, with incident energy between 1 MeV and 50 MeV, interacting at the surface of a large cylindrical homogeneous water phantom. In 2000, a new version of this Monte Carlo code has been made available by the National Research Council of Canada (NRC), which includes various improvements in its electron-transport algorithms. In the present work, we were interested to see if the new physics in this version produces pencil beam dose distributions very different from those calculated with oldest one. The purpose of this study is to quantify as well as to understand these differences. We have compared a series of pencil beam dose distributions scored in cylindrical geometry, for electron energies between 1 MeV and 50 MeV calculated with two versions of the Electron Gamma Shower Monte Carlo Code. Data calculated and compared include isodose distributions, radial dose distributions and fractions of energy deposition. Our results for radial dose distributions show agreement within 10% between doses calculated by the two codes for voxels closer to the pencil beam central axis, while the differences are up to 30% for longer distances. For fractions of energy deposition, the results of the EGS4 are in good agreement (within 2%) with those calculated by EGSnrc at shallow depths for all energies, whereas a slightly worse agreement (15%) is observed at deeper distances. These differences may be mainly attributed to the different multiple scattering for electron transport adopted in these two codes and the inclusion of spin effect, which produces an increase of the effective range of
Theory and application of the RAZOR two-dimensional continuous energy lattice physics code
International Nuclear Information System (INIS)
Zerkle, M.L.; Abu-Shumays, I.K.; Ott, M.W.; Winwood, J.P.
1997-01-01
The theory and application of the RAZOR two-dimensional, continuous energy lattice physics code are discussed. RAZOR solves the continuous energy neutron transport equation in one- and two-dimensional geometries, and calculates equivalent few-group diffusion theory constants that rigorously account for spatial and spectral self-shielding effects. A dual energy resolution slowing down algorithm is used to reduce computer memory and disk storage requirements for the slowing down calculation. Results are presented for a 2D BWR pin cell depletion benchmark problem
Physical-layer network coding for passive optical interconnect in datacenter networks.
Lin, Rui; Cheng, Yuxin; Guan, Xun; Tang, Ming; Liu, Deming; Chan, Chun-Kit; Chen, Jiajia
2017-07-24
We introduce physical-layer network coding (PLNC) technique in a passive optical interconnect (POI) architecture for datacenter networks. The implementation of the PLNC in the POI at 2.5 Gb/s and 10Gb/s have been experimentally validated while the gains in terms of network layer performances have been investigated by simulation. The results reveal that in order to realize negligible packet drop, the wavelengths usage can be reduced by half while a significant improvement in packet delay especially under high traffic load can be achieved by employing PLNC over POI.
Constitutive model development needs for reactor safety thermal-hydraulic codes
International Nuclear Information System (INIS)
Kelly, J.M.
1998-01-01
This paper discusses the constitutive model development needs for our current and future generation of reactor safety thermal-hydraulic analysis codes. Rather than provide a simple 'shopping list' of models to be improved, a detailed description is given of how a constitutive model works within the computational framework of a current reactor safety code employing the two-fluid model of two-phase flow. The intent is to promote a better understanding of both the types of experiments and the instrumentation needs that will be required in the USNRCs code improvement program. First, a summary is given of the modeling considerations that need to be taken into account when developing constitutive models for use in reactor safety thermal-hydraulic codes. Specifically, the two-phase flow model should be applicable to a control volume formulation employing computational volumes with dimensions on the order of meters but containing embedded structure with a dimension on the order of a centimeter. The closure relations are then required to be suitable when averaged over such large volumes containing millions or even tens of millions of discrete fluid particles (bubbles/drops). This implies a space and time averaging procedure that neglects the intermittency observed in slug and chum turbulent two-phase flows. Furthermore, the geometries encountered in reactor systems are complex, the constitutive relations should therefore be component specific (e.g., interfacial shear in a tube does not represent that in a rod bundle nor in the downcomer). When practicable, future modeling efforts should be directed towards resolving the spatial evolution of two-phase flow patterns through the introduction of interfacial area transport equations and by modeling the individual physical processes responsible for the creation or destruction of interfacial area. Then the example of the implementation and assessment of a subcooled boiling model in a two-fluid code is given. The primary parameter
DEFF Research Database (Denmark)
Knudsen, Vibeke Kildegaard; Gille, Maj-Britt; Nielsen, Trine Holmgaard
2011-01-01
Objective: To determine the relative validity of the pre-coded food diary applied in the Danish National Survey of Dietary Habits and Physical Activity. Design: A cross-over study among seventy-two adults (aged 20 to 69 years) recording diet by means of a pre-coded food diary over 4 d and a 4 d...
C code generation applied to nonlinear model predictive control for an artificial pancreas
DEFF Research Database (Denmark)
Boiroux, Dimitri; Jørgensen, John Bagterp
2017-01-01
This paper presents a method to generate C code from MATLAB code applied to a nonlinear model predictive control (NMPC) algorithm. The C code generation uses the MATLAB Coder Toolbox. It can drastically reduce the time required for development compared to a manual porting of code from MATLAB to C...
The effect of hot spots upon swelling of Zircaloy cladding as modelled by the code CANSWEL-2
International Nuclear Information System (INIS)
Haste, T.J.; Gittus, J.H.
1980-12-01
The code CANSWEL-2 models cladding creep deformation under conditions relevant to a loss-of-coolant accident (LOCA) in a pressurised-water reactor (PWR). It can treat azimuthal non-uniformities in cladding thickness and temperature, and model the mechanical restraint imposed by the nearest neighbouring rods, including situations where cladding is forced into non-circular shapes. The physical and mechanical models used in the code are presented. Applications of the code are described, both as a stand-alone version and as part of the PWR LOCA code MABEL-2. Comparison with a limited number of relevant out-of-reactor creep strain experiments has generally shown encouraging agreement with the data. (author)
Kinetic models of gene expression including non-coding RNAs
Energy Technology Data Exchange (ETDEWEB)
Zhdanov, Vladimir P., E-mail: zhdanov@catalysis.r
2011-03-15
In cells, genes are transcribed into mRNAs, and the latter are translated into proteins. Due to the feedbacks between these processes, the kinetics of gene expression may be complex even in the simplest genetic networks. The corresponding models have already been reviewed in the literature. A new avenue in this field is related to the recognition that the conventional scenario of gene expression is fully applicable only to prokaryotes whose genomes consist of tightly packed protein-coding sequences. In eukaryotic cells, in contrast, such sequences are relatively rare, and the rest of the genome includes numerous transcript units representing non-coding RNAs (ncRNAs). During the past decade, it has become clear that such RNAs play a crucial role in gene expression and accordingly influence a multitude of cellular processes both in the normal state and during diseases. The numerous biological functions of ncRNAs are based primarily on their abilities to silence genes via pairing with a target mRNA and subsequently preventing its translation or facilitating degradation of the mRNA-ncRNA complex. Many other abilities of ncRNAs have been discovered as well. Our review is focused on the available kinetic models describing the mRNA, ncRNA and protein interplay. In particular, we systematically present the simplest models without kinetic feedbacks, models containing feedbacks and predicting bistability and oscillations in simple genetic networks, and models describing the effect of ncRNAs on complex genetic networks. Mathematically, the presentation is based primarily on temporal mean-field kinetic equations. The stochastic and spatio-temporal effects are also briefly discussed.
International Nuclear Information System (INIS)
Salko, Robert K.; Schmidt, Rodney C.; Avramova, Maria N.
2015-01-01
parallel, two additional libraries are currently needed: MPI, for inter-processor message passing, and the Parallel Extensible Toolkit for Scientific Computation (PETSc), which is used to solve the global pressure matrix in parallel. Results presented include a set of testing and verification calculations and performance tests assessing parallel scaling characteristics up to a full-core, pincell-resolved model of a PWR core containing 193 17 × 17 assemblies under hot full-power conditions. This model, representative of Watts Bar Unit 1 and containing about 56,000 pins, was modeled with roughly 59,000 subchannels, leading to about 2.8 million thermal–hydraulic control volumes in total. Results demonstrate that CTF can now perform full-core analysis of a PWR (not previously possible owing to excessively long runtimes and memory requirements) on the order of 20 min. This new capability not only is useful to stand-alone CTF users, but also is being leveraged in support of coupled code multi-physics calculations being done in the CASL program
On boundary layer modelling using the ASTEC code
International Nuclear Information System (INIS)
Smith, B.L.
1991-07-01
The modelling of fluid boundary layers adjacent to non-slip, heated surface using the ASTEC code is described. The pricipal boundary layer characteristics are derived using simple dimensional arguments and these are developed into criteria for optimum placement of the computational mesh to achieve realistic simulation. In particular, the need for externally-imposed drag and heat transfer correlations as a function of the local mesh concentration is discussed in the context of both laminar and turbulent flow conditions. Special emphasis is placed in the latter case on the (k-ε) turbulence model, which is standard in the code. As far as possible, the analyses are pursued from first principles, so that no comprehensive knowledge of the history of the subject is required for the general ASTEC user to derive practical advice from the document. Some attention is paid to the use of heat transfer correlations for internal solid/fluid surfaces, whose treatment is not straightforward in ASTEC. It is shown that three formulations are possible to effect the heat transfer, called Explicit, Jacobian and Implicit. The particular advantages and disadvantages of each are discussed with regard to numerical stability and computational efficiency. (author) 18 figs., 1 tab., 39 refs
Physicochemical analog for modeling superimposed and coded memories
Ensanian, Minas
1992-07-01
The mammalian brain is distinguished by a life-time of memories being stored within the same general region of physicochemical space, and having two extraordinary features. First, memories to varying degrees are superimposed, as well as coded. Second, instantaneous recall of past events can often be affected by relatively simple, and seemingly unrelated sensory clues. For the purposes of attempting to mathematically model such complex behavior, and for gaining additional insights, it would be highly advantageous to be able to simulate or mimic similar behavior in a nonbiological entity where some analogical parameters of interest can reasonably be controlled. It has recently been discovered that in nonlinear accumulative metal fatigue memories (related to mechanical deformation) can be superimposed and coded in the crystal lattice, and that memory, that is, the total number of stress cycles can be recalled (determined) by scanning not the surfaces but the `edges' of the objects. The new scanning technique known as electrotopography (ETG) now makes the state space modeling of metallic networks possible. The author provides an overview of the new field and outlines the areas that are of immediate interest to the science of artificial neural networks.
International Nuclear Information System (INIS)
Hainoun, A.; Alhabit, F.; Ghazi, N.
2008-01-01
Two new modifications have been included in the current PARET code that is widely applied in the dynamic and safety analysis of research reactors. A new model was implemented for the simulation of void formation in the subcooled boiling regime, the other modification dealt with the implementation of a new approach to improve the prediction of heat transfer coefficient under natural circulation condition. The modified code was successfully validated using adequate single effect tests covering the physical phenomena of interest for both natural circulation and subcooled void formation at low pressure and low heat flux. The validation results indicate significant improvement of the code compared to the default version. Additionally, to simplify the code application an interactive user interface was developed enabling pre and post-processing of the code predictions. (author)
Phenomenological modeling of critical heat flux: The GRAMP code and its validation
International Nuclear Information System (INIS)
Ahmad, M.; Chandraker, D.K.; Hewitt, G.F.; Vijayan, P.K.; Walker, S.P.
2013-01-01
Highlights: ► Assessment of CHF limits is vital for LWR optimization and safety analysis. ► Phenomenological modeling is a valuable adjunct to pure empiricism. ► It is based on empirical representations of the (several, competing) phenomena. ► Phenomenological modeling codes making ‘aggregate’ predictions need careful assessment against experiments. ► The physical and mathematical basis of a phenomenological modeling code GRAMP is presented. ► The GRAMP code is assessed against measurements from BARC (India) and Harwell (UK), and the Look Up Tables. - Abstract: Reliable knowledge of the critical heat flux is vital for the design of light water reactors, for both safety and optimization. The use of wholly empirical correlations, or equivalently “Look Up Tables”, can be very effective, but is generally less so in more complex cases, and in particular cases where the heat flux is axially non-uniform. Phenomenological models are in principle more able to take into account of a wider range of conditions, with a less comprehensive coverage of experimental measurements. These models themselves are in part based upon empirical correlations, albeit of the more fundamental individual phenomena occurring, rather than the aggregate behaviour, and as such they too require experimental validation. In this paper we present the basis of a general-purpose phenomenological code, GRAMP, and then use two independent ‘direct’ sets of measurement, from BARC in India and from Harwell in the United Kingdom, and the large dataset embodied in the Look Up Tables, to perform a validation exercise on it. Very good agreement between predictions and experimental measurements is observed, adding to the confidence with which the phenomenological model can be used. Remaining important uncertainties in the phenomenological modeling of CHF, namely the importance of the initial entrained fraction on entry to annular flow, and the influence of the heat flux on entrainment rate
7 CFR Exhibit E to Subpart A of... - Voluntary National Model Building Codes
2010-01-01
... 7 Agriculture 12 2010-01-01 2010-01-01 false Voluntary National Model Building Codes E Exhibit E... National Model Building Codes The following documents address the health and safety aspects of buildings and related structures and are voluntary national model building codes as defined in § 1924.4(h)(2) of...
International Nuclear Information System (INIS)
Simon-Cornu, Marie; Mourlon, Christophe; Bordy, J.M.; Daures, J.; Dusiac, D.; Moignau, F.; Gouriou, J.; Million, M.; Moreno, B.; Chabert, I.; Lazaro, D.; Barat, E.; Dautremer, T.; Montagu, T.; Agelou, M.; De Carlan, L.; Patin, D.; Le Loirec, C.; Dupuis, P.; Gassa, F.; Guerin, L.; Batalla, A.; Leni, Pierre-Emmanuel; Laurent, Remy; Gschwind, Regine; Makovicka, Libor; Henriet, Julien; Salomon, Michel; Vivier, Alain; Lopez, Gerald; Dossat, C.; Pourrouquet, P.; Thomas, J.C.; Sarie, I.; Peyrard, P.F.; Chatry, N.; Lavielle, D.; Loze, R.; Brun, E.; Damian, F.; Diop, C.; Dumonteil, E.; Hugot, F.X.; Jouanne, C.; Lee, Y.K.; Malvagi, F.; Mazzolo, A.; Petit, O.; Trama, J.C.; Visonneau, T.; Zoia, A.; Courageot, Estelle; Gaillard-Lecanu, Emmanuelle; Kutschera, Reinald; Le Meur, Gaelle; Uzio, Fabien; De Conto, Celine; Gschwind, Regine; Makovicka, Libor; Farah, Jad; Martinetti, Florent; Sayah, Rima; Donadille, Laurent; Herault, Joel; Delacroix, Sabine; Nauraye, Catherine; Lee, Choonsik; Bolch, Wesley; Clairand, Isabelle; Horodynski, Jean-Michel; Pauwels, Nicolas; Robert, Pierre; VOLLAIRE, Joachim; Nicoletti, C.; Kitsos, S.; Tardy, M.; Marchaud, G.; Stankovskiy, Alexey; Van Den Eynde, Gert; Fiorito, Luca; Malambu, Edouard; Dreuil, Serge; Mougeot, X.; Be, M.M.; Bisch, C.; Villagrasa, C.; Dos Santos, M.; Clairand, I.; Karamitros, M.; Incerti, S.; Petitguillaume, Alice; Franck, Didier; Desbree, Aurelie; Bernardini, Michela; Labriolle-Vaylet, Claire de; Gnesin, Silvano; Leadermann, Jean-Pascal; Paterne, Loic; Bochud, Francois O.; Verdun, Francis R.; Baechler, Sebastien; Prior, John O.; Thomassin, Alain; Arial, Emmanuelle; Laget, Michael; Masse, Veronique; Saldarriaga Vargas, Clarita; Struelens, Lara; Vanhavere, Filip; Perier, Aurelien; Courageot, Estelle; Gaillard-Lecanu, Emmanuelle; Le-Meur, Gaelle; Monier, Catherine; Thers, Dominique; Le-Guen, Bernard; Blond, Serge; Cordier, Gerard; Le Roy, Maiwenn; De Carlan, Loic; Bordy, Jean-Marc; Caccia, Barbara; Andenna, Claudio; Charimadurai, Arun; Selvam, T Palani; Czarnecki, Damian; Zink, Klemens; Gschwind, Regine; Martin, Eric; Huot, Nicolas; Zoubair, Mariam; El Bardouni, Tarek; Lazaro, Delphine; Barat, Eric; Dautremer, Thomas; Montagu, Thierry; Chabert, Isabelle; Guerin, Lucie; Batalla, Alain; Moignier, C.; Huet, C.; Bassinet, C.; Baumann, M.; Barraux, V.; Sebe-Mercier, K.; Loiseau, C.; Batalla, A.; Makovicka, L.; Desnoyers, Yvon; Juhel, Gabriel; Mattera, Christophe; Tempier, Maryline
2014-03-01
These scientific days were organised by the 'technical protection' Section of the French Society of Radiation Protection (SFRP) in cooperation with the French society of medical physicists (SFPM), the Swiss Romandie association of radioprotection (ARRAD) and the associated laboratories of radio-physics and dosimetry (LARD). The objective of these days was to review the existing calculation codes used in radiation transport, source estimation and dose management, and to identify some future prospects. This document brings together the available presentations (slides) together with their corresponding abstracts (in French) and dealing with: 1 - Presentation of the conference days (L. De Carlan); 2 - Simulating radionuclide transfers in the environment: what calculation codes and for what? (C. Mourlon); 3 - Contribution of Monte-Carlo calculation to the theoretical foundation analysis of calibration procedures and dosemeters design for radioprotection photon dosimetry (J.M. Bordy); 4 - Use of calculation codes in R and D for the development of a new passive dosemeter for photons and beta radiations (B. Moreno); 5 - Development of a new virtual sources model for the Monte-Carlo prediction of EPID (Electronic Portal Imaging Device) images and implementation in PENELOPE (I. Chabert); 6 - Prediction of high-resolution EPID images for in-vivo dosimetry (D. Patin); 7 - 4D thorax modeling by artificial neural networks (P.E. Leni); 8 - Presentation of the calculation utilities of the book 'Calculation of ionizing radiations generated doses' (Vivier, Lopez, EDP Sciences 2012) (A. Vivier); 9 - RayXpert C : a 3D modeling and Monte-Carlo dose rate calculation software (C. Dossat); 10 - TRIPOLI-4 R Version 9 S Monte-Carlo code for radioprotection (F. Damian); 11 - Realistic radioprotection training with the digital school workshop (E. Courageot); 12 - Use of BEAMNRC code for dental prostheses influence evaluation in ENT cancers treatment by external radiotherapy (C. De Conto); 13
Introduction to the simulation with MCNP Monte Carlo code and its applications in Medical Physics
International Nuclear Information System (INIS)
Parreno Z, F.; Paucar J, R.; Picon C, C.
