Modelling of boundary plasma in TOKES
International Nuclear Information System (INIS)
Igitkhanov, Yu.; Pestchanyi, S.; Landman, I.
2009-12-01
The main purpose of this report is the development of analytical and numerical transport models of tokamak plasmas, suitable for implementation into the integrated transport code TOKES [1-4]. Therefore this work is presented as an executive guideline for numerical implementation. The tokamak edge plasma in reactor configurations is expected to be rather thin outmost area with strong radial plasma gradients inside the separatrix and the area outside the separatrix, a scrape-off layer (SOL), with open magnetic field surfaces, terminated at the divertor plates. The region beyond the separatrix plays an important role because it serves as a shield, protecting the wall from the hot plasma and bulk plasma from the penetration of impurities and because it is mostly affected by transients. The transport model, proposed here, provides plasma density, temperature and velocity distribution along and across the magnetic field lines in bulk and the edge plasma region. It describes the dependence of temperature and density at the separatrix on the plasma conditions at the plate and the efficiency of the divertor operation in detached or attached conditions, depending on power and particle sources. The calculation gives eventually the power and particle loads on the divertor plates and side walls. During numerical implementation some simple models, allowing an analytical solution, were developed and used for comparison and checking. Some parts of the transport models were also benchmarked with experimental data from various tokamaks. In the frame of this work the following tasks have been completed: - The transport model with neoclassical and anomalous coefficients for bulk plasma and 2D transport model for the SOL have been prepared and implemented into the TOKES code. The coefficients are suitable for description of stationary plasma processes in the bulk and edge tokamak plasmas. - The model of pedestal formation at the plasma edge in H-mode operation was implemented in TOKES
Toke Point, Washington Tsunami Forecast Grids for MOST Model
National Oceanic and Atmospheric Administration, Department of Commerce — The Toke Point, Washington Forecast Model Grids provides bathymetric data strictly for tsunami inundation modeling with the Method of Splitting Tsunami (MOST) model....
TOKES studies of the thermal quench heat load reduction in mitigated ITER disruptions
Directory of Open Access Journals (Sweden)
S. Pestchanyi
2017-08-01
Full Text Available Disruption mitigation by massive gas injection (MGI of Ne gas has been simulated using the 3D TOKES code that includes the injectors of the Disruption Mitigation System (DMS as it will be implemented in ITER. The simulations have been done using a quasi-3D approach, which gives an upper limit for the radiation heat load (notwithstanding possible asymmetries in radial heat flux associated with MHD. The heating of the first wall from the radiation flash has been assessed with respect to injection quantity, the number of injectors, and their location for an H-mode ITER discharge with 280MJ of thermal energy. Simulations for the maximum quantity of Ne (8kPam3 have shown that wall melting can be avoided by using solely the three injectors in the upper ports, whereas shallow melting occurred when the midplane injector had been added. With all four injectors, melting had been avoided for a smaller neon quantity of 250Pam3 that provides still a sufficient radiation level for thermal load mitigation.
Modelling of radiation impact on ITER Beryllium wall
Landman, I. S.; Janeschitz, G.
2009-04-01
In the ITER H-Mode confinement regime, edge localized instabilities (ELMs) will perturb the discharge. Plasma lost after each ELM moves along magnetic field lines and impacts on divertor armour, causing plasma contamination by back propagating eroded carbon or tungsten. These impurities produce enhanced radiation flux distributed mainly over the beryllium main chamber wall. The simulation of the complicated processes involved are subject of the integrated tokamak code TOKES that is currently under development. This work describes the new TOKES model for radiation transport through confined plasma. Equations for level populations of the multi-fluid plasma species and the propagation of different kinds of radiation (resonance, recombination and bremsstrahlung photons) are implemented. First simulation results without account of resonance lines are presented.
Modelling of radiation impact on ITER Beryllium wall
Energy Technology Data Exchange (ETDEWEB)
Landman, I.S. [Forschungszentrum Karlsruhe, IHM, FUSION, P.O. Box 3640, 76021 Karlsruhe (Germany)], E-mail: igor.landman@ihm.fzk.de; Janeschitz, G. [Forschungszentrum Karlsruhe, IHM, FUSION, P.O. Box 3640, 76021 Karlsruhe (Germany)
2009-04-30
In the ITER H-Mode confinement regime, edge localized instabilities (ELMs) will perturb the discharge. Plasma lost after each ELM moves along magnetic field lines and impacts on divertor armour, causing plasma contamination by back propagating eroded carbon or tungsten. These impurities produce enhanced radiation flux distributed mainly over the beryllium main chamber wall. The simulation of the complicated processes involved are subject of the integrated tokamak code TOKES that is currently under development. This work describes the new TOKES model for radiation transport through confined plasma. Equations for level populations of the multi-fluid plasma species and the propagation of different kinds of radiation (resonance, recombination and bremsstrahlung photons) are implemented. First simulation results without account of resonance lines are presented.
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
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.
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
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
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.
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,
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...
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.
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)
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
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.
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
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)
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.
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
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
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
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...
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.
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.
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 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
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)
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...
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.
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.
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.
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
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...
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...
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...
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.
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
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
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.
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
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.
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)
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
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
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
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
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
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.
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.
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.
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
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
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
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
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...
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
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
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.
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)
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
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
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.
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)
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)
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 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)
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.
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
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
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)
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.
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
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)
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.
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.
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.
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
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
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)
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
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
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...
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)
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.
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...
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.
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.
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...
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)
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)
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
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.
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
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
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…
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.
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
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...
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.
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...
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
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
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
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.
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.
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....
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)
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
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
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...
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.
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)
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.
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)
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)
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
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.
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
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
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...
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
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
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.
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
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
Modelling of Cold Water Hammer with WAHA code
International Nuclear Information System (INIS)
Gale, J.; Tiselj, I.
2003-01-01
The Cold Water Hammer experiment described in the present paper is a simple facility where overpressure accelerates a column of liquid water into the steam bubble at the closed vertical end of the pipe. Severe water hammer with high pressure peak occurs when the vapor bubble condenses and the liquid column hits the closed end of the pipe. Experimental data of Forschungszentrum Rossendorf are being used to test the newly developed computer code WAHA and the computer code RELAP5. Results show that a small amount of noncondensable air in the steam bubble significantly affects the magnitude of the calculated pressure peak, while the wall friction and condensation rate only slightly affect the simulated phenomena. (author)
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
Modelling of the Rod Control System in the coupled code RELAP5-QUABOX/CUBBOX
International Nuclear Information System (INIS)
Bencik, V.; Feretic, D.; Grgic, D.
1999-01-01
There is a general agreement that for many light water reactor transient calculations, it is necessary to use a multidimensional neutron kinetics model coupled to sophisticated thermal-hydraulic models in order to obtain satisfactory results. These calculations are needed for a variety of applications for licensing safety analyses, probabilistic risk assessment, operational support, and training. At FER, Zagreb, a coupling of 3D neutronics code QUABOX/CUBBOX and system code RELAP5 was performed. In the paper the Rod Control System model in the RELAP5 part of the coupled code is presented. A first testing of the model was performed by calculation of reactor trip from full power for NPP Krsko. Results of 3D neutronics calculation obtained by coupled code QUABOX/CUBBOX were compared with point kinetics calculation performed with RELAP5 stand alone code.(author)
Improved virtual channel noise model for transform domain Wyner-Ziv video coding
DEFF Research Database (Denmark)
Huang, Xin; Forchhammer, Søren
2009-01-01
Distributed video coding (DVC) has been proposed as a new video coding paradigm to deal with lossy source coding using side information to exploit the statistics at the decoder to reduce computational demands at the encoder. A virtual channel noise model is utilized at the decoder to estimate...... the noise distribution between the side information frame and the original frame. This is one of the most important aspects influencing the coding performance of DVC. Noise models with different granularity have been proposed. In this paper, an improved noise model for transform domain Wyner-Ziv video...... coding is proposed, which utilizes cross-band correlation to estimate the Laplacian parameters more accurately. Experimental results show that the proposed noise model can improve the rate-distortion (RD) performance....
A vectorized Monte Carlo code for modeling photon transport in SPECT
International Nuclear Information System (INIS)
Smith, M.F.; Floyd, C.E. Jr.; Jaszczak, R.J.
1993-01-01
A vectorized Monte Carlo computer code has been developed for modeling photon transport in single photon emission computed tomography (SPECT). The code models photon transport in a uniform attenuating region and photon detection by a gamma camera. It is adapted from a history-based Monte Carlo code in which photon history data are stored in scalar variables and photon histories are computed sequentially. The vectorized code is written in FORTRAN77 and uses an event-based algorithm in which photon history data are stored in arrays and photon history computations are performed within DO loops. The indices of the DO loops range over the number of photon histories, and these loops may take advantage of the vector processing unit of our Stellar GS1000 computer for pipelined computations. Without the use of the vector processor the event-based code is faster than the history-based code because of numerical optimization performed during conversion to the event-based algorithm. When only the detection of unscattered photons is modeled, the event-based code executes 5.1 times faster with the use of the vector processor than without; when the detection of scattered and unscattered photons is modeled the speed increase is a factor of 2.9. Vectorization is a valuable way to increase the performance of Monte Carlo code for modeling photon transport in SPECT
Evaluation of the analysis models in the ASTRA nuclear design code system
Energy Technology Data Exchange (ETDEWEB)
Cho, Nam Jin; Park, Chang Jea; Kim, Do Sam; Lee, Kyeong Taek; Kim, Jong Woon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)
2000-11-15
In the field of nuclear reactor design, main practice was the application of the improved design code systems. During the process, a lot of basis and knowledge were accumulated in processing input data, nuclear fuel reload design, production and analysis of design data, et al. However less efforts were done in the analysis of the methodology and in the development or improvement of those code systems. Recently, KEPO Nuclear Fuel Company (KNFC) developed the ASTRA (Advanced Static and Transient Reactor Analyzer) code system for the purpose of nuclear reactor design and analysis. In the code system, two group constants were generated from the CASMO-3 code system. The objective of this research is to analyze the analysis models used in the ASTRA/CASMO-3 code system. This evaluation requires indepth comprehension of the models, which is important so much as the development of the code system itself. Currently, most of the code systems used in domestic Nuclear Power Plant were imported, so it is very difficult to maintain and treat the change of the situation in the system. Therefore, the evaluation of analysis models in the ASTRA nuclear reactor design code system in very important.
Nodal kinetics model upgrade in the Penn State coupled TRAC/NEM codes
International Nuclear Information System (INIS)
Beam, Tara M.; Ivanov, Kostadin N.; Baratta, Anthony J.; Finnemann, Herbert
1999-01-01
The Pennsylvania State University currently maintains and does development and verification work for its own versions of the coupled three-dimensional kinetics/thermal-hydraulics codes TRAC-PF1/NEM and TRAC-BF1/NEM. The subject of this paper is nodal model enhancements in the above mentioned codes. Because of the numerous validation studies that have been performed on almost every aspect of these codes, this upgrade is done without a major code rewrite. The upgrade consists of four steps. The first two steps are designed to improve the accuracy of the kinetics model, based on the nodal expansion method. The polynomial expansion solution of 1D transverse integrated diffusion equation is replaced with a solution, which uses a semi-analytic expansion. Further the standard parabolic polynomial representation of the transverse leakage in the above 1D equations is replaced with an improved approximation. The last two steps of the upgrade address the code efficiency by improving the solution of the time-dependent NEM equations and implementing a multi-grid solver. These four improvements are implemented into the standalone NEM kinetics code. Verification of this code was accomplished based on the original verification studies. The results show that the new methods improve the accuracy and efficiency of the code. The verification of the upgraded NEM model in the TRAC-PF1/NEM and TRAC-BF1/NEM coupled codes is underway
Parallelization of simulation code for liquid-gas model of lattice-gas fluid
International Nuclear Information System (INIS)
Kawai, Wataru; Ebihara, Kenichi; Kume, Etsuo; Watanabe, Tadashi
2000-03-01
A simulation code for hydrodynamical phenomena which is based on the liquid-gas model of lattice-gas fluid is parallelized by using MPI (Message Passing Interface) library. The parallelized code can be applied to the larger size of the simulations than the non-parallelized code. The calculation times of the parallelized code on VPP500 (Vector-Parallel super computer with dispersed memory units), AP3000 (Scalar-parallel server with dispersed memory units), and a workstation cluster decreased in inverse proportion to the number of processors. (author)
International Nuclear Information System (INIS)
Blanchet, Y.; Obry, P.; Louvet, J.; Deshayes, M.; Phalip, C.
1979-01-01
The explicit 2-D axisymmetric Langrangian code SIRIUS, developed at the CEA/DRNR, Cadarache, deals with transient compressive flows in deformable primary tanks with more or less complex internal component geometries. This code has been subjected to a two-year intensive validation program on scale model experiments and a number of improvements have been incorporated. This paper presents a recent calculation of one of these experiments using the SIRIUS code, and the comparison with experimental results shows the encouraging possibilities of this Lagrangian code
A Test of Two Alternative Cognitive Processing Models: Learning Styles and Dual Coding
Cuevas, Joshua; Dawson, Bryan L.
2018-01-01
This study tested two cognitive models, learning styles and dual coding, which make contradictory predictions about how learners process and retain visual and auditory information. Learning styles-based instructional practices are common in educational environments despite a questionable research base, while the use of dual coding is less…
Coupling of 3D neutronics models with the system code ATHLET
International Nuclear Information System (INIS)
Langenbuch, S.; Velkov, K.
1999-01-01
The system code ATHLET for plant transient and accident analysis has been coupled with 3D neutronics models, like QUABOX/CUBBOX, for the realistic evaluation of some specific safety problems under discussion. The considerations for the coupling approach and its realization are discussed. The specific features of the coupled code system established are explained and experience from first applications is presented. (author)
HADES. A computer code for fast neutron cross section from the Optical Model
International Nuclear Information System (INIS)
Guasp, J.; Navarro, C.
1973-01-01
A FORTRAN V computer code for UNIVAC 1108/6 using a local Optical Model with spin-orbit interaction is described. The code calculates fast neutron cross sections, angular distribution, and Legendre moments for heavy and intermediate spherical nuclei. It allows for the possibility of automatic variation of potential parameters for experimental data fitting. (Author) 55 refs
Implementation of the critical points model in a SFM-FDTD code working in oblique incidence
Energy Technology Data Exchange (ETDEWEB)
Hamidi, M; Belkhir, A; Lamrous, O [Laboratoire de Physique et Chimie Quantique, Universite Mouloud Mammeri, Tizi-Ouzou (Algeria); Baida, F I, E-mail: omarlamrous@mail.ummto.dz [Departement d' Optique P.M. Duffieux, Institut FEMTO-ST UMR 6174 CNRS Universite de Franche-Comte, 25030 Besancon Cedex (France)
2011-06-22
We describe the implementation of the critical points model in a finite-difference-time-domain code working in oblique incidence and dealing with dispersive media through the split field method. Some tests are presented to validate our code in addition to an application devoted to plasmon resonance of a gold nanoparticles grating.
International Nuclear Information System (INIS)
Chambers, F.W.; Masamitsu, J.A.; Lee, E.P.
1982-01-01
RINGBEARER II is a linearized monopole/dipole particle simulation code for studying intense relativistic electron beam propagation in gas. In this report the mathematical models utilized for beam particle dynamics and pinch field computation are delineated. Difficulties encountered in code operations and some remedies are discussed. Sample output is presented detailing the diagnostics and the methods of display and analysis utilized
FISIC - a full-wave code to model ion cyclotron resonance heating of tokamak plasmas
International Nuclear Information System (INIS)
Kruecken, T.
1988-08-01
We present a user manual for the FISIC code which solves the integrodifferential wave equation in the finite Larmor radius approximation in fully toroidal geometry to simulate ICRF heating experiments. The code models the electromagnetic wave field as well as antenna coupling and power deposition profiles in axisymmetric plasmas. (orig.)
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.
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)
A combined N-body and hydrodynamic code for modeling disk galaxies
International Nuclear Information System (INIS)
Schroeder, M.C.
1989-01-01
A combined N-body and hydrodynamic computer code for the modeling of two dimensional galaxies is described. The N-body portion of the code is used to calculate the motion of the particle component of a galaxy, while the hydrodynamics portion of the code is used to follow the motion and evolution of the fluid component. A complete description of the numerical methods used for each portion of the code is given. Additionally, the proof tests of the separate and combined portions of the code are presented and discussed. Finally, a discussion of the topics researched with the code and results obtained is presented. These include: the measurement of stellar relaxation times in disk galaxy simulations; the effects of two-armed spiral perturbations on stable axisymmetric disks; the effects of the inclusion of an instellar medium (ISM) on the stability of disk galaxies; and the effect of the inclusion of stellar evolution on disk galaxy simulations
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.
Development of thermal hydraulic models for the reliable regulatory auditing code
Energy Technology Data Exchange (ETDEWEB)
Chung, B. D.; Song, C. H.; Lee, Y. J.; Kwon, T. S.; Lee, S. W. [Korea Automic Energy Research Institute, Taejon (Korea, Republic of)
2004-02-15
The objective of this project is to develop thermal hydraulic models for use in improving the reliability of the regulatory auditing codes. The current year fall under the second step of the 3 year project, and the main researches were focused on the development of downcorner boiling model. During the current year, the bubble stream model of downcorner has been developed and installed in he auditing code. The model sensitivity analysis has been performed for APR1400 LBLOCA scenario using the modified code. The preliminary calculation has been performed for the experimental test facility using FLUENT and MARS code. The facility for air bubble experiment has been installed. The thermal hydraulic phenomena for VHTR and super critical reactor have been identified for the future application and model development.
Realistic edge field model code REFC for designing and study of isochronous cyclotron
International Nuclear Information System (INIS)
Ismail, M.
1989-01-01
The focussing properties and the requirements for isochronism in cyclotron magnet configuration are well-known in hard edge field model. The fact that they quite often change considerably in realistic field can be attributed mainly to the influence of the edge field. A solution to this problem requires a field model which allows a simple construction of equilibrium orbit and yield simple formulae. This can be achieved by using a fitted realistic edge field (Hudson et al 1975) in the region of the pole edge and such a field model is therefore called a realistic edge field model. A code REFC based on realistic edge field model has been developed to design the cyclotron sectors and the code FIELDER has been used to study the beam properties. In this report REFC code has been described along with some relevant explaination of the FIELDER code. (author). 11 refs., 6 figs
Energy Technology Data Exchange (ETDEWEB)
Farmer, M. T. [Argonne National Lab. (ANL), Argonne, IL (United States)
2017-09-01
MELTSPREAD3 is a transient one-dimensional computer code that has been developed to predict the gravity-driven flow and freezing behavior of molten reactor core materials (corium) in containment geometries. Predictions can be made for corium flowing across surfaces under either dry or wet cavity conditions. The spreading surfaces that can be selected are steel, concrete, a user-specified material (e.g., a ceramic), or an arbitrary combination thereof. The corium can have a wide range of compositions of reactor core materials that includes distinct oxide phases (predominantly Zr, and steel oxides) plus metallic phases (predominantly Zr and steel). The code requires input that describes the containment geometry, melt “pour” conditions, and cavity atmospheric conditions (i.e., pressure, temperature, and cavity flooding information). For cases in which the cavity contains a preexisting water layer at the time of RPV failure, melt jet breakup and particle bed formation can be calculated mechanistically given the time-dependent melt pour conditions (input data) as well as the heatup and boiloff of water in the melt impingement zone (calculated). For core debris impacting either the containment floor or previously spread material, the code calculates the transient hydrodynamics and heat transfer which determine the spreading and freezing behavior of the melt. The code predicts conditions at the end of the spreading stage, including melt relocation distance, depth and material composition profiles, substrate ablation profile, and wall heatup. Code output can be used as input to other models such as CORQUENCH that evaluate long term core-concrete interaction behavior following the transient spreading stage. MELTSPREAD3 was originally developed to investigate BWR Mark I liner vulnerability, but has been substantially upgraded and applied to other reactor designs (e.g., the EPR), and more recently to the plant accidents at Fukushima Daiichi. The most recent round of
Improvement on reaction model for sodium-water reaction jet code and application analysis
International Nuclear Information System (INIS)
Itooka, Satoshi; Saito, Yoshinori; Okabe, Ayao; Fujimata, Kazuhiro; Murata, Shuuichi
2000-03-01
In selecting the reasonable DBL on steam generator (SG), it is necessary to improve analytical method for estimating the sodium temperature on failure propagation due to overheating. Improvement on sodium-water reaction (SWR) jet code (LEAP-JET ver.1.30) and application analysis to the water injection tests for confirmation of code propriety were performed. On the improvement of the code, a gas-liquid interface area density model was introduced to develop a chemical reaction model with a little dependence on calculation mesh size. The test calculation using the improved code (LEAP-JET ver.1.40) were carried out with conditions of the SWAT-3·Run-19 test and an actual scale SG. It is confirmed that the SWR jet behavior on the results and the influence to analysis result of a model are reasonable. For the application analysis to the water injection tests, water injection behavior and SWR jet behavior analyses on the new SWAT-1 (SWAT-1R) and SWAT-3 (SWAT-3R) tests were performed using the LEAP-BLOW code and the LEAP-JET code. In the application analysis of the LEAP-BLOW code, parameter survey study was performed. As the results, the condition of the injection nozzle diameter needed to simulate the water leak rate was confirmed. In the application analysis of the LEAP-JET code, temperature behavior of the SWR jet was investigated. (author)
International Nuclear Information System (INIS)
Serfontein, Dawid E.; Mulder, Eben J.; Reitsma, Frederik
2014-01-01
A computer code was developed for the semi-automatic translation of input models for the VSOP-A diffusion neutronics simulation code to the format of the newer VSOP 99/05 code. In this paper, this algorithm is presented as a generic method for producing codes for the automatic translation of input models from the format of one code version to another, or even to that of a completely different code. Normally, such translations are done manually. However, input model files, such as for the VSOP codes, often are very large and may consist of many thousands of numeric entries that make no particular sense to the human eye. Therefore the task, of for instance nuclear regulators, to verify the accuracy of such translated files can be very difficult and cumbersome. This may cause translation errors not to be picked up, which may have disastrous consequences later on when a reactor with such a faulty design is built. Therefore a generic algorithm for producing such automatic translation codes may ease the translation and verification process to a great extent. It will also remove human error from the process, which may significantly enhance the accuracy and reliability of the process. The developed algorithm also automatically creates a verification log file which permanently record the names and values of each variable used, as well as the list of meanings of all the possible values. This should greatly facilitate reactor licensing applications
Energy Technology Data Exchange (ETDEWEB)
Serfontein, Dawid E., E-mail: Dawid.Serfontein@nwu.ac.za [School of Mechanical and Nuclear Engineering, North West University (PUK-Campus), PRIVATE BAG X6001 (Internal Post Box 360), Potchefstroom 2520 (South Africa); Mulder, Eben J. [School of Mechanical and Nuclear Engineering, North West University (South Africa); Reitsma, Frederik [Calvera Consultants (South Africa)
2014-05-01
A computer code was developed for the semi-automatic translation of input models for the VSOP-A diffusion neutronics simulation code to the format of the newer VSOP 99/05 code. In this paper, this algorithm is presented as a generic method for producing codes for the automatic translation of input models from the format of one code version to another, or even to that of a completely different code. Normally, such translations are done manually. However, input model files, such as for the VSOP codes, often are very large and may consist of many thousands of numeric entries that make no particular sense to the human eye. Therefore the task, of for instance nuclear regulators, to verify the accuracy of such translated files can be very difficult and cumbersome. This may cause translation errors not to be picked up, which may have disastrous consequences later on when a reactor with such a faulty design is built. Therefore a generic algorithm for producing such automatic translation codes may ease the translation and verification process to a great extent. It will also remove human error from the process, which may significantly enhance the accuracy and reliability of the process. The developed algorithm also automatically creates a verification log file which permanently record the names and values of each variable used, as well as the list of meanings of all the possible values. This should greatly facilitate reactor licensing applications.
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
Development of seismic analysis model for HTGR core on commercial FEM code
International Nuclear Information System (INIS)
Tsuji, Nobumasa; Ohashi, Kazutaka
2015-01-01
The aftermath of the Great East Japan Earthquake prods to revise the design basis earthquake intensity severely. In aseismic design of block-type HTGR, the securement of structural integrity of core blocks and other structures which are made of graphite become more important. For the aseismic design of block-type HTGR, it is necessary to predict the motion of core blocks which are collided with adjacent blocks. Some seismic analysis codes have been developed in 1970s, but these codes are special purpose-built codes and have poor collaboration with other structural analysis code. We develop the vertical 2 dimensional analytical model on multi-purpose commercial FEM code, which take into account the multiple impacts and friction between block interfaces and rocking motion on contact with dowel pins of the HTGR core by using contact elements. This model is verified by comparison with the experimental results of 12 column vertical slice vibration test. (author)
Turbine Internal and Film Cooling Modeling For 3D Navier-Stokes Codes
DeWitt, Kenneth; Garg Vijay; Ameri, Ali
2005-01-01
The aim of this research project is to make use of NASA Glenn on-site computational facilities in order to develop, validate and apply aerodynamic, heat transfer, and turbine cooling models for use in advanced 3D Navier-Stokes Computational Fluid Dynamics (CFD) codes such as the Glenn-" code. Specific areas of effort include: Application of the Glenn-HT code to specific configurations made available under Turbine Based Combined Cycle (TBCC), and Ultra Efficient Engine Technology (UEET) projects. Validating the use of a multi-block code for the time accurate computation of the detailed flow and heat transfer of cooled turbine airfoils. The goal of the current research is to improve the predictive ability of the Glenn-HT code. This will enable one to design more efficient turbine components for both aviation and power generation. The models will be tested against specific configurations provided by NASA Glenn.
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)
Context discovery using attenuated Bloom codes: model description and validation
Liu, F.; Heijenk, Geert
A novel approach to performing context discovery in ad-hoc networks based on the use of attenuated Bloom filters is proposed in this report. In order to investigate the performance of this approach, a model has been developed. This document describes the model and its validation. The model has been
Intercept Centering and Time Coding in Latent Difference Score Models
Grimm, Kevin J.
2012-01-01
Latent difference score (LDS) models combine benefits derived from autoregressive and latent growth curve models allowing for time-dependent influences and systematic change. The specification and descriptions of LDS models include an initial level of ability or trait plus an accumulation of changes. A limitation of this specification is that the…
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.
a model for quantity estimation for multi-coded team events
African Journals Online (AJOL)
Participation in multi-coded sports events often involves travel to international ... Medication use by Team south africa during the XXVIIIth olympiad: a model .... individual sports included in the programme (e.g. athletes involved in contact sports ...
Documentation for grants equal to tax model: Volume 3, Source code
International Nuclear Information System (INIS)
Boryczka, M.K.
1986-01-01
The GETT model is capable of forecasting the amount of tax liability associated with all property owned and all activities undertaken by the US Department of Energy (DOE) in site characterization and repository development. The GETT program is a user-friendly, menu-driven model developed using dBASE III/trademark/, a relational data base management system. The data base for GETT consists primarily of eight separate dBASE III/trademark/ files corresponding to each of the eight taxes (real property, personal property, corporate income, franchise, sales, use, severance, and excise) levied by State and local jurisdictions on business property and activity. Additional smaller files help to control model inputs and reporting options. Volume 3 of the GETT model documentation is the source code. The code is arranged primarily by the eight tax types. Other code files include those for JURISDICTION, SIMULATION, VALIDATION, TAXES, CHANGES, REPORTS, GILOT, and GETT. The code has been verified through hand calculations
Field-based tests of geochemical modeling codes: New Zealand hydrothermal systems
International Nuclear Information System (INIS)
Bruton, C.J.; Glassley, W.E.; Bourcier, W.L.
1993-12-01
Hydrothermal systems in the Taupo Volcanic Zone, North Island, New Zealand are being used as field-based modeling exercises for the EQ3/6 geochemical modeling code package. Comparisons of the observed state and evolution of the hydrothermal systems with predictions of fluid-solid equilibria made using geochemical modeling codes will determine how the codes can be used to predict the chemical and mineralogical response of the environment to nuclear waste emplacement. Field-based exercises allow us to test the models on time scales unattainable in the laboratory. Preliminary predictions of mineral assemblages in equilibrium with fluids sampled from wells in the Wairakei and Kawerau geothermal field suggest that affinity-temperature diagrams must be used in conjunction with EQ6 to minimize the effect of uncertainties in thermodynamic and kinetic data on code predictions
Field-based tests of geochemical modeling codes usign New Zealand hydrothermal systems
International Nuclear Information System (INIS)
Bruton, C.J.; Glassley, W.E.; Bourcier, W.L.
