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Sample records for simulation code mcnpx

  1. Computed radiography simulation using the Monte Carlo code MCNPX

    International Nuclear Information System (INIS)

    Correa, S.C.A.; Souza, E.M.; Silva, A.X.; Lopes, R.T.

    2009-01-01

    Simulating x-ray images has been of great interest in recent years as it makes possible an analysis of how x-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed. The sensitivity curve of the BaFBr image plate detector as well as the characteristic noise of a 16-bit computed radiography system were considered during the methodology's development. The results obtained confirm that the proposed procedure for simulating computed radiographic images is satisfactory, as it allows obtaining results comparable with experimental data. (author)

  2. Computed radiography simulation using the Monte Carlo code MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Correa, S.C.A. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Centro Universitario Estadual da Zona Oeste (CCMAT)/UEZO, Av. Manuel Caldeira de Alvarenga, 1203, Campo Grande, 23070-200, Rio de Janeiro, RJ (Brazil); Souza, E.M. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Silva, A.X., E-mail: ademir@con.ufrj.b [PEN/COPPE-DNC/Poli CT, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil); Cassiano, D.H. [Instituto de Radioprotecao e Dosimetria/CNEN Av. Salvador Allende, s/n, Recreio, 22780-160, Rio de Janeiro, RJ (Brazil); Lopes, R.T. [Programa de Engenharia Nuclear/COPPE, Universidade Federal do Rio de Janeiro, Ilha do Fundao, Caixa Postal 68509, 21945-970, Rio de Janeiro, RJ (Brazil)

    2010-09-15

    Simulating X-ray images has been of great interest in recent years as it makes possible an analysis of how X-ray images are affected owing to relevant operating parameters. In this paper, a procedure for simulating computed radiographic images using the Monte Carlo code MCNPX is proposed. The sensitivity curve of the BaFBr image plate detector as well as the characteristic noise of a 16-bit computed radiography system were considered during the methodology's development. The results obtained confirm that the proposed procedure for simulating computed radiographic images is satisfactory, as it allows obtaining results comparable with experimental data.

  3. Dose calculations for a simplified Mammosite system with the Monte Carlo Penelope and MCNPX simulation codes

    International Nuclear Information System (INIS)

    Rojas C, E.L.; Varon T, C.F.; Pedraza N, R.

    2007-01-01

    The treatment of the breast cancer at early stages is of vital importance. For that, most of the investigations are dedicated to the early detection of the suffering and their treatment. As investigation consequence and clinical practice, in 2002 it was developed in U.S.A. an irradiation system of high dose rate known as Mammosite. In this work we carry out dose calculations for a simplified Mammosite system with the Monte Carlo Penelope simulation code and MCNPX, varying the concentration of the contrast material that it is used in the one. (Author)

  4. Simulation for photon detection in spectrometric system of high purity (HPGe) using MCNPX code

    International Nuclear Information System (INIS)

    Correa, Guilherme Jorge de Souza

    2013-01-01

    The Brazilian National Commission of Nuclear Energy defines parameters for classification and management of radioactive waste in accordance with the activity of materials. The efficiency of a detection system is crucial to determine the real activity of a radioactive source. When it's possible, the system's calibration should be performed using a standard source. Unfortunately, there are only a few cases that it can be done this way, considering the difficulty of obtaining appropriate standard sources for each type of measurement. So, computer simulations can be performed to assist in calculating of the efficiency of the system and, consequently, also auxiliary the classification of radioactive waste. This study aims to model a high purity germanium (HPGe) detector with MCNPX code, approaching the spectral values computationally obtained of the values experimentally obtained for the photopeak of 137 Cs. The approach will be made through changes in outer dead layer of the germanium crystal modeled. (author)

  5. Computational simulation of Argonauta/IEN nuclear reactor using MCNPX code

    International Nuclear Information System (INIS)

    Cunha, Victor Lusis Lassance; Silva Junior, Wilson F. Rebello da

    2011-01-01

    The study consisted of developing a computer simulation of a nuclear research reactor using the MCNPX. The reactor modeled is the Argonauta located at IEN (Rio de Janeiro) designed by Argonne National Laboratory (USA), which is primarily used for non-destructive testing with neutron beam and teaching purposes. It was entirely modeled with geometric fidelity, including detailed material description, shielding and irradiation channels. When available, the model was based on the as-built drawings. Four different simulations were made, the first set of two for criticality calculations and the other set for flux measurement. The first simulation set consisted of estimating the reactors reactivity. The second set consisted of placing detectors on specific places where the reactor is monitored and on the fuel axis covering the multiplicative and non-multiplicative media. Based on this data, the thermal neutron flux profile was plotted. All the outputs were compared with experimental data. Since it is a stochastic method, the statistical convergence was successfully checked for all simulations. The results were in good agreement with the experimental values. For the criticality calculations, the relative error was smaller then 1%. The flux measurements were also very well reproduced. The values were normalized for a reference point and the proportionality between the different spots was respected. The neutron flux profile along the core had the expected shape and values. Based on the good results, it can be said that the model is validated. (author)