1998-01-01
The simulation by Monte Carlo is tool which Medical Physics counts with it for the development of its research, the interest by this tool is growing, as we may observe in the main scientific journals for the years 1995-1997 where more than 27 % of the papers treat over Monte Carlo and/or its applications in the radiation transport.In the Peruvian Institute of Nuclear Energy we are implementing and making use of the MCNP4 and EGS4 codes. In this work are presented the general features of the Monte Carlo method and its more useful applications in Medical Physics. Likewise, it is made a simulation of the calculation of isodose curves in an interstitial treatment with Ir-192 wires in a mammary gland carcinoma. (Author)
Benchmarking of epithermal methods in the lattice-physics code EPRI-CELL
International Nuclear Information System (INIS)
Williams, M.L.; Wright, R.Q.; Barhen, J.; Rothenstein, W.; Toney, B.
1982-01-01
The epithermal cross section shielding methods used in the lattice physics code EPRI-CELL (E-C) have been extensively studied to determine its major approximations and to examine the sensitivity of computed results to these approximations. The study has resulted in several improvements in the original methodology. These include: treatment of the external moderator source with intermediate resonance (IR) theory, development of a new Dancoff factor expression to account for clad interactions, development of a new method for treating resonance interference, and application of a generalized least squares method to compute best-estimate values for the Bell factor and group-dependent IR parameters. The modified E-C code with its new ENDF/B-V cross section library is tested for several numerical benchmark problems. Integral parameters computed by EC are compared with those obtained with point-cross section Monte Carlo calculations, and E-C fine group cross sections are benchmarked against point-cross section descrete ordinates calculations. It is found that the code modifications improve agreement between E-C and the more sophisticated methods. E-C shows excellent agreement on the integral parameters and usually agrees within a few percent on fine-group, shielded cross sections
Physical and numerical modeling of Joule-heated melters
Energy Technology Data Exchange (ETDEWEB)
Eyler, L.L.; Skarda, R.J.; Crowder, R.S. III; Trent, D.S.; Reid, C.R.; Lessor, D.L.
1985-10-01
The Joule-heated ceramic-lined melter is an integral part of the high level waste immobilization process under development by the US Department of Energy. Scaleup and design of this waste glass melting furnace requires an understanding of the relationships between melting cavity design parameters and the furnace performance characteristics such as mixing, heat transfer, and electrical requirements. Developing empirical models of these relationships through actual melter testing with numerous designs would be a very costly and time consuming task. Additionally, the Pacific Northwest Laboratory (PNL) has been developing numerical models that simulate a Joule-heated melter for analyzing melter performance. This report documents the method used and results of this modeling effort. Numerical modeling results are compared with the more conventional, physical modeling results to validate the approach. Also included are the results of numerically simulating an operating research melter at PNL. Physical Joule-heated melters modeling results used for qualiying the simulation capabilities of the melter code included: (1) a melter with a single pair of electrodes and (2) a melter with a dual pair (two pairs) of electrodes. The physical model of the melter having two electrode pairs utilized a configuration with primary and secondary electrodes. The principal melter parameters (the ratio of power applied to each electrode pair, modeling fluid depth, electrode spacing) were varied in nine tests of the physical model during FY85. Code predictions were made for five of these tests. Voltage drops, temperature field data, and electric field data varied in their agreement with the physical modeling results, but in general were judged acceptable. 14 refs., 79 figs., 17 tabs.
Physical and numerical modeling of Joule-heated melters
International Nuclear Information System (INIS)
Eyler, L.L.; Skarda, R.J.; Crowder, R.S. III; Trent, D.S.; Reid, C.R.; Lessor, D.L.
1985-10-01
The Joule-heated ceramic-lined melter is an integral part of the high level waste immobilization process under development by the US Department of Energy. Scaleup and design of this waste glass melting furnace requires an understanding of the relationships between melting cavity design parameters and the furnace performance characteristics such as mixing, heat transfer, and electrical requirements. Developing empirical models of these relationships through actual melter testing with numerous designs would be a very costly and time consuming task. Additionally, the Pacific Northwest Laboratory (PNL) has been developing numerical models that simulate a Joule-heated melter for analyzing melter performance. This report documents the method used and results of this modeling effort. Numerical modeling results are compared with the more conventional, physical modeling results to validate the approach. Also included are the results of numerically simulating an operating research melter at PNL. Physical Joule-heated melters modeling results used for qualiying the simulation capabilities of the melter code included: (1) a melter with a single pair of electrodes and (2) a melter with a dual pair (two pairs) of electrodes. The physical model of the melter having two electrode pairs utilized a configuration with primary and secondary electrodes. The principal melter parameters (the ratio of power applied to each electrode pair, modeling fluid depth, electrode spacing) were varied in nine tests of the physical model during FY85. Code predictions were made for five of these tests. Voltage drops, temperature field data, and electric field data varied in their agreement with the physical modeling results, but in general were judged acceptable. 14 refs., 79 figs., 17 tabs
Larmat, C. S.; Rougier, E.; Delorey, A.; Steedman, D. W.; Bradley, C. R.
2016-12-01
The goal of the Source Physics Experiment (SPE) is to bring empirical and theoretical advances to the problem of detection and identification of underground nuclear explosions. For this, the SPE program includes a strong modeling effort based on first principles calculations with the challenge to capture both the source and near-source processes and those taking place later in time as seismic waves propagate within complex 3D geologic environments. In this paper, we report on results of modeling that uses hydrodynamic simulation codes (Abaqus and CASH) coupled with a 3D full waveform propagation code, SPECFEM3D. For modeling the near source region, we employ a fully-coupled Euler-Lagrange (CEL) modeling capability with a new continuum-based visco-plastic fracture model for simulation of damage processes, called AZ_Frac. These capabilities produce high-fidelity models of various factors believed to be key in the generation of seismic waves: the explosion dynamics, a weak grout-filled borehole, the surrounding jointed rock, and damage creation and deformations happening around the source and the free surface. SPECFEM3D, based on the Spectral Element Method (SEM) is a direct numerical method for full wave modeling with mathematical accuracy. The coupling interface consists of a series of grid points of the SEM mesh situated inside of the hydrodynamic code's domain. Displacement time series at these points are computed using output data from CASH or Abaqus (by interpolation if needed) and fed into the time marching scheme of SPECFEM3D. We will present validation tests with the Sharpe's model and comparisons of waveforms modeled with Rg waves (2-8Hz) that were recorded up to 2 km for SPE. We especially show effects of the local topography, velocity structure and spallation. Our models predict smaller amplitudes of Rg waves for the first five SPE shots compared to pure elastic models such as Denny &Johnson (1991).
Isotopic modelling using the ENIGMA-B fuel performance code
International Nuclear Information System (INIS)
Rossiter, G.D.; Cook, P.M.A.; Weston, R.
2001-01-01
A number of experimental programmes by BNFL and other MOX fabricators have now shown that the in-pile performance of MOX fuel is generally similar to that of conventional UO 2 fuel. Models based on UO 2 fuel experience form a good basis for a description of MOX fuel behaviour. However, an area where the performance of MOX fuel is sufficiently different from that of UO 2 to warrant model changes is in the radial power and burnup profile. The differences in radial power and burnup profile arise from the presence of significant concentrations of plutonium in MOX fuel, at beginning of life, and their subsequent evolution with burnup. Amongst other effects, plutonium has a greater neutron absorption cross-section than uranium. This paper focuses on the development of a new model for the radial power and burnup profile within a UO 2 or MOX fuel rod, in which the underlying fissile isotope concentration distributions are tracked during irradiation. The new model has been incorporated into the ENIGMA-B fuel performance code and has been extended to track the isotopic concentrations of the fission gases, xenon and krypton. The calculated distributions have been validated against results from rod puncture measurements and electron probe micro-analysis (EPMA) linescans, performed during the M501 post irradiation examination (PIE) programme. The predicted gas inventory of the fuel/clad gap is compared with the isotopic composition measured during rod puncture and the measured radial distributions of burnup (from neodymium measurements) and plutonium in the fuel are compared with the calculated distributions. It is shown that there is good agreement between the code predictions and the measurements. (author)
Atmospheric radiative transfer modeling: a summary of the AER codes
Energy Technology Data Exchange (ETDEWEB)
Clough, S.A. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States); Shephard, M.W. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States)]. E-mail: mshephar@aer.com; Mlawer, E.J. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States); Delamere, J.S. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States); Iacono, M.J. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States); Cady-Pereira, K. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States); Boukabara, S. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States); Brown, P.D. [Atmospheric and Environmental Research (AER) Inc., 131 Hartwell Avenue, Lexington, MA 02421-3126 (United States)
2005-03-01
The radiative transfer models developed at AER are being used extensively for a wide range of applications in the atmospheric sciences. This communication is intended to provide a coherent summary of the various radiative transfer models and associated databases publicly available from AER (http://www.rtweb.aer.com). Among the communities using the models are the remote sensing community (e.g. TES, IASI), the numerical weather prediction community (e.g. ECMWF, NCEP GFS, WRF, MM5), and the climate community (e.g. ECHAM5). Included in this communication is a description of the central features and recent updates for the following models: the line-by-line radiative transfer model (LBLRTM); the line file creation program (LNFL); the longwave and shortwave rapid radiative transfer models, RRTM{sub L}W and RRTM{sub S}W; the Monochromatic Radiative Transfer Model (MonoRTM); the MT{sub C}KD Continuum; and the Kurucz Solar Source Function. LBLRTM and the associated line parameter database (e.g. HITRAN 2000 with 2001 updates) play a central role in the suite of models. The physics adopted for LBLRTM has been extensively analyzed in the context of closure experiments involving the evaluation of the model inputs (e.g. atmospheric state), spectral radiative measurements and the spectral model output. The rapid radiative transfer models are then developed and evaluated using the validated LBLRTM model.
Release of WIMS10: a versatile reactor physics code for thermal and fast systems - 15467
International Nuclear Information System (INIS)
Lindley, B.A.; Newton, T.D.; Hosking, J.G.; Smith, P.N.; Powney, D.J.; Tollit, B.; Smith, P.J.
2015-01-01
the WIMS code provides a versatile software package for neutronic calculations, which can be applied to all thermal reactor types including mixed moderator systems. It can provide lattice cell and supercell calculations using a range of flux solutions methods to produce the neutronic libraries for use in PANTHER or other whole core analysis codes. With the release of WIMS10, the range of problems which WIMS can solve has been greatly extended. A WIMS/PANTHER calculation route has been developed and validated for part MOX-fuelled PWRs, with calculations showing excellent agreement with 2D core deterministic and Monte Carlo transport solutions. A flexible geometry 3D method of characteristics transport solver, CACTUS3D has been added to the code. CACTUS3D has been benchmarked for a 3D BWR assembly model, and was in good agreement with a direct 172-group solution in the Monte Carlo code MONK. Fast reactor calculations using the ECCO deterministic calculation route have been validated using experimental data from the ZEBRA reactor. Power deposition can be treated through following neutrons and/or photons to their point of interaction. The improved methodology is shown to give more accurate calculation of heat deposition and improve agreement between calculated and measured detector responses for part MOX-fuelled cores. (authors)
Development of Integrated Code for Risk Assessment (INCORIA) for Physical Protection System
International Nuclear Information System (INIS)
Jang, Sung Soon; Seo, Hyung Min; Yoo, Ho Sik
2010-01-01
A physical protection system (PPS) integrates people, procedures and equipment for the protection of assets or facilities against theft, sabotage or other malevolent human attacks. Among critical facilities, nuclear facilities and nuclear weapon sites require the highest level of PPS. After the September 11, 2001 terrorist attacks, international communities, including the IAEA, have made substantial efforts to protect nuclear material and nuclear facilities. The international flow on nuclear security is using the concept or risk assessment. The concept of risk assessment is firstly devised by nuclear safety people. They considered nuclear safety including its possible risk, which is the frequency of failure and possible consequence. Nuclear security people also considers security risk, which is the frequency of threat action, vulnerability, and consequences. The concept means that we should protect more when the credible threat exists and the possible radiological consequence is high. Even if there are several risk assessment methods of nuclear security, the application needs the help of tools because of a lot of calculation. It's also hard to find tools for whole phase of risk assessment. Several codes exist for the part of risk assessment. SAVI are used for vulnerability of PPS. Vital area identification code is used for consequence analysis. We are developing Integrated Code for Risk Assessment (INCORIA) to apply risk assessment methods for nuclear facilities. INCORIA evaluates PP-KINAC measures and generation tools for threat scenario. PP-KINAC is risk assessment measures for physical protection system developed by Hosik Yoo and is easy to apply. A threat scenario tool is used to generate threat scenario, which is used as one of input value to PP-KINAC measures
Literature Review of Dredging Physical Models
This U.S. Army Engineer Research and Development Center, Coastal and Hydraulics Laboratory, special report presents a review of dredging physical ...model studies with the goal of understanding the most current state of dredging physical modeling, understanding conditions of similitude used in past...studies, and determining whether the flow field around a dredging operation has been quantified. Historical physical modeling efforts have focused on
TU Electric reactor physics model verification: Power reactor benchmark
International Nuclear Information System (INIS)
Willingham, C.E.; Killgore, M.R.
1988-01-01
Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors
International Nuclear Information System (INIS)
Arriaga R, L.; Del Valle G, E.; Gomez T, A. M.
2013-10-01
For the Serpent code was developed the three-dimensional model corresponding to the nuclear subcritical assembly (S A) Chicago 9000 of the Escuela Superior de Fisica y Matematicas del Instituto Politecnico Nacional (ESFM-IPN). The model includes: a) the core, formed by 312 aluminum pipes that contain 5 nuclear fuel rods (natural uranium in metallic form), b) the multi-perforated plates where they penetrate the inferior part of each pipe to be able to remain in vertical form, c) water, acting as moderator and reflector, and d) the recipient lodging to the core. The pipes arrangement is hexagonal although the transversal section of the recipient that lodges to the core is circular. The entrance file for the Serpent code was generated with the data provided by the manual of the S A use about the composition and density of the fuel rods and others obtained in direct form of the rods, as the interior and external diameter, mass and height. Of the obtained physical parameters, those more approached to that reported in the manual of the subcritical assembly are the effective multiplication factor and the reproduction factor η. The differences can be because the description of the fuel rods provided by the manual of the S A use do not correspond those that are physically in the S A core. This difference consists on the presence of a circular central channel of 1.245 diameter centimeters in each fuel rod. The fuel rods reported in the mentioned manual do not have that channel. Although the obtained results are encouraging, we want to continue improving the model to incorporate in this the detectors, defined this way by the Serpent code, which could determine the existent neutrons flux in diverse points of interest like the axial or radial aligned points and to compare these with those that are obtained in an experimental way when a generating neutrons source (Pu-Be) is introduced. Added to this effort the cross sections for each unitary cell will be determined, so that
Modelling guidelines for core exit temperature simulations with system codes
Energy Technology Data Exchange (ETDEWEB)
Freixa, J., E-mail: jordi.freixa-terradas@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Martínez-Quiroga, V., E-mail: victor.martinez@nortuen.com [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Zerkak, O., E-mail: omar.zerkak@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Reventós, F., E-mail: francesc.reventos@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain)
2015-05-15
Highlights: • Core exit temperature is used in PWRs as an indication of core heat up. • Modelling guidelines of CET response with system codes. • Modelling of heat transfer processes in the core and UP regions. - Abstract: Core exit temperature (CET) measurements play an important role in the sequence of actions under accidental conditions in pressurized water reactors (PWR). Given the difficulties in placing measurements in the core region, CET readings are used as criterion for the initiation of accident management (AM) procedures because they can indicate a core heat up scenario. However, the CET responses have some limitation in detecting inadequate core cooling and core uncovery simply because the measurement is not placed inside the core. Therefore, it is of main importance in the field of nuclear safety for PWR power plants to assess the capabilities of system codes for simulating the relation between the CET and the peak cladding temperature (PCT). The work presented in this paper intends to address this open question by making use of experimental work at integral test facilities (ITF) where experiments related to the evolution of the CET and the PCT during transient conditions have been carried out. In particular, simulations of two experiments performed at the ROSA/LSTF and PKL facilities are presented. The two experiments are part of a counterpart exercise between the OECD/NEA ROSA-2 and OECD/NEA PKL-2 projects. The simulations are used to derive guidelines in how to correctly reproduce the CET response during a core heat up scenario. Three aspects have been identified to be of main importance: (1) the need for a 3-dimensional representation of the core and Upper Plenum (UP) regions in order to model the heterogeneity of the power zones and axial areas, (2) the detailed representation of the active and passive heat structures, and (3) the use of simulated thermocouples instead of steam temperatures to represent the CET readings.
CODE's new solar radiation pressure model for GNSS orbit determination
Arnold, D.; Meindl, M.; Beutler, G.; Dach, R.; Schaer, S.; Lutz, S.; Prange, L.; Sośnica, K.; Mervart, L.; Jäggi, A.