1994-06-01
Hydrothermal systems in the Taupo Volcanic Zone, North Island, New Zealand are being used as field-based modeling exercises for the EQ3/6 geochemical modeling code package. Comparisons of the observed state and evolution of the hydrothermal systems with predictions of fluid-solid equilibria made using geochemical modeling codes will determine how the codes can be used to predict the chemical and mineralogical response of the environment to nuclear waste emplacement. Field-based exercises allow us to test the models on time scales unattainable in the laboratory. Preliminary predictions of mineral assemblages in equilibrium with fluids sampled from wells in the Wairakei and Kawerau geothermal field suggest that affinity-temperature diagrams must be used in conjunction with EQ6 to minimize the effect of uncertainties in thermodynamic and kinetic data on code predictions
A mathematical model, and code HADES, for migration of radionuclides from a shallow repository
International Nuclear Information System (INIS)
Fraser, J.L.; Jarvis, R.G.
1985-06-01
The mathematical model is one-dimensional, to describe migration of radionuclides, by diffusion and advection, through several consecutive layers that can represent vault materials and surrounding ground. The solutions are evaluated by a computer code HADES
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)
Linear-Time Non-Malleable Codes in the Bit-Wise Independent Tampering Model
DEFF Research Database (Denmark)
Cramer, Ronald; Damgård, Ivan Bjerre; Döttling, Nico
Non-malleable codes were introduced by Dziembowski et al. (ICS 2010) as coding schemes that protect a message against tampering attacks. Roughly speaking, a code is non-malleable if decoding an adversarially tampered encoding of a message m produces the original message m or a value m' (eventuall...... non-malleable codes of Agrawal et al. (TCC 2015) and of Cher- aghchi and Guruswami (TCC 2014) and improves the previous result in the bit-wise tampering model: it builds the first non-malleable codes with linear-time complexity and optimal-rate (i.e. rate 1 - o(1)).......Non-malleable codes were introduced by Dziembowski et al. (ICS 2010) as coding schemes that protect a message against tampering attacks. Roughly speaking, a code is non-malleable if decoding an adversarially tampered encoding of a message m produces the original message m or a value m' (eventually...... abort) completely unrelated with m. It is known that non-malleability is possible only for restricted classes of tampering functions. Since their introduction, a long line of works has established feasibility results of non-malleable codes against different families of tampering functions. However...
Radiation heat transfer model for the SCDAP code
International Nuclear Information System (INIS)
Sohal, M.S.
1984-01-01
A radiation heat transfer model has been developed for severe fuel damage analysis which accounts for anisotropic effects of reflected radiation. The model simplifies the view factor calculation which results in significant savings in computational cost with little loss of accuracy. Radiation heat transfer rates calculated by the isotropic and anisotropic models compare reasonably well with those calculated by other models. The model is applied to an experimental nuclear rod bundle during a slow boiloff of the coolant liquid, a situation encountered during a loss of coolant accident with severe fuel damage. At lower temperatures and also lower temperature gradients in the core, the anisotropic effect was not found to be significant
Assessment of critical flow models of RELAP5-MOD2 and CATHARE codes
International Nuclear Information System (INIS)
Hao Laomi; Zhu Zhanchuan
1992-01-01
The critical flow tests for the long and short nozzles conducted on the SUPER MOBY-DICK facility were analyzed using the RELAP5-MOD2 and CATHARE 1.3 codes to assess the critical flow models of two codes. The critical mass flux calculated for two nozzles are given. The CATHARE code has used the thermodynamic nonequilibrium sound velocity of the two-phase fluid as the critical flow criterion, and has the better interphase transfer models and calculates the critical flow velocities with the completely implicit solution. Therefore, it can well calculate the critical flowrate and can describe the effect of the geometry L/D on the critical flowrate
MIGFRAC - a code for modelling of radionuclide transport in fracture media
International Nuclear Information System (INIS)
Satyanarayana, S.V.M.; Mohankumar, N.; Sasidhar, P.
2002-05-01
Radionuclides migrate through diffusion process from radioactive waste disposal facilities into fractures present in the host rock. The transport phenomenon is aided by the circulating ground waters. To model the transport of radionuclides in the charnockite rock formations present at Kalpakkam, a numerical code - MIGFRAC has been developed at SHINE Group, IGCAR. The code has been subjected to rigorous tests and the results of the build up of radionuclide concentrations are validated with a test case up to a distance of 100 meter along the fracture. The report discusses the model, code features and the results obtained up to a distance of 400 meter are presented. (author)
Modelling of blackout sequence at Atucha-1 using the MARCH3 code
International Nuclear Information System (INIS)
Baron, J.; Bastianelli, B.
1997-01-01
This paper presents the modelling of a complete blackout at the Atucha-1 NPP as preliminary phase for a Level II safety probabilistic analysis. The MARCH3 code of the STCP (Source Term Code Package) is used, based on a plant model made in accordance with particularities of the plant design. The analysis covers all the severe accident phases. The results allow to view the time sequence of the events, and provide the basis for source term studies. (author). 6 refs., 2 figs
RELAP5/MOD3 code manual. Volume 4, Models and correlations
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 presents modeling theory and associated numerical schemes; Volume II details instructions for code application and input data preparation; Volume III presents the results of developmental assessment cases that demonstrate and verify the models used in the code; Volume IV discusses in detail RELAP5 models and correlations; Volume V presents guidelines that have evolved over the past several years through the use of the RELAP5 code; Volume VI discusses the numerical scheme used in RELAP5; and Volume VII presents a collection of independent assessment calculations
An object-oriented framework for magnetic-fusion modeling and analysis codes
International Nuclear Information System (INIS)
Cohen, R H; Yang, T Y Brian.
1999-01-01
The magnetic-fusion energy (MFE) program, like many other scientific and engineering activities, has a need to efficiently develop complex modeling codes which combine detailed models of components to make an integrated model of a device, as well as a rich supply of legacy code that could provide the component models. There is also growing recognition in many technical fields of the desirability of steerable software: computer programs whose functionality can be changed by the user as it is run. This project had as its goals the development of two key pieces of infrastructure that are needed to combine existing code modules, written mainly in Fortran, into flexible, steerable, object-oriented integrated modeling codes for magnetic- fusion applications. These two pieces are (1) a set of tools to facilitate the interfacing of Fortran code with a steerable object-oriented framework (which we have chosen to be based on PythonlW3, an object-oriented interpreted language), and (2) a skeleton for the integrated modeling code which defines the relationships between the modules. The first of these activities obviously has immediate applicability to a spectrum of projects; the second is more focussed on the MFE application, but may be of value as an example for other applications
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 predictive coding account of bistable perception - a model-based fMRI study.
Directory of Open Access Journals (Sweden)
Veith Weilnhammer
2017-05-01
Full Text Available In bistable vision, subjective perception wavers between two interpretations of a constant ambiguous stimulus. This dissociation between conscious perception and sensory stimulation has motivated various empirical studies on the neural correlates of bistable perception, but the neurocomputational mechanism behind endogenous perceptual transitions has remained elusive. Here, we recurred to a generic Bayesian framework of predictive coding and devised a model that casts endogenous perceptual transitions as a consequence of prediction errors emerging from residual evidence for the suppressed percept. Data simulations revealed close similarities between the model's predictions and key temporal characteristics of perceptual bistability, indicating that the model was able to reproduce bistable perception. Fitting the predictive coding model to behavioural data from an fMRI-experiment on bistable perception, we found a correlation across participants between the model parameter encoding perceptual stabilization and the behaviourally measured frequency of perceptual transitions, corroborating that the model successfully accounted for participants' perception. Formal model comparison with established models of bistable perception based on mutual inhibition and adaptation, noise or a combination of adaptation and noise was used for the validation of the predictive coding model against the established models. Most importantly, model-based analyses of the fMRI data revealed that prediction error time-courses derived from the predictive coding model correlated with neural signal time-courses in bilateral inferior frontal gyri and anterior insulae. Voxel-wise model selection indicated a superiority of the predictive coding model over conventional analysis approaches in explaining neural activity in these frontal areas, suggesting that frontal cortex encodes prediction errors that mediate endogenous perceptual transitions in bistable perception. Taken together
A predictive coding account of bistable perception - a model-based fMRI study.
Weilnhammer, Veith; Stuke, Heiner; Hesselmann, Guido; Sterzer, Philipp; Schmack, Katharina
2017-05-01
In bistable vision, subjective perception wavers between two interpretations of a constant ambiguous stimulus. This dissociation between conscious perception and sensory stimulation has motivated various empirical studies on the neural correlates of bistable perception, but the neurocomputational mechanism behind endogenous perceptual transitions has remained elusive. Here, we recurred to a generic Bayesian framework of predictive coding and devised a model that casts endogenous perceptual transitions as a consequence of prediction errors emerging from residual evidence for the suppressed percept. Data simulations revealed close similarities between the model's predictions and key temporal characteristics of perceptual bistability, indicating that the model was able to reproduce bistable perception. Fitting the predictive coding model to behavioural data from an fMRI-experiment on bistable perception, we found a correlation across participants between the model parameter encoding perceptual stabilization and the behaviourally measured frequency of perceptual transitions, corroborating that the model successfully accounted for participants' perception. Formal model comparison with established models of bistable perception based on mutual inhibition and adaptation, noise or a combination of adaptation and noise was used for the validation of the predictive coding model against the established models. Most importantly, model-based analyses of the fMRI data revealed that prediction error time-courses derived from the predictive coding model correlated with neural signal time-courses in bilateral inferior frontal gyri and anterior insulae. Voxel-wise model selection indicated a superiority of the predictive coding model over conventional analysis approaches in explaining neural activity in these frontal areas, suggesting that frontal cortex encodes prediction errors that mediate endogenous perceptual transitions in bistable perception. Taken together, our current work
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.
An Eulerian transport-dispersion model of passive effluents: the Difeul code
International Nuclear Information System (INIS)
Wendum, D.
1994-11-01
R and D has decided to develop an Eulerian diffusion model easy to adapt to meteorological data coming from different sources: for instance the ARPEGE code of Meteo-France or the MERCURE code of EDF. We demand this in order to be able to apply the code in independent cases: a posteriori studies of accidental releases from nuclear power plants ar large or medium scale, simulation of urban pollution episodes within the ''Reactive Atmospheric Flows'' research project. For simplicity reasons, the numerical formulation of our code is the same as the one used in Meteo-France's MEDIA model. The numerical tests presented in this report show the good performance of those schemes. In order to illustrate the method by a concrete example a fictitious release from Saint-Laurent has been simulated at national scale: the results of this simulation agree quite well with those of the trajectory model DIFTRA. (author). 6 figs., 4 tabs
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
Welter, David E.; Doherty, John E.; Hunt, Randall J.; Muffels, Christopher T.; Tonkin, Matthew J.; Schreuder, Willem A.
2012-01-01
An object-oriented parameter estimation code was developed to incorporate benefits of object-oriented programming techniques for solving large parameter estimation modeling problems. The code is written in C++ and is a formulation and expansion of the algorithms included in PEST, a widely used parameter estimation code written in Fortran. The new code is called PEST++ and is designed to lower the barriers of entry for users and developers while providing efficient algorithms that can accommodate large, highly parameterized problems. This effort has focused on (1) implementing the most popular features of PEST in a fashion that is easy for novice or experienced modelers to use and (2) creating a software design that is easy to extend; that is, this effort provides a documented object-oriented framework designed from the ground up to be modular and extensible. In addition, all PEST++ source code and its associated libraries, as well as the general run manager source code, have been integrated in the Microsoft Visual Studio® 2010 integrated development environment. The PEST++ code is designed to provide a foundation for an open-source development environment capable of producing robust and efficient parameter estimation tools for the environmental modeling community into the future.
Description and assessment of structural and temperature models in the FRAP-T6 code
International Nuclear Information System (INIS)
Siefken, L.J.
1983-01-01
The FRAP-T6 code was developed at the Idaho National Engineering Laboratory (INEL) for the purpose of calculating the transient performance of light water reactor fuel rods during reactor transients ranging from mild operational transients to severe hypothetical loss-of-coolant accidents. An important application of the FRAP-T6 code is to calculate the structural performance of fuel rod cladding. The capabilities of the FRAP-T6 code are assessed by comparisons of code calculations with the measurements of several hundred in-pile experiments on fuel rods. The results of the assessments show that the code accurately and efficiently models the structural and thermal response of fuel rods
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.
Stimulus-dependent maximum entropy models of neural population codes.
Directory of Open Access Journals (Sweden)
Einat Granot-Atedgi
Full Text Available Neural populations encode information about their stimulus in a collective fashion, by joint activity patterns of spiking and silence. A full account of this mapping from stimulus to neural activity is given by the conditional probability distribution over neural codewords given the sensory input. For large populations, direct sampling of these distributions is impossible, and so we must rely on constructing appropriate models. We show here that in a population of 100 retinal ganglion cells in the salamander retina responding to temporal white-noise stimuli, dependencies between cells play an important encoding role. We introduce the stimulus-dependent maximum entropy (SDME model-a minimal extension of the canonical linear-nonlinear model of a single neuron, to a pairwise-coupled neural population. We find that the SDME model gives a more accurate account of single cell responses and in particular significantly outperforms uncoupled models in reproducing the distributions of population codewords emitted in response to a stimulus. We show how the SDME model, in conjunction with static maximum entropy models of population vocabulary, can be used to estimate information-theoretic quantities like average surprise and information transmission in a neural population.
International Nuclear Information System (INIS)
Wilson, R.D.; Price, R.K.; Kosanke, K.L.
1983-03-01
As part of the Department of Energy's National Uranium Resource Evaluation (NURE) project's Technology Development effort, a number of computer codes and accompanying data bases were assembled for use in modeling responses of nuclear borehole logging Sondes. The logging methods include fission neutron, active and passive gamma-ray, and gamma-gamma. These CDC-compatible computer codes and data bases are available on magnetic tape from the DOE Technical Library at its Grand Junction Area Office. Some of the computer codes are standard radiation-transport programs that have been available to the radiation shielding community for several years. Other codes were specifically written to model the response of borehole radiation detectors or are specialized borehole modeling versions of existing Monte Carlo transport programs. Results from several radiation modeling studies are available as two large data bases (neutron and gamma-ray). These data bases are accompanied by appropriate processing programs that permit the user to model a wide range of borehole and formation-parameter combinations for fission-neutron, neutron-, activation and gamma-gamma logs. The first part of this report consists of a brief abstract for each code or data base. The abstract gives the code name and title, short description, auxiliary requirements, typical running time (CDC 6600), and a list of references. The next section gives format specifications and/or directory for the tapes. The final section of the report presents listings for programs used to convert data bases between machine floating-point and EBCDIC
Modeling radiation belt dynamics using a 3-D layer method code
Wang, C.; Ma, Q.; Tao, X.; Zhang, Y.; Teng, S.; Albert, J. M.; Chan, A. A.; Li, W.; Ni, B.; Lu, Q.; Wang, S.
2017-08-01
A new 3-D diffusion code using a recently published layer method has been developed to analyze radiation belt electron dynamics. The code guarantees the positivity of the solution even when mixed diffusion terms are included. Unlike most of the previous codes, our 3-D code is developed directly in equatorial pitch angle (α0), momentum (p), and L shell coordinates; this eliminates the need to transform back and forth between (α0,p) coordinates and adiabatic invariant coordinates. Using (α0,p,L) is also convenient for direct comparison with satellite data. The new code has been validated by various numerical tests, and we apply the 3-D code to model the rapid electron flux enhancement following the geomagnetic storm on 17 March 2013, which is one of the Geospace Environment Modeling Focus Group challenge events. An event-specific global chorus wave model, an AL-dependent statistical plasmaspheric hiss wave model, and a recently published radial diffusion coefficient formula from Time History of Events and Macroscale Interactions during Substorms (THEMIS) statistics are used. The simulation results show good agreement with satellite observations, in general, supporting the scenario that the rapid enhancement of radiation belt electron flux for this event results from an increased level of the seed population by radial diffusion, with subsequent acceleration by chorus waves. Our results prove that the layer method can be readily used to model global radiation belt dynamics in three dimensions.
Development of the three dimensional flow model in the SPACE code
International Nuclear Information System (INIS)
Oh, Myung Taek; Park, Chan Eok; Kim, Shin Whan
2014-01-01
SPACE (Safety and Performance Analysis CodE) is a nuclear plant safety analysis code, which has been developed in the Republic of Korea through a joint research between the Korean nuclear industry and research institutes. The SPACE code has been developed with multi-dimensional capabilities as a requirement of the next generation safety code. It allows users to more accurately model the multi-dimensional flow behavior that can be exhibited in components such as the core, lower plenum, upper plenum and downcomer region. Based on generalized models, the code can model any configuration or type of fluid system. All the geometric quantities of mesh are described in terms of cell volume, centroid, face area, and face center, so that it can naturally represent not only the one dimensional (1D) or three dimensional (3D) Cartesian system, but also the cylindrical mesh system. It is possible to simulate large and complex domains by modelling the complex parts with a 3D approach and the rest of the system with a 1D approach. By 1D/3D co-simulation, more realistic conditions and component models can be obtained, providing a deeper understanding of complex systems, and it is expected to overcome the shortcomings of 1D system codes. (author)
A new computer code for discrete fracture network modelling
Xu, Chaoshui; Dowd, Peter
2010-03-01
The authors describe a comprehensive software package for two- and three-dimensional stochastic rock fracture simulation using marked point processes. Fracture locations can be modelled by a Poisson, a non-homogeneous, a cluster or a Cox point process; fracture geometries and properties are modelled by their respective probability distributions. Virtual sampling tools such as plane, window and scanline sampling are included in the software together with a comprehensive set of statistical tools including histogram analysis, probability plots, rose diagrams and hemispherical projections. The paper describes in detail the theoretical basis of the implementation and provides a case study in rock fracture modelling to demonstrate the application of the software.
Provisional safety analyses for SGT stage 2 -- Models, codes and general modelling approach
International Nuclear Information System (INIS)
2014-12-01
In the framework of the provisional safety analyses for Stage 2 of the Sectoral Plan for Deep Geological Repositories (SGT), deterministic modelling of radionuclide release from the barrier system along the groundwater pathway during the post-closure period of a deep geological repository is carried out. The calculated radionuclide release rates are interpreted as annual effective dose for an individual and assessed against the regulatory protection criterion 1 of 0.1 mSv per year. These steps are referred to as dose calculations. Furthermore, from the results of the dose calculations so-called characteristic dose intervals are determined, which provide input to the safety-related comparison of the geological siting regions in SGT Stage 2. Finally, the results of the dose calculations are also used to illustrate and to evaluate the post-closure performance of the barrier systems under consideration. The principal objective of this report is to describe comprehensively the technical aspects of the dose calculations. These aspects comprise: · the generic conceptual models of radionuclide release from the solid waste forms, of radionuclide transport through the system of engineered and geological barriers, of radionuclide transfer in the biosphere, as well as of the potential radiation exposure of the population, · the mathematical models for the explicitly considered release and transport processes, as well as for the radiation exposure pathways that are included, · the implementation of the mathematical models in numerical codes, including an overview of these codes and the most relevant verification steps, · the general modelling approach when using the codes, in particular the generic assumptions needed to model the near field and the geosphere, along with some numerical details, · a description of the work flow related to the execution of the calculations and of the software tools that are used to facilitate the modelling process, and · an overview of the
Provisional safety analyses for SGT stage 2 -- Models, codes and general modelling approach
Energy Technology Data Exchange (ETDEWEB)
NONE
2014-12-15
In the framework of the provisional safety analyses for Stage 2 of the Sectoral Plan for Deep Geological Repositories (SGT), deterministic modelling of radionuclide release from the barrier system along the groundwater pathway during the post-closure period of a deep geological repository is carried out. The calculated radionuclide release rates are interpreted as annual effective dose for an individual and assessed against the regulatory protection criterion 1 of 0.1 mSv per year. These steps are referred to as dose calculations. Furthermore, from the results of the dose calculations so-called characteristic dose intervals are determined, which provide input to the safety-related comparison of the geological siting regions in SGT Stage 2. Finally, the results of the dose calculations are also used to illustrate and to evaluate the post-closure performance of the barrier systems under consideration. The principal objective of this report is to describe comprehensively the technical aspects of the dose calculations. These aspects comprise: · the generic conceptual models of radionuclide release from the solid waste forms, of radionuclide transport through the system of engineered and geological barriers, of radionuclide transfer in the biosphere, as well as of the potential radiation exposure of the population, · the mathematical models for the explicitly considered release and transport processes, as well as for the radiation exposure pathways that are included, · the implementation of the mathematical models in numerical codes, including an overview of these codes and the most relevant verification steps, · the general modelling approach when using the codes, in particular the generic assumptions needed to model the near field and the geosphere, along with some numerical details, · a description of the work flow related to the execution of the calculations and of the software tools that are used to facilitate the modelling process, and · an overview of the
Comparing the line broadened quasilinear model to Vlasov code
International Nuclear Information System (INIS)
Ghantous, K.; Berk, H. L.; Gorelenkov, N. N.
2014-01-01
The Line Broadened Quasilinear (LBQ) model is revisited to study its predicted saturation level as compared with predictions of a Vlasov solver BOT [Lilley et al., Phys. Rev. Lett. 102, 195003 (2009) and M. Lilley, BOT Manual. The parametric dependencies of the model are modified to achieve more accuracy compared to the results of the Vlasov solver both in regards to a mode amplitude's time evolution to a saturated state and its final steady state amplitude in the parameter space of the model's applicability. However, the regions of stability as predicted by LBQ model and BOT are found to significantly differ from each other. The solutions of the BOT simulations are found to have a larger region of instability than the LBQ simulations
Comparing the line broadened quasilinear model to Vlasov code
Energy Technology Data Exchange (ETDEWEB)
Ghantous, K. [Laboratoire de Physique des Plasmas, Ecole Polytechnique, 91128 Palaiseau Cedex (France); Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543-0451 (United States); Berk, H. L. [Institute for Fusion Studies, University of Texas, 2100 San Jacinto Blvd, Austin, Texas 78712-1047 (United States); Gorelenkov, N. N. [Princeton Plasma Physics Laboratory, P.O. Box 451, Princeton, New Jersey 08543-0451 (United States)
2014-03-15
The Line Broadened Quasilinear (LBQ) model is revisited to study its predicted saturation level as compared with predictions of a Vlasov solver BOT [Lilley et al., Phys. Rev. Lett. 102, 195003 (2009) and M. Lilley, BOT Manual. The parametric dependencies of the model are modified to achieve more accuracy compared to the results of the Vlasov solver both in regards to a mode amplitude's time evolution to a saturated state and its final steady state amplitude in the parameter space of the model's applicability. However, the regions of stability as predicted by LBQ model and BOT are found to significantly differ from each other. The solutions of the BOT simulations are found to have a larger region of instability than the LBQ simulations.
Comparing the line broadened quasilinear model to Vlasov code
Ghantous, K.; Berk, H. L.; Gorelenkov, N. N.
2014-03-01
The Line Broadened Quasilinear (LBQ) model is revisited to study its predicted saturation level as compared with predictions of a Vlasov solver BOT [Lilley et al., Phys. Rev. Lett. 102, 195003 (2009) and M. Lilley, BOT Manual. The parametric dependencies of the model are modified to achieve more accuracy compared to the results of the Vlasov solver both in regards to a mode amplitude's time evolution to a saturated state and its final steady state amplitude in the parameter space of the model's applicability. However, the regions of stability as predicted by LBQ model and BOT are found to significantly differ from each other. The solutions of the BOT simulations are found to have a larger region of instability than the LBQ simulations.
The aeroelastic code HawC - model and comparisons
Energy Technology Data Exchange (ETDEWEB)
Thirstrup Petersen, J. [Risoe National Lab., The Test Station for Wind Turbines, Roskilde (Denmark)
1996-09-01
A general aeroelastic finite element model for simulation of the dynamic response of horizontal axis wind turbines is presented. The model has been developed with the aim to establish an effective research tool, which can support the general investigation of wind turbine dynamics and research in specific areas of wind turbine modelling. The model concentrates on the correct representation of the inertia forces in a form, which makes it possible to recognize and isolate effects originating from specific degrees of freedom. The turbine structure is divided into substructures, and nonlinear kinematic terms are retained in the equations of motion. Moderate geometric nonlinearities are allowed for. Gravity and a full wind field including 3-dimensional 3-component turbulence are included in the loading. Simulation results for a typical three bladed, stall regulated wind turbine are presented and compared with measurements. (au)
Improved gap conductance model for the TRAC code
International Nuclear Information System (INIS)
Hatch, S.W.; Mandell, D.A.
1980-01-01
The purpose of the present work, as indicated earlier, is to improve the present constant fuel clad spacing in TRAC-P1A without significantly increasing the computer costs. It is realized that the simple model proposed may not be accurate enough for some cases, but for the initial calculations made the DELTAR model improves the predictions over the constant Δr results of TRAC-P1A and the additional computing costs are negligible
Code modernization and modularization of APEX and SWAT watershed simulation models
SWAT (Soil and Water Assessment Tool) and APEX (Agricultural Policy / Environmental eXtender) are respectively large and small watershed simulation models derived from EPIC Environmental Policy Integrated Climate), a field-scale agroecology simulation model. All three models are coded in FORTRAN an...
PWR hot leg natural circulation modeling with MELCOR code
Energy Technology Data Exchange (ETDEWEB)
Park, Jae Hong; Lee, Jong In [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)
1997-12-31
Previous MELCOR and SCDAP/RELAP5 nodalizations for simulating the counter-current, natural circulation behavior of vapor flow within the RCS hot legs and SG U-tubes when core damage progress can not be applied to the steady state and water-filled conditions during the initial period of accident progression because of the artificially high loss coefficients in the hot legs and SG U-tubes which were chosen from results of COMMIX calculation and the Westinghouse natural circulation experiments in a 1/7-scale facility for simulating steam natural circulation behavior in the vessel and circulation modeling which can be used both for the liquid flow condition at steady state and for the vapor flow condition at the later period of in-vessel core damage. For this, the drag forces resulting from the momentum exchange effects between the two vapor streams in the hot leg was modeled as a pressure drop by pump model. This hot leg natural circulation modeling of MELCOR was able to reproduce similar mass flow rates with those predicted by previous models. 6 refs., 2 figs. (Author)
PWR hot leg natural circulation modeling with MELCOR code
Energy Technology Data Exchange (ETDEWEB)
Park, Jae Hong; Lee, Jong In [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)
1998-12-31
Previous MELCOR and SCDAP/RELAP5 nodalizations for simulating the counter-current, natural circulation behavior of vapor flow within the RCS hot legs and SG U-tubes when core damage progress can not be applied to the steady state and water-filled conditions during the initial period of accident progression because of the artificially high loss coefficients in the hot legs and SG U-tubes which were chosen from results of COMMIX calculation and the Westinghouse natural circulation experiments in a 1/7-scale facility for simulating steam natural circulation behavior in the vessel and circulation modeling which can be used both for the liquid flow condition at steady state and for the vapor flow condition at the later period of in-vessel core damage. For this, the drag forces resulting from the momentum exchange effects between the two vapor streams in the hot leg was modeled as a pressure drop by pump model. This hot leg natural circulation modeling of MELCOR was able to reproduce similar mass flow rates with those predicted by previous models. 6 refs., 2 figs. (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
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
The improvement of the heat transfer model for sodium-water reaction jet code
International Nuclear Information System (INIS)
Hashiguchi, Yoshirou; Yamamoto, Hajime; Kamoshida, Norio; Murata, Shuuichi
2001-02-01
For confirming the reasonable DBL (Design Base Leak) on steam generator (SG), it is necessary to evaluate phenomena of sodium-water reaction (SWR) in an actual steam generator realistically. The improvement of a heat transfer model on sodium-water reaction (SWR) jet code (LEAP-JET ver.1.40) and application analysis to the water injection tests for confirmation of propriety for the code were performed. On the improvement of the code, the heat transfer model between a inside fluid and a tube wall was introduced instead of the prior model which was heat capacity model including both heat capacity of the tube wall and inside fluid. And it was considered that the fluid of inside the heat exchange tube was able to treat as water or sodium and typical heat transfer equations used in SG design were also introduced in the new heat transfer model. Further additional work was carried out in order to improve the stability of the calculation for long calculation time. The test calculation using the improved code (LEAP-JET ver.1.50) were carried out with conditions of the SWAT-IR·Run-HT-2 test. It was confirmed that the SWR jet behavior on the result and the influence to the result of the heat transfer model were reasonable. And also on the improved code (LEAP-JET ver.1.50), user's manual was revised with additional I/O manual and explanation of the heat transfer model and new variable name. (author)
Numerical and modeling techniques used in the EPIC code
International Nuclear Information System (INIS)
Pizzica, P.A.; Abramson, P.B.
1977-01-01
EPIC models fuel and coolant motion which result from internal fuel pin pressure (from fission gas or fuel vapor) and/or from the generation of sodium vapor pressures in the coolant channel subsequent to pin failure in an LMFBR. The modeling includes the ejection of molten fuel from the pin into a coolant channel with any amount of voiding through a clad rip which may be of any length or which may expand with time. One-dimensional Eulerian hydrodynamics is used to model both the motion of fuel and fission gas inside a molten fuel cavity and the mixture of two-phase sodium and fission gas in the channel. Motion of molten fuel particles in the coolant channel is tracked with a particle-in-cell technique
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.)