  6. Computational system to create an entry file for replicating I-125 seeds simulating brachytherapy case studies using the MCNPX code

    Directory of Open Access Journals (Sweden)

    Leonardo da Silva Boia

    2014-03-01

    Full Text Available Purpose: A computational system was developed for this paper in the C++ programming language, to create a 125I radioactive seed entry file, based on the positioning of a virtual grid (template in voxel geometries, with the purpose of performing prostate cancer treatment simulations using the MCNPX code.Methods: The system is fed with information from the planning system with regard to each seed’s location and its depth, and an entry file is automatically created with all the cards (instructions for each seed regarding their cell blocks and surfaces spread out spatially in the 3D environment. The system provides with precision a reproduction of the clinical scenario for the MCNPX code’s simulation environment, thereby allowing the technique’s in-depth study.Results and Conclusion: The preliminary results from this study showed that the lateral penumbra of uniform scanning proton beams was less sensitive In order to validate the computational system, an entry file was created with 88 125I seeds that were inserted in the phantom’s MAX06 prostate region with initial activity determined for the seeds at the 0.27 mCi value. Isodose curves were obtained in all the prostate slices in 5 mm steps in the 7 to 10 cm interval, totaling 7 slices. Variance reduction techniques were applied in order to optimize computational time and the reduction of uncertainties such as photon and electron energy interruptions in 4 keV and forced collisions regarding cells of interest. Through the acquisition of isodose curves, the results obtained show that hot spots have values above 300 Gy, as anticipated in literature, stressing the importance of the sources’ correct positioning, in which the computational system developed provides, in order not to release excessive doses in adjacent risk organs. The 144 Gy prescription curve showed in the validation process that it covers perfectly a large percentage of the volume, at the same time that it demonstrates a large

  7. Dose calculations for a simplified Mammosite system with the Monte Carlo Penelope and MCNPX simulation codes; Calculos de dosis para un sistema Mammosite simplificado con los codigos de simulacion Monte Carlo PENELOPE y MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Rojas C, E.L.; Varon T, C.F.; Pedraza N, R. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: elrc@nuclear.inin.mx

    2007-07-01

    The treatment of the breast cancer at early stages is of vital importance. For that, most of the investigations are dedicated to the early detection of the suffering and their treatment. As investigation consequence and clinical practice, in 2002 it was developed in U.S.A. an irradiation system of high dose rate known as Mammosite. In this work we carry out dose calculations for a simplified Mammosite system with the Monte Carlo Penelope simulation code and MCNPX, varying the concentration of the contrast material that it is used in the one. (Author)

  8. Burnup simulations of different fuel grades using the MCNPX Monte Carlo code

    OpenAIRE

    Asah-Opoku Fiifi; Liang Zhihua; Huque Ziaul; Kommalapati Raghava R.

    2014-01-01

    Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX), uranium oxide ...

  9. Burnup simulations of different fuel grades using the MCNPX Monte Carlo code

    Directory of Open Access Journals (Sweden)

    Asah-Opoku Fiifi

    2014-01-01

    Full Text Available Global energy problems range from the increasing cost of fuel to the unequal distribution of energy resources and the potential climate change resulting from the burning of fossil fuels. A sustainable nuclear energy would augment the current world energy supply and serve as a reliable future energy source. This research focuses on Monte Carlo simulations of pressurized water reactor systems. Three different fuel grades - mixed oxide fuel (MOX, uranium oxide fuel (UOX, and commercially enriched uranium or uranium metal (CEU - are used in this simulation and their impact on the effective multiplication factor (Keff and, hence, criticality and total radioactivity of the reactor core after fuel burnup analyzed. The effect of different clad materials on Keff is also studied. Burnup calculation results indicate a buildup of plutonium isotopes in UOX and CEU, as opposed to a decline in plutonium radioisotopes for MOX fuel burnup time. For MOX fuel, a decrease of 31.9% of the fissile plutonium isotope is observed, while for UOX and CEU, fissile plutonium isotopes increased by 82.3% and 83.8%, respectively. Keff results show zircaloy as a much more effective clad material in comparison to zirconium and stainless steel.