2015-08-01
The Empirical CODE Orbit Model (ECOM) of the Center for Orbit Determination in Europe (CODE), which was developed in the early 1990s, is widely used in the International GNSS Service (IGS) community. For a rather long time, spurious spectral lines are known to exist in geophysical parameters, in particular in the Earth Rotation Parameters (ERPs) and in the estimated geocenter coordinates, which could recently be attributed to the ECOM. These effects grew creepingly with the increasing influence of the GLONASS system in recent years in the CODE analysis, which is based on a rigorous combination of GPS and GLONASS since May 2003. In a first step we show that the problems associated with the ECOM are to the largest extent caused by the GLONASS, which was reaching full deployment by the end of 2011. GPS-only, GLONASS-only, and combined GPS/GLONASS solutions using the observations in the years 2009-2011 of a global network of 92 combined GPS/GLONASS receivers were analyzed for this purpose. In a second step we review direct solar radiation pressure (SRP) models for GNSS satellites. We demonstrate that only even-order short-period harmonic perturbations acting along the direction Sun-satellite occur for GPS and GLONASS satellites, and only odd-order perturbations acting along the direction perpendicular to both, the vector Sun-satellite and the spacecraft's solar panel axis. Based on this insight we assess in the third step the performance of four candidate orbit models for the future ECOM. The geocenter coordinates, the ERP differences w. r. t. the IERS 08 C04 series of ERPs, the misclosures for the midnight epochs of the daily orbital arcs, and scale parameters of Helmert transformations for station coordinates serve as quality criteria. The old and updated ECOM are validated in addition with satellite laser ranging (SLR) observations and by comparing the orbits to those of the IGS and other analysis centers. Based on all tests, we present a new extended ECOM which
A Realistic Model under which the Genetic Code is Optimal
Buhrman, H.; van der Gulik, P.T.S.; Klau, G.W.; Schaffner, C.; Speijer, D.; Stougie, L.
2013-01-01
The genetic code has a high level of error robustness. Using values of hydrophobicity scales as a proxy for amino acid character, and the mean square measure as a function quantifying error robustness, a value can be obtained for a genetic code which reflects the error robustness of that code. By
Evaluating a Model of Youth Physical Activity
Heitzler, Carrie D.; Lytle, Leslie A.; Erickson, Darin J.; Barr-Anderson, Daheia; Sirard, John R.; Story, Mary
2010-01-01
Objective: To explore the relationship between social influences, self-efficacy, enjoyment, and barriers and physical activity. Methods: Structural equation modeling examined relationships between parent and peer support, parent physical activity, individual perceptions, and objectively measured physical activity using accelerometers among a…
American Inst. of Architects, Washington, DC.
A MODEL BUILDING CODE FOR FALLOUT SHELTERS WAS DRAWN UP FOR INCLUSION IN FOUR NATIONAL MODEL BUILDING CODES. DISCUSSION IS GIVEN OF FALLOUT SHELTERS WITH RESPECT TO--(1) NUCLEAR RADIATION, (2) NATIONAL POLICIES, AND (3) COMMUNITY PLANNING. FALLOUT SHELTER REQUIREMENTS FOR SHIELDING, SPACE, VENTILATION, CONSTRUCTION, AND SERVICES SUCH AS ELECTRICAL…
Development of computer code models for analysis of subassembly voiding in the LMFBR
International Nuclear Information System (INIS)
Hinkle, W.
1979-12-01
The research program discussed in this report was started in FY1979 under the combined sponsorship of the US Department of Energy (DOE), General Electric (GE) and Hanford Engineering Development Laboratory (HEDL). The objective of the program is to develop multi-dimensional computer codes which can be used for the analysis of subassembly voiding incoherence under postulated accident conditions in the LMFBR. Two codes are being developed in parallel. The first will use a two fluid (6 equation) model which is more difficult to develop but has the potential for providing a code with the utmost in flexibility and physical consistency for use in the long term. The other will use a mixture (< 6 equation) model which is less general but may be more amenable to interpretation and use of experimental data and therefore, easier to develop for use in the near term. To assure that the models developed are not design dependent, geometries and transient conditions typical of both foreign and US designs are being considered
Modelling RF sources using 2-D PIC codes
Energy Technology Data Exchange (ETDEWEB)
Eppley, K.R.
1993-03-01
In recent years, many types of RF sources have been successfully modelled using 2-D PIC codes. Both cross field devices (magnetrons, cross field amplifiers, etc.) and pencil beam devices (klystrons, gyrotrons, TWT'S, lasertrons, etc.) have been simulated. All these devices involve the interaction of an electron beam with an RF circuit. For many applications, the RF structure may be approximated by an equivalent circuit, which appears in the simulation as a boundary condition on the electric field ( port approximation''). The drive term for the circuit is calculated from the energy transfer between beam and field in the drift space. For some applications it may be necessary to model the actual geometry of the structure, although this is more expensive. One problem not entirely solved is how to accurately model in 2-D the coupling to an external waveguide. Frequently this is approximated by a radial transmission line, but this sometimes yields incorrect results. We also discuss issues in modelling the cathode and injecting the beam into the PIC simulation.
Modelling RF sources using 2-D PIC codes
Energy Technology Data Exchange (ETDEWEB)
Eppley, K.R.
1993-03-01
In recent years, many types of RF sources have been successfully modelled using 2-D PIC codes. Both cross field devices (magnetrons, cross field amplifiers, etc.) and pencil beam devices (klystrons, gyrotrons, TWT`S, lasertrons, etc.) have been simulated. All these devices involve the interaction of an electron beam with an RF circuit. For many applications, the RF structure may be approximated by an equivalent circuit, which appears in the simulation as a boundary condition on the electric field (``port approximation``). The drive term for the circuit is calculated from the energy transfer between beam and field in the drift space. For some applications it may be necessary to model the actual geometry of the structure, although this is more expensive. One problem not entirely solved is how to accurately model in 2-D the coupling to an external waveguide. Frequently this is approximated by a radial transmission line, but this sometimes yields incorrect results. We also discuss issues in modelling the cathode and injecting the beam into the PIC simulation.
Modelling RF sources using 2-D PIC codes
International Nuclear Information System (INIS)
Eppley, K.R.
1993-03-01
In recent years, many types of RF sources have been successfully modelled using 2-D PIC codes. Both cross field devices (magnetrons, cross field amplifiers, etc.) and pencil beam devices (klystrons, gyrotrons, TWT'S, lasertrons, etc.) have been simulated. All these devices involve the interaction of an electron beam with an RF circuit. For many applications, the RF structure may be approximated by an equivalent circuit, which appears in the simulation as a boundary condition on the electric field (''port approximation''). The drive term for the circuit is calculated from the energy transfer between beam and field in the drift space. For some applications it may be necessary to model the actual geometry of the structure, although this is more expensive. One problem not entirely solved is how to accurately model in 2-D the coupling to an external waveguide. Frequently this is approximated by a radial transmission line, but this sometimes yields incorrect results. We also discuss issues in modelling the cathode and injecting the beam into the PIC simulation
The use of Monte Carlo radiation transport codes in radiation physics and dosimetry
CERN. Geneva; Ferrari, Alfredo; Silari, Marco
2006-01-01
Transport and interaction of electromagnetic radiation Interaction models and simulation schemes implemented in modern Monte Carlo codes for the simulation of coupled electron-photon transport will be briefly reviewed. In these codes, photon transport is simulated by using the detailed scheme, i.e., interaction by interaction. Detailed simulation is easy to implement, and the reliability of the results is only limited by the accuracy of the adopted cross sections. Simulations of electron and positron transport are more difficult, because these particles undergo a large number of interactions in the course of their slowing down. Different schemes for simulating electron transport will be discussed. Condensed algorithms, which rely on multiple-scattering theories, are comparatively fast, but less accurate than mixed algorithms, in which hard interactions (with energy loss or angular deflection larger than certain cut-off values) are simulated individually. The reliability, and limitations, of electron-interacti...
Maximizing entropy of image models for 2-D constrained coding
DEFF Research Database (Denmark)
Forchhammer, Søren; Danieli, Matteo; Burini, Nino
2010-01-01
This paper considers estimating and maximizing the entropy of two-dimensional (2-D) fields with application to 2-D constrained coding. We consider Markov random fields (MRF), which have a non-causal description, and the special case of Pickard random fields (PRF). The PRF are 2-D causal finite...... context models, which define stationary probability distributions on finite rectangles and thus allow for calculation of the entropy. We consider two binary constraints and revisit the hard square constraint given by forbidding neighboring 1s and provide novel results for the constraint that no uniform 2...... £ 2 squares contains all 0s or all 1s. The maximum values of the entropy for the constraints are estimated and binary PRF satisfying the constraint are characterized and optimized w.r.t. the entropy. The maximum binary PRF entropy is 0.839 bits/symbol for the no uniform squares constraint. The entropy...
Thermal modeling: at the crossroads of several subjects of physics
International Nuclear Information System (INIS)
1997-01-01
The modeling of thermal phenomena is of prime importance for the dimensioning of industrial facilities. However, the understanding of thermal processes requires to refer to other subjects of physics like electromagnetism, matter transformation, fluid mechanics, chemistry etc.. The aim of this workshop organized by the industrial electro-thermal engineering section of the French society of thermal engineers is to take stock of current or forthcoming advances in the coupling of thermal engineering codes with electromagnetic, fluid mechanics, chemical and mechanical engineering codes. The modeling of phenomena remains the essential link between the laboratory research of new processes and their industrial developments. From the 9 talks given during this workshop, 2 of them deal with thermal processes in nuclear reactors and fall into the INIS scope and the others concern the modeling of industrial heating or electrical processes and were selected for ETDE. (J.S.)
Dynamic modeling of physical phenomena for PRAs using neural networks
International Nuclear Information System (INIS)
Benjamin, A.S.; Brown, N.N.; Paez, T.L.
1998-04-01
In most probabilistic risk assessments, there is a set of accident scenarios that involves the physical responses of a system to environmental challenges. Examples include the effects of earthquakes and fires on the operability of a nuclear reactor safety system, the effects of fires and impacts on the safety integrity of a nuclear weapon, and the effects of human intrusions on the transport of radionuclides from an underground waste facility. The physical responses of the system to these challenges can be quite complex, and their evaluation may require the use of detailed computer codes that are very time consuming to execute. Yet, to perform meaningful probabilistic analyses, it is necessary to evaluate the responses for a large number of variations in the input parameters that describe the initial state of the system, the environments to which it is exposed, and the effects of human interaction. Because the uncertainties of the system response may be very large, it may also be necessary to perform these evaluations for various values of modeling parameters that have high uncertainties, such as material stiffnesses, surface emissivities, and ground permeabilities. The authors have been exploring the use of artificial neural networks (ANNs) as a means for estimating the physical responses of complex systems to phenomenological events such as those cited above. These networks are designed as mathematical constructs with adjustable parameters that can be trained so that the results obtained from the networks will simulate the results obtained from the detailed computer codes. The intent is for the networks to provide an adequate simulation of the detailed codes over a significant range of variables while requiring only a small fraction of the computer processing time required by the detailed codes. This enables the authors to integrate the physical response analyses into the probabilistic models in order to estimate the probabilities of various responses
Nuclear model codes available at the Nuclear Energy Agency Computer Program Library (NEA-CPL)
International Nuclear Information System (INIS)
Sartori, E.; Garcia Viedma, L. de
1976-01-01
This paper briefly outlines the objectives of the NEA-CPL and its activities in the field of Nuclear Model Computer Codes. A short description of the computer codes available from the CPL in this field is also presented. (author)
Large-Signal Code TESLA: Improvements in the Implementation and in the Model
National Research Council Canada - National Science Library
Chernyavskiy, Igor A; Vlasov, Alexander N; Anderson, Jr., Thomas M; Cooke, Simon J; Levush, Baruch; Nguyen, Khanh T
2006-01-01
We describe the latest improvements made in the large-signal code TESLA, which include transformation of the code to a Fortran-90/95 version with dynamical memory allocation and extension of the model...
Poli, E.; Bock, A.; Lochbrunner, M.; Maj, O.; Reich, M.; Snicker, A.; Stegmeir, A.; Volpe, F.; Bertelli, N.; Bilato, R.; Conway, G. D.; Farina, D.; Felici, F.; Figini, L.; Fischer, R.; Galperti, C.; Happel, T.; Lin-Liu, Y. R.; Marushchenko, N. B.; Mszanowski, U.; Poli, F. M.; Stober, J.; Westerhof, E.; Zille, R.; Peeters, A. G.; Pereverzev, G. V.
2018-04-01
The paraxial WKB code TORBEAM (Poli, 2001) is widely used for the description of electron-cyclotron waves in fusion plasmas, retaining diffraction effects through the solution of a set of ordinary differential equations. With respect to its original form, the code has undergone significant transformations and extensions, in terms of both the physical model and the spectrum of applications. The code has been rewritten in Fortran 90 and transformed into a library, which can be called from within different (not necessarily Fortran-based) workflows. The models for both absorption and current drive have been extended, including e.g. fully-relativistic calculation of the absorption coefficient, momentum conservation in electron-electron collisions and the contribution of more than one harmonic to current drive. The code can be run also for reflectometry applications, with relativistic corrections for the electron mass. Formulas that provide the coupling between the reflected beam and the receiver have been developed. Accelerated versions of the code are available, with the reduced physics goal of inferring the location of maximum absorption (including or not the total driven current) for a given setting of the launcher mirrors. Optionally, plasma volumes within given flux surfaces and corresponding values of minimum and maximum magnetic field can be provided externally to speed up the calculation of full driven-current profiles. These can be employed in real-time control algorithms or for fast data analysis.
MINIMARS interim report appendix halo model and computer code
International Nuclear Information System (INIS)
Santarius, J.F.; Barr, W.L.; Deng, B.Q.; Emmert, G.A.
1985-01-01
A tenuous, cool plasma called the halo shields the core plasma in a tandem mirror from neutral gas and impurities. The neutral particles are ionized and then pumped by the halo to the end tanks of the device, since flow of plasma along field lines is much faster than radial flow. Plasma reaching the end tank walls recombines, and the resulting neutral gas is vacuum pumped. The basic geometry of the MINIMARS halo is shown. For halo modeling purposes, the core plasma and cold gas regions may be treated as single radial zones leading to halo source and sink terms. The halo itself is differential into two major radial zones: halo scraper and halo dump. The halo scraper zone is defined by the radial distance required for the ion end plugging potential to drop to the central cell value, and thus have no effect on axial confinement; this distance is typically a sloshing plug ion Larmor diameter. The outer edge of the halo dump zone is defined by the last central cell flux tube to pass through the choke coil. This appendix will summarize the halo model that has been developed for MINIMARS and the methodology used in implementing that model as a computer code
A validated physical model of greenhouse climate.
Bot, G.P.A.
1989-01-01
In the greenhouse model the momentaneous environmental crop growth factors are calculated as output, together with the physical behaviour of the crop. The boundary conditions for this model are the outside weather conditions; other inputs are the physical characteristics of the crop, of the
International Nuclear Information System (INIS)
Kandiev, Y.Z.; Zatsepin, O.V.
2013-01-01
At RFNC-VNIITF, the PRIZMA code which has been developed for more than 30 years, is used to model radiation transport by the Monte Carlo method. The code implements individual and coupled tracking of neutrons, photons, electrons, positrons and ions in one dimensional (1D), 2D or 3D geometry. Attendance estimators are used for tallying, i.e., the estimators whose scores are only nonzero from particles which cross a region or surface of interest. Importance sampling is used to make deep penetration and detection calculations more effective. However, its application to reactor analysis appeared peculiar and required further development. The paper reviews methods used for deep penetration and detection calculations by PRIZMA. It describes in what these calculations differ when applied to reactor analysis and how we compute approximated importance functions and parameters for biased distributions. Methods to control the statistical weight of particles are also discussed. A number of test and applied calculations which were done for the purpose of verification are provided. They are shown to agree either with asymptotic solutions if exist, or with results of analog calculations or predictions by other codes. The applied calculations include the estimation of ex-core detector response from neutron sources arranged in the core, and the estimation of in-core detector response. (authors)
Modelling of Aquitaine II pipe whipping test with EUROPLEXUS fast dynamics code
International Nuclear Information System (INIS)
Potapov, S.
2003-01-01
To validate the modelling of multi-physics phenomena with EUROPLEXUS code we considered a pipe whipping problem occurring in thermal hydraulic conditions of a Loss of Coolant Accident in PWR primary circuit. Two numerical fluid-structure interaction (FSI) models, a simplified 'pipe-like' model and a mixed 1D/3D model, were used to simulate both the conditioning phase and a phase of whipping. The results of calculations were compared with existing experimental data. Analysis of numerical results shows that both models give a good prediction of global behaviour of the coupled fluid-structure system, namely for pipe displacements and stresses in the pipe walls, as well as for pressure and velocity in the fluid. By comparison with experimental data, we show that only the mixed EUROPLEXUS model, where the pipe elbow is discretized with shells, allows us to estimate correctly the time history and maximum value of the contact force between the pipe and the obstacle. The 1D model with reduced kinematics (rigid cross section hypothesis) does not allow the correct detection of contact phenomenon. This study shows that the use of mixed numerical models containing simplified and totally 3D parts duly interconnected allows a very efficient and CPU inexpensive numerical analysis which is able to take into account different global and local physical phenomena. (author)
PCCS model development for SBWR using the CONTAIN code
International Nuclear Information System (INIS)
Tills, J.; Murata, K.K.; Washington, K.E.