Development and validation of the ENIGMA code for MOX fuel performance modelling
International Nuclear Information System (INIS)
Palmer, I.; Rossiter, G.; White, R.J.
2000-01-01
The ENIGMA fuel performance code has been under development in the UK since the mid-1980s with contributions made by both the fuel vendor (BNFL) and the utility (British Energy). In recent years it has become the principal code for UO 2 fuel licensing for both PWR and AGR reactor systems in the UK and has also been used by BNFL in support of overseas UO 2 and MOX fuel business. A significant new programme of work has recently been initiated by BNFL to further develop the code specifically for MOX fuel application. Model development is proceeding hand in hand with a major programme of MOX fuel testing and PIE studies, with the objective of producing a fuel modelling code suitable for mechanistic analysis, as well as for licensing applications. This paper gives an overview of the model developments being undertaken and of the experimental data being used to underpin and to validate the code. The paper provides a summary of the code development programme together with specific examples of new models produced. (author)
International Nuclear Information System (INIS)
Lee, Suk Ho; You, Sung Chang; Kim, Han Gon
2011-01-01
The SBLOCA (Small Break Loss-of-Coolant Accident) evaluation methodology for the APR1400 (Advanced Power Reactor 1400) is under development using the SPACE code. The goal of the development of this methodology is to set up a conservative evaluation methodology in accordance with Appendix K of 10CFR50 by the end of 2012. In order to develop the Appendix K version of the SPACE code, the code modification is considered through implementation of the code on the required evaluation models. For the conservative models required in the SPACE code, the metal-water reaction (MWR) model, the critical flow model, the Critical Heat Flux (CHF) model and the post-CHF model must be implemented in the code. At present, the integration of the model to generate the Appendix K version of SPACE is in its preliminary stage. Among them, the conservative MWR model and its code applicability are introduced in this paper
Recommendations for computer modeling codes to support the UMTRA groundwater restoration project
Energy Technology Data Exchange (ETDEWEB)
Tucker, M.D. [Sandia National Labs., Albuquerque, NM (United States); Khan, M.A. [IT Corp., Albuquerque, NM (United States)
1996-04-01
The Uranium Mill Tailings Remediation Action (UMTRA) Project is responsible for the assessment and remedial action at the 24 former uranium mill tailings sites located in the US. The surface restoration phase, which includes containment and stabilization of the abandoned uranium mill tailings piles, has a specific termination date and is nearing completion. Therefore, attention has now turned to the groundwater restoration phase, which began in 1991. Regulated constituents in groundwater whose concentrations or activities exceed maximum contaminant levels (MCLs) or background levels at one or more sites include, but are not limited to, uranium, selenium, arsenic, molybdenum, nitrate, gross alpha, radium-226 and radium-228. The purpose of this report is to recommend computer codes that can be used to assist the UMTRA groundwater restoration effort. The report includes a survey of applicable codes in each of the following areas: (1) groundwater flow and contaminant transport modeling codes, (2) hydrogeochemical modeling codes, (3) pump and treat optimization codes, and (4) decision support tools. Following the survey of the applicable codes, specific codes that can best meet the needs of the UMTRA groundwater restoration program in each of the four areas are recommended.
Energy Technology Data Exchange (ETDEWEB)
Di Maio, Francesco, E-mail: francesco.dimaio@polimi.it [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Nicola, Giancarlo [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Zio, Enrico [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Chair on System Science and Energetic Challenge Fondation EDF, Ecole Centrale Paris and Supelec, Paris (France); Yu, Yu [School of Nuclear Science and Engineering, North China Electric Power University, 102206 Beijing (China)
2015-08-15
Highlights: • Uncertainties of TH codes affect the system failure probability quantification. • We present Finite Mixture Models (FMMs) for sensitivity analysis of TH codes. • FMMs approximate the pdf of the output of a TH code with a limited number of simulations. • The approach is tested on a Passive Containment Cooling System of an AP1000 reactor. • The novel approach overcomes the results of a standard variance decomposition method. - Abstract: For safety analysis of Nuclear Power Plants (NPPs), Best Estimate (BE) Thermal Hydraulic (TH) codes are used to predict system response in normal and accidental conditions. The assessment of the uncertainties of TH codes is a critical issue for system failure probability quantification. In this paper, we consider passive safety systems of advanced NPPs and present a novel approach of Sensitivity Analysis (SA). The approach is based on Finite Mixture Models (FMMs) to approximate the probability density function (i.e., the uncertainty) of the output of the passive safety system TH code with a limited number of simulations. We propose a novel Sensitivity Analysis (SA) method for keeping the computational cost low: an Expectation Maximization (EM) algorithm is used to calculate the saliency of the TH code input variables for identifying those that most affect the system functional failure. The novel approach is compared with a standard variance decomposition method on a case study considering a Passive Containment Cooling System (PCCS) of an Advanced Pressurized reactor AP1000.
Recommendations for computer modeling codes to support the UMTRA groundwater restoration project
International Nuclear Information System (INIS)
Tucker, M.D.; Khan, M.A.
1996-04-01
The Uranium Mill Tailings Remediation Action (UMTRA) Project is responsible for the assessment and remedial action at the 24 former uranium mill tailings sites located in the US. The surface restoration phase, which includes containment and stabilization of the abandoned uranium mill tailings piles, has a specific termination date and is nearing completion. Therefore, attention has now turned to the groundwater restoration phase, which began in 1991. Regulated constituents in groundwater whose concentrations or activities exceed maximum contaminant levels (MCLs) or background levels at one or more sites include, but are not limited to, uranium, selenium, arsenic, molybdenum, nitrate, gross alpha, radium-226 and radium-228. The purpose of this report is to recommend computer codes that can be used to assist the UMTRA groundwater restoration effort. The report includes a survey of applicable codes in each of the following areas: (1) groundwater flow and contaminant transport modeling codes, (2) hydrogeochemical modeling codes, (3) pump and treat optimization codes, and (4) decision support tools. Following the survey of the applicable codes, specific codes that can best meet the needs of the UMTRA groundwater restoration program in each of the four areas are recommended
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)
International Nuclear Information System (INIS)
Ho, C.K.; Altman, S.J.; Arnold, B.W.
1995-09-01
Groundwater travel time (GWTT) calculations will play an important role in addressing site-suitability criteria for the potential high-level nuclear waste repository at Yucca Mountain,Nevada. In support of these calculations, Preliminary assessments of the candidate codes and models are presented in this report. A series of benchmark studies have been designed to address important aspects of modeling flow through fractured media representative of flow at Yucca Mountain. Three codes (DUAL, FEHMN, and TOUGH 2) are compared in these benchmark studies. DUAL is a single-phase, isothermal, two-dimensional flow simulator based on the dual mixed finite element method. FEHMN is a nonisothermal, multiphase, multidimensional simulator based primarily on the finite element method. TOUGH2 is anon isothermal, multiphase, multidimensional simulator based on the integral finite difference method. Alternative conceptual models of fracture flow consisting of the equivalent continuum model (ECM) and the dual permeability (DK) model are used in the different codes
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
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.
Nuclear model codes and related software distributed by the OECD/NEA Data Bank
International Nuclear Information System (INIS)
Sartori, E.
1993-01-01
Software and data for nuclear energy applications is acquired, tested and distributed by several information centres; in particular, relevant computer codes are distributed internationally by the OECD/NEA Data Bank (France) and by ESTSC and EPIC/RSIC (United States). This activity is coordinated among the centres and is extended outside the OECD area through an arrangement with the IAEA. This article covers more specifically the availability of nuclear model codes and also those codes which further process their results into data sets needed for specific nuclear application projects. (author). 2 figs
Energy Technology Data Exchange (ETDEWEB)
Veshchunov, M.S.; Kisselev, A.E.; Palagin, A.V. [Nuclear Safety Institute, Moscow (Russian Federation)] [and others
1995-09-01
The code package SVECHA for the modeling of in-vessel core degradation (CD) phenomena in severe accidents is being developed in the Nuclear Safety Institute, Russian Academy of Science (NSI RAS). The code package presents a detailed mechanistic description of the phenomenology of severe accidents in a reactor core. The modules of the package were developed and validated on separate effect test data. These modules were then successfully implemented in the ICARE2 code and validated against a wide range of integral tests. Validation results have shown good agreement with separate effect tests data and with the integral tests CORA-W1/W2, CORA-13, PHEBUS-B9+.
Implementing the WebSocket Protocol Based on Formal Modelling and Automated Code Generation
DEFF Research Database (Denmark)
Simonsen, Kent Inge; Kristensen, Lars Michael
2014-01-01
with pragmatic annotations for automated code generation of protocol software. The contribution of this paper is an application of the approach as implemented in the PetriCode tool to obtain protocol software implementing the IETF WebSocket protocol. This demonstrates the scalability of our approach to real...... protocols. Furthermore, we perform formal verification of the CPN model prior to code generation, and test the implementation for interoperability against the Autobahn WebSocket test-suite resulting in 97% and 99% success rate for the client and server implementation, respectively. The tests show...
A Secure Network Coding-based Data Gathering Model and Its Protocol in Wireless Sensor Networks
Directory of Open Access Journals (Sweden)
Qian Xiao
2012-09-01
Full Text Available To provide security for data gathering based on network coding in wireless sensor networks (WSNs, a secure network coding-based data gathering model is proposed, and a data-privacy preserving and pollution preventing (DPPaamp;PP protocol using network coding is designed. DPPaamp;PP makes use of a new proposed pollution symbol selection and pollution (PSSP scheme based on a new obfuscation idea to pollute existing symbols. Analyses of DPPaamp;PP show that it not only requires low overhead on computation and communication, but also provides high security on resisting brute-force attacks.
Comparison for the interfacial and wall friction models in thermal-hydraulic system analysis codes
International Nuclear Information System (INIS)
Hwang, Moon Kyu; Park, Jee Won; Chung, Bub Dong; Kim, Soo Hyung; Kim, See Dal
2007-07-01
The average equations employed in the current thermal hydraulic analysis codes need to be closed with the appropriate models and correlations to specify the interphase phenomena along with fluid/structure interactions. This includes both thermal and mechanical interactions. Among the closure laws, an interfacial and wall frictions, which are included in the momentum equations, not only affect pressure drops along the fluid flow, but also have great effects for the numerical stability of the codes. In this study, the interfacial and wall frictions are reviewed for the commonly applied thermal-hydraulic system analysis codes, i.e. RELAP5-3D, MARS-3D, TRAC-M, and CATHARE
Description of codes and models to be used in risk assessment
International Nuclear Information System (INIS)
1991-09-01
Human health and environmental risk assessments will be performed as part of the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) remedial investigation/feasibility study (RI/FS) activities at the Hanford Site. Analytical and computer encoded numerical models are commonly used during both the remedial investigation (RI) and feasibility study (FS) to predict or estimate the concentration of contaminants at the point of exposure to humans and/or the environment. This document has been prepared to identify the computer codes that will be used in support of RI/FS human health and environmental risk assessments at the Hanford Site. In addition to the CERCLA RI/FS process, it is recommended that these computer codes be used when fate and transport analyses is required for other activities. Additional computer codes may be used for other purposes (e.g., design of tracer tests, location of observation wells, etc.). This document provides guidance for unit managers in charge of RI/FS activities. Use of the same computer codes for all analytical activities at the Hanford Site will promote consistency, reduce the effort required to develop, validate, and implement models to simulate Hanford Site conditions, and expedite regulatory review. The discussion provides a description of how models will likely be developed and utilized at the Hanford Site. It is intended to summarize previous environmental-related modeling at the Hanford Site and provide background for future model development. The modeling capabilities that are desirable for the Hanford Site and the codes that were evaluated. The recommendations include the codes proposed to support future risk assessment modeling at the Hanford Site, and provides the rational for the codes selected. 27 refs., 3 figs., 1 tab
International Nuclear Information System (INIS)
King, C.M.; Wilhite, E.L.; Root, R.W. Jr.
1985-01-01
The Savannah River Laboratory DOSTOMAN code has been used since 1978 for environmental pathway analysis of potential migration of radionuclides and hazardous chemicals. The DOSTOMAN work is reviewed including a summary of historical use of compartmental models, the mathematical basis for the DOSTOMAN code, examples of exact analytical solutions for simple matrices, methods for numerical solution of complex matrices, and mathematical validation/calibration of the SRL code. The review includes the methodology for application to nuclear and hazardous chemical waste disposal, examples of use of the model in contaminant transport and pathway analysis, a user's guide for computer implementation, peer review of the code, and use of DOSTOMAN at other Department of Energy sites. 22 refs., 3 figs
Mechanical modelling of PCI with FRAGEMA and CEA finite element codes
International Nuclear Information System (INIS)
Joseph, J.; Bernard, Ph.; Atabek, R.; Chantant, M.
1983-01-01
In the framework of their common program, CEA and FRAGEMA have undertaken the mechanical modelization of PCI. In the first step two different codes, TITUS and VERDON, have been tested by FRAGEMA and CEA respectively. Whereas the two codes use a finite element method to describe the thermomechanical behaviour of a fuel element, input models are not the same for the two codes: to take into account the presence of cracks in UO 2 , an axisymmetric two dimensional mesh pattern and the Druecker-Prager criterion are used in VERDON and a 3D equivalent method in TITUS. Two rods have been studied with these two methods: PRISCA 04bis and PRISCA 104 which were ramped in SILOE. The results show that the stresses and strains are the same with the two codes. These methods are further applied to the complete series of the common ramp test rods program of FRAGEMA and CEA. (author)
A model of a code of ethics for tissue banks operating in developing countries.
Morales Pedraza, Jorge
2012-12-01
Ethical practice in the field of tissue banking requires the setting of principles, the identification of possible deviations and the establishment of mechanisms that will detect and hinder abuses that may occur during the procurement, processing and distribution of tissues for transplantation. This model of a Code of Ethics has been prepared with the purpose of being used for the elaboration of a Code of Ethics for tissue banks operating in the Latin American and the Caribbean, Asia and the Pacific and the African regions in order to guide the day-to-day operation of these banks. The purpose of this model of Code of Ethics is to assist interested tissue banks in the preparation of their own Code of Ethics towards ensuring that the tissue bank staff support with their actions the mission and values associated with tissue banking.
A Dual Coding Theoretical Model of Decoding in Reading: Subsuming the LaBerge and Samuels Model
Sadoski, Mark; McTigue, Erin M.; Paivio, Allan
2012-01-01
In this article we present a detailed Dual Coding Theory (DCT) model of decoding. The DCT model reinterprets and subsumes The LaBerge and Samuels (1974) model of the reading process which has served well to account for decoding behaviors and the processes that underlie them. However, the LaBerge and Samuels model has had little to say about…
Energy Technology Data Exchange (ETDEWEB)
Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)
2011-06-01
This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the
Model of nuclear reactor type VVER-1000/V-320 built by computer code ATHLET-CD
International Nuclear Information System (INIS)
Georgiev, Yoto; Filipov, Kalin; Velev, Vladimir
2014-01-01
A model of nuclear reactor type VVER-1000 V-320 developed for computer code ATHLET-CD2.1A is presented. Validation of the has been made, in the analysis of the station blackout scenario with LOCA on fourth cold leg is shown. As the calculation has been completed, the results are checked through comparison with the results from the computer codes ATHLET-2.1A, ASTEC-2.1 and RELAP5mod3.2
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
Domain-specific modeling enabling full code generation
Kelly, Steven
2007-01-01
Domain-Specific Modeling (DSM) is the latest approach tosoftware development, promising to greatly increase the speed andease of software creation. Early adopters of DSM have been enjoyingproductivity increases of 500–1000% in production for over adecade. This book introduces DSM and offers examples from variousfields to illustrate to experienced developers how DSM can improvesoftware development in their teams. Two authorities in the field explain what DSM is, why it works,and how to successfully create and use a DSM solution to improveproductivity and quality. Divided into four parts, the book covers:background and motivation; fundamentals; in-depth examples; andcreating DSM solutions. There is an emphasis throughout the book onpractical guidelines for implementing DSM, including how toidentify the nece sary language constructs, how to generate fullcode from models, and how to provide tool support for a new DSMlanguage. The example cases described in the book are available thebook's Website, www.dsmbook....
Model-Driven Engineering: Automatic Code Generation and Beyond
2015-03-01
herein to any specific commercial product, process, or service by trade name, trade mark, manufacturer , or otherwise, does not necessarily constitute or...export of an Extensible Markup Language (XML) representation of the model. The XML Metadata Interchange (XMI) is an OMG standard for representing...overall company financial results for the past 3 years. What financial re- sults are you projecting for the next year? 1.2.5.2 Percentage of Gross
Installation of aerosol behavior model into multi-dimensional thermal hydraulic analysis code AQUA
International Nuclear Information System (INIS)
Kisohara, Naoyuki; Yamaguchi, Akira
1997-12-01
The safety analysis of FBR plant system for sodium leak phenomena needs to evaluate the deposition of the aerosol particle to the components in the plant, the chemical reaction of aerosol to humidity in the air and the effect of the combustion heat through aerosol to the structural component. For this purpose, ABC-INTG (Aerosol Behavior in Containment-INTeGrated Version) code has been developed and used until now. This code calculates aerosol behavior in the gas area of uniform temperature and pressure by 1 cell-model. Later, however, more detailed calculation of aerosol behavior requires the installation of aerosol model into multi-cell thermal hydraulic analysis code AQUA. AQUA can calculate the carrier gas flow, temperature and the distribution of the aerosol spatial concentration. On the other hand, ABC-INTG can calculate the generation, deposition to the wall and flower, agglomeration of aerosol particle and figure out the distribution of the aerosol particle size. Thus, the combination of these two codes enables to deal with aerosol model coupling the distribution of the aerosol spatial concentration and that of the aerosol particle size. This report describes aerosol behavior model, how to install the aerosol model to AQUA and new subroutine equipped to the code. Furthermore, the test calculations of the simple structural model were executed by this code, appropriate results were obtained. Thus, this code has prospect to predict aerosol behavior by the introduction of coupling analysis with multi-dimensional gas thermo-dynamics for sodium combustion evaluation. (J.P.N.)
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)
Learning-Based Just-Noticeable-Quantization- Distortion Modeling for Perceptual Video Coding.
Ki, Sehwan; Bae, Sung-Ho; Kim, Munchurl; Ko, Hyunsuk
2018-07-01
Conventional predictive video coding-based approaches are reaching the limit of their potential coding efficiency improvements, because of severely increasing computation complexity. As an alternative approach, perceptual video coding (PVC) has attempted to achieve high coding efficiency by eliminating perceptual redundancy, using just-noticeable-distortion (JND) directed PVC. The previous JNDs were modeled by adding white Gaussian noise or specific signal patterns into the original images, which were not appropriate in finding JND thresholds due to distortion with energy reduction. In this paper, we present a novel discrete cosine transform-based energy-reduced JND model, called ERJND, that is more suitable for JND-based PVC schemes. Then, the proposed ERJND model is extended to two learning-based just-noticeable-quantization-distortion (JNQD) models as preprocessing that can be applied for perceptual video coding. The two JNQD models can automatically adjust JND levels based on given quantization step sizes. One of the two JNQD models, called LR-JNQD, is based on linear regression and determines the model parameter for JNQD based on extracted handcraft features. The other JNQD model is based on a convolution neural network (CNN), called CNN-JNQD. To our best knowledge, our paper is the first approach to automatically adjust JND levels according to quantization step sizes for preprocessing the input to video encoders. In experiments, both the LR-JNQD and CNN-JNQD models were applied to high efficiency video coding (HEVC) and yielded maximum (average) bitrate reductions of 38.51% (10.38%) and 67.88% (24.91%), respectively, with little subjective video quality degradation, compared with the input without preprocessing applied.
International Nuclear Information System (INIS)
Katalin Eged; Zoltan Kis; Natalia Semioschkina; Gabriele Voigt
2004-01-01
One of the objectives of the EC project EVANET-TERRA is to provide suitable inputs to the RODOS system. This study gives an overview on urban dose calculation models with special emphasis on the RECLAIM-EDEM2M and TEMAS-urban codes. The TEMAS-urban code is more complex compared to the RECLAIM-EDEM2M code although both models use similar and some times even same model parameters. The database and the way of its data collection as used in RECLAIM-EDEM2M is recommended as a preferred option because it contains many data from local and regional measurements. However in a decision situation the outputs of the TEMASurban model may better help stake holders by providing a ranking of the surfaces to be decontaminated. (author)
Pre-engineering Spaceflight Validation of Environmental Models and the 2005 HZETRN Simulation Code
Nealy, John E.; Cucinotta, Francis A.; Wilson, John W.; Badavi, Francis F.; Dachev, Ts. P.; Tomov, B. T.; Walker, Steven A.; DeAngelis, Giovanni; Blattnig, Steve R.; Atwell, William
2006-01-01
The HZETRN code has been identified by NASA for engineering design in the next phase of space exploration highlighting a return to the Moon in preparation for a Mars mission. In response, a new series of algorithms beginning with 2005 HZETRN, will be issued by correcting some prior limitations and improving control of propagated errors along with established code verification processes. Code validation processes will use new/improved low Earth orbit (LEO) environmental models with a recently improved International Space Station (ISS) shield model to validate computational models and procedures using measured data aboard ISS. These validated models will provide a basis for flight-testing the designs of future space vehicles and systems of the Constellation program in the LEO environment.
Testing of the PELSHIE shielding code using Benchmark problems and other special shielding models
International Nuclear Information System (INIS)
Language, A.E.; Sartori, D.E.; De Beer, G.P.
1981-08-01
The PELSHIE shielding code for gamma rays from point and extended sources was written in 1971 and a revised version was published in October 1979. At Pelindaba the program is used extensively due to its flexibility and ease of use for a wide range of problems. The testing of PELSHIE results with the results of a range of models and so-called Benchmark problems is desirable to determine possible weaknesses in PELSHIE. Benchmark problems, experimental data, and shielding models, some of which were resolved by the discrete-ordinates method with the ANISN and DOT 3.5 codes, were used for the efficiency test. The description of the models followed the pattern of a classical shielding problem. After the intercomparison with six different models, the usefulness of the PELSHIE code was quantitatively determined [af
The Nuremberg Code subverts human health and safety by requiring animal modeling
Directory of Open Access Journals (Sweden)
Greek Ray
2012-07-01
Full Text Available Abstract Background The requirement that animals be used in research and testing in order to protect humans was formalized in the Nuremberg Code and subsequent national and international laws, codes, and declarations. Discussion We review the history of these requirements and contrast what was known via science about animal models then with what is known now. We further analyze the predictive value of animal models when used as test subjects for human response to drugs and disease. We explore the use of animals for models in toxicity testing as an example of the problem with using animal models. Summary We conclude that the requirements for animal testing found in the Nuremberg Code were based on scientifically outdated principles, compromised by people with a vested interest in animal experimentation, serve no useful function, increase the cost of drug development, and prevent otherwise safe and efficacious drugs and therapies from being implemented.
Modeling of BWR core meltdown accidents - for application in the MELRPI. MOD2 computer code
Energy Technology Data Exchange (ETDEWEB)
Koh, B R; Kim, S H; Taleyarkhan, R P; Podowski, M Z; Lahey, Jr, R T
1985-04-01
This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing.
Modelling Chemical Equilibrium Partitioning with the GEMS-PSI Code
Energy Technology Data Exchange (ETDEWEB)
Kulik, D.; Berner, U.; Curti, E
2004-03-01
Sorption, co-precipitation and re-crystallisation are important retention processes for dissolved contaminants (radionuclides) migrating through the sub-surface. The retention of elements is usually measured by empirical partition coefficients (Kd), which vary in response to many factors: temperature, solid/liquid ratio, total contaminant loading, water composition, host-mineral composition, etc. The Kd values can be predicted for in-situ conditions from thermodynamic modelling of solid solution, aqueous solution or sorption equilibria, provided that stoichiometry, thermodynamic stability and mixing properties of the pure components are known (Example 1). Unknown thermodynamic properties can be retrieved from experimental Kd values using inverse modelling techniques (Example 2). An efficient, advanced tool for performing both tasks is the Gibbs Energy Minimization (GEM) approach, implemented in the user-friendly GEM-Selector (GEMS) program package, which includes the Nagra-PSI chemical thermodynamic database. The package is being further developed at PSI and used extensively in studies relating to nuclear waste disposal. (author)
Modelling Chemical Equilibrium Partitioning with the GEMS-PSI Code
International Nuclear Information System (INIS)
Kulik, D.; Berner, U.; Curti, E.
2004-01-01
Sorption, co-precipitation and re-crystallisation are important retention processes for dissolved contaminants (radionuclides) migrating through the sub-surface. The retention of elements is usually measured by empirical partition coefficients (Kd), which vary in response to many factors: temperature, solid/liquid ratio, total contaminant loading, water composition, host-mineral composition, etc. The Kd values can be predicted for in-situ conditions from thermodynamic modelling of solid solution, aqueous solution or sorption equilibria, provided that stoichiometry, thermodynamic stability and mixing properties of the pure components are known (Example 1). Unknown thermodynamic properties can be retrieved from experimental Kd values using inverse modelling techniques (Example 2). An efficient, advanced tool for performing both tasks is the Gibbs Energy Minimization (GEM) approach, implemented in the user-friendly GEM-Selector (GEMS) program package, which includes the Nagra-PSI chemical thermodynamic database. The package is being further developed at PSI and used extensively in studies relating to nuclear waste disposal. (author)
International Nuclear Information System (INIS)
Zhang, Z.Y.; Hu, E.B.; Meng, X.C.; Zhang, Y.; Yao, R.T.
1993-01-01
According to requirement of accident emergency plan for Qinshan Nuclear Power Plant, a Gaussian straight-line model was adopted for estimating radionuclide concentration in surface air. In addition, the effects of mountain body on atmospheric dispersion was considered. By combination of field atmospheric dispersion experiment and wind tunnel modeling test, necessary modifications have been done for some models and parameters. A computer code for assessment was written in Quick BASIC (V4.5) language. The radius of assessment region is 10 km and the code is applicable to early accident assessment. (1 tab.)
Models and Correlations of Interfacial and Wall Frictions for the SPACE code
International Nuclear Information System (INIS)
Kim, Soo Hyung; Hwang, Moon Kyu; Chung, Bub Dong
2010-04-01
This report describes models and correlations for the interfacial and wall frictions implemented in the SPACE code which has the capability to predict thermal-hydraulic behavior of nuclear power plants. The interfacial and wall frictions are essential to solve the momentum conservation equations of gas, continuous liquid and droplet. The interfacial and wall frictions are dealt in the Chapter 2 and 3, respectively. In Chapter 4, selection criteria for models and correlations are explained. In Chapter 5, the origins of the selected models and correlations used in this code are examined to check whether they are in confliction with intellectual proprietary rights
A new balance-of-plant model for the SASSYS-1 LMR systems analysis code
International Nuclear Information System (INIS)
Briggs, L.L.
1991-01-01
In this paper, a balance-of-plant model is developed for the SASSYS code. This model represents the balance of plant as a network of components. It interfaces with the existing SASSYS code through the water side of the steam generator. The network representation provides a discretization of the mass, momentum, and energy equations and the equation of state and allows a simultaneous solution for the changes in pressure, flow, and enthalpy throughout the waterside system. The model has been tested for several types of transients and been found to perform both accurately and efficiently
An improved UO2 thermal conductivity model in the ELESTRES computer code
International Nuclear Information System (INIS)
Chassie, G.G.; Tochaie, M.; Xu, Z.