  10. Brachytherapy treatment simulation of strontium-90 and ruthenium-106 plaques on small size posterior uveal melanoma using MCNPX code

    Science.gov (United States)

    Barbosa, N. A.; da Rosa, L. A. R.; Facure, A.; Braz, D.

    2014-02-01

    Concave eye applicators with 90Sr/90Y and 106Ru/106Rh beta-ray sources are usually used in brachytherapy for the treatment of superficial intraocular tumors as uveal melanoma with thickness up to 5 mm. The aim of this work consisted in using the Monte Carlo code MCNPX to calculate the 3D dose distribution on a mathematical model of the human eye, considering 90Sr/90Y and 160Ru/160Rh beta-ray eye applicators, in order to treat a posterior uveal melanoma with a thickness 3.8 mm from the choroid surface. Mathematical models were developed for the two ophthalmic applicators, CGD produced by BEBIG Company and SIA.6 produced by the Amersham Company, with activities 1 mCi and 4.23 mCi respectively. They have a concave form. These applicators' mathematical models were attached to the eye model and the dose distributions were calculated using the MCNPX *F8 tally. The average doses rates were determined in all regions of the eye model. The *F8 tally results showed that the deposited energy due to the applicator with the radionuclide 106Ru/106Rh is higher in all eye regions, including tumor. However the average dose rate in the tumor region is higher for the applicator with 90Sr/90Y, due to its high activity. Due to the dosimetric characteristics of these applicators, the PDD value for 3 mm water is 73% for the 106Ru/106Rh applicator and 60% for 90Sr/90Y applicator. For a better choice of the applicator type and radionuclide it is important to know the thickness of the tumor and its location.

  11. Calculation of absorbed doses in sphere volumes around the Mammosite using the Monte Carlo simulation code MCNPX; Calculo de dosis absorbida en volumenes esfericos alrededor del Mammosite utilizando el codigo de simulacion Monte Carlo MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Rojas C, E. L. [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)

    2008-07-01

    The objective of this study is to investigate the changes observed in the absorbed doses in mammary gland tissue when irradiated with a equipment of high dose rate known as Mammosite and introducing material resources contrary to the tissue that constitutes the mammary gland. The modeling study is performed with the code MCNPX, 2005 version, the equipment and the mammary gland and calculating the absorbed doses in tissue when introduced small volumes of air or calcium in the system. (Author)

  12. 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)

  13. Characterization of the MCNPX computer code in micro processed architectures

    International Nuclear Information System (INIS)

    Almeida, Helder C.; Dominguez, Dany S.; Orellana, Esbel T.V.; Milian, Felix M.

    2009-01-01

    The MCNPX (Monte Carlo N-Particle extended) can be used to simulate the transport of several types of nuclear particles, using probabilistic methods. The technique used for MCNPX is to follow the history of each particle from its origin to its extinction that can be given by absorption, escape or other reasons. To obtain accurate results in simulations performed with the MCNPX is necessary to process a large number of histories, which demand high computational cost. Currently the MCNPX can be installed in virtually all computing platforms available, however there is virtually no information on the performance of the application in each. This paper studies the performance of MCNPX, to work with electrons and photons in phantom Faux on two platforms used by most researchers, Windows and Li nux. Both platforms were tested on the same computer to ensure the reliability of the hardware in the measures of performance. The performance of MCNPX was measured by time spent to run a simulation, making the variable time the main measure of comparison. During the tests the difference in performance between the two platforms MCNPX was evident. In some cases we were able to gain speed more than 10% only with the exchange platforms, without any specific optimization. This shows the relevance of the study to optimize this tool on the platform most appropriate for its use. (author)

  14. Design of CONAS Station for MCNPX Simulation

    International Nuclear Information System (INIS)

    Tuan, Hoang Sy Minh; Sun, Gwang Min

    2011-01-01

    The HANARO Cold neutron Research Facility (CNRF) construction project was formulated since early 2003. During the feasibility study, a user survey on cold neutron instruments, which includes SANS, TAS, TOF spectrometer, spin-echo spectrometer, neutron interferometer, powder-Laue diffractometer, etc., were performed. Beside these instruments, CNRF especially includes the neutron activation analysis instruments using the cold neutron beams. At present, a Cold Neutron Activation Station (CONAS) has been under construction with a support by the National Research Foundation of Korea (NRF) since May 2010. CONAS consists of Neutron Depth Profiling (NDP) and Prompt Gamma Activation Analysis (PGAA) instruments. This project will finish in April 2012. In order to estimate the CONAS, the MCNPX code was adopted in this study for estimating and optimizing the setup parameters of these instruments

  15. Relative efficiency calculation of a HPGe detector using MCNPX code

    International Nuclear Information System (INIS)

    Medeiros, Marcos P.C.; Rebello, Wilson F.; Lopes, Jose M.; Silva, Ademir X.