1994-01-01
The General Electric Simplified Boiling Water Reactor (SBWR) employs a passive containment cooling system (PCCS) to maintain long-term containment gas pressure and temperature below design limits during accidents. This system consists of a steam supply line that connects the upper portion of the drywell with a vertical shell-and-tube single pass heat exchanger located in an open water pool outside of the containment safety envelope. The heat exchanger tube outlet is connected to a vent line that is submerged below the suppression pool surface but above the main suppression pool horizontal vents. Steam generated in the post-shutdown period flows into the heat exchanger tubes as the result of suction and/or a low pressure differential between the drywell and suppression chamber. Operation of the PCCS is complicated by the presence of noncondensables in the flow stream. Build-up of noncondensables in the exchanger and vent line for the periods when the vent is not cleared causes a reduction in the exchanger heat removal capacity. As flow to the exchanger is reduced due to the noncondensable gas build-up, the drywell pressure increases until the vent line is cleared and the noncondensables are purged into the suppression chamber, restoring the heat removal capability of the PCCS. This paper reports on progress made in modeling SBWR containment loads using the CONTAIN code. As a central part of this effort, a PCCS model development effort has recently been undertaken to implement an appropriate model in CONTAIN. The CONTAIN PCCS modeling approach is discussed and validated. A full SBWR containment input deck has also been developed for CONTAIN. The plant response to a postulated design basis accident (DBA) has been calculated with the CONTAIN PCCS model and plant deck, and the preliminary results are discussed
Code-Hopping Based Transmission Scheme for Wireless Physical-Layer Security
Directory of Open Access Journals (Sweden)
Liuguo Yin
2018-01-01
Full Text Available Due to the broadcast and time-varying natures of wireless channels, traditional communication systems that provide data encryption at the application layer suffer many challenges such as error diffusion. In this paper, we propose a code-hopping based secrecy transmission scheme that uses dynamic nonsystematic low-density parity-check (LDPC codes and automatic repeat-request (ARQ mechanism to jointly encode and encrypt source messages at the physical layer. In this scheme, secret keys at the transmitter and the legitimate receiver are generated dynamically upon the source messages that have been transmitted successfully. During the transmission, each source message is jointly encoded and encrypted by a parity-check matrix, which is dynamically selected from a set of LDPC matrices based on the shared dynamic secret key. As for the eavesdropper (Eve, the uncorrectable decoding errors prevent her from generating the same secret key as the legitimate parties. Thus she cannot select the correct LDPC matrix to recover the source message. We demonstrate that our scheme can be compatible with traditional cryptosystems and enhance the security without sacrificing the error-correction performance. Numerical results show that the bit error rate (BER of Eve approaches 0.5 as the number of transmitted source messages increases and the security gap of the system is small.
Methods tuned on the physical problem. A way to improve numerical codes
International Nuclear Information System (INIS)
Ixaru, L.Gr.
2010-01-01
We consider the problem on how the numerical methods tuned on the physical problem can contribute to the enhancement of the performance of the codes. We illustrate this on two simple cases: solution of time independent one-dimensional Schroedinger equation, and the computation of integrals with oscillatory integrands. In both cases the tuned versions bring a massive gain in accuracy at negligible extra cost. We presented two simple problems where successive levels of tuning enhance significantly the accuracy at negligible extra cost. These problems should be seen as representing only some illustrations on how the codes can be improved but we must also mention that in many cases tuned versions still have to be developed. Just for a suggestion, quadrature formulae which involve the integrand and a number of successive derivatives of this exist, but no formula is available when some of these derivatives are missing, for example when we dispose of y and y'' but not of y'. A direct application will be on the case when the integrand involves the solution of the Schrodinger equation by the method of Numerov. (author)
Joint Power Allocation for Multicast Systems with Physical-Layer Network Coding
Directory of Open Access Journals (Sweden)
Zhu Wei-Ping
2010-01-01
Full Text Available This paper addresses the joint power allocation issue in physical-layer network coding (PLNC of multicast systems with two sources and two destinations communicating via a large number of distributed relays. By maximizing the achievable system rate, a constrained optimization problem is first formulated to jointly allocate powers for the source and relay terminals. Due to the nonconvex nature of the cost function, an iterative algorithm with guaranteed convergence is developed to solve the joint power allocation problem. As an alternative, an upper bound of the achievable rate is also derived to modify the original cost function in order to obtain a convex optimization solution. This approximation is shown to be asymptotically optimal in the sense of maximizing the achievable rate. It is confirmed through Monte Carlo simulations that the proposed joint power allocation schemes are superior to the existing schemes in terms of achievable rate and cumulative distribution function (CDF.
Hotspot health physics codes used as a tool for managing excess risk on radiological emergencies
International Nuclear Information System (INIS)
Andrade, Edson Ramos de; Alves, Nelson Mendes; Rocha, Joao B.T.; Cruz, Ivana B. Manica da; Santos, Greice F. Fey dos; Machado, Michel Mansur; Rossato, Veronica Venturini; Bauermann, Liliane Freitas
2008-01-01
This work is aimed to use the Hotspot Health Physics codes in acute mode in order to estimate the immediate radiological impact associated with high acute radiation doses, which is applied to special target organs such as lung, small intestine wall, and red bone marrow. Organic compounds such as Diphenyl Diselenide (C 6 H 5 Se 2 C 6 H 5 ) and Ebselen (C 13 H 9 NOSe), an antioxidants selenium containing compounds, were used over irradiated phospholipids extracted from chicken yolk eggs, in vitro in order to reduce lipo-peroxidation. Experimental data were measured by Thiobarbituric Acid Reactive Substance (TBARS) assay which is able to measure the production of oxidative stress in the sample. Experimental data were extrapolated and applied as a reduction factors over equations for cancer excess risk calculation from BEIR V, for helping the decisonmaking process on Radiological Emergency Scenarios. (author)
Numerical modelling in material physics
International Nuclear Information System (INIS)
Proville, L.
2004-12-01
The author first briefly presents his past research activities: investigation of a dislocation sliding in solid solution by molecular dynamics, modelling of metal film growth by phase field and Monte Carlo kinetics, phase field model for surface self-organisation, phase field model for the Al 3 Zr alloy, calculation of anharmonic photons, mobility of bipolarons in superconductors. Then, he more precisely reports the mesoscopic modelling in phase field, and some atomistic modelling (dislocation sliding, Monte Carlo simulation of metal surface growth, anharmonic network optical spectrum modelling)
Energy Technology Data Exchange (ETDEWEB)
Liao, J.; Cao, L.; Ohkawa, K.; Frepoli, C. [LOCA Integrated Services I, Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)
2012-07-01
The non-condensable gases condensation suppression model is important for a realistic LOCA safety analysis code. A condensation suppression model for direct contact condensation was previously developed by Westinghouse using first principles. The model is believed to be an accurate description of the direct contact condensation process in the presence of non-condensable gases. The Westinghouse condensation suppression model is further revised by applying a more physical model. The revised condensation suppression model is thus implemented into the WCOBRA/TRAC-TF2 LOCA safety evaluation code for both 3-D module (COBRA-TF) and 1-D module (TRAC-PF1). Parametric study using the revised Westinghouse condensation suppression model is conducted. Additionally, the performance of non-condensable gases condensation suppression model is examined in the ACHILLES (ISP-25) separate effects test and LOFT L2-5 (ISP-13) integral effects test. (authors)
UCODE, a computer code for universal inverse modeling
Poeter, E.P.; Hill, M.C.
1999-01-01
This article presents the US Geological Survey computer program UCODE, which was developed in collaboration with the US Army Corps of Engineers Waterways Experiment Station and the International Ground Water Modeling Center of the Colorado School of Mines. UCODE performs inverse modeling, posed as a parameter-estimation problem, using nonlinear regression. Any application model or set of models can be used; the only requirement is that they have numerical (ASCII or text only) input and output files and that the numbers in these files have sufficient significant digits. Application models can include preprocessors and postprocessors as well as models related to the processes of interest (physical, chemical and so on), making UCODE extremely powerful for model calibration. Estimated parameters can be defined flexibly with user-specified functions. Observations to be matched in the regression can be any quantity for which a simulated equivalent value can be produced, thus simulated equivalent values are calculated using values that appear in the application model output files and can be manipulated with additive and multiplicative functions, if necessary. Prior, or direct, information on estimated parameters also can be included in the regression. The nonlinear regression problem is solved by minimizing a weighted least-squares objective function with respect to the parameter values using a modified Gauss-Newton method. Sensitivities needed for the method are calculated approximately by forward or central differences and problems and solutions related to this approximation are discussed. Statistics are calculated and printed for use in (1) diagnosing inadequate data or identifying parameters that probably cannot be estimated with the available data, (2) evaluating estimated parameter values, (3) evaluating the model representation of the actual processes and (4) quantifying the uncertainty of model simulated values. UCODE is intended for use on any computer operating
Modeling ion exchange in clinoptilolite using the EQ3/6 geochemical modeling code
International Nuclear Information System (INIS)
Viani, B.E.; Bruton, C.J.
1992-06-01
Assessing the suitability of Yucca Mtn., NV as a potential repository for high-level nuclear waste requires the means to simulate ion-exchange behavior of zeolites. Vanselow and Gapon convention cation-exchange models have been added to geochemical modeling codes EQ3NR/EQ6, allowing exchange to be modeled for up to three exchangers or a single exchanger with three independent sites. Solid-solution models that are numerically equivalent to the ion-exchange models were derived and also implemented in the code. The Gapon model is inconsistent with experimental adsorption isotherms of trace components in clinoptilolite. A one-site Vanselow model can describe adsorption of Cs or Sr on clinoptilolite, but a two-site Vanselow exchange model is necessary to describe K contents of natural clinoptilolites
Methodology Using MELCOR Code to Model Proposed Hazard Scenario
Energy Technology Data Exchange (ETDEWEB)
Gavin Hawkley
2010-07-01
This study demonstrates a methodology for using the MELCOR code to model a proposed hazard scenario within a building containing radioactive powder, and the subsequent evaluation of a leak path factor (LPF) (or the amount of respirable material which that escapes a facility into the outside environment), implicit in the scenario. This LPF evaluation will analyzes the basis and applicability of an assumed standard multiplication of 0.5 × 0.5 (in which 0.5 represents the amount of material assumed to leave one area and enter another), for calculating an LPF value. The outside release is dependsent upon the ventilation/filtration system, both filtered and un-filtered, and from other pathways from the building, such as doorways (, both open and closed). This study is presents ed to show how the multiple leak path factorsLPFs from the interior building can be evaluated in a combinatory process in which a total leak path factorLPF is calculated, thus addressing the assumed multiplication, and allowing for the designation and assessment of a respirable source term (ST) for later consequence analysis, in which: the propagation of material released into the environmental atmosphere can be modeled and the dose received by a receptor placed downwind can be estimated and the distance adjusted to maintains such exposures as low as reasonably achievableALARA.. Also, this study will briefly addresses particle characteristics thatwhich affect atmospheric particle dispersion, and compares this dispersion with leak path factorLPF methodology.
Lost opportunities: Modeling commercial building energy code adoption in the United States
International Nuclear Information System (INIS)
Nelson, Hal T.
2012-01-01
This paper models the adoption of commercial building energy codes in the US between 1977 and 2006. Energy code adoption typically results in an increase in aggregate social welfare by cost effectively reducing energy expenditures. Using a Cox proportional hazards model, I test if relative state funding, a new, objective, multivariate regression-derived measure of government capacity, as well as a vector of control variables commonly used in comparative state research, predict commercial building energy code adoption. The research shows little political influence over historical commercial building energy code adoption in the sample. Colder climates and higher electricity prices also do not predict more frequent code adoptions. I do find evidence of high government capacity states being 60 percent more likely than low capacity states to adopt commercial building energy codes in the following year. Wealthier states are also more likely to adopt commercial codes. Policy recommendations to increase building code adoption include increasing access to low cost capital for the private sector and providing noncompetitive block grants to the states from the federal government. - Highlights: ► Model the adoption of commercial building energy codes from 1977–2006 in the US. ► Little political influence over historical building energy code adoption. ► High capacity states are over 60 percent more likely than low capacity states to adopt codes. ► Wealthier states are more likely to adopt commercial codes. ► Access to capital and technical assistance is critical to increase code adoption.
Problems in physical modeling of magnetic materials
International Nuclear Information System (INIS)
Della Torre, E.
2004-01-01
Physical modeling of magnetic materials should give insights into the basic processes involved and should be able to extrapolate results to new situations that the models were not necessarily intended to solve. Thus, for example, if a model is designed to describe a static magnetization curve, it should also be able to describe aspects of magnetization dynamics. Both micromagnetic modeling and Preisach modeling, the two most popular magnetic models, fulfill this requirement, but in the process of fulfilling this requirement, they both had to be modified in some ways. Hence, we should view physical modeling as an iterative process whereby we start with some simple assumptions and refine them as reality requires. In the process of refining these assumptions, we should try to appeal to physical arguments for the modifications, if we are to come up with good models. If we consider phenomenological models, on the other hand, that is as axiomatic models requiring no physical justification, we can follow them logically to see the end and examine the consequences of their assumptions. In this way, we can learn the properties, limitations and achievements of the particular model. Physical and phenomenological models complement each other in furthering our understanding of the behavior of magnetic materials
Coding conventions and principles for a National Land-Change Modeling Framework
Donato, David I.
2017-07-14
This report establishes specific rules for writing computer source code for use with the National Land-Change Modeling Framework (NLCMF). These specific rules consist of conventions and principles for writing code primarily in the C and C++ programming languages. Collectively, these coding conventions and coding principles create an NLCMF programming style. In addition to detailed naming conventions, this report provides general coding conventions and principles intended to facilitate the development of high-performance software implemented with code that is extensible, flexible, and interoperable. Conventions for developing modular code are explained in general terms and also enabled and demonstrated through the appended templates for C++ base source-code and header files. The NLCMF limited-extern approach to module structure, code inclusion, and cross-module access to data is both explained in the text and then illustrated through the module templates. Advice on the use of global variables is provided.
High precision Standard Model Physics
International Nuclear Information System (INIS)
Magnin, J.
2009-01-01
The main goal of the LHCb experiment, one of the four large experiments of the Large Hadron Collider, is to try to give answers to the question of why Nature prefers matter over antimatter? This will be done by studying the decay of b quarks and their antimatter partners, b-bar, which will be produced by billions in 14 TeV p-p collisions by the LHC. In addition, as 'beauty' particles mainly decay in charm particles, an interesting program of charm physics will be carried on, allowing to measure quantities as for instance the D 0 -D-bar 0 mixing, with incredible precision.
A simplified treatment of the boundary conditions of the k- ε model in coarse-mesh CFD-type codes
International Nuclear Information System (INIS)
Analytis, G.Th.; Andreani, M.
1999-01-01
In coarse-mesh, CFD-type codes such as the containment analysis code GOTHIC, one of the options that can be used for modelling of turbulence is the k - ε model. However, in contrast to most other CFD codes which are designed to perform detailed CFD calculations with a large number of spatial meshes, codes such as GOTHIC are primarily aimed at simplified calculation of transients in large spaces (e.g., reactor containments), and generally use coarse meshes. The solution of the two parabolic equations for the k - ε model requires the definition of boundary conditions at physical boundaries and this, in turn, requires very small spatial meshes near these boundaries. Hence, while in codes like CFX this is done in a rigorous and consistent manner, codes like GOTHIC adopt an indirect and heuristic approach, due to the fact that the spatial meshes are usually large. This can have adverse consequences during the calculation of a transient and in this work, we shall give some examples of this and outline a method by which this problem can be avoided. (author)
Physics Based Modeling of Compressible Turbulance
2016-11-07
AFRL-AFOSR-VA-TR-2016-0345 PHYSICS -BASED MODELING OF COMPRESSIBLE TURBULENCE PARVIZ MOIN LELAND STANFORD JUNIOR UNIV CA Final Report 09/13/2016...on the AFOSR project (FA9550-11-1-0111) entitled: Physics based modeling of compressible turbulence. The period of performance was, June 15, 2011...by ANSI Std. Z39.18 Page 1 of 2FORM SF 298 11/10/2016https://livelink.ebs.afrl.af.mil/livelink/llisapi.dll PHYSICS -BASED MODELING OF COMPRESSIBLE
International Nuclear Information System (INIS)
Calloo, A.A.
2012-01-01
In reactor physics, calculation schemes with deterministic codes are validated with respect to a reference Monte Carlo code. The remaining biases are attributed to the approximations and models induced by the multigroup theory (self-shielding models and expansion of the scattering law using Legendre polynomials) to represent physical phenomena (resonant absorption and scattering anisotropy respectively). This work focuses on the relevance of a polynomial expansion to model the scattering law. Since the outset of reactor physics, the latter has been expanded on a truncated Legendre polynomial basis. However, the transfer cross sections are highly anisotropic, with non-zero values for a very small range of the cosine of the scattering angle. Besides, the finer the energy mesh and the lighter the scattering nucleus, the more exacerbated is the peaked shape of this cross section. As such, the Legendre expansion is less suited to represent the scattering law. Furthermore, this model induces negative values which are non-physical. In this work, various scattering laws are briefly described and the limitations of the existing model are pointed out. Hence, piecewise-constant functions have been used to represent the multigroup scattering cross section. This representation requires a different model for the diffusion source. The discrete ordinates method which is widely employed to solve the transport equation has been adapted. Thus, the finite volume method for angular discretization has been developed and implemented in Paris environment which hosts the S n solver, Snatch. The angular finite volume method has been compared to the collocation method with Legendre moments to ensure its proper performance. Moreover, unlike the latter, this method is adapted for both the Legendre moments and the piecewise-constant functions representations of the scattering cross section. This hybrid-source method has been validated for different cases: fuel cell in infinite lattice
Energy Technology Data Exchange (ETDEWEB)
Jaeger, Wadim; Manes, Jorge Perez; Imke, Uwe; Escalante, Javier Jimenez; Espinoza, Victor Sanchez, E-mail: victor.sanchez@kit.edu
2013-10-15
Highlights: • Simulation of BFBT turbine and pump transients at multiple scales. • CFD, sub-channel and system codes are used for the comparative study. • Heat transfer models are compared to identify difference between the code predictions. • All three scales predict results in good agreement to experiment. • Sub cooled boiling models are identified as field for future research. -- Abstract: The Institute for Neutron Physics and Reactor Technology (INR) at the Karlsruhe Institute of Technology (KIT) is involved in the validation and qualification of modern thermo hydraulic simulations tools at various scales. In the present paper, the prediction capabilities of four codes from three different scales – NEPTUNE{sub C}FD as fine mesh computational fluid dynamics code, SUBCHANFLOW and COBRA-TF as sub channels codes and TRACE as system code – are assessed with respect to their two-phase flow modeling capabilities. The subject of the investigations is the well-known and widely used data base provided within the NUPEC BFBT benchmark related to BWRs. Void fraction measurements simulating a turbine and a re-circulation pump trip are provided at several axial levels of the bundle. The prediction capabilities of the codes for transient conditions with various combinations of boundary conditions are validated by comparing the code predictions with the experimental data. In addition, the physical models of the different codes are described and compared to each other in order to explain the different results and to identify areas for further improvements.