2010-01-01
This paper describes the improved UO 2 thermal conductivity model for use in the ELESTRES (ELEment Simulation and sTRESses) computer code. The ELESTRES computer code models the thermal, mechanical and microstructural behaviour of a CANDU® fuel element under normal operating conditions. The main purpose of the code is to calculate fuel temperatures, fission gas release, internal gas pressure, fuel pellet deformation, and fuel sheath strains for fuel element design and assessment. It is also used to provide initial conditions for evaluating fuel behaviour during high temperature transients. The thermal conductivity of UO 2 fuel is one of the key parameters that affect ELESTRES calculations. The existing ELESTRES thermal conductivity model has been assessed and improved based on a large amount of thermal conductivity data from measurements of irradiated and un-irradiated UO 2 fuel with different densities. The UO 2 thermal conductivity data cover 90% to 99% theoretical density of UO 2 , temperature up to 3027 K, and burnup up to 1224 MW·h/kg U. The improved thermal conductivity model, which is recommended for a full implementation in the ELESTRES computer code, has reduced the ELESTRES code prediction biases of temperature, fission gas release, and fuel sheath strains when compared with the available experimental data. This improved thermal conductivity model has also been checked with a test version of ELESTRES over the full ranges of fuel temperature, fuel burnup, and fuel density expected in CANDU fuel. (author)
Two-phase wall friction model for the trace computer code
International Nuclear Information System (INIS)
Wang Weidong
2005-01-01
The wall drag model in the TRAC/RELAP5 Advanced Computational Engine computer code (TRACE) has certain known deficiencies. For example, in an annular flow regime, the code predicts an unphysical high liquid velocity compared to the experimental data. To address those deficiencies, a new wall frictional drag package has been developed and implemented in the TRACE code to model the wall drag for two-phase flow system code. The modeled flow regimes are (1) annular/mist, (2) bubbly/slug, and (3) bubbly/slug with wall nucleation. The new models use void fraction (instead of flow quality) as the correlating variable to minimize the calculation oscillation. In addition, the models allow for transitions between the three regimes. The annular/mist regime is subdivided into three separate regimes for pure annular flow, annular flow with entrainment, and film breakdown. For adiabatic two-phase bubbly/slug flows, the vapor phase primarily exists outside of the boundary layer, and the wall shear uses single-phase liquid velocity for friction calculation. The vapor phase wall friction drag is set to zero for bubbly/slug flows. For bubbly/slug flows with wall nucleation, the bubbles are presented within the hydrodynamic boundary layer, and the two-phase wall friction drag is significantly higher with a pronounced mass flux effect. An empirical correlation has been studied and applied to account for nucleate boiling. Verification and validation tests have been performed, and the test results showed a significant code improvement. (authors)
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
Evaluation of Advanced Models for PAFS Condensation Heat Transfer in SPACE Code
Energy Technology Data Exchange (ETDEWEB)
Bae, Byoung-Uhn; Kim, Seok; Park, Yu-Sun; Kang, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahn, Tae-Hwan; Yun, Byong-Jo [Pusan National University, Busan (Korea, Republic of)
2015-10-15
The PAFS (Passive Auxiliary Feedwater System) is operated by the natural circulation to remove the core decay heat through the PCHX (Passive Condensation Heat Exchanger) which is composed of the nearly horizontal tubes. For validation of the cooling and operational performance of the PAFS, PASCAL (PAFS Condensing Heat Removal Assessment Loop) facility was constructed and the condensation heat transfer and natural convection phenomena in the PAFS was experimentally investigated at KAERI (Korea Atomic Energy Research Institute). From the PASCAL experimental result, it was found that conventional system analysis code underestimated the condensation heat transfer. In this study, advanced condensation heat transfer models which can treat the heat transfer mechanisms with the different flow regimes in the nearly horizontal heat exchanger tube were analyzed. The models were implemented in a thermal hydraulic safety analysis code, SPACE (Safety and Performance Analysis Code for Nuclear Power Plant), and it was evaluated with the PASCAL experimental data. With an aim of enhancing the prediction capability for the condensation phenomenon inside the PCHX tube of the PAFS, advanced models for the condensation heat transfer were implemented into the wall condensation model of the SPACE code, so that the PASCAL experimental result was utilized to validate the condensation models. Calculation results showed that the improved model for the condensation heat transfer coefficient enhanced the prediction capability of the SPACE code. This result confirms that the mechanistic modeling for the film condensation in the steam phase and the convection in the condensate liquid contributed to enhance the prediction capability of the wall condensation model of the SPACE code and reduce conservatism in prediction of condensation heat transfer.
Code Generation by Model Transformation : A Case Study in Transformation Modularity
Hemel, Z.; Kats, L.C.L.; Visser, E.
2008-01-01
Preprint of paper published in: Theory and Practice of Model Transformations (ICMT 2008), Lecture Notes in Computer Science 5063; doi:10.1007/978-3-540-69927-9_13 The realization of model-driven software development requires effective techniques for implementing code generators for domain-specific
Fine-Grained Energy Modeling for the Source Code of a Mobile Application
DEFF Research Database (Denmark)
Li, Xueliang; Gallagher, John Patrick
2016-01-01
The goal of an energy model for source code is to lay a foundation for the application of energy-aware programming techniques. State of the art solutions are based on source-line energy information. In this paper, we present an approach to constructing a fine-grained energy model which is able...
Improvements to the nuclear model code GNASH for cross section calculations at higher energies
International Nuclear Information System (INIS)
Young, P.G.; Chadwick, M.B.
1994-01-01
The nuclear model code GNASH, which in the past has been used predominantly for incident particle energies below 20 MeV, has been modified extensively for calculations at higher energies. The model extensions and improvements are described in this paper, and their significance is illustrated by comparing calculations with experimental data for incident energies up to 160 MeV
Time-domain modeling of electromagnetic diffusion with a frequency-domain code
Mulder, W.A.; Wirianto, M.; Slob, E.C.
2007-01-01
We modeled time-domain EM measurements of induction currents for marine and land applications with a frequency-domain code. An analysis of the computational complexity of a number of numerical methods shows that frequency-domain modeling followed by a Fourier transform is an attractive choice if a
OWL: A code for the two-center shell model with spherical Woods-Saxon potentials
Diaz-Torres, Alexis
2018-03-01
A Fortran-90 code for solving the two-center nuclear shell model problem is presented. The model is based on two spherical Woods-Saxon potentials and the potential separable expansion method. It describes the single-particle motion in low-energy nuclear collisions, and is useful for characterizing a broad range of phenomena from fusion to nuclear molecular structures.
The modelling of wall condensation with noncondensable gases for the containment codes
Energy Technology Data Exchange (ETDEWEB)
Leduc, C.; Coste, P.; Barthel, V.; Deslandes, H. [Commissariat a l`Energi Atomique, Grenoble (France)
1995-09-01
This paper presents several approaches in the modelling of wall condensation in the presence of noncondensable gases for containment codes. The lumped-parameter modelling and the local modelling by 3-D codes are discussed. Containment analysis codes should be able to predict the spatial distributions of steam, air, and hydrogen as well as the efficiency of cooling by wall condensation in both natural convection and forced convection situations. 3-D calculations with a turbulent diffusion modelling are necessary since the diffusion controls the local condensation whereas the wall condensation may redistribute the air and hydrogen mass in the containment. A fine mesh modelling of film condensation in forced convection has been in the developed taking into account the influence of the suction velocity at the liquid-gas interface. It is associated with the 3-D model of the TRIO code for the gas mixture where a k-{xi} turbulence model is used. The predictions are compared to the Huhtiniemi`s experimental data. The modelling of condensation in natural convection or mixed convection is more complex. As no universal velocity and temperature profile exist for such boundary layers, a very fine nodalization is necessary. More simple models integrate equations over the boundary layer thickness, using the heat and mass transfer analogy. The model predictions are compared with a MIT experiment. For the containment compartments a two node model is proposed using the lumped parameter approach. Heat and mass transfer coefficients are tested on separate effect tests and containment experiments. The CATHARE code has been adapted to perform such calculations and shows a reasonable agreement with data.
Modelling Brazilian tests with FRACOD2D (FRActure propagation CODe)
International Nuclear Information System (INIS)
Lanaro, Flavio; Sato, Toshinori; Rinne, Mikael; Stephansson, Ove
2008-01-01
This study focuses on the influence of initiated cracks on the stress distribution within rock samples subjected to tensile loading by traditional Brazilian testing. The numerical analyses show that the stress distribution is only marginally affected by the considered loading boundary conditions. On the other hand, the initiation and propagation of cracks produce a stress field that is very different from that assumed by considering the rock material as continuous, homogeneous, isotropic and elastic. In the models, stress concentrations at the bridges between the cracks were found to have tensile stresses much higher than the macroscopic direct tensile strength of the intact rock. This was possible thanks to the development of large stress gradients that can be carried by the rock between the cracks. The analysis of the deformation along the sample diameter perpendicular to the loading direction might enable one to determine the macroscopic direct tensile strength of the rock or, in a real case, of the weakest grains. The strength is indicated by the point where the stress-strain curves depart from linearity. (author)
A review of MAAP4 code structure and core T/H model
International Nuclear Information System (INIS)
Song, Yong Mann; Park, Soo Yong
1998-03-01
The modular accident analysis program (MAAP) version 4 is a computer code that can simulate the response of LWR plants during severe accident sequences and includes models for all of the important phenomena which might occur during accident sequences. In this report, MAAP4 code structure and core thermal hydraulic (T/H) model which models the T/H behavior of the reactor core and the response of core components during all accident phases involving degraded cores are specifically reviewed and then reorganized. This reorganization is performed via getting the related models together under each topic whose contents and order are same with other two reports for MELCOR and SCDAP/RELAP5 to be simultaneously published. Major purpose of the report is to provide information about the characteristics of MAAP4 core T/H models for an integrated severe accident computer code development being performed under the one of on-going mid/long-term nuclear developing project. The basic characteristics of the new integrated severe accident code includes: 1) Flexible simulation capability of primary side, secondary side, and the containment under severe accident conditions, 2) Detailed plant simulation, 3) Convenient user-interfaces, 4) Highly modularization for easy maintenance/improvement, and 5) State-of-the-art model selection. In conclusion, MAAP4 code has appeared to be superior for 3) and 4) items but to be somewhat inferior for 1) and 2) items. For item 5), more efforts should be made in the future to compare separated models in detail with not only other codes but also recent world-wide work. (author). 17 refs., 1 tab., 12 figs
Thermal-hydraulic and aerosol containment phenomena modelling in ASTEC severe accident computer code
International Nuclear Information System (INIS)
Kljenak, Ivo; Dapper, Maik; Dienstbier, Jiri; Herranz, Luis E.; Koch, Marco K.; Fontanet, Joan
2010-01-01
Transients in containment systems of different scales (Phebus.FP containment, KAEVER vessel, Battelle Model Containment, LACE vessel and VVER-1000 nuclear power plant containment) involving thermal-hydraulic phenomena and aerosol behaviour, were simulated with the computer integral code ASTEC. The results of the simulations in the first four facilities were compared with experimental results, whereas the results of the simulated accident in the VVER-1000 containment were compared to results obtained with the MELCOR code. The main purpose of the simulations was the validation of the CPA module of the ASTEC code. The calculated results support the applicability of the code for predicting in-containment thermal-hydraulic and aerosol phenomena during a severe accident in a nuclear power plant.
Gap conductance model validation in the TASS/SMR-S code
International Nuclear Information System (INIS)
Ahn, Sang-Jun; Yang, Soo-Hyung; Chung, Young-Jong; Bae, Kyoo-Hwan; Lee, Won-Jae
2011-01-01
An advanced integral pressurized water reactor, SMART (System-Integrated Modular Advanced ReacTor) has been developed by KAERI (Korea Atomic Energy Research and Institute). The purposes of the SMART are sea water desalination and an electricity generation. For the safety evaluation and performance analysis of the SMART, TASS/SMR-S (Transient And Setpoint Simulation/System-integrated Modular Reactor) code, has been developed. In this paper, the gap conductance model for the calculation of gap conductance has been validated by using another system code, MARS code, and experimental results. In the validation, the behaviors of fuel temperature and gap width are selected as the major parameters. According to the evaluation results, the TASS/SMR-S code predicts well the behaviors of fuel temperatures and gap width variation, compared to the MARS calculation results and experimental data. (author)
The implementation of a toroidal limiter model into the gyrokinetic code ELMFIRE
Energy Technology Data Exchange (ETDEWEB)
Leerink, S.; Janhunen, S.J.; Kiviniemi, T.P.; Nora, M. [Euratom-Tekes Association, Helsinki University of Technology (Finland); Heikkinen, J.A. [Euratom-Tekes Association, VTT, P.O. Box 1000, FI-02044 VTT (Finland); Ogando, F. [Universidad Nacional de Educacion a Distancia, Madrid (Spain)
2008-03-15
The ELMFIRE full nonlinear gyrokinetic simulation code has been developed for calculations of plasma evolution and dynamics of turbulence in tokamak geometry. The code is applicable for calculations of strong perturbations in particle distribution function, rapid transients and steep gradients in plasma. Benchmarking against experimental reflectometry data from the FT2 tokamak is being discussed and in this paper a model for comparison and studying poloidal velocity is presented. To make the ELMFIRE code suitable for scrape-off layer simulations a simplified toroidal limiter model has been implemented. The model is be discussed and first results are presented. (copyright 2008 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)
Implementation of a dry process fuel cycle model into the DYMOND code
International Nuclear Information System (INIS)
Park, Joo Hwan; Jeong, Chang Joon; Choi, Hang Bok
2004-01-01
For the analysis of a dry process fuel cycle, new modules were implemented into the fuel cycle analysis code DYMOND, which was developed by the Argonne National Laboratory. The modifications were made to the energy demand prediction model, a Canada Deuterium Uranium (CANDU) reactor, direct use of spent Pressurized Water Reactor (PWR) fuel in CANDU reactors (DUPIC) fuel cycle model, the fuel cycle calculation module, and the input/output modules. The performance of the modified DYMOND code was assessed for the postulated once-through fuel cycle models including both the PWR and CANDU reactor. This paper presents modifications of the DYMOND code and the results of sample calculations for the PWR once-through and DUPIC fuel cycles
Inclusion of models to describe severe accident conditions in the fuel simulation code DIONISIO
Energy Technology Data Exchange (ETDEWEB)
Lemes, Martín; Soba, Alejandro [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Daverio, Hernando [Gerencia Reactores y Centrales Nucleares, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Denis, Alicia [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina)
2017-04-15
The simulation of fuel rod behavior is a complex task that demands not only accurate models to describe the numerous phenomena occurring in the pellet, cladding and internal rod atmosphere but also an adequate interconnection between them. In the last years several models have been incorporated to the DIONISIO code with the purpose of increasing its precision and reliability. After the regrettable events at Fukushima, the need for codes capable of simulating nuclear fuels under accident conditions has come forth. Heat removal occurs in a quite different way than during normal operation and this fact determines a completely new set of conditions for the fuel materials. A detailed description of the different regimes the coolant may exhibit in such a wide variety of scenarios requires a thermal-hydraulic formulation not suitable to be included in a fuel performance code. Moreover, there exist a number of reliable and famous codes that perform this task. Nevertheless, and keeping in mind the purpose of building a code focused on the fuel behavior, a subroutine was developed for the DIONISIO code that performs a simplified analysis of the coolant in a PWR, restricted to the more representative situations and provides to the fuel simulation the boundary conditions necessary to reproduce accidental situations. In the present work this subroutine is described and the results of different comparisons with experimental data and with thermal-hydraulic codes are offered. It is verified that, in spite of its comparative simplicity, the predictions of this module of DIONISIO do not differ significantly from those of the specific, complex codes.
International Nuclear Information System (INIS)
Hiergesell, R.; Taylor, G.
2010-01-01
An investigation was conducted to compare and evaluate contaminant transport results of two model codes, GoldSim and Porflow, using a simple 1-D string of elements in each code. Model domains were constructed to be identical with respect to cell numbers and dimensions, matrix material, flow boundary and saturation conditions. One of the codes, GoldSim, does not simulate advective movement of water; therefore the water flux term was specified as a boundary condition. In the other code, Porflow, a steady-state flow field was computed and contaminant transport was simulated within that flow-field. The comparisons were made solely in terms of the ability of each code to perform contaminant transport. The purpose of the investigation was to establish a basis for, and to validate follow-on work that was conducted in which a 1-D GoldSim model developed by abstracting information from Porflow 2-D and 3-D unsaturated and saturated zone models and then benchmarked to produce equivalent contaminant transport results. A handful of contaminants were selected for the code-to-code comparison simulations, including a non-sorbing tracer and several long- and short-lived radionuclides exhibiting both non-sorbing to strongly-sorbing characteristics with respect to the matrix material, including several requiring the simulation of in-growth of daughter radionuclides. The same diffusion and partitioning coefficients associated with each contaminant and the half-lives associated with each radionuclide were incorporated into each model. A string of 10-elements, having identical spatial dimensions and properties, were constructed within each code. GoldSim's basic contaminant transport elements, Mixing cells, were utilized in this construction. Sand was established as the matrix material and was assigned identical properties (e.g. bulk density, porosity, saturated hydraulic conductivity) in both codes. Boundary conditions applied included an influx of water at the rate of 40 cm/yr at one
Aquelarre. A computer code for fast neutron cross sections from the statistical model
International Nuclear Information System (INIS)
Guasp, J.
1974-01-01
A Fortran V computer code for Univac 1108/6 using the partial statistical (or compound nucleus) model is described. The code calculates fast neutron cross sections for the (n, n'), (n, p), (n, d) and (n, α reactions and the angular distributions and Legendre moments.for the (n, n) and (n, n') processes in heavy and intermediate spherical nuclei. A local Optical Model with spin-orbit interaction for each level is employed, allowing for the width fluctuation and Moldauer corrections, as well as the inclusion of discrete and continuous levels. (Author) 67 refs
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
Development of the Monju core safety analysis numerical models by super-COPD code
International Nuclear Information System (INIS)
Yamada, Fumiaki; Minami, Masaki
2010-12-01
Japan Atomic Energy Agency constructed a computational model for safety analysis of Monju reactor core to be built into a modularized plant dynamics analysis code Super-COPD code, for the purpose of heat removal capability evaluation at the in total 21 defined transients in the annex to the construction permit application. The applicability of this model to core heat removal capability evaluation has been estimated by back to back result comparisons of the constituent models with conventionally applied codes and by application of the unified model. The numerical model for core safety analysis has been built based on the best estimate model validated by the actually measured plant behavior up to 40% rated power conditions, taking over safety analysis models of conventionally applied COPD and HARHO-IN codes, to be capable of overall calculations of the entire plant with the safety protection and control systems. Among the constituents of the analytical model, neutronic-thermal model, heat transfer and hydraulic models of PHTS, SHTS, and water/steam system are individually verified by comparisons with the conventional calculations. Comparisons are also made with the actually measured plant behavior up to 40% rated power conditions to confirm the calculation adequacy and conservativeness of the input data. The unified analytical model was applied to analyses of in total 8 anomaly events; reactivity insertion, abnormal power distribution, decrease and increase of coolant flow rate in PHTS, SHTS and water/steam systems. The resulting maximum values and temporal variations of the key parameters in safety evaluation; temperatures of fuel, cladding, in core sodium coolant and RV inlet and outlet coolant have negligible discrepancies against the existing analysis result in the annex to the construction permit application, verifying the unified analytical model. These works have enabled analytical evaluation of Monju core heat removal capability by Super-COPD utilizing the
Off-take Model of the SPACE Code and Its Validation
International Nuclear Information System (INIS)
Oh, Myung Taek; Park, Chan Eok; Sohn, Jong Joo
2011-01-01
Liquid entrainment and vapor pull-through models of horizontal pipe have been implemented in the SPACE code. The model of SPACE accounts for the phase separation phenomena and computes the flux of mass and energy through an off-take attached to a horizontal pipe when stratified conditions occur in the horizontal pipe. This model is referred to as the off-take model. The importance of predicting the fluid conditions through an off-take in a small-break LOCA has been well known. In this case, the occurrence of the stratification can affect the break node void fraction and thus the break flow discharged from the primary system. In order to validate the off-take model newly developed for the SPACE code, a simulation of the HDU experiments has been performed. The main feature of the off-take model and its application results will be presented in this paper
A Perceptual Model for Sinusoidal Audio Coding Based on Spectral Integration
Directory of Open Access Journals (Sweden)
Jensen Søren Holdt
2005-01-01
Full Text Available 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 audio signals. In this paper, we present a new perceptual model that predicts masked thresholds for sinusoidal distortions. The model relies on signal detection theory and incorporates more recent insights about spectral and temporal integration in auditory masking. As a consequence, the model is able to predict the distortion detectability. In fact, the distortion detectability defines a (perceptually relevant norm on the underlying signal space which is beneficial for optimisation algorithms such as rate-distortion optimisation or linear predictive coding. We evaluate the merits of the model by combining it with a sinusoidal extraction method and compare the results with those obtained with the ISO MPEG-1 Layer I-II recommended model. Listening tests show a clear preference for the new model. More specifically, the model presented here leads to a reduction of more than 20% in terms of number of sinusoids needed to represent signals at a given quality level.
Basic Pilot Code Development for Two-Fluid, Three-Field Model
International Nuclear Information System (INIS)
Jeong, Jae Jun; Bae, S. W.; Lee, Y. J.; Chung, B. D.; Hwang, M.; Ha, K. S.; Kang, D. H.
2006-03-01
A basic pilot code for one-dimensional, transient, two-fluid, three-field model has been developed. Using 9 conceptual problems, the basic pilot code has been verified. The results of the verification are summarized below: - It was confirmed that the basic pilot code can simulate various flow conditions (such as single-phase liquid flow, bubbly flow, slug/churn turbulent flow, annular-mist flow, and single-phase vapor flow) and transitions of the flow conditions. A mist flow was not simulated, but it seems that the basic pilot code can simulate mist flow conditions. - The pilot code was programmed so that the source terms of the governing equations and numerical solution schemes can be easily tested. - The mass and energy conservation was confirmed for single-phase liquid and single-phase vapor flows. - It was confirmed that the inlet pressure and velocity boundary conditions work properly. - It was confirmed that, for single- and two-phase flows, the velocity and temperature of non-existing phase are calculated as intended. - During the simulation of a two-phase flow, the calculation reaches a quasisteady state with small-amplitude oscillations. The oscillations seem to be induced by some numerical causes. The research items for the improvement of the basic pilot code are listed in the last section of this report
Basic Pilot Code Development for Two-Fluid, Three-Field Model
Energy Technology Data Exchange (ETDEWEB)
Jeong, Jae Jun; Bae, S. W.; Lee, Y. J.; Chung, B. D.; Hwang, M.; Ha, K. S.; Kang, D. H
2006-03-15
A basic pilot code for one-dimensional, transient, two-fluid, three-field model has been developed. Using 9 conceptual problems, the basic pilot code has been verified. The results of the verification are summarized below: - It was confirmed that the basic pilot code can simulate various flow conditions (such as single-phase liquid flow, bubbly flow, slug/churn turbulent flow, annular-mist flow, and single-phase vapor flow) and transitions of the flow conditions. A mist flow was not simulated, but it seems that the basic pilot code can simulate mist flow conditions. - The pilot code was programmed so that the source terms of the governing equations and numerical solution schemes can be easily tested. - The mass and energy conservation was confirmed for single-phase liquid and single-phase vapor flows. - It was confirmed that the inlet pressure and velocity boundary conditions work properly. - It was confirmed that, for single- and two-phase flows, the velocity and temperature of non-existing phase are calculated as intended. - During the simulation of a two-phase flow, the calculation reaches a quasisteady state with small-amplitude oscillations. The oscillations seem to be induced by some numerical causes. The research items for the improvement of the basic pilot code are listed in the last section of this report.
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)
Energy Technology Data Exchange (ETDEWEB)
Burke, G.J.
1988-04-08
Computer modeling of antennas, since its start in the late 1960's, has become a powerful and widely used tool for antenna design. Computer codes have been developed based on the Method-of-Moments, Geometrical Theory of Diffraction, or integration of Maxwell's equations. Of such tools, the Numerical Electromagnetics Code-Method of Moments (NEC) has become one of the most widely used codes for modeling resonant sized antennas. There are several reasons for this including the systematic updating and extension of its capabilities, extensive user-oriented documentation and accessibility of its developers for user assistance. The result is that there are estimated to be several hundred users of various versions of NEC world wide. 23 refs., 10 figs.
Implementation of a structural dependent model for the superalloy IN738LC in ABAQUS-code
International Nuclear Information System (INIS)
Wolters, J.; Betten, J.; Penkalla, H.J.
1994-05-01
Superalloys, mainly consisting of nickel, are used for applications in aerospace as well as in stationary gas turbines. In the temperature range above 800 C the blades, which are manufactured of these superalloys, are subjected to high centrifugal forces and thermal induced loads. For computer based analysis of the thermo-mechanical behaviour of the blades models for the stress-strain behaviour are necessary. These models have to give a reliable description of the stress-strain behaviour, with emphasis on inelastic affects. The implementation of the model in finite element codes requires a numerical treatment of the constitutive equations with respect to the given interface of the used code. In this paper constitutive equations for the superalloy IN738LC are presented and the implementation in the finite element code ABAQUS with the numerical preparation of the model is described. In order to validate the model calculations were performed for simple uniaxial loading conditions as well as for a complete cross section of a turbine blade under combined thermal and mechanical loading. The achieved results were compared with those of additional calculations by using ABAQUS, including Norton's law, which was already implemented in this code. (orig.) [de
Energy Technology Data Exchange (ETDEWEB)
Jeon, Seong-Su [Department of Engineering Project, FNC Technology Co., Ltd., Bldg. 135-308, Seoul National University, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Department of Nuclear Engineering, Seoul National University, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Hong, Soon-Joon, E-mail: sjhong90@fnctech.com [Department of Engineering Project, FNC Technology Co., Ltd., Bldg. 135-308, Seoul National University, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Park, Ju-Yeop; Seul, Kwang-Won [Korea Institute of Nuclear Safety, 19 Kuseong-dong, Yuseong-gu, Daejon (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University, Gwanak-gu, Seoul 151-744 (Korea, Republic of)
2013-01-15
Highlights: Black-Right-Pointing-Pointer This study collected 11 horizontal in-tube condensation models for stratified flow. Black-Right-Pointing-Pointer This study assessed the predictive capability of the models for steam condensation. Black-Right-Pointing-Pointer Purdue-PCCS experiments were simulated using MARS code incorporated with models. Black-Right-Pointing-Pointer Cavallini et al. (2006) model predicts well the data for stratified flow condition. Black-Right-Pointing-Pointer Results of this study can be used to improve condensation model in RELAP5 or MARS. - Abstract: The accurate prediction of the horizontal in-tube condensation heat transfer is a primary concern in the optimum design and safety analysis of horizontal heat exchangers of passive safety systems such as the passive containment cooling system (PCCS), the emergency condenser system (ECS) and the passive auxiliary feed-water system (PAFS). It is essential to analyze and assess the predictive capability of the previous horizontal in-tube condensation models for each flow regime using various experimental data. This study assessed totally 11 condensation models for the stratified flow, one of the main flow regime encountered in the horizontal condenser, with the heat transfer data from the Purdue-PCCS experiment using the multi-dimensional analysis of reactor safety (MARS) code. From the assessments, it was found that the models by Akers and Rosson, Chato, Tandon et al., Sweeney and Chato, and Cavallini et al. (2002) under-predicted the data in the main condensation heat transfer region, on the contrary to this, the models by Rosson and Meyers, Jaster and Kosky, Fujii, Dobson and Chato, and Thome et al. similarly- or over-predicted the data, and especially, Cavallini et al. (2006) model shows good predictive capability for all test conditions. The results of this study can be used importantly to improve the condensation models in thermal hydraulic code, such as RELAP5 or MARS code.
Energy Technology Data Exchange (ETDEWEB)
Yoon, C.; Rhee, B. W.; Chung, B. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahn, S. H.; Kim, M. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2009-05-15
Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to a lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use a vendor's code for a regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed the RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of the existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of the RELAP5/MOD3/CANDU code to the MARS code including a quality assurance of the developed models.
International Nuclear Information System (INIS)
Yoon, C.; Rhee, B. W.; Chung, B. D.; Ahn, S. H.; Kim, M. W.
2009-01-01
Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to a lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use a vendor's code for a regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed the RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of the existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of the RELAP5/MOD3/CANDU code to the MARS code including a quality assurance of the developed models
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
Overlaid Alice: a statistical model computer code including fission and preequilibrium models
International Nuclear Information System (INIS)
Blann, M.
1976-01-01
The most recent edition of an evaporation code originally written previously with frequent updating and improvement. This version replaces the version Alice described previously. A brief summary is given of the types of calculations which can be done. A listing of the code and the results of several sample calculations are presented
A study on the dependency between turbulent models and mesh configurations of CFD codes
International Nuclear Information System (INIS)
Bang, Jungjin; Heo, Yujin; Jerng, Dong-Wook
2015-01-01
This paper focuses on the analysis of the behavior of hydrogen mixing and hydrogen stratification, using the GOTHIC code and the CFD code. Specifically, we examined the mesh sensitivity and how the turbulence model affects hydrogen stratification or hydrogen mixing, depending on the mesh configuration. In this work, sensitivity analyses for the meshes and the turbulence models were conducted for missing and stratification phenomena. During severe accidents in a nuclear power plants, the generation of hydrogen may occur and this will complicate the atmospheric condition of the containment by causing stratification of air, steam, and hydrogen. This could significantly impact containment integrity analyses, as hydrogen could be accumulated in local region. From this need arises the importance of research about stratification of gases in the containment. Two computation fluid dynamics code, i.e. GOTHIC and STAR-CCM+ were adopted and the computational results were benchmarked against the experimental data from PANDA facility. The main findings observed through the present work can be summarized as follows: 1) In the case of the GOTHIC code, it was observed that the aspect ratio of the mesh was found more important than the mesh size. Also, if the number of the mesh is over 3,000, the effects of the turbulence models were marginal. 2) For STAR-CCM+, the tendency is quite different from the GOTHIC code. That is, the effects of the turbulence models were small for fewer number of the mesh, however, as the number of mesh increases, the effects of the turbulence models becomes significant. Another observation is that away from the injection orifice, the role of the turbulence models tended to be important due to the nature of mixing process and inducted jet stream
A study on the dependency between turbulent models and mesh configurations of CFD codes
Energy Technology Data Exchange (ETDEWEB)
Bang, Jungjin; Heo, Yujin; Jerng, Dong-Wook [CAU, Seoul (Korea, Republic of)
2015-10-15
This paper focuses on the analysis of the behavior of hydrogen mixing and hydrogen stratification, using the GOTHIC code and the CFD code. Specifically, we examined the mesh sensitivity and how the turbulence model affects hydrogen stratification or hydrogen mixing, depending on the mesh configuration. In this work, sensitivity analyses for the meshes and the turbulence models were conducted for missing and stratification phenomena. During severe accidents in a nuclear power plants, the generation of hydrogen may occur and this will complicate the atmospheric condition of the containment by causing stratification of air, steam, and hydrogen. This could significantly impact containment integrity analyses, as hydrogen could be accumulated in local region. From this need arises the importance of research about stratification of gases in the containment. Two computation fluid dynamics code, i.e. GOTHIC and STAR-CCM+ were adopted and the computational results were benchmarked against the experimental data from PANDA facility. The main findings observed through the present work can be summarized as follows: 1) In the case of the GOTHIC code, it was observed that the aspect ratio of the mesh was found more important than the mesh size. Also, if the number of the mesh is over 3,000, the effects of the turbulence models were marginal. 2) For STAR-CCM+, the tendency is quite different from the GOTHIC code. That is, the effects of the turbulence models were small for fewer number of the mesh, however, as the number of mesh increases, the effects of the turbulence models becomes significant. Another observation is that away from the injection orifice, the role of the turbulence models tended to be important due to the nature of mixing process and inducted jet stream.