International Nuclear Information System (INIS)
Guillermier, Pierre; Daniel, Lucile; Gauthier, Laurent
2009-01-01
generally in the fuel element (pebble or compact) aims at estimating the source term of the fission products release outside the fuel element in normal operation or accidental conditions. In ATLAS, the transport mechanisms are modelled in a single transport law using effective diffusion coefficients for the fission product species in the different constitutive materials. The Verification and Validation of ATLAS code rests on two main steps: - Testing plan on fuel particle thermal mechanical behaviour has been carried out regarding sensitivity on dimensional parameters and physical properties such as kernel diameter, density and layer thicknesses and pyrocarbon layer anisotropy. The obtained results allow justifying and specifying the design for the manufacture. - Regarding fission product release under core heat-up accident conditions, the IAEA Coordinated Research Project 6 on 'Advanced in HTGR Fuel Technology Development' benchmark is the basis of the ATLAS code verification step. The ATLAS results obtained on IAEA benchmark cases with analytical solutions demonstrate that the models used fit the physical, chemical and mathematical laws. Regarding past irradiation tests and heating tests, ATLAS results show good agreement with the experimental database measurements. Comparison between ATLAS code results with analytical and experimental data allows defining confidence zones where ATLAS code gives accurate results and critical limits. These limits show where R and D efforts on models and material properties are needed to refine laws and models. (author)
Electron and ion cyclotron heating calculations in the tandem-mirror modeling code MERTH
International Nuclear Information System (INIS)
Smith, G.R.
1985-01-01
To better understand and predict tandem-mirror experiments, we are building a comprehensive Mirror Equilibrium Radial Transport and Heating (MERTH) code. In this paper we first describe our method for developing the code. Then we report our plans for the installation of physics packages for electron- and ion-cyclotron heating of the plasma
International Nuclear Information System (INIS)
Kalin, J.; Petkovsek, B.; Montarnal, Ph.; Genty, A.; Deville, E.; Krivic, J.; Ratej, J.
2011-01-01
In the past years the Slovenian Performance Analysis/Safety Assessment team has performed many generic studies for the future Slovenian low and intermediate level waste repository, most recently a Special Safety Analysis for the Krsko site. The modelling approach was to split the problem into three parts: near-field (detailed model of the repository), far-field (i.e., geosphere) and biosphere. In the Special Safety Analysis the code used to perform the near-field calculations was Hydrus2D. Recently the team has begun a cooperation with the French Commisariat al'Energie Atomique/Saclay (CEA/Saclay) and, as a part of this cooperation, began investigations into using the Alliances numerical platform for near-field calculations in order to compare the overall approach and calculated results. The article presents the comparison between these two codes for a silo-type repository that was considered in the Special Safety Analysis. The physical layout and characteristics of the repository are presented and a hydraulic and transport model of the repository is developed and implemented in Alliances. Some analysis of sensitivity to mesh fineness and to simulation timestep has been preformed and is also presented. The compared quantity is the output flux of radionuclides on the boundary of the model. Finally the results from Hydrus2D and Alliances are compared and the differences and similarities are commented.
Energy Technology Data Exchange (ETDEWEB)
Kalin, J., E-mail: jan.kalin@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Petkovsek, B., E-mail: borut.petkovsek@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Montarnal, Ph., E-mail: philippe.montarnal@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Genty, A., E-mail: alain.genty@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Deville, E., E-mail: estelle.deville@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Krivic, J., E-mail: jure.krivic@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia); Ratej, J., E-mail: joze.ratej@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia)
2011-04-15
In the past years the Slovenian Performance Analysis/Safety Assessment team has performed many generic studies for the future Slovenian low and intermediate level waste repository, most recently a Special Safety Analysis for the Krsko site. The modelling approach was to split the problem into three parts: near-field (detailed model of the repository), far-field (i.e., geosphere) and biosphere. In the Special Safety Analysis the code used to perform the near-field calculations was Hydrus2D. Recently the team has begun a cooperation with the French Commisariat al'Energie Atomique/Saclay (CEA/Saclay) and, as a part of this cooperation, began investigations into using the Alliances numerical platform for near-field calculations in order to compare the overall approach and calculated results. The article presents the comparison between these two codes for a silo-type repository that was considered in the Special Safety Analysis. The physical layout and characteristics of the repository are presented and a hydraulic and transport model of the repository is developed and implemented in Alliances. Some analysis of sensitivity to mesh fineness and to simulation timestep has been preformed and is also presented. The compared quantity is the output flux of radionuclides on the boundary of the model. Finally the results from Hydrus2D and Alliances are compared and the differences and similarities are commented.
Three-field modeling for MARS 1-D code
International Nuclear Information System (INIS)
Hwang, Moonkyu; Lim, Ho-Gon; Jeong, Jae-Jun; Chung, Bub-Dong
2006-01-01
In this study, the three-field modeling of the two-phase mixture is developed. The finite difference equations for the three-field equations thereafter are devised. The solution scheme has been implemented into the MARS 1-D code. The three-field formulations adopted are similar to those for MARS 3-D module, in a sense that the mass and momentum are treated separately for the entrained liquid and continuous liquid. As in the MARS-3D module, the entrained liquid and continuous liquid are combined into one for the energy equation, assuming thermal equilibrium between the two. All the non-linear terms are linearized to arrange the finite difference equation set into a linear matrix form with respect to the unknown arguments. The problems chosen for the assessment of the newly added entrained field consist of basic conceptual tests. Among the tests are gas-only test, liquid-only test, gas-only with supplied entrained liquid test, Edwards pipe problem, and GE level swell problem. The conceptual tests performed confirm the sound integrity of the three-field solver
Cross-band noise model refinement for transform domain Wyner–Ziv video coding
DEFF Research Database (Denmark)
Huang, Xin; Forchhammer, Søren
2012-01-01
TDWZ video coding trails that of conventional video coding solutions, mainly due to the quality of side information, inaccurate noise modeling and loss in the final coding step. The major goal of this paper is to enhance the accuracy of the noise modeling, which is one of the most important aspects...... influencing the coding performance of DVC. A TDWZ video decoder with a novel cross-band based adaptive noise model is proposed, and a noise residue refinement scheme is introduced to successively update the estimated noise residue for noise modeling after each bit-plane. Experimental results show...... that the proposed noise model and noise residue refinement scheme can improve the rate-distortion (RD) performance of TDWZ video coding significantly. The quality of the side information modeling is also evaluated by a measure of the ideal code length....
The Physical Internet and Business Model Innovation
Directory of Open Access Journals (Sweden)
Diane Poulin
2012-06-01
Full Text Available Building on the analogy of data packets within the Digital Internet, the Physical Internet is a concept that dramatically transforms how physical objects are designed, manufactured, and distributed. This approach is open, efficient, and sustainable beyond traditional proprietary logistical solutions, which are often plagued by inefficiencies. The Physical Internet redefines supply chain configurations, business models, and value-creation patterns. Firms are bound to be less dependent on operational scale and scope trade-offs because they will be in a position to offer novel hybrid products and services that would otherwise destroy value. Finally, logistical chains become flexible and reconfigurable in real time, thus becoming better in tune with firm strategic choices. This article focuses on the potential impact of the Physical Internet on business model innovation, both from the perspectives of Physical-Internet enabled and enabling business models.
Energy Technology Data Exchange (ETDEWEB)
Hartmann, Tobias
2013-07-03
The successful development and operation of a demonstration power plant (DEMO) is the next important step on roadmaps for fusion energy after ITER that is currently constructed in France. In the first phase of the development process for such devices, the conceptual design phase, the primary aim is to identify coherent designs that are composed of self-consistent sets of values for all key parameters like machine size, plasma current or magnetic field strength. This multidimensional parameter space can be explored with systems codes in order to identify areas that seem to be suited for more detailed investigation. Systems codes are composed of simplified models for all crucial systems of fusion devices that take into account all requirements and constraints of each component. This thesis is about the development of a new systems code called TREND (Tokamak Reactor code for the Evaluation of Next-step Devices). TREND is implemented with modular code architecture and consists of modules for geometry, core plasma physics, divertor, power flow, technology and costing. The main focus has been on the core physics module, since the development of TREND was done in parallel to work on physics design guidelines for DEMO. Moreover, the validation of TREND in terms of benchmarks with other European and Japanese systems codes is discussed. For these benchmarks, specific parameter sets were selected and the observed deviations were traced back to differences concerning the individual modellings. One of these parameter sets constitutes also the basis for parameter studies that were conducted with TREND. The general idea behind these studies is the analysis of implications that arise from specific assumptions on selected key parameters. Besides constant fusion power and constant additional heating power, the plasma density is fixed with respect to the Greenwald limit. The benchmarks helped particularly to detect shortages in the modellings of all involved systems codes
International Nuclear Information System (INIS)
Hartmann, Tobias
2013-01-01
The successful development and operation of a demonstration power plant (DEMO) is the next important step on roadmaps for fusion energy after ITER that is currently constructed in France. In the first phase of the development process for such devices, the conceptual design phase, the primary aim is to identify coherent designs that are composed of self-consistent sets of values for all key parameters like machine size, plasma current or magnetic field strength. This multidimensional parameter space can be explored with systems codes in order to identify areas that seem to be suited for more detailed investigation. Systems codes are composed of simplified models for all crucial systems of fusion devices that take into account all requirements and constraints of each component. This thesis is about the development of a new systems code called TREND (Tokamak Reactor code for the Evaluation of Next-step Devices). TREND is implemented with modular code architecture and consists of modules for geometry, core plasma physics, divertor, power flow, technology and costing. The main focus has been on the core physics module, since the development of TREND was done in parallel to work on physics design guidelines for DEMO. Moreover, the validation of TREND in terms of benchmarks with other European and Japanese systems codes is discussed. For these benchmarks, specific parameter sets were selected and the observed deviations were traced back to differences concerning the individual modellings. One of these parameter sets constitutes also the basis for parameter studies that were conducted with TREND. The general idea behind these studies is the analysis of implications that arise from specific assumptions on selected key parameters. Besides constant fusion power and constant additional heating power, the plasma density is fixed with respect to the Greenwald limit. The benchmarks helped particularly to detect shortages in the modellings of all involved systems codes
International Nuclear Information System (INIS)
Lemehov, Sergei; Suzuki, Motoe
2000-01-01
This report summarizes the review on the modeling features of FEMAXI code and proposal of a new theoretical equation model of clad creep on the basis of irradiation-induced microstructure change. It was pointed out that plutonium build-up in fuel matrix and non-uniform radial power profile at high burn-up affect significantly fuel behavior through the interconnected effects with such phenomena as clad irradiation-induced creep, fission gas release, fuel thermal conductivity degradation, rim porous band formation and associated fuel swelling. Therefore, these combined effects should be properly incorporated into the models of the FEMAXI code so that the code can carry out numerical analysis at the level of accuracy and elaboration that modern experimental data obtained in test reactors have. Also, the proposed new mechanistic clad creep model has a general formalism which allows the model to be flexibly applied for clad behavior analysis under normal operation conditions and power transients as well for Zr-based clad materials by the use of established out-of-pile mechanical properties. The model has been tested against experimental data, while further verification is needed with specific emphasis on power ramps and transients. (author)
International Nuclear Information System (INIS)
Paul, D.; Makovicka, L.; Ricard, M.
2005-01-01
Synthesis of the scientific French speaking days on numerical codes in radiation protection, in radio-physics and in dosimetry. The paper carries the title of 'French speaking' scientific days co-organized on October 2-3, 2003 in Sochaux by the SFRP, SFPM and FIRAM societies. It has for objective to establish the scientific balance sheet of this international event, to give the synthesis of current tendencies in the field of the development and of the use of the numerical codes in radiation protection, in radio-physics and in dosimetry. (author)
Are Physical Education Majors Models for Fitness?
Kamla, James; Snyder, Ben; Tanner, Lori; Wash, Pamela
2012-01-01
The National Association of Sport and Physical Education (NASPE) (2002) has taken a firm stance on the importance of adequate fitness levels of physical education teachers stating that they have the responsibility to model an active lifestyle and to promote fitness behaviors. Since the NASPE declaration, national initiatives like Let's Move…
Laser-Plasma Modeling Using PERSEUS Extended-MHD Simulation Code for HED Plasmas
Hamlin, Nathaniel; Seyler, Charles
2017-10-01
We discuss the use of the PERSEUS extended-MHD simulation code for high-energy-density (HED) plasmas in modeling the influence of Hall and electron inertial physics on laser-plasma interactions. By formulating the extended-MHD equations as a relaxation system in which the current is semi-implicitly time-advanced using the Generalized Ohm's Law, PERSEUS enables modeling of extended-MHD phenomena (Hall and electron inertial physics) without the need to resolve the smallest electron time scales, which would otherwise be computationally prohibitive in HED plasma simulations. We first consider a laser-produced plasma plume pinched by an applied magnetic field parallel to the laser axis in axisymmetric cylindrical geometry, forming a conical shock structure and a jet above the flow convergence. The Hall term produces low-density outer plasma, a helical field structure, flow rotation, and field-aligned current, rendering the shock structure dispersive. We then model a laser-foil interaction by explicitly driving the oscillating laser fields, and examine the essential physics governing the interaction. This work is supported by the National Nuclear Security Administration stewardship sciences academic program under Department of Energy cooperative agreements DE-FOA-0001153 and DE-NA0001836.
Quark models in hadron physics
International Nuclear Information System (INIS)
Phatak, Shashikant C.
2007-01-01
In this talk, we review the role played by the quark models in the study of interaction of strong, weak and electromagnetic probes with hadrons at intermediate and high momentum transfers. By hadrons, we mean individual nucleons as well as nuclei. We argue that at these momentum transfers, the structure of hadrons plays an important role. The hadron structure of the hadrons is because of the underlying quark structure of hadrons and therefore the quark models play an important role in determining the hadron structure. Further, the properties of hadrons are likely to change when these are placed in nuclear medium and this change should arise from the underlying quark structure. We shall consider some quark models to look into these aspects. (author)
Searching for Physics Beyond the Standard Model
Energy Technology Data Exchange (ETDEWEB)
Catterall, Simon [Syracuse Univ., NY (United States)
2016-12-01
This final report summarizes the work carried out by the Syracuse component of a multi-institutional SciDAC grant led by USQCD. This grant supported software development for theoretical high energy physics. The Syracuse component specifically targeted the development of code for the numerical simulation of N=4 super Yang-Mills theory. The work described in the final report includes this and a summary of results achieve in exploring the structure of this theory. It also describes the personnel - students and a postdoc who were directly or indirectly involved in this project. A list of publication is also described.
International Nuclear Information System (INIS)
Manahan, M.P.
1983-01-01
An improved Zircaloy-steam oxidation reaction model has been incorporated into the MARCH 2 code which includes: (1) improved physical modeling for solid-state process oxidation, (2) improved geometric modeling for gaseous diffusion oxidation, (3) chemisorption/dissociation retardation due to high hydrogen partial pressures, and (4) laminar and turbulent flow conditions. Several accident sequences have been analyzed using the model, and for the sequences considered, the results indicate that the integrated and averaged variables are not significantly altered for the current level of fuel modeling, however, the localized variables such as nodal temperature and oxide thickness are affected
Sodium/water pool-deposit bed model of the CONACS code
International Nuclear Information System (INIS)
Peak, R.D.
1983-01-01
A new Pool-Bed model of the CONACS (Containment Analysis Code System) code represents a major advance over the pool models of other containment analysis code (NABE code of France, CEDAN code of Japan and CACECO and CONTAIN codes of the United States). This new model advances pool-bed modeling because of the number of significant materials and processes which are included with appropriate rigor. This CONACS pool-bed model maintains material balances for eight chemical species (C, H 2 O, Na, NaH, Na 2 O, Na 2 O 2 , Na 2 CO 3 and NaOH) that collect in the stationary liquid pool on the floor and in the desposit bed on the elevated shelf of the standard CONACS analysis cell
Young, Robert D.
1973-01-01
Discusses the charge independence, wavefunctions, magnetic moments, and high-energy scattering of hadrons on the basis of group theory and nonrelativistic quark model with mass spectrum calculated by first-order perturbation theory. The presentation is explainable to advanced undergraduate students. (CC)
A Perceptual Model for Sinusoidal Audio Coding Based on Spectral Integration
Van de Par, S.; Kohlrausch, A.; Heusdens, R.; Jensen, J.; Holdt Jensen, S.
2005-01-01
Psychoacoustical models have been used extensively within audio coding applications over the past decades. Recently, parametric coding techniques have been applied to general audio and this has created the need for a psychoacoustical model that is specifically suited for sinusoidal modelling of
A perceptual model for sinusoidal audio coding based on spectral integration
Van de Par, S.; Kohlrauch, A.; Heusdens, R.; Jensen, J.; Jensen, S.H.
2005-01-01
Psychoacoustical models have been used extensively within audio coding applications over the past decades. Recently, parametric coding techniques have been applied to general audio and this has created the need for a psychoacoustical model that is specifically suited for sinusoidal modelling of
Simplified Models for LHC New Physics Searches
Alves, Daniele; Arora, Sanjay; Bai, Yang; Baumgart, Matthew; Berger, Joshua; Buckley, Matthew; Butler, Bart; Chang, Spencer; Cheng, Hsin-Chia; Cheung, Clifford; Chivukula, R.Sekhar; Cho, Won Sang; Cotta, Randy; D'Alfonso, Mariarosaria; El Hedri, Sonia; Essig, Rouven; Evans, Jared A.; Fitzpatrick, Liam; Fox, Patrick; Franceschini, Roberto; Freitas, Ayres; Gainer, James S.; Gershtein, Yuri; Gray, Richard; Gregoire, Thomas; Gripaios, Ben; Gunion, Jack; Han, Tao; Haas, Andy; Hansson, Per; Hewett, JoAnne; Hits, Dmitry; Hubisz, Jay; Izaguirre, Eder; Kaplan, Jared; Katz, Emanuel; Kilic, Can; Kim, Hyung-Do; Kitano, Ryuichiro; Koay, Sue Ann; Ko, Pyungwon; Krohn, David; Kuflik, Eric; Lewis, Ian; Lisanti, Mariangela; Liu, Tao; Liu, Zhen; Lu, Ran; Luty, Markus; Meade, Patrick; Morrissey, David; Mrenna, Stephen; Nojiri, Mihoko; Okui, Takemichi; Padhi, Sanjay; Papucci, Michele; Park, Michael; Park, Myeonghun; Perelstein, Maxim; Peskin, Michael; Phalen, Daniel; Rehermann, Keith; Rentala, Vikram; Roy, Tuhin; Ruderman, Joshua T.; Sanz, Veronica; Schmaltz, Martin; Schnetzer, Stephen; Schuster, Philip; Schwaller, Pedro; Schwartz, Matthew D.; Schwartzman, Ariel; Shao, Jing; Shelton, Jessie; Shih, David; Shu, Jing; Silverstein, Daniel; Simmons, Elizabeth; Somalwar, Sunil; Spannowsky, Michael; Spethmann, Christian; Strassler, Matthew; Su, Shufang; Tait, Tim; Thomas, Brooks; Thomas, Scott; Toro, Natalia; Volansky, Tomer; Wacker, Jay; Waltenberger, Wolfgang; Yavin, Itay; Yu, Felix; Zhao, Yue; Zurek, Kathryn
2012-01-01
This document proposes a collection of simplified models relevant to the design of new-physics searches at the LHC and the characterization of their results. Both ATLAS and CMS have already presented some results in terms of simplified models, and we encourage them to continue and expand this effort, which supplements both signature-based results and benchmark model interpretations. A simplified model is defined by an effective Lagrangian describing the interactions of a small number of new particles. Simplified models can equally well be described by a small number of masses and cross-sections. These parameters are directly related to collider physics observables, making simplified models a particularly effective framework for evaluating searches and a useful starting point for characterizing positive signals of new physics. This document serves as an official summary of the results from the "Topologies for Early LHC Searches" workshop, held at SLAC in September of 2010, the purpose of which was to develop a...
COCOA Code for Creating Mock Observations of Star Cluster Models
Askar, Abbas; Giersz, Mirek; Pych, Wojciech; Dalessandro, Emanuele
2017-01-01
We introduce and present results from the COCOA (Cluster simulatiOn Comparison with ObservAtions) code that has been developed to create idealized mock photometric observations using results from numerical simulations of star cluster evolution. COCOA is able to present the output of realistic numerical simulations of star clusters carried out using Monte Carlo or \\textit{N}-body codes in a way that is useful for direct comparison with photometric observations. In this paper, we describe the C...
International Nuclear Information System (INIS)
Lázaro, A.; Ammirabile, L.; Bandini, G.; Darmet, G.; Massara, S.; Dufour, Ph.; Tosello, A.; Gallego, E.; Jimenez, G.; Mikityuk, K.; Schikorr, M.; Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Stempniewicz, M.
2014-01-01
Highlights: • Ten system-code models of the ESFR were developed in the frame of the CP-ESFR project. • Eight different thermohydraulic system codes adapted to sodium fast reactor's technology. • Benchmarking exercise settled to check the consistency of the calculations. • Upgraded system codes able to simulate the reactivity feedback and key safety parameters. -- Abstract: The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes
Energy Technology Data Exchange (ETDEWEB)
Lázaro, A., E-mail: aurelio.lazaro-chueca@ec.europa.eu [JRC-IET European Commission—Westerduinweg 3, PO Box-2, 1755 ZG Petten (Netherlands); UPV—Universidad Politecnica de Valencia, Cami de vera s/n-46002, Valencia (Spain); Ammirabile, L. [JRC-IET European Commission—Westerduinweg 3, PO Box-2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Massara, S. [EDF, 1 avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2-28006 Madrid (Spain); Mikityuk, K. [PSI—Paul Scherrer Institut, 5232 Villigen Switzerland (Switzerland); Schikorr, M.; Bubelis, E.; Ponomarev, A.; Kruessmann, R. [KIT—Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen Germany (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, PO Box 9034 6800 ES, Arnhem (Netherlands)
2014-01-15
Highlights: • Ten system-code models of the ESFR were developed in the frame of the CP-ESFR project. • Eight different thermohydraulic system codes adapted to sodium fast reactor's technology. • Benchmarking exercise settled to check the consistency of the calculations. • Upgraded system codes able to simulate the reactivity feedback and key safety parameters. -- Abstract: The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes.
Coding a Weather Model: DOE-FIU Science & Technology Workforce Development Program.
Energy Technology Data Exchange (ETDEWEB)
Bradley, Jon David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2017-12-01
DOE Fellow, Andres Cremisini, completed a 10-week internship with Sandia National Laboratories (SNL) in Albuquerque, New Mexico. Under the management of Kristopher Klingler and the mentorship of Jon Bradley, he was tasked with conceiving and coding a realistic weather model for use in physical security applications. The objective was to make a weather model that could use real data to accurately predict wind and precipitation conditions at any location of interest on the globe at any user-determined time. The intern received guidance on software design, the C++ programming language and clear communication of project goals and ongoing progress. In addition, Mr. Cremisini was given license to structure the program however he best saw fit, an experience that will benefit ongoing research endeavors.
International Nuclear Information System (INIS)
Klos, R.A.; Sinclair, J.E.; Torres, C.; Mobbs, S.F.; Galson, D.A.
1991-01-01
The probabilistic Systems Assessment Code (PSAC) User Group of the OECD Nuclear Energy Agency has organised a series of code intercomparison studies of relevance to the performance assessment of underground repositories for radioactive wastes - known collectively by the name PSACOIN. The latest of these to be undertaken is designated PSACOIN Level 1b, and the case specification provides a complete assessment model of the behaviour of radionuclides following release into the biosphere. PSACOIN Level 1b differs from other biosphere oriented intercomparison exercises in that individual dose is the end point of the calculations as opposed to any other intermediate quantity. The PSACOIN Level 1b case specification describes a simple source term which is used to simulate the release of activity to the biosphere from certain types of near surface waste repository, the transport of radionuclides through the biosphere and their eventual uptake by humankind. The biosphere sub model comprises 4 compartments representing top and deep soil layers, river water and river sediment. The transport of radionuclides between the physical compartments is described by ten transfer coefficients and doses to humankind arise from the simultaneous consumption of water, fish, meat, milk, and grain as well as from dust inhalation and external γ-irradiation. The parameters of the exposure pathway sub model are chosen to be representative of an individual living in a small agrarian community. (13 refs., 3 figs., 2 tabs.)
Modeling Cyber Physical War Gaming
2017-08-07
games share similar constructs. We also provide a game-theoretic approach to mathematically analyze attacker and defender strategies in cyber war...Military Practice of Course-of-Action Analysis 4 2. Game-Theoretic Method 7 2.1 Mathematical Model 7 2.2 Strategy Selection 10 2.2.1 Pure...officers, hundreds of combat and support vehicles, helicopters, sophisticated intelligence and communication equipment and specialists , artillery and
Physics beyond the Standard Model
Lach, Theodore
2011-04-01
Recent discoveries of the excited states of the Bs** meson along with the discovery of the omega-b-minus have brought into popular acceptance the concept of the orbiting quarks predicted by the Checker Board Model (CBM) 14 years ago. Back then the concept of orbiting quarks was not fashionable. Recent estimates of velocities of these quarks inside the proton and neutron are in excess of 90% the speed of light also in agreement with the CBM model. Still a 2D structure of the nucleus has not been accepted nor has it been proven wrong. The CBM predicts masses of the up and dn quarks are 237.31 MeV and 42.392 MeV respectively and suggests that a lighter generation of quarks u and d make up a different generation of quarks that make up light mesons. The CBM also predicts that the T' and B' quarks do exist and are not as massive as might be expected. (this would make it a 5G world in conflict with the SM) The details of the CB model and prediction of quark masses can be found at: http://checkerboard.dnsalias.net/ (1). T.M. Lach, Checkerboard Structure of the Nucleus, Infinite Energy, Vol. 5, issue 30, (2000). (2). T.M. Lach, Masses of the Sub-Nuclear Particles, nucl-th/0008026, @http://xxx.lanl.gov/.
Ladder physics in the spin fermion model
Tsvelik, A. M.
2017-05-01
A link is established between the spin fermion (SF) model of the cuprates and the approach based on the analogy between the physics of doped Mott insulators in two dimensions and the physics of fermionic ladders. This enables one to use nonperturbative results derived for fermionic ladders to move beyond the large-N approximation in the SF model. It is shown that the paramagnon exchange postulated in the SF model has exactly the right form to facilitate the emergence of the fully gapped d -Mott state in the region of the Brillouin zone at the hot spots of the Fermi surface. Hence, the SF model provides an adequate description of the pseudogap.
A student's guide to Python for physical modeling
Kinder, Jesse M
2015-01-01
Python is a computer programming language that is rapidly gaining popularity throughout the sciences. A Student’s Guide to Python for Physical Modeling aims to help you, the student, teach yourself enough of the Python programming language to get started with physical modeling. You will learn how to install an open-source Python programming environment and use it to accomplish many common scientific computing tasks: importing, exporting, and visualizing data; numerical analysis; and simulation. No prior programming experience is assumed. This tutorial focuses on fundamentals and introduces a wide range of useful techniques, including: Basic Python programming and scripting Numerical arrays Two- and three-dimensional graphics Monte Carlo simulations Numerical methods, including solving ordinary differential equations Image processing Animation Numerous code samples and exercises—with solutions—illustrate new ideas as they are introduced. A website that accompanies this guide provides additional resourc...
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.
Ontology modeling in physical asset integrity management
Yacout, Soumaya
2015-01-01
This book presents cutting-edge applications of, and up-to-date research on, ontology engineering techniques in the physical asset integrity domain. Though a survey of state-of-the-art theory and methods on ontology engineering, the authors emphasize essential topics including data integration modeling, knowledge representation, and semantic interpretation. The book also reflects novel topics dealing with the advanced problems of physical asset integrity applications such as heterogeneity, data inconsistency, and interoperability existing in design and utilization. With a distinctive focus on applications relevant in heavy industry, Ontology Modeling in Physical Asset Integrity Management is ideal for practicing industrial and mechanical engineers working in the field, as well as researchers and graduate concerned with ontology engineering in physical systems life cycles. This book also: Introduces practicing engineers, research scientists, and graduate students to ontology engineering as a modeling techniqu...
A physical data model for fields and agents
de Jong, Kor; de Bakker, Merijn; Karssenberg, Derek
2016-04-01
Two approaches exist in simulation modeling: agent-based and field-based modeling. In agent-based (or individual-based) simulation modeling, the entities representing the system's state are represented by objects, which are bounded in space and time. Individual objects, like an animal, a house, or a more abstract entity like a country's economy, have properties representing their state. In an agent-based model this state is manipulated. In field-based modeling, the entities representing the system's state are represented by fields. Fields capture the state of a continuous property within a spatial extent, examples of which are elevation, atmospheric pressure, and water flow velocity. With respect to the technology used to create these models, the domains of agent-based and field-based modeling have often been separate worlds. In environmental modeling, widely used logical data models include feature data models for point, line and polygon objects, and the raster data model for fields. Simulation models are often either agent-based or field-based, even though the modeled system might contain both entities that are better represented by individuals and entities that are better represented by fields. We think that the reason for this dichotomy in kinds of models might be that the traditional object and field data models underlying those models are relatively low level. We have developed a higher level conceptual data model for representing both non-spatial and spatial objects, and spatial fields (De Bakker et al. 2016). Based on this conceptual data model we designed a logical and physical data model for representing many kinds of data, including the kinds used in earth system modeling (e.g. hydrological and ecological models). The goal of this work is to be able to create high level code and tools for the creation of models in which entities are representable by both objects and fields. Our conceptual data model is capable of representing the traditional feature data
International Nuclear Information System (INIS)
Khan, M.J.H.; Alam, A.B.M.K.; Ahsan, M.H.; Mamun, K.A.A.; Islam, S.M.A.
2015-01-01
Highlights: • To validate the reactor physics lattice code WIMSD-5B by this analysis. • To model TRX and BAPL critical experiments using WIMSD-5B. • To compare the calculated results with experiment and MCNP results. • To rely on WIMSD-5B code for TRIGA calculations. - Abstract: The aim of this analysis is to validate the reactor physics lattice transport code WIMSD-5B by TRX (thermal reactor-one region lattice) and BAPL (Bettis Atomic Power Laboratory-one region lattice) critical experiments of light water reactors for neutronics analysis of 3 MW TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh. This analysis is achieved through the analysis of integral parameters of five light water reactor critical experiments TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 based on evaluated nuclear data libraries JEFF-3.1 and ENDF/B-VII.1. In integral measurements, these experiments are considered as standard benchmark lattices for validating the reactor physics lattice transport code WIMSD-5B as well as evaluated nuclear data libraries. The integral parameters of the said critical experiments are calculated using the reactor physics lattice transport code WIMSD-5B. The calculated integral parameters are compared to the measured values as well as the earlier published MCNP results based on the Chinese evaluated nuclear data library CENDL-3.0 for assessment of deterministic calculation. It was found that the calculated integral parameters give mostly reasonable and globally consistent results with the experiment and the MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are well consistent with each other. Therefore, this analysis reveals the validation study of the reactor physics lattice transport code WIMSD-5B based on JEFF-3.1 and ENDF/B-VII.1 libraries and can also be essential to
International Nuclear Information System (INIS)
Fiorina, C.; Mikityuk, K.
2015-01-01
A new multi-physics solver for nuclear reactor analysis, named GeN-Foam (Generalized Nuclear Foam), has been developed by the FAST group at the Paul Scherrer Institut. It is based on OpenFOAM and has been developed for the multi-physics transient analyses of pin-based (e.g., liquid metal Fast Reactors, Light Water Reactors) or homogeneous (e.g., fast spectrum Molten Salt Reactors) nuclear reactors. It includes solutions of coarse or fine mesh thermal-hydraulics, thermal-mechanics and neutron diffusion. In particular, thermal-hydraulics solution can combine on the same mesh both a traditional RANS model and a porous medium model, depending on the desired degree of approximation for each region. In case the active reactor core is modeled as a porous medium, a simple sub-solver computes the sub-scale radial temperature profiles in fuel and cladding. The mesh used for neutronics calculations is deformed according to the displacement field predicted by the thermal-mechanics solver, thus allowing for a direct prediction of expansion-related feedback effects in Fast Reactors. To limit computational requirements, GeN-Foam permits the use of three different unstructured meshes for thermal-hydraulics, thermal-mechanics and neutron diffusion. For the same reason, an adaptive time step is employed. The different equations can be solved altogether or selectively included. In this work, GeN-Foam is applied to the analysis of the European Sodium Fast Reactor (ESFR). In particular, a 3-D model of the ESFR core is set up employing a coarse-mesh porous-medium approach for the thermal-hydraulics. The reactor steady-state and different accidental transients are investigated to offer an overview of GeN-Foam use and capabilities, as well as to preliminarily investigate the impact of a relatively accurate thermal-mechanic treatment on the predicted ESFR behavior. A code-to-code benchmark against the TRACE system code is performed to verify the adequacy of the results provided by the new
Energy Technology Data Exchange (ETDEWEB)
Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N. [Moscow Power Engineering Institute (Technical University), Moscow (Russian Federation)
2007-07-01
The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes.
International Nuclear Information System (INIS)
Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N.
2007-01-01
The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes
A multi-physics code system based on ANC9, VIPRE-W and BOA for CIPS evaluation
International Nuclear Information System (INIS)
Zhang, B.; Sung, Y.; Secker, J.; Beard, C.; Hilton, P.; Wang, G.; Oelrich, R.; Karoutas, Z.; Sung, Y.