INDOSE V2.1.1, Internal Dosimetry Code Using Biokinetics Models
International Nuclear Information System (INIS)
Silverman, Ido
2002-01-01
A - Description of program or function: InDose is an internal dosimetry code developed to enable dose estimations using the new biokinetic models (presented in ICRP-56 to ICRP71) as well as the old ones. The code is written in FORTRAN90 and uses the ICRP-66 respiratory tract model and the ICRP-30 gastrointestinal tract model as well as the new and old biokinetic models. The code has been written in such a way that the user is able to change any of the parameters of any one of the models without recompiling the code. All the parameters are given in well annotated parameters files that the user may change and the code reads during invocation. As default, these files contains the values listed in ICRP publications. The full InDose code is planed to have three parts: 1) the main part includes the uptake and systemic models and is used to calculate the activities in the body tissues and excretion as a function of time for a given intake. 2) An optimization module for automatic estimation of the intake for a specific exposure case. 3) A module to calculate the dose due to the estimated intake. Currently, the code is able to perform only its main task (part 1) while the other two have to be done externally using other tools. In the future we would like to add these modules in order to provide a complete solution for the people in the laboratory. The code has been tested extensively to verify the accuracy of its results. The verification procedure was divided into three parts: 1) verification of the implementation of each model, 2) verification of the integrity of the whole code, and 3) usability test. The first two parts consisted of comparing results obtained with InDose to published results for the same cases. For example ICRP-78 monitoring data. The last part consisted of participating in the 3. EIE-IDA and assessing some of the scenarios provided in this exercise. These tests where presented in a few publications. It has been found that there is very good agreement
Code-switched English pronunciation modeling for Swahili spoken term detection
CSIR Research Space (South Africa)
Kleynhans, N
2016-05-01
Full Text Available Computer Science 81 ( 2016 ) 128 – 135 5th Workshop on Spoken Language Technology for Under-resourced Languages, SLTU 2016, 9-12 May 2016, Yogyakarta, Indonesia Code-switched English Pronunciation Modeling for Swahili Spoken Term Detection Neil...
A model code on co-determination and CSR : The Netherlands: A bottom-up approach
Lambooy, T.E.
2011-01-01
This article discusses the works council’s role in the determination of a company’s CSR strategy and the implementation thereof throughout the organisation. The association of the works councils of multinational companies with a base in the Netherlands has recently developed a ‘Model Code on
Reliability in the performance-based concept of fib Model Code 2010
Bigaj-van Vliet, A.; Vrouwenvelder, T.
2013-01-01
The design philosophy of the new fib Model Code for Concrete Structures 2010 represents the state of the art with regard to performance-based approach to the design and assessment of concrete structures. Given the random nature of quantities determining structural behaviour, the assessment of
Monte Carlo simulation of Ising models by multispin coding on a vector computer
Wansleben, Stephan; Zabolitzky, John G.; Kalle, Claus
1984-11-01
Rebbi's efficient multispin coding algorithm for Ising models is combined with the use of the vector computer CDC Cyber 205. A speed of 21.2 million updates per second is reached. This is comparable to that obtained by special- purpose computers.
Turbomachinery Heat Transfer and Loss Modeling for 3D Navier-Stokes Codes
DeWitt, Kenneth; Ameri, Ali
2005-01-01
This report's contents focus on making use of NASA Glenn on-site computational facilities,to develop, validate, and apply models for use in advanced 3D Navier-Stokes Computational Fluid Dynamics (CFD) codes to enhance the capability to compute heat transfer and losses in turbomachiney.
Developing a universal model of reading necessitates cracking the orthographic code.
Davis, Colin J
2012-10-01
I argue, contra Frost, that when prime lexicality and target density are considered, it is not clear that there are fundamental differences between form priming effects in Semitic and European languages. Furthermore, identifying and naming printed words in these languages raises common theoretical problems. Solving these problems and developing a universal model of reading necessitates "cracking" the orthographic input code.
Wang, Yanqing; Li, Hang; Feng, Yuqiang; Jiang, Yu; Liu, Ying
2012-01-01
The traditional assessment approach, in which one single written examination counts toward a student's total score, no longer meets new demands of programming language education. Based on a peer code review process model, we developed an online assessment system called "EduPCR" and used a novel approach to assess the learning of computer…
A mathematical model, and code LIXY, for leaching of radionuclides from containment
International Nuclear Information System (INIS)
Fraser, J.L.; Jarvis, R.G.
1985-06-01
A mathematical model has been developed to describe the leaching of a radionuclide from an inner region into an outer region, by diffusion processes. Answers have been obtained for the whole range of time values, and have been written into a code LIXY, to calculate the concentration of nuclide at the outer face of the outer region
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
Institute of Scientific and Technical Information of China (English)
呼义翔; 雷天时; 吴撼宇; 郭宁; 韩娟娟; 邱爱慈; 王亮平; 黄涛; 丛培天; 张信军; 李岩; 曾正中; 孙铁平
2011-01-01
The transmission-line-circuit model of the Z accelerator, developed originally by W. A. STYGAR, P. A. CORCORAN, et al., is revised. The revised model uses different calculations for the electron loss and flow impedance in the magnetically insulated transmission line system of the Z accelerator before and after magnetic insulation is established. By including electron pressure and zero electric field at the cathode, a closed set of equations is obtained at each time step, and dynamic shunt resistance （used to represent any electron loss to the anode） and flow impedance are solved, which have been incorporated into the transmission line code for simulations of the vacuum section in the Z accelerator. Finally, the results are discussed in comparison with earlier findings to show the effectiveness and limitations of the model.
A simplified model of Passive Containment Cooling System in a CFD code
International Nuclear Information System (INIS)
Jiang, X.W.; Studer, E.; Kudriakov, S.
2013-01-01
Highlights: ► We have built a condensing model using Navier–Stokes equations in CAST3M code. ► We have done a benchmark work on the condensing model using the COPAIN tests data. ► We have built an evaporating model according to Aiello's model in CAST3M code. ► We used Kang and Park's film evaporation tests data to validate the model. ► An integrated model was derived by coupling two individual models with a steel plate. -- Abstract: In this paper, we built up a simplified model of the Passive Containment Cooling System in a CFD code, including a steel plate, a condensing channel and an evaporating channel. In the inner side of the plate, the condensing channel is supposed to be the source of heat transfer into the steel plate. Along the outer side, an evaporating falling film is used to extract the heat from the steel plate. Upward flow of air is also considered along the evaporating film. In the condensing channel, a flow solver based on an asymptotic model of the Navier–Stokes equations at the low Mach number regime and two turbulence models (Buleev's model and Chien's k–ε model) are considered. The condensing channel model was used to model the COPAIN test, the computed heat flux and condensation rate were compared with the experimental data. In the evaporating channel, a simplified model developed by Aiello and Ciofalo (2009) was used, which considered the heat and mass balance between the falling film and the ascending air flow. The model was validated for two cases: a dry wall case and a completely wet wall case. In the former case, the results were compared with 2D predictions obtained by using the CFX-4 CFD code. In the latter case, the results were compared with experimental data obtained by Kang and Park. The comparison showed a satisfactory agreement on heat transfer rates, despite some overprediction depending on the air velocity. At the end, the condensing channel model and the evaporating channel model were coupled by the steel plate
Modelling of water sump evaporation in a CFD code for nuclear containment studies
Energy Technology Data Exchange (ETDEWEB)
Malet, J., E-mail: jeanne.malet@irsn.f [Institute for Radioprotection and Nuclear Safety, DSU/SERAC/LEMAC, BP68 - 91192 Gif-sur-Yvette cedex (France); Bessiron, M., E-mail: matthieu.bessiron@irsn.f [Institute for Radioprotection and Nuclear Safety, DSU/SERAC/LEMAC, BP68 - 91192 Gif-sur-Yvette cedex (France); Perrotin, C., E-mail: christophe.perrotin@irsn.f [Institute for Radioprotection and Nuclear Safety, DSU/SERAC/LEMAC, BP68 - 91192 Gif-sur-Yvette cedex (France)
2011-05-15
Highlights: We model sump evaporation in the reactor containment for CFD codes. The sump is modelled by an interface temperature and an evaporation mass flow-rate. These two variables are modelled using energy and mass balance. Results are compared with specific experiments in a 7 m3 vessel (Tonus Qualification ANalytique, TOSQAN). A good agreement is observed, for pressure, temperatures, mass flow-rates. - Abstract: During the course of a hypothetical severe accident in a pressurized water reactor (PWR), water can be collected in the sump containment through steam condensation on walls and spray systems activation. This water is generally under evaporation conditions. The objective of this paper is twofold: to present a sump model developed using external user-defined functions for the TONUS-CFD code and to perform a first detailed comparison of the model results with experimental data. The sump model proposed here is based on energy and mass balance and leads to a good agreement between the numerical and the experimental results. Such a model can be rather easily added to any CFD code for which boundary conditions, such as injection temperature and mass flow-rate, can be modified by external user-defined functions, depending on the atmosphere conditions.
Energy Technology Data Exchange (ETDEWEB)
Schultz, Peter Andrew
2011-12-01
The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.
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)
Modeling developments for the SAS4A and SASSYS computer codes
International Nuclear Information System (INIS)
Cahalan, J.E.; Wei, T.Y.C.
1990-01-01
The SAS4A and SASSYS computer codes are being developed at Argonne National Laboratory for transient analysis of liquid metal cooled reactors. The SAS4A code is designed to analyse severe loss-of-coolant flow and overpower accidents involving coolant boiling, Cladding failures, and fuel melting and relocation. Recent SAS4A modeling developments include extension of the coolant boiling model to treat sudden fission gas release upon pin failure, expansion of the DEFORM fuel behavior model to handle advanced cladding materials and metallic fuel, and addition of metallic fuel modeling capability to the PINACLE and LEVITATE fuel relocation models. The SASSYS code is intended for the analysis of operational and beyond-design-basis transients, and provides a detailed transient thermal and hydraulic simulation of the core, the primary and secondary coolant circuits, and the balance-of-plant, in addition to a detailed model of the plant control and protection systems. Recent SASSYS modeling developments have resulted in detailed representations of the balance of plant piping network and components, including steam generators, feedwater heaters and pumps, and the turbine. 12 refs., 2 tabs
International Nuclear Information System (INIS)
Schultz, Peter Andrew
2011-01-01
The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M and S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V and V) is required throughout the system to establish evidence-based metrics for the level of confidence in M and S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V and V challenge at the subcontinuum scale, an approach to incorporate V and V concepts into subcontinuum scale modeling and simulation (M and S), and a plan to incrementally incorporate effective V and V into subcontinuum scale M and S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.
Modeling approach for annular-fuel elements using the ASSERT-PV subchannel code
International Nuclear Information System (INIS)
Dominguez, A.N.; Rao, Y.
2012-01-01
The internally and externally cooled annular fuel (hereafter called annular fuel) is under consideration for a new high burn-up fuel bundle design in Atomic Energy of Canada Limited (AECL) for its current, and its Generation IV reactor. An assessment of different options to model a bundle fuelled with annular fuel elements is presented. Two options are discussed: 1) Modify the subchannel code ASSERT-PV to handle multiple types of elements in the same bundle, and 2) coupling ASSERT-PV with an external application. Based on this assessment, the selected option is to couple ASSERT-PV with the thermalhydraulic system code CATHENA. (author)
Utilization of the Nelkin model in a Hammer computer code for calculation the reactor parameters
International Nuclear Information System (INIS)
Leal, L.C.
1980-07-01
The possibility of modifying the HAMMER code, in the thermal part, by changing the thermal neutron scattering Kernel of its library for another one calculated in a subprogramm which can be incorporated to the code, is studied. This subprogramm uses the original version of the Nelkin model instead of its approximation which is used in the HAMMER. It has the advantage of giving the values of the Kernel for any temperature of the reactor for the approximations P 0 , P 1 , P 2 and P 3 . (Author) [pt
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.
Energy Technology Data Exchange (ETDEWEB)
Coste, P.; Bestion, D. [Commissariat a l Energie Atomique, Grenoble (France)
1995-09-01
This paper presents a simple modelling of mass diffusion effects on condensation. In presence of noncondensable gases, the mass diffusion near the interface is modelled using the heat and mass transfer analogy and requires normally an iterative procedure to calculate the interface temperature. Simplifications of the model and of the solution procedure are used without important degradation of the predictions. The model is assessed on experimental data for both film condensation in vertical tubes and direct contact condensation in horizontal tubes, including air-steam, Nitrogen-steam and Helium-steam data. It is implemented in the Cathare code, a french system code for nuclear reactor thermal hydraulics developed by CEA, EDF, and FRAMATOME.
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)
Validation of the Thermal-Hydraulic Model in the SACAP Code with the ISP Tests
Energy Technology Data Exchange (ETDEWEB)
Park, Soon-Ho; Kim, Dong-Min; Park, Chang-Hwan [FNC Technology Co., Yongin (Korea, Republic of)
2016-10-15
In safety viewpoint, the pressure of the containment is the important parameter, of course, the local hydrogen concentration is also the parameter of the major concern because of its flammability and the risk of the detonation. In Korea, there have been an extensive efforts to develop the computer code which can analyze the severe accident behavior of the pressurized water reactor. The development has been done in a modularized manner and SACAP(Severe Accident Containment Analysis Package) code is now under final stage of development. SACAP code adopts LP(Lumped Parameter) model and is applicable to analyze the synthetic behavior of the containment during severe accident occurred by thermal-hydraulic transient, combustible gas burn, direct containment heating by high pressure melt ejection, steam explosion and molten core-concrete interaction. The analyses of a number of ISP(International Standard Problem) experiments were done as a part of the SACAP code V and V(verification and validation). In this paper, the SACAP analysis results for ISP-35 NUPEC and ISP-47 TOSQAN are presented including comparison with other existing NPP simulation codes. In this paper, we selected and analyzed ISP-35 NUPEC, ISP-47 TOSQAN in order to confirm the computational performance of SACAP code currently under development. Now the multi-node analysis for the ISP-47 is under process. As a result of simulation, SACAP predicts well the thermal-hydraulic variables such as temperature, pressure, etc. Also, we verify that SACAP code is properly equipped to analyze the gas distribution and condensation.
Non-standard model for electron heat transport for multidimensional hydrodynamic codes
Energy Technology Data Exchange (ETDEWEB)
Nicolai, Ph.; Busquet, M.; Schurtz, G. [CEA/DAM-Ile de France, 91 - Bruyeres Le Chatel (France)
2000-07-01
In simulations of laser-produced plasma, modeling of heat transport requires an artificial limitation of standard Spitzer-Haerm fluxes. To improve heat conduction processing, we have developed a multidimensional model which accounts for non-local features of heat transport and effects of self-generated magnetic fields. This consistent treatment of both mechanisms has been implemented in a two-dimensional radiation-hydrodynamic code. First results indicate good agreements between simulations and experimental data. (authors)
Non-standard model for electron heat transport for multidimensional hydrodynamic codes
International Nuclear Information System (INIS)
Nicolai, Ph.; Busquet, M.; Schurtz, G.
2000-01-01
In simulations of laser-produced plasma, modeling of heat transport requires an artificial limitation of standard Spitzer-Haerm fluxes. To improve heat conduction processing, we have developed a multidimensional model which accounts for non-local features of heat transport and effects of self-generated magnetic fields. This consistent treatment of both mechanisms has been implemented in a two-dimensional radiation-hydrodynamic code. First results indicate good agreements between simulations and experimental data. (authors)
Modelling the IRSN's radio-photo-luminescent dosimeter using the MCPNX Monte Carlo code
International Nuclear Information System (INIS)
Hocine, N.; Donadille, L.; Huet, Ch.; Itie, Ch.
2010-01-01
The authors report the modelling of the new radio-photo-luminescent (RPL) dosimeter of the IRSN using the MCPNX Monte Carlo code. The Hp(10) and Hp(0, 07) dose equivalents are computed for different irradiation configurations involving photonic beams (gamma and X) defined according to the ISO 4037-1 standard. Results are compared to experimental measurements performed on the RPL dosimeter. The agreement is good and the model is thus validated
Applications of the 3-dim ICRH global wave code FISIC and comparison with other models
International Nuclear Information System (INIS)
Kruecken, T.; Brambilla, M.
1989-01-01
Numerical simulations of two ICRF heating experiments in ASDEX are presented, using the FISIC code to solve the integrodifferential wave equations in the finite Larmor radius (FLR) approximation model and of ray tracing. The different models show on the whole good agreement; we can however identify a few interesting toroidal effects, in particular on the efficiency of mode conversion and on the propagation of ion Bernstein waves. (author)
Assessment of one dimensional reflood model in REFLA/TRAC code
International Nuclear Information System (INIS)
Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio
1993-12-01
Post-test calculations for twelve selected SSRTF, SCTF and CCTF tests were performed to assess the predictive capability of the one-dimensional reflood model in the REFLA/TRAC code for core thermal behavior during the reflood in a PWR LOCA. Both core void fraction profile and clad temperature transients were predicted excellently by the REFLA/TRAC code including parameter effect of core inlet subcooling, core flooding rate, core configuration, core power, system pressure, initial clad temperature and so on. The peak clad temperature was predicted within an error of 50 K. Based on these assessment results, it is verified that the core thermal hydraulic behaviors during the reflood can be predicted excellently with the REFLA/TRAC code under various conditions where the reflood may occur in a PWR LOCA. (author)
Development of Adiabatic Doppler Feedback Model in 3D space time analysis Code ARCH
International Nuclear Information System (INIS)
Dwivedi, D.K.; Gupta, Anurag
2015-01-01
Integrated 3D space-time neutron kinetics with thermal-hydraulic feedback code system is being developed for transient analysis of Compact High Temperature Reactor (CHTR) and Advanced Heavy Water Reactor (AHWR). ARCH (code for Analysis of Reactor transients in Cartesian and Hexagon geometries) has been developed with IQS module for efficient 3D space time analysis. Recently, an adiabatic Doppler (fuel temperature) feedback module has been incorporated in this ARCH-IQS version of tile code. In the adiabatic model of fuel temperature feedback, the transfer of the excess heat from the fuel to the coolant during transient is neglected. The viability of Doppler feedback in ARCH-IQS with adiabatic heating has been checked with AER benchmark (Dyn002). Analyses of anticipated transient without scram (ATWS) case in CHTR as well as in AHWR have been performed with adiabatic fuel temperature feedback. The methodology and results have been presented in this paper. (author)
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.
MELMRK 2.0: A description of computer models and results of code testing
International Nuclear Information System (INIS)
Wittman, R.S.; Denny, V.; Mertol, A.
1992-01-01
An advanced version of the MELMRK computer code has been developed that provides detailed models for conservation of mass, momentum, and thermal energy within relocating streams of molten metallics during meltdown of Savannah River Site (SRS) reactor assemblies. In addition to a mechanistic treatment of transport phenomena within a relocating stream, MELMRK 2.0 retains the MOD1 capability for real-time coupling of the in-depth thermal response of participating assembly heat structure and, further, augments this capability with models for self-heating of relocating melt owing to steam oxidation of metallics and fission product decay power. As was the case for MELMRK 1.0, the MOD2 version offers state-of-the-art numerics for solving coupled sets of nonlinear differential equations. Principal features include application of multi-dimensional Newton-Raphson techniques to accelerate convergence behavior and direct matrix inversion to advance primitive variables from one iterate to the next. Additionally, MELMRK 2.0 provides logical event flags for managing the broad range of code options available for treating such features as (1) coexisting flow regimes, (2) dynamic transitions between flow regimes, and (3) linkages between heatup and relocation code modules. The purpose of this report is to provide a detailed description of the MELMRK 2.0 computer models for melt relocation. Also included are illustrative results for code testing, as well as an integrated calculation for meltdown of a Mark 31a assembly
International Nuclear Information System (INIS)
Falck, W.E.
1991-01-01
There is a growing need to consider the influence of organic macromolecules on the speciation of ions in natural waters. It is recognized that a simple discrete ligand approach to the binding of protons/cations to organic macromolecules is not appropriate to represent heterogeneities of binding site distributions. A more realistic approach has been incorporated into the speciation code PHREEQE which retains the discrete ligand approach but modifies the binding intensities using an electrostatic (surface complexation) model. To allow for different conformations of natural organic material two alternative concepts have been incorporated: it is assumed that (a) the organic molecules form rigid, impenetrable spheres, and (b) the organic molecules form flat surfaces. The former concept will be more appropriate for molecules in the smaller size range, while the latter will be more representative for larger size molecules or organic surface coatings. The theoretical concept is discussed and the relevant changes to the standard PHREEQE code are explained. The modified codes are called PHREEQEO-RS and PHREEQEO-FS for the rigid-sphere and flat-surface models respectively. Improved output facilities for data transfer to other computers, e.g. the Macintosh, are introduced. Examples where the model is tested against literature data are shown and practical problems are discussed. Appendices contain listings of the modified subroutines GAMMA and PTOT, an example input file and an example command procedure to run the codes on VAX computers
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
Energy Technology Data Exchange (ETDEWEB)
Santos-Villalobos, Hector J [ORNL; Gregor, Jens [University of Tennessee, Knoxville (UTK); Bingham, Philip R [ORNL
2014-01-01
At the present, neutron sources cannot be fabricated small and powerful enough in order to achieve high resolution radiography while maintaining an adequate flux. One solution is to employ computational imaging techniques such as a Magnified Coded Source Imaging (CSI) system. A coded-mask is placed between the neutron source and the object. The system resolution is increased by reducing the size of the mask holes and the flux is increased by increasing the size of the coded-mask and/or the number of holes. One limitation of such system is that the resolution of current state-of-the-art scintillator-based detectors caps around 50um. To overcome this challenge, the coded-mask and object are magnified by making the distance from the coded-mask to the object much smaller than the distance from object to detector. In previous work, we have shown via synthetic experiments that our least squares method outperforms other methods in image quality and reconstruction precision because of the modeling of the CSI system components. However, the validation experiments were limited to simplistic neutron sources. In this work, we aim to model the flux distribution of a real neutron source and incorporate such a model in our least squares computational system. We provide a full description of the methodology used to characterize the neutron source and validate the method with synthetic experiments.
An evaluation of nodalization/decay heat/ volatile fission product release models in ISAAC code
Energy Technology Data Exchange (ETDEWEB)
Song, Yong Mann; Park, Soo Yong; Kim, Dong Ha
2003-03-01
An ISAAC computer code, which was developed for a Level-2 PSA during 1995, has developed mainly with fundamental models for CANDU-specific severe accident progression and also the accident-analyzing experiences are limited to Level-2 PSA purposes. Hence the system nodalization model, decay model and volatile fission product release model, which are known to affect fission product behavior directly or indirectly, are evaluated to both enhance understanding for basic models and accumulate accident-analyzing experiences. As a research strategy, sensitivity studies of model parameters and sensitivity coefficients are performed. According to the results from core nodalization sensitivity study, an original 3x3 nodalization (per loop) method which groups horizontal fuel channels into 12 representative channels, is evaluated to be sufficient for an optimal scheme because detailed nodalization methods have no large effect on fuel thermal-hydraulic behavior, total accident progression and fission product behavior. As ANSI/ANS standard model for decay heat prediction after reactor trip has no needs for further model evaluation due to both wide application on accident analysis codes and good comparison results with the ORIGEN code, ISAAC calculational results of decay heat are used as they are. In addition, fission product revaporization in a containment which is caused by the embedded decay heat, is demonstrated. The results for the volatile fission product release model are analyzed. In case of early release, the IDCOR model with an in-vessel Te release option shows the most conservative results and for the late release case, NUREG-0772 model shows the most conservative results. Considering both early and late release, the IDCOR model with an in-vessel Te bound option shows mitigated conservative results.
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
Using finite mixture models in thermal-hydraulics system code uncertainty analysis
Energy Technology Data Exchange (ETDEWEB)
Carlos, S., E-mail: scarlos@iqn.upv.es [Department d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain); Sánchez, A. [Department d’Estadística Aplicada i Qualitat, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain); Ginestar, D. [Department de Matemàtica Aplicada, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain); Martorell, S. [Department d’Enginyeria Química i Nuclear, Universitat Politècnica de València, Camí de Vera s.n, 46022 València (Spain)
2013-09-15
Highlights: • Best estimate codes simulation needs uncertainty quantification. • The output variables can present multimodal probability distributions. • The analysis of multimodal distribution is performed using finite mixture models. • Two methods to reconstruct output variable probability distribution are used. -- Abstract: Nuclear Power Plant safety analysis is mainly based on the use of best estimate (BE) codes that predict the plant behavior under normal or accidental conditions. As the BE codes introduce uncertainties due to uncertainty in input parameters and modeling, it is necessary to perform uncertainty assessment (UA), and eventually sensitivity analysis (SA), of the results obtained. These analyses are part of the appropriate treatment of uncertainties imposed by current regulation based on the adoption of the best estimate plus uncertainty (BEPU) approach. The most popular approach for uncertainty assessment, based on Wilks’ method, obtains a tolerance/confidence interval, but it does not completely characterize the output variable behavior, which is required for an extended UA and SA. However, the development of standard UA and SA impose high computational cost due to the large number of simulations needed. In order to obtain more information about the output variable and, at the same time, to keep computational cost as low as possible, there has been a recent shift toward developing metamodels (model of model), or surrogate models, that approximate or emulate complex computer codes. In this way, there exist different techniques to reconstruct the probability distribution using the information provided by a sample of values as, for example, the finite mixture models. In this paper, the Expectation Maximization and the k-means algorithms are used to obtain a finite mixture model that reconstructs the output variable probability distribution from data obtained with RELAP-5 simulations. Both methodologies have been applied to a separated
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)
Maxwell: A semi-analytic 4D code for earthquake cycle modeling of transform fault systems
Sandwell, David; Smith-Konter, Bridget
2018-05-01
We have developed a semi-analytic approach (and computational code) for rapidly calculating 3D time-dependent deformation and stress caused by screw dislocations imbedded within an elastic layer overlying a Maxwell viscoelastic half-space. The maxwell model is developed in the Fourier domain to exploit the computational advantages of the convolution theorem, hence substantially reducing the computational burden associated with an arbitrarily complex distribution of force couples necessary for fault modeling. The new aspect of this development is the ability to model lateral variations in shear modulus. Ten benchmark examples are provided for testing and verification of the algorithms and code. One final example simulates interseismic deformation along the San Andreas Fault System where lateral variations in shear modulus are included to simulate lateral variations in lithospheric structure.
An evolutionary model for protein-coding regions with conserved RNA structure
DEFF Research Database (Denmark)
Pedersen, Jakob Skou; Forsberg, Roald; Meyer, Irmtraud Margret
2004-01-01
in the RNA structure. The overlap of these fundamental dependencies is sufficient to cause "contagious" context dependencies which cascade across many nucleotide sites. Such large-scale dependencies challenge the use of traditional phylogenetic models in evolutionary inference because they explicitly assume...... components of traditional phylogenetic models. We applied this to a data set of full-genome sequences from the hepatitis C virus where five RNA structures are mapped within the coding region. This allowed us to partition the effects of selection on different structural elements and to test various hypotheses......Here we present a model of nucleotide substitution in protein-coding regions that also encode the formation of conserved RNA structures. In such regions, apparent evolutionary context dependencies exist, both between nucleotides occupying the same codon and between nucleotides forming a base pair...