2011-01-01
This paper summarizes the development of a multi-physics code system for evaluation of Crud Induced Power Shift (CIPS) phenomenon experienced in some Pressurized Water Reactors (PWR). CIPS is an unexpected change in reactor core axial power distribution, caused by boron compounds in crud deposited in the high power fuel assemblies undergoing subcooled boiling. As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL) sponsored by the US Department of Energy (DOE), this paper describes the initial linkage and application of a multi-physics code system ANC9/VIPRE-W/BOA for evaluating changes in core power distributions due to boron deposited in crud. The initial linkage of the code system along with the application results will be the base for the future CASL development. (author)
A multi-physics code system based on ANC9, VIPRE-W and BOA for CIPS evaluation
Energy Technology Data Exchange (ETDEWEB)
Zhang, B.; Sung, Y.; Secker, J.; Beard, C.; Hilton, P.; Wang, G.; Oelrich, R.; Karoutas, Z.; Sung, Y. [Westinghouse Electric Company LLC, Pittsburgh (United States)
2011-07-01
This paper summarizes the development of a multi-physics code system for evaluation of Crud Induced Power Shift (CIPS) phenomenon experienced in some Pressurized Water Reactors (PWR). CIPS is an unexpected change in reactor core axial power distribution, caused by boron compounds in crud deposited in the high power fuel assemblies undergoing subcooled boiling. As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL) sponsored by the US Department of Energy (DOE), this paper describes the initial linkage and application of a multi-physics code system ANC9/VIPRE-W/BOA for evaluating changes in core power distributions due to boron deposited in crud. The initial linkage of the code system along with the application results will be the base for the future CASL development. (author)
Application of the thermal-hydraulic codes in VVER-440 steam generators modelling
Energy Technology Data Exchange (ETDEWEB)
Matejovic, P.; Vranca, L.; Vaclav, E. [Nuclear Power Plant Research Inst. VUJE (Slovakia)
1995-12-31
Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.
Application of the thermal-hydraulic codes in VVER-440 steam generators modelling
Energy Technology Data Exchange (ETDEWEB)
Matejovic, P; Vranca, L; Vaclav, E [Nuclear Power Plant Research Inst. VUJE (Slovakia)
1996-12-31
Performances with the CATHARE2 V1.3U and RELAP5/MOD3.0 application to the VVER-440 SG modelling during normal conditions and during transient with secondary water lowering are described. Similar recirculation model was chosen for both codes. In the CATHARE calculation, no special measures were taken with the aim to optimize artificially flow rate distribution coefficients for the junction between SG riser and steam dome. Contrary to RELAP code, the CATHARE code is able to predict reasonable the secondary swell level in nominal conditions. Both codes are able to model properly natural phase separation on the SG water level. 6 refs.
ASTEC V2 severe accident integral code: Fission product modelling and validation
International Nuclear Information System (INIS)
Cantrel, L.; Cousin, F.; Bosland, L.; Chevalier-Jabet, K.; Marchetto, C.
2014-01-01
One main goal of the severe accident integral code ASTEC V2, jointly developed since almost more than 15 years by IRSN and GRS, is to simulate the overall behaviour of fission products (FP) in a damaged nuclear facility. ASTEC applications are source term determinations, level 2 Probabilistic Safety Assessment (PSA2) studies including the determination of uncertainties, accident management studies and physical analyses of FP experiments to improve the understanding of the phenomenology. ASTEC is a modular code and models of a part of the phenomenology are implemented in each module: the release of FPs and structural materials from degraded fuel in the ELSA module; the transport through the reactor coolant system approximated as a sequence of control volumes in the SOPHAEROS module; and the radiochemistry inside the containment nuclear building in the IODE module. Three other modules, CPA, ISODOP and DOSE, allow respectively computing the deposition rate of aerosols inside the containment, the activities of the isotopes as a function of time, and the gaseous dose rate which is needed to model radiochemistry in the gaseous phase. In ELSA, release models are semi-mechanistic and have been validated for a wide range of experimental data, and noticeably for VERCORS experiments. For SOPHAEROS, the models can be divided into two parts: vapour phase phenomena and aerosol phase phenomena. For IODE, iodine and ruthenium chemistry are modelled based on a semi-mechanistic approach, these FPs can form some volatile species and are particularly important in terms of potential radiological consequences. The models in these 3 modules are based on a wide experimental database, resulting for a large part from international programmes, and they are considered at the state of the art of the R and D knowledge. This paper illustrates some FPs modelling capabilities of ASTEC and computed values are compared to some experimental results, which are parts of the validation matrix
V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009
International Nuclear Information System (INIS)
Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.
2010-07-01
V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)
V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009
Energy Technology Data Exchange (ETDEWEB)
Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.
2010-07-15
V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)
Progress in nuclear well logging modeling using deterministic transport codes
International Nuclear Information System (INIS)
Kodeli, I.; Aldama, D.L.; Maucec, M.; Trkov, A.
2002-01-01
Further studies in continuation of the work presented in 2001 in Portoroz were performed in order to study and improve the performances, precission and domain of application of the deterministic transport codes with respect to the oil well logging analysis. These codes are in particular expected to complement the Monte Carlo solutions, since they can provide a detailed particle flux distribution in the whole geometry in a very reasonable CPU time. Real-time calculation can be envisaged. The performances of deterministic transport methods were compared to those of the Monte Carlo method. IRTMBA generic benchmark was analysed using the codes MCNP-4C and DORT/TORT. Centric as well as excentric casings were considered using 14 MeV point neutron source and NaI scintillation detectors. Neutron and gamma spectra were compared at two detector positions.(author)
Santos, José; Monteagudo, Angel
2011-02-21
As the canonical code is not universal, different theories about its origin and organization have appeared. The optimization or level of adaptation of the canonical genetic code was measured taking into account the harmful consequences resulting from point mutations leading to the replacement of one amino acid for another. There are two basic theories to measure the level of optimization: the statistical approach, which compares the canonical genetic code with many randomly generated alternative ones, and the engineering approach, which compares the canonical code with the best possible alternative. Here we used a genetic algorithm to search for better adapted hypothetical codes and as a method to guess the difficulty in finding such alternative codes, allowing to clearly situate the canonical code in the fitness landscape. This novel proposal of the use of evolutionary computing provides a new perspective in the open debate between the use of the statistical approach, which postulates that the genetic code conserves amino acid properties far better than expected from a random code, and the engineering approach, which tends to indicate that the canonical genetic code is still far from optimal. We used two models of hypothetical codes: one that reflects the known examples of codon reassignment and the model most used in the two approaches which reflects the current genetic code translation table. Although the standard code is far from a possible optimum considering both models, when the more realistic model of the codon reassignments was used, the evolutionary algorithm had more difficulty to overcome the efficiency of the canonical genetic code. Simulated evolution clearly reveals that the canonical genetic code is far from optimal regarding its optimization. Nevertheless, the efficiency of the canonical code increases when mistranslations are taken into account with the two models, as indicated by the fact that the best possible codes show the patterns of the
Directory of Open Access Journals (Sweden)
Monteagudo Ángel
2011-02-01
Full Text Available Abstract Background As the canonical code is not universal, different theories about its origin and organization have appeared. The optimization or level of adaptation of the canonical genetic code was measured taking into account the harmful consequences resulting from point mutations leading to the replacement of one amino acid for another. There are two basic theories to measure the level of optimization: the statistical approach, which compares the canonical genetic code with many randomly generated alternative ones, and the engineering approach, which compares the canonical code with the best possible alternative. Results Here we used a genetic algorithm to search for better adapted hypothetical codes and as a method to guess the difficulty in finding such alternative codes, allowing to clearly situate the canonical code in the fitness landscape. This novel proposal of the use of evolutionary computing provides a new perspective in the open debate between the use of the statistical approach, which postulates that the genetic code conserves amino acid properties far better than expected from a random code, and the engineering approach, which tends to indicate that the canonical genetic code is still far from optimal. We used two models of hypothetical codes: one that reflects the known examples of codon reassignment and the model most used in the two approaches which reflects the current genetic code translation table. Although the standard code is far from a possible optimum considering both models, when the more realistic model of the codon reassignments was used, the evolutionary algorithm had more difficulty to overcome the efficiency of the canonical genetic code. Conclusions Simulated evolution clearly reveals that the canonical genetic code is far from optimal regarding its optimization. Nevertheless, the efficiency of the canonical code increases when mistranslations are taken into account with the two models, as indicated by the
Modelling Mathematical Reasoning in Physics Education
Uhden, Olaf; Karam, Ricardo; Pietrocola, Maurício; Pospiech, Gesche
2012-04-01
Many findings from research as well as reports from teachers describe students' problem solving strategies as manipulation of formulas by rote. The resulting dissatisfaction with quantitative physical textbook problems seems to influence the attitude towards the role of mathematics in physics education in general. Mathematics is often seen as a tool for calculation which hinders a conceptual understanding of physical principles. However, the role of mathematics cannot be reduced to this technical aspect. Hence, instead of putting mathematics away we delve into the nature of physical science to reveal the strong conceptual relationship between mathematics and physics. Moreover, we suggest that, for both prospective teaching and further research, a focus on deeply exploring such interdependency can significantly improve the understanding of physics. To provide a suitable basis, we develop a new model which can be used for analysing different levels of mathematical reasoning within physics. It is also a guideline for shifting the attention from technical to structural mathematical skills while teaching physics. We demonstrate its applicability for analysing physical-mathematical reasoning processes with an example.
Hybrid microscopic depletion model in nodal code DYN3D
International Nuclear Information System (INIS)
Bilodid, Y.; Kotlyar, D.; Shwageraus, E.; Fridman, E.; Kliem, S.
2016-01-01
Highlights: • A new hybrid method of accounting for spectral history effects is proposed. • Local concentrations of over 1000 nuclides are calculated using micro depletion. • The new method is implemented in nodal code DYN3D and verified. - Abstract: The paper presents a general hybrid method that combines the micro-depletion technique with correction of micro- and macro-diffusion parameters to account for the spectral history effects. The fuel in a core is subjected to time- and space-dependent operational conditions (e.g. coolant density), which cannot be predicted in advance. However, lattice codes assume some average conditions to generate cross sections (XS) for nodal diffusion codes such as DYN3D. Deviation of local operational history from average conditions leads to accumulation of errors in XS, which is referred as spectral history effects. Various methods to account for the spectral history effects, such as spectral index, burnup-averaged operational parameters and micro-depletion, were implemented in some nodal codes. Recently, an alternative method, which characterizes fuel depletion state by burnup and 239 Pu concentration (denoted as Pu-correction) was proposed, implemented in nodal code DYN3D and verified for a wide range of history effects. The method is computationally efficient, however, it has applicability limitations. The current study seeks to improve the accuracy and applicability range of Pu-correction method. The proposed hybrid method combines the micro-depletion method with a XS characterization technique similar to the Pu-correction method. The method was implemented in DYN3D and verified on multiple test cases. The results obtained with DYN3D were compared to those obtained with Monte Carlo code Serpent, which was also used to generate the XS. The observed differences are within the statistical uncertainties.
Utilities for high performance dispersion model PHYSIC
International Nuclear Information System (INIS)
Yamazawa, Hiromi
1992-09-01
The description and usage of the utilities for the dispersion calculation model PHYSIC were summarized. The model was developed in the study of developing high performance SPEEDI with the purpose of introducing meteorological forecast function into the environmental emergency response system. The procedure of PHYSIC calculation consists of three steps; preparation of relevant files, creation and submission of JCL, and graphic output of results. A user can carry out the above procedure with the help of the Geographical Data Processing Utility, the Model Control Utility, and the Graphic Output Utility. (author)
Waste Feed Evaporation Physical Properties Modeling
International Nuclear Information System (INIS)
Daniel, W.E.
2003-01-01
This document describes the waste feed evaporator modeling work done in the Waste Feed Evaporation and Physical Properties Modeling test specification and in support of the Hanford River Protection Project (RPP) Waste Treatment Plant (WTP) project. A private database (ZEOLITE) was developed and used in this work in order to include the behavior of aluminosilicates such a NAS-gel in the OLI/ESP simulations, in addition to the development of the mathematical models. Mathematical models were developed that describe certain physical properties in the Hanford RPP-WTP waste feed evaporator process (FEP). In particular, models were developed for the feed stream to the first ultra-filtration step characterizing its heat capacity, thermal conductivity, and viscosity, as well as the density of the evaporator contents. The scope of the task was expanded to include the volume reduction factor across the waste feed evaporator (total evaporator feed volume/evaporator bottoms volume). All the physical properties were modeled as functions of the waste feed composition, temperature, and the high level waste recycle volumetric flow rate relative to that of the waste feed. The goal for the mathematical models was to predict the physical property to predicted simulation value. The simulation model approximating the FEP process used to develop the correlations was relatively complex, and not possible to duplicate within the scope of the bench scale evaporation experiments. Therefore, simulants were made of 13 design points (a subset of the points used in the model fits) using the compositions of the ultra-filtration feed streams as predicted by the simulation model. The chemistry and physical properties of the supernate (the modeled stream) as predicted by the simulation were compared with the analytical results of experimental simulant work as a method of validating the simulation software
A study on the intrusion model by physical modeling
Energy Technology Data Exchange (ETDEWEB)
Kim, Jung Yul; Kim, Yoo Sung; Hyun, Hye Ja [Korea Inst. of Geology Mining and Materials, Taejon (Korea, Republic of)
1995-12-01
In physical modeling, the actual phenomena of seismic wave propagation are directly measured like field survey and furthermore the structure and physical properties of subsurface can be known. So the measured datasets from physical modeling can be very desirable as input data to test the efficiency of various inversion algorithms. An underground structure formed by intrusion, which can be often seen in seismic section for oil exploration, is investigated by physical modeling. The model is characterized by various types of layer boundaries with steep dip angle. Therefore, this physical modeling data are very available not only to interpret seismic sections for oil exploration as a case history, but also to develop data processing techniques and estimate the capability of software such as migration, full waveform inversion. (author). 5 refs., 18 figs.
Holographic quantum error-correcting codes: toy models for the bulk/boundary correspondence
Energy Technology Data Exchange (ETDEWEB)
Pastawski, Fernando; Yoshida, Beni [Institute for Quantum Information & Matter and Walter Burke Institute for Theoretical Physics,California Institute of Technology,1200 E. California Blvd., Pasadena CA 91125 (United States); Harlow, Daniel [Princeton Center for Theoretical Science, Princeton University,400 Jadwin Hall, Princeton NJ 08540 (United States); Preskill, John [Institute for Quantum Information & Matter and Walter Burke Institute for Theoretical Physics,California Institute of Technology,1200 E. California Blvd., Pasadena CA 91125 (United States)
2015-06-23
We propose a family of exactly solvable toy models for the AdS/CFT correspondence based on a novel construction of quantum error-correcting codes with a tensor network structure. Our building block is a special type of tensor with maximal entanglement along any bipartition, which gives rise to an isometry from the bulk Hilbert space to the boundary Hilbert space. The entire tensor network is an encoder for a quantum error-correcting code, where the bulk and boundary degrees of freedom may be identified as logical and physical degrees of freedom respectively. These models capture key features of entanglement in the AdS/CFT correspondence; in particular, the Ryu-Takayanagi formula and the negativity of tripartite information are obeyed exactly in many cases. That bulk logical operators can be represented on multiple boundary regions mimics the Rindler-wedge reconstruction of boundary operators from bulk operators, realizing explicitly the quantum error-correcting features of AdS/CFT recently proposed in http://dx.doi.org/10.1007/JHEP04(2015)163.
Experimental data bases useful for quantification of model uncertainties in best estimate codes
International Nuclear Information System (INIS)
Wilson, G.E.; Katsma, K.R.; Jacobson, J.L.; Boodry, K.S.
1988-01-01
A data base is necessary for assessment of thermal hydraulic codes within the context of the new NRC ECCS Rule. Separate effect tests examine particular phenomena that may be used to develop and/or verify models and constitutive relationships in the code. Integral tests are used to demonstrate the capability of codes to model global characteristics and sequence of events for real or hypothetical transients. The nuclear industry has developed a large experimental data base of fundamental nuclear, thermal-hydraulic phenomena for code validation. Given a particular scenario, and recognizing the scenario's important phenomena, selected information from this data base may be used to demonstrate applicability of a particular code to simulate the scenario and to determine code model uncertainties. LBLOCA experimental data bases useful to this objective are identified in this paper. 2 tabs
Modelling of the RA-1 reactor using a Monte Carlo code
International Nuclear Information System (INIS)
Quinteiro, Guillermo F.; Calabrese, Carlos R.
2000-01-01
It was carried out for the first time, a model of the Argentine RA-1 reactor using the MCNP Monte Carlo code. This model was validated using data for experimental neutron and gamma measurements at different energy ranges and locations. In addition, the resulting fluxes were compared with the data obtained using a 3D diffusion code. (author)
Modelling of fluid-solid interaction using two stand-alone codes
CSIR Research Space (South Africa)
Grobler, Jan H
2010-01-01
Full Text Available A method is proposed for the modelling of fluid-solid interaction in applications where fluid forces dominate. Data are transferred between two stand-alone codes: a dedicated computational fluid dynamics (CFD) code capable of free surface modelling...
International Nuclear Information System (INIS)
Ricci, Paolo; Theiler, C.; Fasoli, A.; Furno, I.; Labit, B.; Mueller, S. H.; Podesta, M.; Poli, F. M.
2009-01-01
The methodology for plasma-turbulence code validation is discussed, with focus on the quantities to use for the simulation-experiment comparison, i.e., the validation observables, and application to the TORPEX basic plasma physics experiment [A. Fasoli et al., Phys. Plasmas 13, 055902 (2006)]. The considered validation observables are deduced from Langmuir probe measurements and are ordered into a primacy hierarchy, according to the number of model assumptions and to the combinations of measurements needed to form each of them. The lowest levels of the primacy hierarchy correspond to observables that require the lowest number of model assumptions and measurement combinations, such as the statistical and spectral properties of the ion saturation current time trace, while at the highest levels, quantities such as particle transport are considered. The comparison of the observables at the lowest levels in the hierarchy is more stringent than at the highest levels. Examples of the use of the proposed observables are applied to a specific TORPEX plasma configuration characterized by interchange-driven turbulence.
The APS SASE FEL: modeling and code comparison
International Nuclear Information System (INIS)
Biedron, S. G.