Setup of HDRK-Man voxel model in Geant4 Monte Carlo code
Energy Technology Data Exchange (ETDEWEB)
Jeong, Jong Hwi; Cho, Sung Koo; Kim, Chan Hyeong [Hanyang Univ., Seoul (Korea, Republic of); Choi, Sang Hyoun [Inha Univ., Incheon (Korea, Republic of); Cho, Kun Woo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2008-10-15
Many different voxel models, developed using tomographic images of human body, are used in various fields including both ionizing and non-ionizing radiation fields. Recently a high-quality voxel model/ named HDRK-Man, was constructed at Hanyang University and used to calculate the dose conversion coefficients (DCC) values for external photon and neutron beams using the MCNPX Monte Carlo code. The objective of the present study is to set up the HDRK-Man model in Geant4 in order to use it in more advanced calculations such as 4-D Monte Carlo simulations and space dosimetry studies involving very high energy particles. To that end, the HDRK-Man was ported to Geant4 and used to calculate the DCC values for external photon beams. The calculated values were then compared with the results of the MCNPX code. In addition, a computational Linux cluster was built to improve the computing speed in Geant4.
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
NRMC - A GPU code for N-Reverse Monte Carlo modeling of fluids in confined media
Sánchez-Gil, Vicente; Noya, Eva G.; Lomba, Enrique
2017-08-01
NRMC is a parallel code for performing N-Reverse Monte Carlo modeling of fluids in confined media [V. Sánchez-Gil, E.G. Noya, E. Lomba, J. Chem. Phys. 140 (2014) 024504]. This method is an extension of the usual Reverse Monte Carlo method to obtain structural models of confined fluids compatible with experimental diffraction patterns, specifically designed to overcome the problem of slow diffusion that can appear under conditions of tight confinement. Most of the computational time in N-Reverse Monte Carlo modeling is spent in the evaluation of the structure factor for each trial configuration, a calculation that can be easily parallelized. Implementation of the structure factor evaluation in NVIDIA® CUDA so that the code can be run on GPUs leads to a speed up of up to two orders of magnitude.
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)
Incorporation of the capillary hysteresis model HYSTR into the numerical code TOUGH
International Nuclear Information System (INIS)
Niemi, A.; Bodvarsson, G.S.; Pruess, K.
1991-11-01
As part of the work performed to model flow in the unsaturated zone at Yucca Mountain Nevada, a capillary hysteresis model has been developed. The computer program HYSTR has been developed to compute the hysteretic capillary pressure -- liquid saturation relationship through interpolation of tabulated data. The code can be easily incorporated into any numerical unsaturated flow simulator. A complete description of HYSTR, including a brief summary of the previous hysteresis literature, detailed description of the program, and instructions for its incorporation into a numerical simulator are given in the HYSTR user's manual (Niemi and Bodvarsson, 1991a). This report describes the incorporation of HYSTR into the numerical code TOUGH (Transport of Unsaturated Groundwater and Heat; Pruess, 1986). The changes made and procedures for the use of TOUGH for hysteresis modeling are documented
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).
SCDAP/RELAP5/MOD 3.1 code manual: Damage progression model theory. Volume 2
International Nuclear Information System (INIS)
Davis, K.L.
1995-06-01
The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, 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 as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed descriptions of the severe accident models and correlations. It provides the user with the underlying assumptions and simplifications used to generate and implement the basic equations into the code, so an intelligent assessment of the applicability and accuracy of the resulting calculation can be made
SCDAP/RELAP5/MOD 3.1 code manual: Damage progression model theory. Volume 2
Energy Technology Data Exchange (ETDEWEB)
Davis, K.L. [ed.; Allison, C.M.; Berna, G.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)] [and others
1995-06-01
The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, 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 as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed descriptions of the severe accident models and correlations. It provides the user with the underlying assumptions and simplifications used to generate and implement the basic equations into the code, so an intelligent assessment of the applicability and accuracy of the resulting calculation can be made.
Modeling of the WWER-1000 fuel-rod behavior in steady-state condition with FRAPCONE-3 computer code
International Nuclear Information System (INIS)
Andreeva, Marina; Totev, Totju; Stoyanov, Stoyan
2008-01-01
It is presented within the paper the results of the modeling and the assessment of the integral code predictions of the WWER fuel-rod behavior in steady-state condition. The assessments in this paper have used the MASSIH and ANS 5.4 subroutine in the code. The modeling and calculations have been performed with FRAPCONE-3 computer code in Argonne National Laboratory, USA
International Nuclear Information System (INIS)
Heames, T.J.; Khatib-Rahbar, M.; Kelly, J.E.
1995-01-01
The hierarchy-by-interval (HBI) methodology was developed to determine an appropriate phenomena identification and ranking table for an independent peer review of severe-accident computer codes. The methodology is described, and the results of a specific code review are presented. Use of this systematic and structured approach ensures that important code models that need improvement are identified and prioritized, which allows code sponsors to more effectively direct limited resources in future code development. In addition, critical phenomenological areas that need more fundamental work, such as experimentation, are identified
Extension of the EQ3/6 computer codes to geochemical modeling of brines
Energy Technology Data Exchange (ETDEWEB)
Jackson, K.J.; Wolery, T.J.
1984-10-23
Recent modifications to the EQ3/6 geochemical modeling software package provide for the use of Pitzer's equations to calculate the activity coefficients of aqueous species and the activity of water. These changes extend the range of solute concentrations over which the codes can be used to dependably calculate equilibria in geochemical systems, and permit the inclusion of ion pairs, complexes, and undissociated acids and bases as explicit component species in the Pitzer model. Comparisons of calculations made by the EQ3NR and EQ6 compuer codes with experimental data confirm that the modifications not only allow the codes to accurately evaluate activity coefficients in concentrated solutions, but also permit prediction of solubility limits of evaporite minerals in brines at 25/sup 0/C and elevated temperatures. Calculations for a few salts can be made at temperatures up to approx. 300/sup 0/C, but the temperature range for most electrolytes is constrained by the availability of requisite data to values less than or equal to 100/sup 0/C. The implementation of Pitzer's equations in EQ3/6 allows application of these codes to problems involving calculation of geochemical equilibria in brines; such as evaluation of the chemical environment which might be anticipated for nuclear waste canisters located in a salt repository. 26 references, 3 figures, 1 table.
Directory of Open Access Journals (Sweden)
Wang Yanqing
2016-03-01
Full Text Available A good assignment of code reviewers can effectively utilize the intellectual resources, assure code quality and improve programmers’ skills in software development. However, little research on reviewer assignment of code review has been found. In this study, a code reviewer assignment model is created based on participants’ preference to reviewing assignment. With a constraint of the smallest size of a review group, the model is optimized to maximize review outcomes and avoid the negative impact of “mutual admiration society”. This study shows that the reviewer assignment strategies incorporating either the reviewers’ preferences or the authors’ preferences get much improvement than a random assignment. The strategy incorporating authors’ preference makes higher improvement than that incorporating reviewers’ preference. However, when the reviewers’ and authors’ preference matrixes are merged, the improvement becomes moderate. The study indicates that the majority of the participants have a strong wish to work with reviewers and authors having highest competence. If we want to satisfy the preference of both reviewers and authors at the same time, the overall improvement of learning outcomes may be not the best.
What Are They Talking About? Analyzing Code Reviews in Pull-Based Development Model
Institute of Scientific and Technical Information of China (English)
Zhi-Xing Li; Yue Yu; Gang Yin; Tao Wang; Huai-Min Wang
2017-01-01
Code reviews in pull-based model are open to community users on GitHub. Various participants are taking part in the review discussions and the review topics are not only about the improvement of code contributions but also about project evolution and social interaction. A comprehensive understanding of the review topics in pull-based model would be useful to better organize the code review process and optimize review tasks such as reviewer recommendation and pull-request prioritization. In this paper, we first conduct a qualitative study on three popular open-source software projects hosted on GitHub and construct a fine-grained two-level taxonomy covering four level-1 categories (code correctness, pull-request decision-making, project management, and social interaction) and 11 level-2 subcategories (e.g., defect detecting, reviewer assigning, contribution encouraging). Second, we conduct preliminary quantitative analysis on a large set of review comments that were labeled by TSHC (a two-stage hybrid classification algorithm), which is able to automatically classify review comments by combining rule-based and machine-learning techniques. Through the quantitative study, we explore the typical review patterns. We find that the three projects present similar comments distribution on each subcategory. Pull-requests submitted by inexperienced contributors tend to contain potential issues even though they have passed the tests. Furthermore, external contributors are more likely to break project conventions in their early contributions.
Beyond the Business Model: Incentives for Organizations to Publish Software Source Code
Lindman, Juho; Juutilainen, Juha-Pekka; Rossi, Matti
The software stack opened under Open Source Software (OSS) licenses is growing rapidly. Commercial actors have released considerable amounts of previously proprietary source code. These actions beg the question why companies choose a strategy based on giving away software assets? Research on outbound OSS approach has tried to answer this question with the concept of the “OSS business model”. When studying the reasons for code release, we have observed that the business model concept is too generic to capture the many incentives organizations have. Conversely, in this paper we investigate empirically what the companies’ incentives are by means of an exploratory case study of three organizations in different stages of their code release. Our results indicate that the companies aim to promote standardization, obtain development resources, gain cost savings, improve the quality of software, increase the trustworthiness of software, or steer OSS communities. We conclude that future research on outbound OSS could benefit from focusing on the heterogeneous incentives for code release rather than on revenue models.
Bhattacharya, Moumita; Jurkovitz, Claudine; Shatkay, Hagit
2018-04-12
Patients associated with multiple co-occurring health conditions often face aggravated complications and less favorable outcomes. Co-occurring conditions are especially prevalent among individuals suffering from kidney disease, an increasingly widespread condition affecting 13% of the general population in the US. This study aims to identify and characterize patterns of co-occurring medical conditions in patients employing a probabilistic framework. Specifically, we apply topic modeling in a non-traditional way to find associations across SNOMED-CT codes assigned and recorded in the EHRs of >13,000 patients diagnosed with kidney disease. Unlike most prior work on topic modeling, we apply the method to codes rather than to natural language. Moreover, we quantitatively evaluate the topics, assessing their tightness and distinctiveness, and also assess the medical validity of our results. Our experiments show that each topic is succinctly characterized by a few highly probable and unique disease codes, indicating that the topics are tight. Furthermore, inter-topic distance between each pair of topics is typically high, illustrating distinctiveness. Last, most coded conditions grouped together within a topic, are indeed reported to co-occur in the medical literature. Notably, our results uncover a few indirect associations among conditions that have hitherto not been reported as correlated in the medical literature. Copyright © 2018. Published by Elsevier Inc.
International Nuclear Information System (INIS)
Bujan, A.; Adamik, V.; Misak, J.
1986-01-01
A brief description is presented of the expansion of the SICHTA-83 computer code for the analysis of the thermal history of the fuel channel for large LOCAs by modelling the mechanical behaviour of fuel element cladding. The new version of the code has a more detailed treatment of heat transfer in the fuel-cladding gap because it also respects the mechanical (plastic) deformations of the cladding and the fuel-cladding interaction (magnitude of contact pressure). Also respected is the change in pressure of the gas filling of the fuel element, the mechanical criterion is considered of a failure of the cladding and the degree is considered of the blockage of the through-flow cross section for coolant flow in the fuel channel. The LOCA WWER-440 model computation provides a comparison of the new SICHTA-85/MOD 1 code with the results of the original 83 version of SICHTA. (author)
Comprehensive nuclear model calculations: theory and use of the GNASH code
International Nuclear Information System (INIS)
Young, P.G.; Arthur, E.D.; Chadwick, M.B.
1998-01-01
The theory and operation of the nuclear reaction theory computer code GNASH is described, and detailed instructions are presented for code users. The code utilizes statistical Hauser-Feshbach theory with full angular momentum conservation and includes corrections for preequilibrium effects. This version is expected to be applicable for incident particle energies between 1 keV and 150 MeV and for incident photon energies to 140 MeV. General features of the code, the nuclear models that are utilized, input parameters needed to perform calculations, and the output quantities from typical problems are described in detail. A number of new features compared to previous versions are described in this manual, including the following: (1) inclusion of multiple preequilibrium processes, which allows the model calculations to be performed above 50 MeV; (2) a capability to calculate photonuclear reactions; (3) a method for determining the spin distribution of residual nuclei following preequilibrium reactions; and (4) a description of how preequilibrium spectra calculated with the FKK theory can be utilized (the 'FKK-GNASH' approach). The computational structure of the code and the subroutines and functions that are called are summarized as well. Two detailed examples are considered: 14-MeV neutrons incident on 93 Nb and 12-MeV neutrons incident on 238 U. The former example illustrates a typical calculation aimed at determining neutron, proton, and alpha emission spectra from 14-MeV reactions, and the latter example demonstrates use of the fission model in GNASH. Results from a variety of other cases are illustrated. (author)
SCANAIR a transient fuel performance code Part two: Assessment of modelling capabilities
Energy Technology Data Exchange (ETDEWEB)
Georgenthum, Vincent, E-mail: vincent.georgenthum@irsn.fr; Moal, Alain; Marchand, Olivier
2014-12-15
Highlights: • The SCANAIR code is devoted to the study of irradiated fuel rod behaviour during RIA. • The paper deals with the status of the code validation for PWR rods. • During the PCMI stage there is a good agreement between calculations and experiments. • The boiling crisis occurrence is rather well predicted. • The code assessment during the boiling crisis has still to be improved. - Abstract: In the frame of their research programmes on fuel safety, the French Institut de Radioprotection et de Sûreté Nucléaire develops the SCANAIR code devoted to the study of irradiated fuel rod behaviour during reactivity initiated accident. A first paper was focused on detailed modellings and code description. This second paper deals with the status of the code validation for pressurised water reactor rods performed thanks to the available experimental results. About 60 integral tests carried out in CABRI and NSRR experimental reactors and 24 separated tests performed in the PATRICIA facility (devoted to the thermal-hydraulics study) have been recalculated and compared to experimental data. During the first stage of the transient, the pellet clad mechanical interaction phase, there is a good agreement between calculations and experiments: the clad residual elongation and hoop strain of non failed tests but also the failure occurrence and failure enthalpy of failed tests are correctly calculated. After this first stage, the increase of cladding temperature can lead to the Departure from Nucleate Boiling. During the film boiling regime, the clad temperature can reach a very high temperature (>700 °C). If the boiling crisis occurrence is rather well predicted, the calculation of the clad temperature and the clad hoop strain during this stage have still to be improved.
Comparative study of boron transport models in NRC Thermal-Hydraulic Code Trace
Energy Technology Data Exchange (ETDEWEB)
Olmo-Juan, Nicolás; Barrachina, Teresa; Miró, Rafael; Verdú, Gumersindo; Pereira, Claubia, E-mail: nioljua@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es, E-mail: claubia@nuclear.ufmg.br [Institute for Industrial, Radiophysical and Environmental Safety (ISIRYM). Universitat Politècnica de València (Spain); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear
2017-07-01
Recently, the interest in the study of various types of transients involving changes in the boron concentration inside the reactor, has led to an increase in the interest of developing and studying new models and tools that allow a correct study of boron transport. Therefore, a significant variety of different boron transport models and spatial difference schemes are available in the thermal-hydraulic codes, as TRACE. According to this interest, in this work it will be compared the results obtained using the different boron transport models implemented in the NRC thermal-hydraulic code TRACE. To do this, a set of models have been created using the different options and configurations that could have influence in boron transport. These models allow to reproduce a simple event of filling or emptying the boron concentration in a long pipe. Moreover, with the aim to compare the differences obtained when one-dimensional or three-dimensional components are chosen, it has modeled many different cases using only pipe components or a mix of pipe and vessel components. In addition, the influence of the void fraction in the boron transport has been studied and compared under close conditions to BWR commercial model. A final collection of the different cases and boron transport models are compared between them and those corresponding to the analytical solution provided by the Burgers equation. From this comparison, important conclusions are drawn that will be the basis of modeling the boron transport in TRACE adequately. (author)
Modeling compositional dynamics based on GC and purine contents of protein-coding sequences
Zhang, Zhang; Yu, Jun
2010-01-01
Background: Understanding the compositional dynamics of genomes and their coding sequences is of great significance in gaining clues into molecular evolution and a large number of publically-available genome sequences have allowed us to quantitatively predict deviations of empirical data from their theoretical counterparts. However, the quantification of theoretical compositional variations for a wide diversity of genomes remains a major challenge.Results: To model the compositional dynamics of protein-coding sequences, we propose two simple models that take into account both mutation and selection effects, which act differently at the three codon positions, and use both GC and purine contents as compositional parameters. The two models concern the theoretical composition of nucleotides, codons, and amino acids, with no prerequisite of homologous sequences or their alignments. We evaluated the two models by quantifying theoretical compositions of a large collection of protein-coding sequences (including 46 of Archaea, 686 of Bacteria, and 826 of Eukarya), yielding consistent theoretical compositions across all the collected sequences.Conclusions: We show that the compositions of nucleotides, codons, and amino acids are largely determined by both GC and purine contents and suggest that deviations of the observed from the expected compositions may reflect compositional signatures that arise from a complex interplay between mutation and selection via DNA replication and repair mechanisms.Reviewers: This article was reviewed by Zhaolei Zhang (nominated by Mark Gerstein), Guruprasad Ananda (nominated by Kateryna Makova), and Daniel Haft. 2010 Zhang and Yu; licensee BioMed Central Ltd.
Directory of Open Access Journals (Sweden)
Marcus Völp
2012-11-01
Full Text Available Reliability in terms of functional properties from the safety-liveness spectrum is an indispensable requirement of low-level operating-system (OS code. However, with evermore complex and thus less predictable hardware, quantitative and probabilistic guarantees become more and more important. Probabilistic model checking is one technique to automatically obtain these guarantees. First experiences with the automated quantitative analysis of low-level operating-system code confirm the expectation that the naive probabilistic model checking approach rapidly reaches its limits when increasing the numbers of processes. This paper reports on our work-in-progress to tackle the state explosion problem for low-level OS-code caused by the exponential blow-up of the model size when the number of processes grows. We studied the symmetry reduction approach and carried out our experiments with a simple test-and-test-and-set lock case study as a representative example for a wide range of protocols with natural inter-process dependencies and long-run properties. We quickly see a state-space explosion for scenarios where inter-process dependencies are insignificant. However, once inter-process dependencies dominate the picture models with hundred and more processes can be constructed and analysed.
Modeling compositional dynamics based on GC and purine contents of protein-coding sequences
Zhang, Zhang
2010-11-08
Background: Understanding the compositional dynamics of genomes and their coding sequences is of great significance in gaining clues into molecular evolution and a large number of publically-available genome sequences have allowed us to quantitatively predict deviations of empirical data from their theoretical counterparts. However, the quantification of theoretical compositional variations for a wide diversity of genomes remains a major challenge.Results: To model the compositional dynamics of protein-coding sequences, we propose two simple models that take into account both mutation and selection effects, which act differently at the three codon positions, and use both GC and purine contents as compositional parameters. The two models concern the theoretical composition of nucleotides, codons, and amino acids, with no prerequisite of homologous sequences or their alignments. We evaluated the two models by quantifying theoretical compositions of a large collection of protein-coding sequences (including 46 of Archaea, 686 of Bacteria, and 826 of Eukarya), yielding consistent theoretical compositions across all the collected sequences.Conclusions: We show that the compositions of nucleotides, codons, and amino acids are largely determined by both GC and purine contents and suggest that deviations of the observed from the expected compositions may reflect compositional signatures that arise from a complex interplay between mutation and selection via DNA replication and repair mechanisms.Reviewers: This article was reviewed by Zhaolei Zhang (nominated by Mark Gerstein), Guruprasad Ananda (nominated by Kateryna Makova), and Daniel Haft. 2010 Zhang and Yu; licensee BioMed Central Ltd.
Source-term model for the SYVAC3-NSURE performance assessment code
International Nuclear Information System (INIS)
Rowat, J.H.; Rattan, D.S.; Dolinar, G.M.
1996-11-01
Radionuclide contaminants in wastes emplaced in disposal facilities will not remain in those facilities indefinitely. Engineered barriers will eventually degrade, allowing radioactivity to escape from the vault. The radionuclide release rate from a low-level radioactive waste (LLRW) disposal facility, the source term, is a key component in the performance assessment of the disposal system. This report describes the source-term model that has been implemented in Ver. 1.03 of the SYVAC3-NSURE (Systems Variability Analysis Code generation 3-Near Surface Repository) code. NSURE is a performance assessment code that evaluates the impact of near-surface disposal of LLRW through the groundwater pathway. The source-term model described here was developed for the Intrusion Resistant Underground Structure (IRUS) disposal facility, which is a vault that is to be located in the unsaturated overburden at AECL's Chalk River Laboratories. The processes included in the vault model are roof and waste package performance, and diffusion, advection and sorption of radionuclides in the vault backfill. The model presented here was developed for the IRUS vault; however, it is applicable to other near-surface disposal facilities. (author). 40 refs., 6 figs
International Nuclear Information System (INIS)
Dominguez, L.; Camargo, C.T.M.
1985-01-01
A two loop simulation capability was developed to the ALMOD 3W2 code through modelling the steam header line connecting the secondary side of steam generators. A brief description of the model is presented and two test cases are shown. Basic code changes are addressed. (Author) [pt
Introduction of the bubble rise dynamic model into the ALMOD 3 code pressurizer
International Nuclear Information System (INIS)
Madeira, A.A.; Camargo, C.T.M.
1985-01-01
A new evaporation model for the ALMOD 3 code pressurizer is implemented in order to estimate more accurately the water level behaviour and its influence in the pressure transient for very fast depressurization cases. For the inclusion of the bubble rise dynamic model it was necessary to consider a two-phase mixture in the water volume. The modifications don't require additional input data and virtually had not modified the processing time. The results and processing time for the original and the new models are presented. (F.E.) [pt
Modelling of QUENCH-03 and QUENCH-06 Experiments Using RELAP/SCDAPSIM and ASTEC Codes
Directory of Open Access Journals (Sweden)
Tadas Kaliatka
2014-01-01
Full Text Available To prevent total meltdown of the uncovered and overheated core, the reflooding with water is a necessary accident management measure. Because these actions lead to the generation of hydrogen, which can cause further problems, the related phenomena are investigated performing experiments and computer simulations. In this paper, for the experiments of loss of coolant accidents, performed in Forschungszentrum Karlsruhe, QUENCH-03 and QUENCH-06 are modelled using RELAP5/SCDAPSIM and ASTEC codes. The performed benchmark allowed analysing different modelling features. The recommendations for the model development are presented.
Improved SAFARI-1 research reactor irradiation position modeling in OSCAR-3 code system
International Nuclear Information System (INIS)
Moloko, L. E.; Belal, M. G. A. H.
2009-01-01
The demand on the availability of irradiation positions in the SAFARI-1 reactor is continuously increasing due to the commercial pressure to produce isotopes more efficiently. This calls for calculational techniques and modeling methods to be improved regularly to optimize irradiation services. The irradiation position models are improved using the OSCAR-3 code system, and results are compared to experimental measurements. It is concluded that the irradiation position models are essential if realistic core follow and reload studies are to be performed and most importantly, for the realization of improved agreement between experimental data and calculated results. (authors)
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
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
Behaviors of impurity in ITER and DEMOs using BALDUR integrated predictive modeling code
International Nuclear Information System (INIS)
Onjun, Thawatchai; Buangam, Wannapa; Wisitsorasak, Apiwat
2015-01-01
The behaviors of impurity are investigated using self-consistent modeling of 1.5D BALDUR integrated predictive modeling code, in which theory-based models are used for both core and edge region. In these simulations, a combination of NCLASS neoclassical transport and Multi-mode (MMM95) anomalous transport model is used to compute a core transport. The boundary is taken to be at the top of the pedestal, where the pedestal values are described using a theory-based pedestal model. This pedestal temperature model is based on a combination of magnetic and flow shear stabilization pedestal width scaling and an infinite-n ballooning pressure gradient model. The time evolution of plasma current, temperature and density profiles is carried out for ITER and DEMOs plasmas. As a result, the impurity behaviors such as impurity accumulation and impurity transport can be investigated. (author)
Development of thermal hydraulic models for the reliable regulatory auditing code
Energy Technology Data Exchange (ETDEWEB)
Chung, B. D.; Song, C. H.; Lee, Y. J.; Kwon, T. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2003-04-15
The objective of this project is to develop thermal hydraulic models for use in improving the reliability of the regulatory auditing codes. The current year fall under the first step of the 3 year project, and the main researches were focused on identifying the candidate thermal hydraulic models for improvement and to develop prototypical model development. During the current year, the verification calculations submitted for the APR 1400 design certification have been reviewed, the experimental data from the MIDAS DVI experiment facility in KAERI have been analyzed and evaluated, candidate thermal hydraulic models for improvement have been identified, prototypical models for the improved thermal hydraulic models have been developed, items for experiment in connection with the model development have been identified, and preliminary design of the experiment has been carried out.
Development of thermal hydraulic models for the reliable regulatory auditing code
International Nuclear Information System (INIS)
Chung, B. D.; Song, C. H.; Lee, Y. J.; Kwon, T. S.
2003-04-01
The objective of this project is to develop thermal hydraulic models for use in improving the reliability of the regulatory auditing codes. The current year fall under the first step of the 3 year project, and the main researches were focused on identifying the candidate thermal hydraulic models for improvement and to develop prototypical model development. During the current year, the verification calculations submitted for the APR 1400 design certification have been reviewed, the experimental data from the MIDAS DVI experiment facility in KAERI have been analyzed and evaluated, candidate thermal hydraulic models for improvement have been identified, prototypical models for the improved thermal hydraulic models have been developed, items for experiment in connection with the model development have been identified, and preliminary design of the experiment has been carried out
Blackout sequence modeling for Atucha-I with MARCH3 code
International Nuclear Information System (INIS)
Baron, J.; Bastianelli, B.
1997-01-01
The modeling of a blackout sequence in Atucha I nuclear power plant is presented in this paper, as a preliminary phase for a level II probabilistic safety assessment. Such sequence is analyzed with the code MARCH3 from STCP (Source Term Code Package), based on a specific model developed for Atucha, that takes into accounts it peculiarities. The analysis includes all the severe accident phases, from the initial transient (loss of heat sink), loss of coolant through the safety valves, core uncovered, heatup, metal-water reaction, melting and relocation, heatup and failure of the pressure vessel, core-concrete interaction in the reactor cavity, heatup and failure of the containment building (multi-compartmented) due to quasi-static overpressurization. The results obtained permit to visualize the time sequence of these events, as well as provide the basis for source term studies. (author) [es
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)
Comparison of different methods used in integral codes to model coagulation of aerosols
Beketov, A. I.; Sorokin, A. A.; Alipchenkov, V. M.; Mosunova, N. A.
2013-09-01
The methods for calculating coagulation of particles in the carrying phase that are used in the integral codes SOCRAT, ASTEC, and MELCOR, as well as the Hounslow and Jacobson methods used to model aerosol processes in the chemical industry and in atmospheric investigations are compared on test problems and against experimental results in terms of their effectiveness and accuracy. It is shown that all methods are characterized by a significant error in modeling the distribution function for micrometer particles if calculations are performed using rather "coarse" spectra of particle sizes, namely, when the ratio of the volumes of particles from neighboring fractions is equal to or greater than two. With reference to the problems considered, the Hounslow method and the method applied in the aerosol module used in the ASTEC code are the most efficient ones for carrying out calculations.
Analytical model for bottom reflooding heat transfer in light water reactors (the UCFLOOD code)
International Nuclear Information System (INIS)
Arrieta, L.; Yadigaroglu, G.
1978-08-01
The UCFLOOD code is based on mechanistic models developed to analyze bottom reflooding of a single flow channel and its associated fuel rod, or a tubular test section with internal flow. From the hydrodynamic point of view the flow channel is divided into a single-phase liquid region, a continuous-liquid two-phase region, and a dispersed-liquid region. The void fraction is obtained from drift flux models. For heat transfer calculations, the channel is divided into regions of single-phase-liquid heat transfer, nucleate boiling and forced-convection vaporization, inverted-annular film boiling, and dispersed-flow film boiling. The heat transfer coefficients are functions of the local flow conditions. Good agreement of calculated and experimental results has been obtained. A code user's manual is appended
Athaudage, Chandranath R. N.; Bradley, Alan B.; Lech, Margaret
2003-12-01
A dynamic programming-based optimization strategy for a temporal decomposition (TD) model of speech and its application to low-rate speech coding in storage and broadcasting is presented. In previous work with the spectral stability-based event localizing (SBEL) TD algorithm, the event localization was performed based on a spectral stability criterion. Although this approach gave reasonably good results, there was no assurance on the optimality of the event locations. In the present work, we have optimized the event localizing task using a dynamic programming-based optimization strategy. Simulation results show that an improved TD model accuracy can be achieved. A methodology of incorporating the optimized TD algorithm within the standard MELP speech coder for the efficient compression of speech spectral information is also presented. The performance evaluation results revealed that the proposed speech coding scheme achieves 50%-60% compression of speech spectral information with negligible degradation in the decoded speech quality.
MINI-TRAC code: a driver program for assessment of constitutive equations of two-fluid model
International Nuclear Information System (INIS)
Akimoto, Hajime; Abe, Yutaka; Ohnuki, Akira; Murao, Yoshio
1991-05-01
MINI-TRAC code, a driver program for assessment of constitutive equations of two-fluid model, has been developed to perform assessment and improvement of constitutive equations of two-fluid model widely and efficiently. The MINI-TRAC code uses one-dimensional conservation equations for mass, momentum and energy based on the two-fluid model. The code can work on a personal computer because it can be operated with a core memory size less than 640 KB. The MINI-TRAC code includes constitutive equations of TRAC-PF1/MOD1 code, TRAC-BF1 code and RELAP5/MOD2 code. The code is modulated so that one can easily change constitutive equations to perform a test calculation. This report is a manual of the MINI-TRAC code. The basic equations, numerics, constitutive, equations included in the MINI-TRAC code will be described. The user's manual such as input description will be presented. The program structure and contents of main variables will also be mentioned in this report. (author)
Development of multidimensional two-fluid model code ACE-3D for evaluation of constitutive equations
Energy Technology Data Exchange (ETDEWEB)
Ohnuki, Akira; Akimoto, Hajime [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kamo, Hideki
1996-11-01
In order to perform design calculations for a passive safety reactor with good accuracy by a multidimensional two-fluid model, we developed an analysis code, ACE-3D, which can apply for evaluation of constitutive equations. The developed code has the following features: 1. The basic equations are based on 3-dimensional two-fluid model and the orthogonal or the cylindrical coordinate system can be selected. The fluid system is air-water or steam-water. 2. The basic equations are formulated by the finite-difference scheme of staggered mesh. The convection term is formulated by an upwind scheme and the diffusion term by a center-difference scheme. 3. Semi-implicit numerical scheme is adopted and the mass and the energy equations are treated equally in convergent steps for Jacobi equations. 4. The interfacial stress term consists of drag force, life force, turbulent dispersion force, wall force and virtual mass force. 5. A {kappa}-{epsilon} turbulent model for bubbly flow is incorporated as the turbulent model. The predictive capability of ACE-3D has been verified using a data-base for bubbly flow in a small-scale vertical pipe. In future, the constitutive equations will be improved with a data-base in a large vertical pipe developed in our laboratory and we have a plan to construct a reliable analytical tool through the improvement work, the progress of calculational speed with vector and parallel processing, the assessments for phase change terms and so on. This report describes the outline for the basic equations and the finite-difference equations in ACE-3D code and also the outline for the program structure. Besides, the results for the assessments of ACE-3D code for the small-scale pipe are summarized. (author)
Verification of simulation model with COBRA-IIIP code by confrontment of experimental results
International Nuclear Information System (INIS)
Silva Galetti, M.R. da; Pontedeiro, A.C.; Oliveira Barroso, A.C. de
1985-01-01
It is presented an evaluation of the COBRA IIIP/MIT code (of thermal hydraulic analysis by subchannels), comparing their results with experimental data obtained in stationary and transient regimes. It was done a study to calculate the spatial and temporal critical heat flux. It is presented a sensitivity study of simulation model related to the turbulent mixture and the number of axial intervals. (M.C.K.) [pt
The Karlsruhe code MODINA for model independent analysis of elastic scattering of spinless particles
International Nuclear Information System (INIS)
Gils, H.J.
1983-12-01
The Karlsruhe code MODINA (KfK 3063, published November 1980) has been extended in particular with respect to new approximations in the folding models and to the calculation of errors in the fourier-Bessel potentials. The corresponding subroutines replacing previous ones are compiled in this first supplement. The listings of the fit-routine-package FITEX missing in the first publication of MODINA are also included now. (orig.) [de
PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance
International Nuclear Information System (INIS)
Vondy, D.R.
1979-10-01
The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined
PREMOR: a point reactor exposure model computer code for survey analysis of power plant performance
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.
1979-10-01
The PREMOR computer code was written to exploit a simple, two-group point nuclear reactor power plant model for survey analysis. Up to thirteen actinides, fourteen fission products, and one lumped absorber nuclide density are followed over a reactor history. Successive feed batches are accounted for with provision for from one to twenty batches resident. The effect of exposure of each of the batches to the same neutron flux is determined.
Development of multidimensional two-fluid model code ACE-3D for evaluation of constitutive equations
International Nuclear Information System (INIS)
Ohnuki, Akira; Akimoto, Hajime; Kamo, Hideki.
1996-11-01
In order to perform design calculations for a passive safety reactor with good accuracy by a multidimensional two-fluid model, we developed an analysis code, ACE-3D, which can apply for evaluation of constitutive equations. The developed code has the following features: 1. The basic equations are based on 3-dimensional two-fluid model and the orthogonal or the cylindrical coordinate system can be selected. The fluid system is air-water or steam-water. 2. The basic equations are formulated by the finite-difference scheme of staggered mesh. The convection term is formulated by an upwind scheme and the diffusion term by a center-difference scheme. 3. Semi-implicit numerical scheme is adopted and the mass and the energy equations are treated equally in convergent steps for Jacobi equations. 4. The interfacial stress term consists of drag force, life force, turbulent dispersion force, wall force and virtual mass force. 5. A κ-ε turbulent model for bubbly flow is incorporated as the turbulent model. The predictive capability of ACE-3D has been verified using a data-base for bubbly flow in a small-scale vertical pipe. In future, the constitutive equations will be improved with a data-base in a large vertical pipe developed in our laboratory and we have a plan to construct a reliable analytical tool through the improvement work, the progress of calculational speed with vector and parallel processing, the assessments for phase change terms and so on. This report describes the outline for the basic equations and the finite-difference equations in ACE-3D code and also the outline for the program structure. Besides, the results for the assessments of ACE-3D code for the small-scale pipe are summarized. (author)
Directory of Open Access Journals (Sweden)
Michael J. Pelosi
2014-12-01
Full Text Available Development teams and programmers must retain critical information about their work during work intervals and gaps in order to improve future performance when work resumes. Despite time lapses, project managers want to maximize coding efficiency and effectiveness. By developing a mathematically justified, practically useful, and computationally tractable quantitative and cognitive model of learning and memory retention, this study establishes calculations designed to maximize scheduling payoff and optimize developer efficiency and effectiveness.
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
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.
QEFSM model and Markov Algorithm for translating Quran reciting rules into Braille code
Directory of Open Access Journals (Sweden)
Abdallah M. Abualkishik
2015-07-01
Full Text Available The Holy Quran is the central religious verbal text of Islam. Muslims are expected to read, understand, and apply the teachings of the Holy Quran. The Holy Quran was translated to Braille code as a normal Arabic text without having its reciting rules included. It is obvious that the users of this transliteration will not be able to recite the Quran the right way. Through this work, Quran Braille Translator (QBT presents a specific translator to translate Quran verses and their reciting rules into the Braille code. Quran Extended Finite State Machine (QEFSM model is proposed through this study as it is able to detect the Quran reciting rules (QRR from the Quran text. Basis path testing was used to evaluate the inner work for the model by checking all the test cases for the model. Markov Algorithm (MA was used for translating the detected QRR and Quran text into the matched Braille code. The data entries for QBT are Arabic letters and diacritics. The outputs of this study are seen in the double lines of Braille symbols; the first line is the proposed Quran reciting rules and the second line is for the Quran scripts.
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
Development and validation of corium oxidation model for the VAPEX code
International Nuclear Information System (INIS)
Blinkov, V.N.; Melikhov, V.I.; Davydov, M.V.; Melikhov, O.I.; Borovkova, E.M.
2011-01-01
In light water reactor core melt accidents, the molten fuel (corium) can be brought into contact with coolant water in the course of the melt relocation in-vessel and ex-vessel as well as in an accident mitigation action of water addition. Mechanical energy release from such an interaction is of interest in evaluating the structural integrity of the reactor vessel as well as of the containment. Usually, the source for the energy release is considered to be the rapid transfer of heat from the molten fuel to the water ('vapor explosion'). When the fuel contains a chemically reactive metal component, there could be an additional source for the energy release, which is the heat release and hydrogen production due to the metal-water chemical reaction. In Electrogorsk Research and Engineering Center the computer code VAPEX (VAPor EXplosion) has been developed for analysis of the molten fuel coolant interaction. Multifield approach is used for modeling of dynamics of following phases: water, steam, melt jet, melt droplets, debris. The VAPEX code was successfully validated on FARO experimental data. Hydrogen generation was observed in FARO tests even though corium didn't contain metal component. The reason for hydrogen generation was not clear, so, simplified empirical model of hydrogen generation was implemented in the VAPEX code to take into account input of hydrogen into pressure increase. This paper describes new more detailed model of hydrogen generation due to the metal-water chemical reaction and results of its validation on ZREX experiments. (orig.)
Integration of CFD codes and advanced combustion models for quantitative burnout determination
Energy Technology Data Exchange (ETDEWEB)
Javier Pallares; Inmaculada Arauzo; Alan Williams [University of Zaragoza, Zaragoza (Spain). Centre of Research for Energy Resources and Consumption (CIRCE)
2007-10-15
CFD codes and advanced kinetics combustion models are extensively used to predict coal burnout in large utility boilers. Modelling approaches based on CFD codes can accurately solve the fluid dynamics equations involved in the problem but this is usually achieved by including simple combustion models. On the other hand, advanced kinetics combustion models can give a detailed description of the coal combustion behaviour by using a simplified description of the flow field, this usually being obtained from a zone-method approach. Both approximations describe correctly general trends on coal burnout, but fail to predict quantitative values. In this paper a new methodology which takes advantage of both approximations is described. In the first instance CFD solutions were obtained of the combustion conditions in the furnace in the Lamarmora power plant (ASM Brescia, Italy) for a number of different conditions and for three coals. Then, these furnace conditions were used as inputs for a more detailed chemical combustion model to predict coal burnout. In this, devolatilization was modelled using a commercial macromolecular network pyrolysis model (FG-DVC). For char oxidation an intrinsic reactivity approach including thermal annealing, ash inhibition and maceral effects, was used. Results from the simulations were compared against plant experimental values, showing a reasonable agreement in trends and quantitative values. 28 refs., 4 figs., 4 tabs.
Self-dual random-plaquette gauge model and the quantum toric code
Takeda, Koujin; Nishimori, Hidetoshi
2004-05-01
We study the four-dimensional Z2 random-plaquette lattice gauge theory as a model of topological quantum memory, the toric code in particular. In this model, the procedure of quantum error correction works properly in the ordered (Higgs) phase, and phase boundary between the ordered (Higgs) and disordered (confinement) phases gives the accuracy threshold of error correction. Using self-duality of the model in conjunction with the replica method, we show that this model has exactly the same mathematical structure as that of the two-dimensional random-bond Ising model, which has been studied very extensively. This observation enables us to derive a conjecture on the exact location of the multicritical point (accuracy threshold) of the model, pc=0.889972…, and leads to several nontrivial results including bounds on the accuracy threshold in three dimensions.
Self-dual random-plaquette gauge model and the quantum toric code
International Nuclear Information System (INIS)
Takeda, Koujin; Nishimori, Hidetoshi
2004-01-01
We study the four-dimensional Z 2 random-plaquette lattice gauge theory as a model of topological quantum memory, the toric code in particular. In this model, the procedure of quantum error correction works properly in the ordered (Higgs) phase, and phase boundary between the ordered (Higgs) and disordered (confinement) phases gives the accuracy threshold of error correction. Using self-duality of the model in conjunction with the replica method, we show that this model has exactly the same mathematical structure as that of the two-dimensional random-bond Ising model, which has been studied very extensively. This observation enables us to derive a conjecture on the exact location of the multicritical point (accuracy threshold) of the model, p c =0.889972..., and leads to several nontrivial results including bounds on the accuracy threshold in three dimensions
Comparison of a Coupled Near and Far Wake Model With a Free Wake Vortex Code
DEFF Research Database (Denmark)
Pirrung, Georg; Riziotis, Vasilis; Aagaard Madsen, Helge
2016-01-01
to be updated during the computation. Further, the effect of simplifying the exponential function approximation of the near wake model to increase the computation speed is investigated in this work. A modification of the dynamic inflow weighting factors of the far wake model is presented that ensures good...... computations performed using a free wake panel code. The focus of the description of the aerodynamics model is on the numerical stability, the computation speed and the accuracy of 5 unsteady simulations. To stabilize the near wake model, it has to be iterated to convergence, using a relaxation factor that has...... and a BEM model is centered around the NREL 5 MW reference turbine. The response to pitch steps at different pitching speeds is compared. By means of prescribed vibration cases, the effect of the aerodynamic model on the predictions of the aerodynamic work is investigated. The validation shows that a BEM...
Energy Technology Data Exchange (ETDEWEB)
Park, Ju Yeop; In, Wang Kee; Chun, Tae Hyun; Oh, Dong Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)
2000-02-01
The development of orthogonal 2-dimensional numerical code is made. The present code contains 9 kinds of turbulence models that are widely used. They include a standard k-{epsilon} model and 8 kinds of low Reynolds number ones. They also include 6 kinds of numerical schemes including 5 kinds of low order schemes and 1 kind of high order scheme such as QUICK. To verify the present numerical code, pipe flow, channel flow and expansion pipe flow are solved by this code with various options of turbulence models and numerical schemes and the calculated outputs are compared to experimental data. Furthermore, the discretization error that originates from the use of standard k-{epsilon} turbulence model with wall function is much more diminished by introducing a new grid system than a conventional one in the present code. 23 refs., 58 figs., 6 tabs. (Author)
International Nuclear Information System (INIS)
Jarnicki, R.; Sobiesiak, A.
2002-01-01
In order to solve the averaged conservation equations for turbulent reacting flow one is faced with a task of specifying the averaged chemical reaction rate. This is due to turbulence influence on the mean reaction rates that appear in the species concentration Reynolds-averaged equation. In order to investigate the Partially Stirred Reactor (PaSR) combustion model capabilities, a CFD modeling using KIVA3V Code with the PaSR model of two very different combustion processes, was performed. Experimental results were compared with modeling
A wide-range model of two-group gross sections in the dynamics code HEXTRAN
International Nuclear Information System (INIS)
Kaloinen, E.; Peltonen, J.
2002-01-01
In dynamic analyses the thermal hydraulic conditions within the reactor core may have a large variation, which sets a special requirement on the modeling of cross sections. The standard model in the dynamics code HEXTRAN is the same as in the static design code HEXBU-3D/MODS. It is based on a linear and second order fitting of two-group cross sections on fuel and moderator temperature, moderator density and boron density. A new, wide-range model of cross sections developed in Fortum Nuclear Services for HEXBU-3D/MOD6 has been included as an option into HEXTRAN. In this model the nodal cross sections are constructed from seven state variables in a polynomial of more than 40 terms. Coefficients of the polynomial are created by a least squares fitting to the results of a large number of fuel assembly calculations. Depending on the choice of state variables for the spectrum calculations, the new cross section model is capable to cover local conditions from cold zero power to boiling at full power. The 5. dynamic benchmark problem of AER is analyzed with the new option and results are compared to calculations with the standard model of cross sections in HEXTRAN (Authors)
Software Code Smell Prediction Model Using Shannon, Rényi and Tsallis Entropies
Directory of Open Access Journals (Sweden)
Aakanshi Gupta
2018-05-01
Full Text Available The current era demands high quality software in a limited time period to achieve new goals and heights. To meet user requirements, the source codes undergo frequent modifications which can generate the bad smells in software that deteriorate the quality and reliability of software. Source code of the open source software is easily accessible by any developer, thus frequently modifiable. In this paper, we have proposed a mathematical model to predict the bad smells using the concept of entropy as defined by the Information Theory. Open-source software Apache Abdera is taken into consideration for calculating the bad smells. Bad smells are collected using a detection tool from sub components of the Apache Abdera project, and different measures of entropy (Shannon, Rényi and Tsallis entropy. By applying non-linear regression techniques, the bad smells that can arise in the future versions of software are predicted based on the observed bad smells and entropy measures. The proposed model has been validated using goodness of fit parameters (prediction error, bias, variation, and Root Mean Squared Prediction Error (RMSPE. The values of model performance statistics ( R 2 , adjusted R 2 , Mean Square Error (MSE and standard error also justify the proposed model. We have compared the results of the prediction model with the observed results on real data. The results of the model might be helpful for software development industries and future researchers.
Sodium spray and jet fire model development within the CONTAIN-LMR code
International Nuclear Information System (INIS)
Scholtyssek, W.
1993-01-01
An assessment was made of the sodium spray fire model implemented in the CONTAIN code. The original droplet burn model, which was based on the NACOM code, was improved in several aspects, especially concerning evaluation of the droplet burning rate, reaction chemistry and heat balance, spray geometry and droplet motion, and consistency with CONTAIN standards of gas property evaluation. An additional droplet burning model based on a proposal by Krolikowski was made available to include the effect of the chemical equilibrium conditions at the flame temperature. The models were validated against single-droplet burn experiments as well as spray and jet fire experiments. Reasonable agreement was found between the two burn models and experimental data. When the gas temperature in the burning compartment reaches high values, the Krolikowski model seems to be preferable. Critical parameters for spray fire evaluation were found to be the spray characterization, especially the droplet size, which largely determines the burning efficiency, and heat transfer conditions at the interface between the atmosphere and structures, which controls the thermal hydraulic behavior in the burn compartment
International Nuclear Information System (INIS)
Bezrukov, Yu.A.; Shchekoldin, V.I.
2002-01-01
On the basis of the Gidropress OKB (Special Design Bureau) experimental data bank one verified the KORSAR code design models and correlations as to heat exchange crisis and overcrisis heat transfer as applied to the WWER reactor normal and emergency conditions. The VI.006.000 version of KORSAR code base calculations is shown to describe adequately the conducted experiments and to deviate insignificantly towards the conservative approach. So it may be considered as one of the codes ensuring more precise estimation [ru
Description of the Fokker-Plank code used to model ECRH of the Constance 2 plasma
International Nuclear Information System (INIS)
Mauel, M.E.
1982-01-01
The time-dependent Fokker-Plank code which is used to model the development of the electron velocity distribution during ECRH of the Constance 2 mirror-confined plasma is described in this report. The ECRH is modeled by the bounce-averaged quasilinear theory derived by Mauel. The effect of collisions are found by taking the appropriate gradients of the Rosenbluth potentials, and the electron distribution is advanced in time by using a modified alternating direction implicit (ADI) technique as explained by Killeen and Marx. The program was written in LISP to be run in the MACSYMA environment of the MACSYMA Consortium's PDP-10 computer
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
International nuclear model and code comparison on pre-equilibrium effects
International Nuclear Information System (INIS)
Gruppelaar, H.; van der Kamp, H.A.J.; Nagel, P.
1983-01-01
This paper gives the specification of an intercomparison of statistical nuclear models and codes with emphasis on pre-equilibrium effects. It is partly based upon the conclusions of a meeting of an ad-hoc working group on this subject. The parameters studied are: masses, Q values, level scheme data, optical model parameters, X-ray competition parameters, total level-density specifications, for 86 Rb, 89 Sr, 90 Y, 92 Y, 92 Zr, 93 Zr, 89 Y, 91 Nb, 92 Nb and 93 Nb
Coding Model and Mapping Method of Spherical Diamond Discrete Grids Based on Icosahedron
Directory of Open Access Journals (Sweden)
LIN Bingxian
2016-12-01
Full Text Available Discrete Global Grid(DGG provides a fundamental environment for global-scale spatial data's organization and management. DGG's encoding scheme, which blocks coordinate transformation between different coordination reference frames and reduces the complexity of spatial analysis, contributes a lot to the multi-scale expression and unified modeling of spatial data. Compared with other kinds of DGGs, Diamond Discrete Global Grid(DDGG based on icosahedron is beneficial to the spherical spatial data's integration and expression for much better geometric properties. However, its structure seems more complicated than DDGG on octahedron due to its initial diamond's edges cannot fit meridian and parallel. New challenges are posed when it comes to the construction of hierarchical encoding system and mapping relationship with geographic coordinates. On this issue, this paper presents a DDGG's coding system based on the Hilbert curve and designs conversion methods between codes and geographical coordinates. The study results indicate that this encoding system based on the Hilbert curve can express space scale and location information implicitly with the similarity between DDG and planar grid put into practice, and balances efficiency and accuracy of conversion between codes and geographical coordinates in order to support global massive spatial data's modeling, integrated management and all kinds of spatial analysis.
An Efficient Code-Based Threshold Ring Signature Scheme with a Leader-Participant Model
Directory of Open Access Journals (Sweden)
Guomin Zhou
2017-01-01
Full Text Available Digital signature schemes with additional properties have broad applications, such as in protecting the identity of signers allowing a signer to anonymously sign a message in a group of signers (also known as a ring. While these number-theoretic problems are still secure at the time of this research, the situation could change with advances in quantum computing. There is a pressing need to design PKC schemes that are secure against quantum attacks. In this paper, we propose a novel code-based threshold ring signature scheme with a leader-participant model. A leader is appointed, who chooses some shared parameters for other signers to participate in the signing process. This leader-participant model enhances the performance because every participant including the leader could execute the decoding algorithm (as a part of signing process upon receiving the shared parameters from the leader. The time complexity of our scheme is close to Courtois et al.’s (2001 scheme. The latter is often used as a basis to construct other types of code-based signature schemes. Moreover, as a threshold ring signature scheme, our scheme is as efficient as the normal code-based ring signature.
Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code
Energy Technology Data Exchange (ETDEWEB)
Stefanova, S; Vitkova, M; Simeonova, V; Passage, G; Manolova, M [Institute for Nuclear Research and Nuclear Energy, Sofia (Bulgaria); Haralampieva, Z [National Electric Company Ltd., Kozloduy (Bulgaria); Scheglov, A; Proselkov, V [Institute of Nuclear Reactors, RSC Kurchatov Inst., Moscow (Russian Federation)
1997-08-01
The report summarizes the first practical experience obtained by fuel rod performance modelling at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences. The results of application of the PIN-micro code and the code modification PINB1 for thermomechanical analysis of WWER-440 fuel assemblies (FAs) are presented. The aim of this analysis is to study the fuel rod behaviour of the operating WWER reactors. The performance of two FAs with maximal linear power and varying geometrical and technological parameters is analyzed. On the basis of recent publications on WWER fuel performance modelling at extended burnup, a modified PINB1 version of the standard PIN-micro code is shortly described and applied for the selected FAs. Comparison of the calculated results is performed. The PINB1 version predicts higher fuel temperatures and more adequate FGR rate, accounting for the extended burnup. The results presented in this paper prove the existence of sufficient safety margins, for the fuel performance limiting parameters during the whole considered period of core operation. (author). 8 refs, 16 figs, 1 tab.
Evolutionary modeling and prediction of non-coding RNAs in Drosophila.
Directory of Open Access Journals (Sweden)
Robert K Bradley
2009-08-01
Full Text Available We performed benchmarks of phylogenetic grammar-based ncRNA gene prediction, experimenting with eight different models of structural evolution and two different programs for genome alignment. We evaluated our models using alignments of twelve Drosophila genomes. We find that ncRNA prediction performance can vary greatly between different gene predictors and subfamilies of ncRNA gene. Our estimates for false positive rates are based on simulations which preserve local islands of conservation; using these simulations, we predict a higher rate of false positives than previous computational ncRNA screens have reported. Using one of the tested prediction grammars, we provide an updated set of ncRNA predictions for D. melanogaster and compare them to previously-published predictions and experimental data. Many of our predictions show correlations with protein-coding genes. We found significant depletion of intergenic predictions near the 3' end of coding regions and furthermore depletion of predictions in the first intron of protein-coding genes. Some of our predictions are colocated with larger putative unannotated genes: for example, 17 of our predictions showing homology to the RFAM family snoR28 appear in a tandem array on the X chromosome; the 4.5 Kbp spanned by the predicted tandem array is contained within a FlyBase-annotated cDNA.
Modelling of WWER-440 fuel rod behaviour under operational conditions with the PIN-micro code
International Nuclear Information System (INIS)
Stefanova, S.; Vitkova, M.; Simeonova, V.; Passage, G.; Manolova, M.; Haralampieva, Z.; Scheglov, A.; Proselkov, V.
1997-01-01
The report summarizes the first practical experience obtained by fuel rod performance modelling at the Institute for Nuclear Research and Nuclear Energy, Bulgarian Academy of Sciences. The results of application of the PIN-micro code and the code modification PINB1 for thermomechanical analysis of WWER-440 fuel assemblies (FAs) are presented. The aim of this analysis is to study the fuel rod behaviour of the operating WWER reactors. The performance of two FAs with maximal linear power and varying geometrical and technological parameters is analyzed. On the basis of recent publications on WWER fuel performance modelling at extended burnup, a modified PINB1 version of the standard PIN-micro code is shortly described and applied for the selected FAs. Comparison of the calculated results is performed. The PINB1 version predicts higher fuel temperatures and more adequate FGR rate, accounting for the extended burnup. The results presented in this paper prove the existence of sufficient safety margins, for the fuel performance limiting parameters during the whole considered period of core operation. (author). 8 refs, 16 figs, 1 tab
Progress modelling of aqueous electrons and hydroxyl radicals in RAIM code
Energy Technology Data Exchange (ETDEWEB)
Kim, A Yeong; Kim, Han-Chul; Lee, Jongseong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2015-10-15
In this paper, the RAIM code was revised minutely with regards to aqueous electrons and hydroxyl radicals, and simulated the P10T2 test. The recent study indicated that the RAIM had the potential for improvement of simulating the iodine behavior influenced by water radiolysis products such as aqueous electrons and hydroxyl radicals. In the existing RAIM modelling, it was considered that aqueous electrons only interacted with oxygen as a consumption reaction, but the reaction with hydrogen peroxide also could be major contributor to the iodine behavior as well as the consumption reaction of aqueous electrons. In case of hydroxyl radicals, RAIM took no notice of the pH impact. In other words, it dealt with the consumption reaction constants but not as a variable of pH. In this communication, the procedures to develop the model related to aqueous electrons and hydroxyl radicals in RAIM will be addressed. And the upgraded RAIM (RAIM-1, 2, 3) codes were applied to OECD-BIP P10T2 test which showed the effect of pH on the iodine behavior and compared with the existing RAIM1.8.3 code. Comparing with the existing RAIM, the improvement reduced the difference about 10%. However, the absolute difference values that is about one order at pH 10 could not be reduced by this approach.
Energy Technology Data Exchange (ETDEWEB)
Yoon, C.; Rhee, B. W.; Chung, B. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Y. J.; Kim, M. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2008-05-15
Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use vendor's code for regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of RELAP5/MOD3/CANDU code to MARS code including quality assurance of the developed models. This first part of the research series presents the implementation and verification of the Wolsong pump model, the pressure tube deformation model, and the off-take model for arbitrary-angled branch pipes.
International Nuclear Information System (INIS)
Yoon, C.; Rhee, B. W.; Chung, B. D.; Cho, Y. J.; Kim, M. W.
2008-01-01
Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use vendor's code for regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of RELAP5/MOD3/CANDU code to MARS code including quality assurance of the developed models. This first part of the research series presents the implementation and verification of the Wolsong pump model, the pressure tube deformation model, and the off-take model for arbitrary-angled branch pipes
A new balance-of-plant model for the SASSYS-1 LMR systems analysis code
International Nuclear Information System (INIS)
Briggs, L.L.
1989-01-01
A balance-of-plant model has been added to the SASSYS-1 liquid metal reactor systems analysis code. Until this addition, the only waterside component which SASSYS-1 could explicitly model was the water side of a steam generator, with the remainder of the water side represented by boundary conditions on the steam generator. The balance-of-plant model is based on the model used for the sodium side of the plant. It will handle subcooled liquid water, superheated steam, and saturated two-phase fluid. With the exception of heated flow paths in heaters, the model assumes adiabatic conditions along flow paths; this assumption simplifies the solution procedure while introducing very little error for a wide range of reactor plant problems. Only adiabatic flow is discussed in this report. 3 refs., 4 figs
Modelling and description of PHEBUS FPT1 experiment with the computer code ASTEC
International Nuclear Information System (INIS)
Tusheva, P.; Kalchev, B.
2005-01-01
The PHEBUS Fission Product (FP) programme was initiated in 1988 after major severe reactor accidents (at Three Mile Island and Chernobyl). The main objective of the programme is to study the release, transport and retention of fission products in an in-pile facility under severe accident conditions in a LWR. This paper covers the FPT 1 experiment description and modelling by the ASTEC code. The main calculated events, the temperature evolution at the middle part of the test bundle, the state of the bundle degradation at the end of calculation and the calculated Hydrogen production are presented and discussed. The obtained results from the ASTEC code calculation show a good agreement between experimental and calculated results. The calculated Hydrogen production is slightly overestimated in comparison with the experimental results
Modeling of pipe break accident in a district heating system using RELAP5 computer code
International Nuclear Information System (INIS)
Kaliatka, A.; Valinčius, M.