1999-01-01
A self-amplified spontaneous emission (SASE) free-electron laser (FEL) is under construction at the Advanced Photon Source (APS). Five FEL simulation codes were used in the design phase: GENESIS, GINGER, MEDUSA, RON, and TDA3D. Initial comparisons between each of these independent formulations show good agreement for the parameters of the APS SASE FEL
Physically realistic modeling of maritime training simulation
Cieutat , Jean-Marc
2003-01-01
Maritime training simulation is an important matter of maritime teaching, which requires a lot of scientific and technical skills.In this framework, where the real time constraint has to be maintained, all physical phenomena cannot be studied; the most visual physical phenomena relating to the natural elements and the ship behaviour are reproduced only. Our swell model, based on a surface wave simulation approach, permits to simulate the shape and the propagation of a regular train of waves f...
The CORSYS neutronics code system
International Nuclear Information System (INIS)
Caner, M.; Krumbein, A.D.; Saphier, D.; Shapira, M.
1994-01-01
The purpose of this work is to assemble a code package for LWR core physics including coupled neutronics, burnup and thermal hydraulics. The CORSYS system is built around the cell code WIMS (for group microscopic cross section calculations) and 3-dimension diffusion code CITATION (for burnup and fuel management). We are implementing such a system on an IBM RS-6000 workstation. The code was rested with a simplified model of the Zion Unit 2 PWR. (authors). 6 refs., 8 figs., 1 tabs
Code development of total sensitivity and uncertainty analysis for reactor physics calculations
International Nuclear Information System (INIS)
Wan, C.; Cao, L.; Wu, H.; Zu, T.; Shen, W.
2015-01-01
Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)
Code development of total sensitivity and uncertainty analysis for reactor physics calculations
Energy Technology Data Exchange (ETDEWEB)
Wan, C.; Cao, L.; Wu, H.; Zu, T., E-mail: chenghuiwan@stu.xjtu.edu.cn, E-mail: caolz@mail.xjtu.edu.cn, E-mail: hongchun@mail.xjtu.edu.cn, E-mail: tiejun@mail.xjtu.edu.cn [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Shen, W., E-mail: Wei.Shen@cnsc-ccsn.gc.ca [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Canadian Nuclear Safety Commission, Ottawa, ON (Canada)
2015-07-01
Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)
Computational models in physics teaching: a framework
Directory of Open Access Journals (Sweden)
Marco Antonio Moreira
2012-08-01
Full Text Available The purpose of the present paper is to present a theoretical framework to promote and assist meaningful physics learning through computational models. Our proposal is based on the use of a tool, the AVM diagram, to design educational activities involving modeling and computer simulations. The idea is to provide a starting point for the construction and implementation of didactical approaches grounded in a coherent epistemological view about scientific modeling.
International Nuclear Information System (INIS)
Horak, W.C.; Guppy, J.G.
1984-01-01
The Super System Code (SSC) was developed at the Brookhaven National Laboratory (BNL) for the thermal hydraulic analysis of natural circulation transients, operational transients, and other system wide transients in nuclear power plants. SSC is a generic, best estimate code that models the in-vessel components, heat transport loops, plant protection systems and plant control systems. SSC also simulates the balance of plant when interfaced with the MINET code. SSC has been validated against both numerical and experimental data bases and is now used by several outside users. An important area of interest in LMFBR transient analysis is the prediction of the response of the reactor core under low flow conditions, such as experienced during a natural circulation event. Under these circumstances there are many physical phenomena which must be modeled to provide an adequate representation by a computer code simulation. The present version of SSC contains numerous models which account for most of the major phenomena. However, one area where the present model in SSC is being improved is in the representation of heat transfer and buoyancy effects under low flow operation. To properly improve the present version, the addition of models to represent certain inter-assembly effects is required
Modeling of the YALINA booster facility by the Monte Carlo code MONK
International Nuclear Information System (INIS)
Talamo, A.; Gohar, Y.; Kondev, F.; Kiyavitskaya, H.; Serafimovich, I.; Bournos, V.; Fokov, Y.; Routkovskaya, C.
2007-01-01
The YALINA-Booster facility has been modeled according to the benchmark specifications defined for the IAEA activity without any geometrical homogenization using the Monte Carlo codes MONK and MCNP/MCNPX/MCB. The MONK model perfectly matches the MCNP one. The computational analyses have been extended through the MCB code, which is an extension of the MCNP code with burnup capability because of its additional feature for analyzing source driven multiplying assemblies. The main neutronics arameters of the YALINA-Booster facility were calculated using these computer codes with different nuclear data libraries based on ENDF/B-VI-0, -6, JEF-2.2, and JEF-3.1.
Energy Technology Data Exchange (ETDEWEB)
Joseph, Ilon [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
2014-05-27
Jacobian-free Newton-Krylov (JFNK) algorithms are a potentially powerful class of methods for solving the problem of coupling codes that address dfferent physics models. As communication capability between individual submodules varies, different choices of coupling algorithms are required. The more communication that is available, the more possible it becomes to exploit the simple sparsity pattern of the Jacobian, albeit of a large system. The less communication that is available, the more dense the Jacobian matrices become and new types of preconditioners must be sought to efficiently take large time steps. In general, methods that use constrained or reduced subsystems can offer a compromise in complexity. The specific problem of coupling a fluid plasma code to a kinetic neutrals code is discussed as an example.
Mosunova, N. A.
2018-05-01
The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium-plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal-hydraulic, neutronics, and thermal-mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal-hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code's thermal-hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.
Simplified Models for LHC New Physics Searches
International Nuclear Information System (INIS)
Alves, Daniele; Arkani-Hamed, Nima; Arora, Sanjay; Bai, Yang; Baumgart, Matthew; Berger, Joshua; Butler, Bart; Chang, Spencer; Cheng, Hsin-Chia; Cheung, Clifford; Chivukula, R. Sekhar; Cho, Won Sang; Cotta, Randy; D'Alfonso, Mariarosaria; El Hedri, Sonia; Essig, Rouven; Fitzpatrick, Liam; Fox, Patrick; Franceschini, Roberto
2012-01-01
This document proposes a collection of simplified models relevant to the design of new-physics searches at the LHC and the characterization of their results. Both ATLAS and CMS have already presented some results in terms of simplified models, and we encourage them to continue and expand this effort, which supplements both signature-based results and benchmark model interpretations. A simplified model is defined by an effective Lagrangian describing the interactions of a small number of new particles. Simplified models can equally well be described by a small number of masses and cross-sections. These parameters are directly related to collider physics observables, making simplified models a particularly effective framework for evaluating searches and a useful starting point for characterizing positive signals of new physics. This document serves as an official summary of the results from the 'Topologies for Early LHC Searches' workshop, held at SLAC in September of 2010, the purpose of which was to develop a set of representative models that can be used to cover all relevant phase space in experimental searches. Particular emphasis is placed on searches relevant for the first ∼ 50-500 pb -1 of data and those motivated by supersymmetric models. This note largely summarizes material posted at http://lhcnewphysics.org/, which includes simplified model definitions, Monte Carlo material, and supporting contacts within the theory community. We also comment on future developments that may be useful as more data is gathered and analyzed by the experiments.
Simplified Models for LHC New Physics Searches
Energy Technology Data Exchange (ETDEWEB)
Alves, Daniele; /SLAC; Arkani-Hamed, Nima; /Princeton, Inst. Advanced Study; Arora, Sanjay; /Rutgers U., Piscataway; Bai, Yang; /SLAC; Baumgart, Matthew; /Johns Hopkins U.; Berger, Joshua; /Cornell U., Phys. Dept.; Buckley, Matthew; /Fermilab; Butler, Bart; /SLAC; Chang, Spencer; /Oregon U. /UC, Davis; Cheng, Hsin-Chia; /UC, Davis; Cheung, Clifford; /UC, Berkeley; Chivukula, R.Sekhar; /Michigan State U.; Cho, Won Sang; /Tokyo U.; Cotta, Randy; /SLAC; D' Alfonso, Mariarosaria; /UC, Santa Barbara; El Hedri, Sonia; /SLAC; Essig, Rouven, (ed.); /SLAC; Evans, Jared A.; /UC, Davis; Fitzpatrick, Liam; /Boston U.; Fox, Patrick; /Fermilab; Franceschini, Roberto; /LPHE, Lausanne /Pittsburgh U. /Argonne /Northwestern U. /Rutgers U., Piscataway /Rutgers U., Piscataway /Carleton U. /CERN /UC, Davis /Wisconsin U., Madison /SLAC /SLAC /SLAC /Rutgers U., Piscataway /Syracuse U. /SLAC /SLAC /Boston U. /Rutgers U., Piscataway /Seoul Natl. U. /Tohoku U. /UC, Santa Barbara /Korea Inst. Advanced Study, Seoul /Harvard U., Phys. Dept. /Michigan U. /Wisconsin U., Madison /Princeton U. /UC, Santa Barbara /Wisconsin U., Madison /Michigan U. /UC, Davis /SUNY, Stony Brook /TRIUMF; /more authors..
2012-06-01
This document proposes a collection of simplified models relevant to the design of new-physics searches at the LHC and the characterization of their results. Both ATLAS and CMS have already presented some results in terms of simplified models, and we encourage them to continue and expand this effort, which supplements both signature-based results and benchmark model interpretations. A simplified model is defined by an effective Lagrangian describing the interactions of a small number of new particles. Simplified models can equally well be described by a small number of masses and cross-sections. These parameters are directly related to collider physics observables, making simplified models a particularly effective framework for evaluating searches and a useful starting point for characterizing positive signals of new physics. This document serves as an official summary of the results from the 'Topologies for Early LHC Searches' workshop, held at SLAC in September of 2010, the purpose of which was to develop a set of representative models that can be used to cover all relevant phase space in experimental searches. Particular emphasis is placed on searches relevant for the first {approx} 50-500 pb{sup -1} of data and those motivated by supersymmetric models. This note largely summarizes material posted at http://lhcnewphysics.org/, which includes simplified model definitions, Monte Carlo material, and supporting contacts within the theory community. We also comment on future developments that may be useful as more data is gathered and analyzed by the experiments.
PHYSICAL EDUCATION - PHYSICAL CULTURE. TWO MODELS, TWO DIDACTIC
Directory of Open Access Journals (Sweden)
Manuel Vizuete Carrizosa
2014-11-01
The survival of these conflicting positions and their interests and different views on education, in a lengthy space of time, as a consequence threw two teaching approaches and two different educational models, in which the objectives and content of education differ , and with them the forms and methods of teaching. The need to define the cultural and educational approach, in every time and place, is now a pressing need and challenge the processes of teacher training, as responsible for shaping an advanced physical education, adjusted to the time and place, the interests and needs of citizens and the democratic values of modern society.
Composing Models of Geographic Physical Processes
Hofer, Barbara; Frank, Andrew U.
Processes are central for geographic information science; yet geographic information systems (GIS) lack capabilities to represent process related information. A prerequisite to including processes in GIS software is a general method to describe geographic processes independently of application disciplines. This paper presents such a method, namely a process description language. The vocabulary of the process description language is derived formally from mathematical models. Physical processes in geography can be described in two equivalent languages: partial differential equations or partial difference equations, where the latter can be shown graphically and used as a method for application specialists to enter their process models. The vocabulary of the process description language comprises components for describing the general behavior of prototypical geographic physical processes. These process components can be composed by basic models of geographic physical processes, which is shown by means of an example.
Energy Technology Data Exchange (ETDEWEB)
Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto, E-mail: alan@lmp.ufrj.br, E-mail: andressa@lmp.ufrj.br, E-mail: schirru@lmp.ufrj.br, E-mail: ffreire@eletronuclear.gov.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil)
2017-11-01
The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10{sup 13} combinations and 10{sup 11} great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)
International Nuclear Information System (INIS)
Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto
2017-01-01
The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10"1"3 combinations and 10"1"1 great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)
Modular Modeling System (MMS) code: a versatile power plant analysis package
International Nuclear Information System (INIS)
Divakaruni, S.M.; Wong, F.K.L.
1987-01-01
The basic version of the Modular Modeling System (MMS-01), a power plant systems analysis computer code jointly developed by the Nuclear Power and the Coal Combustion Systems Divisions of the Electric Power Research Institute (EPRI), has been released to the utility power industry in April 1983 at a code release workshop held in Charlotte, North Carolina. Since then, additional modules have been developed to analyze the Pressurized Water Reactors (PWRs) and the Boiling Water Reactors (BWRs) when the safety systems are activated. Also, a selected number of modules in the MMS-01 library have been modified to allow the code users more flexibility in constructing plant specific systems for analysis. These new PWR and BWR modules constitute the new MMS library, and it includes the modifications to the MMS-01 library. A year and half long extensive code qualification program of this new version of the MMS code at EPRI and the contractor sites, back by further code testing in an user group environment is culminating in the MMS-02 code release announcement seminar. At this seminar, the results of user group efforts and the code qualification program will be presented in a series of technical sessions. A total of forty-nine papers will be presented to describe the new code features and the code qualification efforts. For the sake of completion, an overview of the code is presented to include the history of the code development, description of the MMS code and its structure, utility engineers involvement in MMS-01 and MMS-02 validations, the enhancements made in the last 18 months to the code, and finally the perspective on the code future in the fossil and nuclear industry
PHYSICAL EDUCATION - PHYSICAL CULTURE. TWO MODELS, TWO DIDACTIC
Directory of Open Access Journals (Sweden)
Manuel Vizuete Carrizosa
2014-10-01
Full Text Available Physical Education is currently facing a number of problems that are rooted in the identity crisis prompted by the spread of the professional group, the confrontation of ideas from the scientific community and the competing interests of different political and social areas, compared to which physical education has failed, or unable, to react in time. The political and ideological confrontation that characterized the twentieth century gave us two forms, each with a consistent ideological position, in which the body as a subject of education was understood from two different positions: one set from the left and communism and another, from Western democratic societies.The survival of these conflicting positions and their interests and different views on education, in a lengthy space of time, as a consequence threw two teaching approaches and two different educational models, in which the objectives and content of education differ , and with them the forms and methods of teaching. The need to define the cultural and educational approach, in every time and place, is now a pressing need and challenge the processes of teacher training, as responsible for shaping an advanced physical education, adjusted to the time and place, the interests and needs of citizens and the democratic values of modern society.
The modeling of core melting and in-vessel corium relocation in the APRIL code
Energy Technology Data Exchange (ETDEWEB)
Kim. S.W.; Podowski, M.Z.; Lahey, R.T. [Rensselaer Polytechnic Institute, Troy, NY (United States)] [and others
1995-09-01
This paper is concerned with the modeling of severe accident phenomena in boiling water reactors (BWR). New models of core melting and in-vessel corium debris relocation are presented, developed for implementation in the APRIL computer code. The results of model testing and validations are given, including comparisons against available experimental data and parametric/sensitivity studies. Also, the application of these models, as parts of the APRIL code, is presented to simulate accident progression in a typical BWR reactor.
Development of CAP code for nuclear power plant containment: Lumped model
Energy Technology Data Exchange (ETDEWEB)
Hong, Soon Joon, E-mail: sjhong90@fnctech.com [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Ha, Sang Jun [Central Research Institute, Korea Hydro & Nuclear Power Company, Ltd., 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of)
2015-09-15
Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP.
Development of CAP code for nuclear power plant containment: Lumped model
International Nuclear Information System (INIS)
Hong, Soon Joon; Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul; Ha, Sang Jun
2015-01-01
Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP
Applications of the lots computer code to laser fusion systems and other physical optics problems
International Nuclear Information System (INIS)
Lawrence, G.; Wolfe, P.N.
1979-01-01
The Laser Optical Train Simulation (LOTS) code has been developed at the Optical Sciences Center, University of Arizona under contract to Los Alamos Scientific Laboratory (LASL). LOTS is a diffraction based code designed to beam quality and energy of the laser fusion system in an end-to-end calculation
PHYSICS OF ECLIPSING BINARIES. II. TOWARD THE INCREASED MODEL FIDELITY
Energy Technology Data Exchange (ETDEWEB)
Prša, A.; Conroy, K. E.; Horvat, M.; Kochoska, A.; Hambleton, K. M. [Villanova University, Dept. of Astrophysics and Planetary Sciences, 800 E Lancaster Avenue, Villanova PA 19085 (United States); Pablo, H. [Université de Montréal, Pavillon Roger-Gaudry, 2900, boul. Édouard-Montpetit Montréal QC H3T 1J4 (Canada); Bloemen, S. [Radboud University Nijmegen, Department of Astrophysics, IMAPP, P.O. Box 9010, 6500 GL, Nijmegen (Netherlands); Giammarco, J. [Eastern University, Dept. of Astronomy and Physics, 1300 Eagle Road, St. Davids, PA 19087 (United States); Degroote, P. [KU Leuven, Instituut voor Sterrenkunde, Celestijnenlaan 200D, B-3001 Heverlee (Belgium)
2016-12-01
The precision of photometric and spectroscopic observations has been systematically improved in the last decade, mostly thanks to space-borne photometric missions and ground-based spectrographs dedicated to finding exoplanets. The field of eclipsing binary stars strongly benefited from this development. Eclipsing binaries serve as critical tools for determining fundamental stellar properties (masses, radii, temperatures, and luminosities), yet the models are not capable of reproducing observed data well, either because of the missing physics or because of insufficient precision. This led to a predicament where radiative and dynamical effects, insofar buried in noise, started showing up routinely in the data, but were not accounted for in the models. PHOEBE (PHysics Of Eclipsing BinariEs; http://phoebe-project.org) is an open source modeling code for computing theoretical light and radial velocity curves that addresses both problems by incorporating missing physics and by increasing the computational fidelity. In particular, we discuss triangulation as a superior surface discretization algorithm, meshing of rotating single stars, light travel time effects, advanced phase computation, volume conservation in eccentric orbits, and improved computation of local intensity across the stellar surfaces that includes the photon-weighted mode, the enhanced limb darkening treatment, the better reflection treatment, and Doppler boosting. Here we present the concepts on which PHOEBE is built and proofs of concept that demonstrate the increased model fidelity.