2012-01-01
Reliability of a district heat supply system is a very important factor. However, accidents are inevitable and they occur due to various reasons, therefore it is necessary to have possibility to evaluate the consequences of possible accidents. This paper demonstrated the capabilities of developed district heating network model (for RELAP5 code) to analyze dynamic processes taking place in the network. A pipe break in a water supply line accident scenario in Kaunas city (Lithuania) heating network is presented in this paper. The results of this case study were used to demonstrate a possibility of the break location identification by pressure decrease propagation in the network. -- Highlights: ► Nuclear reactor accident analysis code RELAP5 was applied for accident analysis in a district heating network. ► Pipe break accident scenario in Kaunas city (Lithuania) district heating network has been analyzed. ► An innovative method of pipe break location identification by pressure-time data is proposed.
International Nuclear Information System (INIS)
Guillermier, Pierre; Daniel, Lucile; Gauthier, Laurent
2009-01-01
To support AREVA NP in its design on HTR reactor and its HTR fuel R and D program, the Commissariat a l'Energie Atomique developed the ATLAS code (Advanced Thermal mechanicaL Analysis Software) with the objectives: - to quantify, with a statistical approach, the failed particle fraction and fission product release of a HTR fuel core under normal and accidental conditions (compact or pebble design). - to simulate irradiation tests or benchmark in order to compare measurements or others code results with ATLAS evaluation. These two objectives aim at qualifying the code in order to predict fuel behaviour and to design fuel according to core performance and safety requirements. A statistical calculation uses numerous deterministic calculations. The finite element method is used for these deterministic calculations, in order to be able to choose among three types of meshes, depending on what must be simulated: - One-dimensional calculation of one single particle, for intact particles or particles with fully debonded layers. - Two-dimensional calculations of one single particle, in the case of particles which are cracked, partially debonded or shaped in various ways. - Three-dimensional calculations of a whole compact slice, in order to simulate the interactions between the particles, the thermal gradient and the transport of fission products up to the coolant. - Some calculations of a whole pebble, using homogenization methods are being studied. The temperatures, displacements, stresses, strains and fission product concentrations are calculated on each mesh of the model. Statistical calculations are done using these results, taking into account ceramic failure mode, but also fabrication tolerances and material property uncertainties, variations of the loads (fluence, temperature, burn-up) and core data parameters. The statistical method used in ATLAS is the importance sampling. The model of migration of long-lived fission products in the coated particle and more
Trading speed and accuracy by coding time: a coupled-circuit cortical model.
Directory of Open Access Journals (Sweden)
Dominic Standage
2013-04-01
Full Text Available Our actions take place in space and time, but despite the role of time in decision theory and the growing acknowledgement that the encoding of time is crucial to behaviour, few studies have considered the interactions between neural codes for objects in space and for elapsed time during perceptual decisions. The speed-accuracy trade-off (SAT provides a window into spatiotemporal interactions. Our hypothesis is that temporal coding determines the rate at which spatial evidence is integrated, controlling the SAT by gain modulation. Here, we propose that local cortical circuits are inherently suited to the relevant spatial and temporal coding. In simulations of an interval estimation task, we use a generic local-circuit model to encode time by 'climbing' activity, seen in cortex during tasks with a timing requirement. The model is a network of simulated pyramidal cells and inhibitory interneurons, connected by conductance synapses. A simple learning rule enables the network to quickly produce new interval estimates, which show signature characteristics of estimates by experimental subjects. Analysis of network dynamics formally characterizes this generic, local-circuit timing mechanism. In simulations of a perceptual decision task, we couple two such networks. Network function is determined only by spatial selectivity and NMDA receptor conductance strength; all other parameters are identical. To trade speed and accuracy, the timing network simply learns longer or shorter intervals, driving the rate of downstream decision processing by spatially non-selective input, an established form of gain modulation. Like the timing network's interval estimates, decision times show signature characteristics of those by experimental subjects. Overall, we propose, demonstrate and analyse a generic mechanism for timing, a generic mechanism for modulation of decision processing by temporal codes, and we make predictions for experimental verification.
Directory of Open Access Journals (Sweden)
Peter eVuust
2014-10-01
Full Text Available Musical rhythm, consisting of apparently abstract intervals of accented temporal events, has a remarkable capacity to move our minds and bodies. How does the cognitive system enable our experiences of rhythmically complex music? In this paper, we describe some common forms of rhythmic complexity in music and propose the theory of predictive coding as a framework for understanding how rhythm and rhythmic complexity are processed in the brain. We also consider why we feel so compelled by rhythmic tension in music. First, we consider theories of rhythm and meter perception, which provide hierarchical and computational approaches to modeling. Second, we present the theory of predictive coding, which posits a hierarchical organization of brain responses reflecting fundamental, survival-related mechanisms associated with predicting future events. According to this theory, perception and learning is manifested through the brain’s Bayesian minimization of the error between the input to the brain and the brain’s prior expectations. Third, we develop a predictive coding model of musical rhythm, in which rhythm perception is conceptualized as an interaction between what is heard (‘rhythm’ and the brain’s anticipatory structuring of music (‘meter’. Finally, we review empirical studies of the neural and behavioral effects of syncopation, polyrhythm and groove, and propose how these studies can be seen as special cases of the predictive coding theory. We argue that musical rhythm exploits the brain’s general principles of prediction and propose that pleasure and desire for sensorimotor synchronization from musical rhythm may be a result of such mechanisms.
International Nuclear Information System (INIS)
Dominguez, L.; Camargo, C.T.M.
1984-09-01
The first step of the project for implementation of two non-symmetric cooling loops modeled by the ALMOD3 computer code is presented. This step consists of the introduction of a simplified model for simulating the steam generator. This model is the GEVAP computer code, integrant part of LOOP code, which simulates the primary coolant circuit of PWR nuclear power plants during transients. The ALMOD3 computer code has a model for the steam generator, called UTSG, which is very detailed. This model has spatial dependence, correlations for 2-phase flow, distinguished correlations for different heat transfer process. The GEVAP model has thermal equilibrium between phases (gaseous and liquid homogeneous mixture), no spatial dependence and uses only one generalized correlation to treat several heat transfer processes. (Author) [pt
Kaptein, S.P.; Schwartz, M.S.
2007-01-01
textabstractBusiness codes are a widely used management instrument. Research into the effectiveness of business codes has, however, produced conflicting results. The main reasons for the divergent findings are: varying definitions of key terms; deficiencies in the empirical data and methodologies used; and a lack of theory. In this paper, we propose an integrated research model and suggest directions for future research.
Hickok, Gregory
2012-01-01
Speech recognition is an active process that involves some form of predictive coding. This statement is relatively uncontroversial. What is less clear is the source of the prediction. The dual-stream model of speech processing suggests that there are two possible sources of predictive coding in speech perception: the motor speech system and the…
Validation of full core geometry model of the NODAL3 code in the PWR transient Benchmark problems
International Nuclear Information System (INIS)
T-M Sembiring; S-Pinem; P-H Liem
2015-01-01
The coupled neutronic and thermal-hydraulic (T/H) code, NODAL3 code, has been validated in some PWR static benchmark and the NEACRP PWR transient benchmark cases. However, the NODAL3 code have not yet validated in the transient benchmark cases of a control rod assembly (CR) ejection at peripheral core using a full core geometry model, the C1 and C2 cases. By this research work, the accuracy of the NODAL3 code for one CR ejection or the unsymmetrical group of CRs ejection case can be validated. The calculations by the NODAL3 code have been carried out by the adiabatic method (AM) and the improved quasistatic method (IQS). All calculated transient parameters by the NODAL3 code were compared with the reference results by the PANTHER code. The maximum relative difference of 16 % occurs in the calculated time of power maximum parameter by using the IQS method, while the relative difference of the AM method is 4 % for C2 case. All calculation results by the NODAL3 code shows there is no systematic difference, it means the neutronic and T/H modules are adopted in the code are considered correct. Therefore, all calculation results by using the NODAL3 code are very good agreement with the reference results. (author)
Rivier, Leonard Gilles
Using an efficient parallel code solving the primitive equations of atmospheric dynamics, the jet structure of a Jupiter like atmosphere is modeled. In the first part of this thesis, a parallel spectral code solving both the shallow water equations and the multi-level primitive equations of atmospheric dynamics is built. The implementation of this code called BOB is done so that it runs effectively on an inexpensive cluster of workstations. A one dimensional decomposition and transposition method insuring load balancing among processes is used. The Legendre transform is cache-blocked. A "compute on the fly" of the Legendre polynomials used in the spectral method produces a lower memory footprint and enables high resolution runs on relatively small memory machines. Performance studies are done using a cluster of workstations located at the National Center for Atmospheric Research (NCAR). BOB performances are compared to the parallel benchmark code PSTSWM and the dynamical core of NCAR's CCM3.6.6. In both cases, the comparison favors BOB. In the second part of this thesis, the primitive equation version of the code described in part I is used to study the formation of organized zonal jets and equatorial superrotation in a planetary atmosphere where the parameters are chosen to best model the upper atmosphere of Jupiter. Two levels are used in the vertical and only large scale forcing is present. The model is forced towards a baroclinically unstable flow, so that eddies are generated by baroclinic instability. We consider several types of forcing, acting on either the temperature or the momentum field. We show that only under very specific parametric conditions, zonally elongated structures form and persist resembling the jet structure observed near the cloud level top (1 bar) on Jupiter. We also study the effect of an equatorial heat source, meant to be a crude representation of the effect of the deep convective planetary interior onto the outer atmospheric layer. We
Learning to Act Like a Lawyer: A Model Code of Professional Responsibility for Law Students
Directory of Open Access Journals (Sweden)
David M. Tanovich
2009-02-01
incidents at law schools that raise serious issues about the professionalism of law students. They include, for example, the UofT marks scandal, the Windsor first year blog and the proliferation of blogs like www.lawstudents.ca and www.lawbuzz.ca with gratuitous, defamatory and offensive entries. It is not clear that all of this conduct would be caught by University codes of conduct which often limit their reach to on campus behaviour or University sanctioned events. What should a law school code of professional responsibility look like and what ethical responsibilities should it identify? For example, should there be a mandatory pro bono obligation on students or a duty to report misconduct. The last part of the article addresses this question by setting out a model code of professional responsibility for law students. Les étudiants et étudiantes en droit constituent l’avenir de la profession juridique. Comment bien préparés sont-ils lorsqu’ils quittent la faculté de droit pour assumer leurs obligations professionnelles et éthiques envers eux-mêmes, envers la profession et envers le public? Cette question a mené à un intérêt grandissant au Canada à l’enseignement de l’éthique juridique. Elle a aussi mené à plus d’emphase sur le développement de formation clinique et expérientielle tel que l’exemplifie le savoir et l’enseignement de la professeure Rose Voyvodic. Toutefois, moins d’attention a été consacrée à identifier les responsabilités éthiques générales d’étudiants et étudiantes en droit lorsqu’ils n’oeuvrent pas dans une clinique ou dans un autre contexte légal. Cela se voit dans les faits qu’il y a très peu d’articles canadiens qui portent sur la question, et, de plus grande importance, qu’il y a pénurie, au sein de facultés de droit, de politiques disciplinaires ou de codes déontologiques qui présentent les obligations professionnelles d’étudiants et étudiantes en droit. Cet article développe une id
An Empirical Model for Vane-Type Vortex Generators in a Navier-Stokes Code
Dudek, Julianne C.
2005-01-01
An empirical model which simulates the effects of vane-type vortex generators in ducts was incorporated into the Wind-US Navier-Stokes computational fluid dynamics code. The model enables the effects of the vortex generators to be simulated without defining the details of the geometry within the grid, and makes it practical for researchers to evaluate multiple combinations of vortex generator arrangements. The model determines the strength of each vortex based on the generator geometry and the local flow conditions. Validation results are presented for flow in a straight pipe with a counter-rotating vortex generator arrangement, and the results are compared with experimental data and computational simulations using a gridded vane generator. Results are also presented for vortex generator arrays in two S-duct diffusers, along with accompanying experimental data. The effects of grid resolution and turbulence model are also examined.
The effect of turbulent mixing models on the predictions of subchannel codes
International Nuclear Information System (INIS)
Tapucu, A.; Teyssedou, A.; Tye, P.; Troche, N.
1994-01-01
In this paper, the predictions of the COBRA-IV and ASSERT-4 subchannel codes have been compared with experimental data on void fraction, mass flow rate, and pressure drop obtained for two interconnected subchannels. COBRA-IV is based on a one-dimensional separated flow model with the turbulent intersubchannel mixing formulated as an extension of the single-phase mixing model, i.e. fluctuating equal mass exchange. ASSERT-4 is based on a drift flux model with the turbulent mixing modelled by assuming an exchange of equal volumes with different densities thus allowing a net fluctuating transverse mass flux from one subchannel to the other. This feature is implemented in the constitutive relationship for the relative velocity required by the conservation equations. It is observed that the predictions of ASSERT-4 follow the experimental trends better than COBRA-IV; therefore the approach of equal volume exchange constitutes an improvement over that of the equal mass exchange. ((orig.))
Pitchcontrol of wind turbines using model free adaptivecontrol based on wind turbine code
DEFF Research Database (Denmark)
Zhang, Yunqian; Chen, Zhe; Cheng, Ming
2011-01-01
value is only based on I/O data of the wind turbine is identified and then the wind turbine system is replaced by a dynamic linear time-varying model. In order to verify the correctness and robustness of the proposed model free adaptive pitch controller, the wind turbine code FAST which can predict......As the wind turbine is a nonlinear high-order system, to achieve good pitch control performance, model free adaptive control (MFAC) approach which doesn't need the mathematical model of the wind turbine is adopted in the pitch control system in this paper. A pseudo gradient vector whose estimation...... the wind turbine loads and response in high accuracy is used. The results show that the controller produces good dynamic performance, good robustness and adaptability....
Validation of film dryout model in a three-fluid code FIDAS
International Nuclear Information System (INIS)
Sugawara, Satoru
1989-11-01
Analytical prediction model of critical heat flux (CHF) has been developed on the basis of film dryout criterion due to droplets deposition and entrainment in annular mist flow. CHF in round tubes were analyzed by the Film Dryout Analysis Code in Subchannels, FIDAS, which is based on the three-fluid, three-field and newly developed film dryout model. Predictions by FIDAS were compared with the world-wide experimental data on CHF obtained in water and Freon for uniformly and non-uniformly heated tubes under vertical upward flow condition. Furthermore, CHF prediction capability of FIDAS was compared with those of other film dryout models for annular flow and Katto's CHF correlation. The predictions of FIDAS are in sufficient agreement with the experimental CHF data, and indicate better agreement than the other film dryout models and empirical correlation of Katto. (author)
International Nuclear Information System (INIS)
Simmons, C.S.; Cole, C.R.
1985-05-01
This document was written to provide guidance to managers and site operators on how ground-water transport codes should be selected for assessing burial site performance. There is a need for a formal approach to selecting appropriate codes from the multitude of potentially useful ground-water transport codes that are currently available. Code selection is a problem that requires more than merely considering mathematical equation-solving methods. These guidelines are very general and flexible and are also meant for developing systems simulation models to be used to assess the environmental safety of low-level waste burial facilities. Code selection is only a single aspect of the overall objective of developing a systems simulation model for a burial site. The guidance given here is mainly directed toward applications-oriented users, but managers and site operators need to be familiar with this information to direct the development of scientifically credible and defensible transport assessment models. Some specific advice for managers and site operators on how to direct a modeling exercise is based on the following five steps: identify specific questions and study objectives; establish costs and schedules for achieving answers; enlist the aid of professional model applications group; decide on approach with applications group and guide code selection; and facilitate the availability of site-specific data. These five steps for managers/site operators are discussed in detail following an explanation of the nine systems model development steps, which are presented first to clarify what code selection entails
H2-O2 supercritical combustion modeling using a CFD code
Directory of Open Access Journals (Sweden)
Benarous Abdallah
2009-01-01
Full Text Available The characteristics of propellant injection, mixing, and combustion have a profound effect on liquid rocket engine performance. The necessity of raising rocket engines performance requires a combustion chamber operation often in a supercritical regime. A supercritical combustion model based on a one-phase multi-components approach is developed and tested on a non-premixed H2-O2 flame configuration. A two equations turbulence model is used for describing the jet dynamics where a limited Pope correction is added to account for the oxidant spreading rate. Transport properties of the mixture are calculated using extended high pressure forms of the mixing rules. An equilibrium chemistry scheme is adopted in this combustion case, with both algebraic and stochastic expressions for the chemistry/turbulence coupling. The model was incorporated into a computational fluid dynamics commercial code (Fluent 6.2.16. The validity of the present model was investigated by comparing predictions of temperature, species mass fractions, recirculation zones and visible flame length to the experimental data measured on the Mascotte test rig. The results were confronted also with advanced code simulations. It appears that the agreement between the results was fairly good in the chamber regions situated downstream the near injection zone.
RELAP5-3D Code Includes ATHENA Features and Models
International Nuclear Information System (INIS)
Riemke, Richard A.; Davis, Cliff B.; Schultz, Richard R.
2006-01-01
Version 2.3 of the RELAP5-3D computer program includes all features and models previously available only in the ATHENA version of the code. These include the addition of new working fluids (i.e., ammonia, blood, carbon dioxide, glycerol, helium, hydrogen, lead-bismuth, lithium, lithium-lead, nitrogen, potassium, sodium, and sodium-potassium) and a magnetohydrodynamic model that expands the capability of the code to model many more thermal-hydraulic systems. In addition to the new working fluids along with the standard working fluid water, one or more noncondensable gases (e.g., air, argon, carbon dioxide, carbon monoxide, helium, hydrogen, krypton, nitrogen, oxygen, SF 6 , xenon) can be specified as part of the vapor/gas phase of the working fluid. These noncondensable gases were in previous versions of RELAP5-3D. Recently four molten salts have been added as working fluids to RELAP5-3D Version 2.4, which has had limited release. These molten salts will be in RELAP5-3D Version 2.5, which will have a general release like RELAP5-3D Version 2.3. Applications that use these new features and models are discussed in this paper. (authors)
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.
A self-organized internal models architecture for coding sensory-motor schemes
Directory of Open Access Journals (Sweden)
Esaú eEscobar Juárez
2016-04-01
Full Text Available Cognitive robotics research draws inspiration from theories and models on cognition, as conceived by neuroscience or cognitive psychology, to investigate biologically plausible computational models in artificial agents. In this field, the theoretical framework of Grounded Cognition provides epistemological and methodological grounds for the computational modeling of cognition. It has been stressed in the literature that textit{simulation}, textit{prediction}, and textit{multi-modal integration} are key aspects of cognition and that computational architectures capable of putting them into play in a biologically plausible way are a necessity.Research in this direction has brought extensive empirical evidencesuggesting that textit{Internal Models} are suitable mechanisms forsensory-motor integration. However, current Internal Models architectures show several drawbacks, mainly due to the lack of a unified substrate allowing for a true sensory-motor integration space, enabling flexible and scalable ways to model cognition under the embodiment hypothesis constraints.We propose the Self-Organized Internal ModelsArchitecture (SOIMA, a computational cognitive architecture coded by means of a network of self-organized maps, implementing coupled internal models that allow modeling multi-modal sensory-motor schemes. Our approach addresses integrally the issues of current implementations of Internal Models.We discuss the design and features of the architecture, and provide empirical results on a humanoid robot that demonstrate the benefits and potentialities of the SOIMA concept for studying cognition in artificial agents.
A fifth equation to model the relative velocity the 3-D thermal-hydraulic code THYC
International Nuclear Information System (INIS)
Jouhanique, T.; Rascle, P.
1995-11-01
E.D.F. has developed, since 1986, a general purpose code named THYC (Thermal HYdraulic Code) designed to study three-dimensional single and two-phase flows in rod tube bundles (pressurised water reactor cores, steam generators, condensers, heat exchangers). In these studies, the relative velocity was calculated by a drift-flux correlation. However, the relative velocity between vapor and liquid is an important parameter for the accuracy of a two-phase flow modelling in a three-dimensional code. The range of application of drift-flux correlations is mainly limited by the characteristic of the flow pattern (counter current flow ...) and by large 3-D effects. The purpose of this paper is to describe a numerical scheme which allows the relative velocity to be computed in a general case. Only the methodology is investigated in this paper which is not a validation work. The interfacial drag force is an important factor of stability and accuracy of the results. This force, closely dependent on the flow pattern, is not entirely established yet, so a range of multiplicator of its expression is used to compare the numerical results with the VATICAN test section measurements. (authors). 13 refs., 6 figs
International Nuclear Information System (INIS)
Hong, In Seob; Suk, Ho Chun; Kim, Soon Young; Jo, Chang Keun
2002-06-01
The major objective of this research is to validate the incremental cross section property of DRAGON code in connection with WIMS-AECL/DRAGON/RFSP code system with ENDF/B-VI library and full 2G calculation model. The direct comparison between the incremental cross section results calculated by DRAGON with ENDF/B-VI and ENDF/B-V and MULTICELL with ENDF/B-V indicate that there are not much differences between the incremental cross sections of DRAGON with ENDF/B-V and ENDF/B-VI, but there exists large discrepancies between the results of DRAGON and those of MULTICELL. In the analysis of the difference between calculated and measured reactivity worths of various types of control devices during Phase-B Post-Simulation of Wolsong Units 2, 3 and 4, WIMS-AECL/DRAGON/RFSP analysis well agrees with those of previous WIMS-AECL /MULTICELL/RFSP analysis within very small differences. From those results, we can conclude that DRAGON code can be used as a general purpose incremental cross section generation tool for not only the natural uranium fuel but also slightly enriched fuel such as RU or SEU, to cover the shortcomings of natural uranium based MULTICELL code
International Nuclear Information System (INIS)
Lee, J.H.; Park, G.C.; Cho, H.K.
2015-01-01
In the containment of a nuclear reactor, the wall condensation occurs when containment cooling system and structures remove the mass and energy release and this phenomenon is of great importance to ensure containment integrity. If the phenomenon occurs in the presence of non-condensable gases, their accumulation near the condensate film leads to significant reduction in heat transfer during the condensation. This study aims at simulating the wall film condensation in the presence of non-condensable gas using CUPID, a computational multi-fluid dynamics code, which is developed by the Korea Atomic Energy Research Institute (KAERI) for the analysis of transient two-phase flows in nuclear reactor components. In order to simulate the wall film condensation in containment, the code requires a proper wall condensation model and liquid film model applicable to the analysis of the large scale system. In the present study, the liquid film model and wall film condensation model were implemented in the two-fluid model of CUPID. For the condensation simulation, a wall function approach with heat and mass transfer analogy was applied in order to save computational time without considerable refinement for the boundary layer. This paper presents the implemented wall film condensation model and then, introduces the simulation result using CUPID with the model for a conceptual condensation problem in a large system. (authors)
International Nuclear Information System (INIS)
Bott, E.; Frepoli, C.; Monti, R.; Notini, V.; Carcassi, M.; Fineschi, F.; Heitsch, M.
1999-01-01
Large amounts of hydrogen can be generated in the containment of a nuclear power plant following a postulated accident with significant fuel damage. Different strategies have been proposed and implemented to prevent violent hydrogen combustion. An attractive one aims to eliminate hydrogen without burning processes; it is based on the use of catalytic hydrogen recombiners. This paper describes a simulation methodology which is being developed by Ansaldo, to support the application of the above strategy, in the frame of two projects sponsored by the Commission of the European Communities within the IV Framework Program on Reactor Safety. Involved organizations also include the DCMN of Pisa University (Italy), Battelle Institute and GRS (Germany), Politechnical University of Madrid (Spain). The aims to make available a simulation approach, suitable for use for containment design at industrial level (i.e. with reasonable computer running time) and capable to correctly capture the relevant phenomenologies (e.g. multiflow convective flow patterns, hydrogen, air and steam distribution in the containment atmosphere as determined by containment structures and geometries as well as by heat and mass sources and sinks). Eulerian algorithms provide the capability of three dimensional modelling with a fairly accurate prediction, however lower than CFD codes with a full Navier Stokes formulation. Open linking of an Eulerian code as GOTHIC to a full Navier Stokes CFD code as CFX 4.1 allows to dynamically tune the solving strategies of the Eulerian code itself. The effort in progress is an application of this innovative methodology to detailed hydrogen recombination simulation and a validation of the approach itself by reproducing experimental data. (author)
International Nuclear Information System (INIS)
Carvalho, F. de A.T. de.
1985-01-01
Some antecipated transients without scram (ATWS) for a pressurized water cooled reactor, model KWU 1300 MWe, are studied using coupling of the containment code CORAN to the system model code ALMOD, under severe random conditions. This coupling has the objective of including containment model as part of a unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle a failure in the closure of the pressurizer relief valve was also investigated. For the beginning of the cycle, the containment participates actively during the transient. It is noted that the effect of the burn-up in the fuel is to reduce the seriousness of these transients. On the other hand, the failure in the closure of the pressurized relief valve makes this transients more severe. Moreover, the containment safety or radiological public safety is not affected in any of the cases. (Author) [pt
Compressor Modeling for Transient Analysis of Supercritical CO2 Brayton Cycle by using MARS code
Energy Technology Data Exchange (ETDEWEB)
Park, Joo Hyun; Park, Hyun Sun; Kim, Tae Ho; Kwon, Jin Gyu [POSTECH, Pohang (Korea, Republic of); Bae, Sung Won; Cha, Jae Eun [KAERI, Daejeon (Korea, Republic of)
2016-05-15
In this study, SCIEL (Supercritical CO{sub 2} Integral Experimental Loop) was chosen as a reference loop and the MARS code was as the transient cycle analysis code. As a result, the compressor homologous curve was developed from the SCIEL experimental data and MARS analysis was performed and presented in the paper. The advantages attract SCO{sub 2}BC as a promising next generation power cycles. The high thermal efficiency comes from the operation of compressor near the critical point where the properties of SCO{sub 2}. The approaches to those of liquid phase, leading drastically lower the compression work loss. However, the advantage requires precise and smooth operation of the cycle near the critical point. However, it is one of the key technical challenges. The experimental data was steady state at compressor rotating speed of 25,000 rpm. The time, 3133 second, was starting point of steady state. Numerical solutions were well matched with the experimental data. The mass flow rate from the MARS analysis of approximately 0.7 kg/s was close to the experimental result of 0.9 kg/s. It is expected that the difference come from the measurement error in the experiment. In this study, the compressor model was developed and implemented in MARS to study the transient analysis of SCO{sub 2}BC in SCIEL. We obtained the homologous curves for the SCIEL compressor using experimental data and performed nodalization of the compressor model using MARS code. In conclusions, it was found that numerical solutions from the MARS model were well matched with experimental data.
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
International Nuclear Information System (INIS)
Park, Hyun Sik; Choi, Ki Yong; Moon, Sang Ki; Kim, Jung Woo; Kim, Kyung Doo
2009-01-01
The wall condensation heat transfer models are developed for the SPACE code and are assessed for various condensation conditions. Both default and alternative models were selected through an extensive literature survey. For a pure steam condensation, a maximum value among the Nusselt, Chato, and Shah's correlations is used in order to consider the geometric and turbulent effects. In the presence of non-condensable gases, the Colburn-Hougen's diffusion model was used as a default model and a non-iterative condensation model proposed by No and Park was selected as an alternative model. The wall condensation heat transfer models were assessed preliminarily by using arbitrary test conditions. Both wall condensation models could simulate the heat transfer coefficients and heat fluxes in the vertical, horizontal and turbulent conditions quite reasonably for a pure steam condensation. Both the default and alternative wall condensation models were also verified for the condensation heat transfer coefficient and heat flux in the presence of noncondensable gas. However, some improvements and further detailed verification are necessary for the condensation phenomena in the presence of noncondensable gas
International Nuclear Information System (INIS)
Strenge, D.L.; Watson, E.C.; Droppo, J.G.
1976-06-01
The development of technological bases for siting nuclear fuel cycle facilities requires calculational models and computer codes for the evaluation of risks and the assessment of environmental impact of radioactive effluents. A literature search and review of available computer programs revealed that no one program was capable of performing all of the great variety of calculations (i.e., external dose, internal dose, population dose, chronic release, accidental release, etc.). Available literature on existing computer programs has been reviewed and a description of each program reviewed is given
Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code
International Nuclear Information System (INIS)
Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P
2004-01-01
Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z eff . Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values
Modelling of Ohmic discharges in ADITYA tokamak using the Tokamak Simulation Code
Energy Technology Data Exchange (ETDEWEB)
Bandyopadhyay, I; Ahmed, S M; Atrey, P K; Bhatt, S B; Bhattacharya, R; Chaudhury, M B; Deshpande, S P; Gupta, C N; Jha, R; Joisa, Y Shankar; Kumar, Vinay; Manchanda, R; Raju, D; Rao, C V S; Vasu, P [Institute for Plasma Research, Bhat, Gandhinagar 382428 (India)
2004-09-01
Several Ohmic discharges of the ADITYA tokamak are simulated using the Tokamak Simulation Code (TSC), similar to that done earlier for the TFTR tokamak. Unlike TFTR, the dominant radiation process in ADITYA is through impurity line radiation. TSC can follow the experimental plasma current and position to very good accuracy. The thermal transport model of TSC including impurity line radiation gives a good match of the simulated results with experimental data for the Ohmic flux consumption, electron temperature and Z{sub eff}. Even the simulated magnetic probe signals are in reasonably good agreement with the experimental values.