    2015-01-01

    High-purity germanium detectors (HPGe) are mandatory tools for spectrometry because of their excellent energy resolution. The efficiency of such detectors, quoted in the list of specifications by the manufacturer, frequently refers to the relative full-energy peak efficiency, related to the absolute full-energy peak efficiency of a 7.6 cm x 7.6 cm (diameter x height) NaI(Tl) crystal, based on the 1.33 MeV peak of a 60 Co source positioned 25 cm from the detector. In this study, we used MCNPX code to simulate a HPGe detector (Canberra GC3020), from Real-Time Neutrongraphy Laboratory of UFRJ, to survey the spectrum of a 60 Co source located 25 cm from the detector in order to calculate and confirm the efficiency declared by the manufacturer. Agreement between experimental and simulated data was achieved. The model under development will be used for calculating and comparison purposes with the detector calibration curve from software Genie2000™, also serving as a reference for future studies. (author)

  16. Isodose curve determination of prostate for the treatment of brachytherapy using MCNPX code

    International Nuclear Information System (INIS)

    Reis Junior, J.P.; Menezes, A.F.; Medeiros, J.A.C.C.; Facure, A.N.S.S.; Silva, A.X.

    2011-01-01

    Using voxel phantom MAX 06 coupled to the code MCNPX it possible to plot the isodose curves for the main levels involved in the treatment of prostate brachytherapy, V100 and V150 which are, respectively corresponding curves 144 and 216 Gy to curves are indicative of the quality of the existing implant of prostate brachytherapy. The number of 79 seeds 125 I, were placed in the voxels simulator MAX 06, in the slices x = 7.0, 7.5, 8.0, 8.5, 9.0, 9.5, 10.0 with the calculation model used in MCNPX in all voxels present in a matrix, it was possible to trace the isodose curves for MATLAB. For comparison and using own routines MCNPX it was possible to trace the same curves using mesh tallies. The results showed agreement with predicted values in the planning system prowess 3D. (author)

  17. Interfacing MCNPX and McStas for simulation of neutron transport

    DEFF Research Database (Denmark)

    Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik

    2013-01-01

    Stas[4, 5, 6, 7]. The coupling between the two simulation suites typically consists of providing analytical fits of MCNPX neutron spectra to McStas. This method is generally successful but has limitations, as it e.g. does not allow for re-entry of neutrons into the MCNPX regime. Previous work to resolve......Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using Monte Carlo codes such as MCNPX[1] or FLUKA[2, 3] whereas simulations of neutron transport from the moderator and the instrument response are performed by neutron ray tracing codes such as Mc...... geometries, backgrounds, interference between beam-lines as well as shielding requirements along the neutron guides....

  18. Primary and scattered component analyses of simulated images with MCNPX

    International Nuclear Information System (INIS)

    Souza, Edmilson M. de; Correa, Samanda C.A.; Silva, Ademir X. da; Lopes, Ricardo T.

    2007-01-01

    MCNPX radiography tallies are an array of point detectors (tally F5) close to one another to generate an image based on the point detector fluxes. The goal of this work is to study the primary and scattered radiation behavior in simulated images with MCNPX radiography tally. Monte Carlo simulations were performed for calculations of the PSF for different materials, thickness and distances between sample and detector. PSF analyses of simulated images showed that the radiography tallies to reproduce with perfection geometric and physical characteristics of experimental images. (author)

  19. MCNPX simulation of proton dose distribution in homogeneous and CT phantoms

    International Nuclear Information System (INIS)

    Lee, C.C.; Lee, Y.J.; Tung, C.J.; Cheng, H.W.; Chao, T.C.

    2014-01-01

    A dose simulation system was constructed based on the MCNPX Monte Carlo package to simulate proton dose distribution in homogeneous and CT phantoms. Conversion from Hounsfield unit of a patient CT image set to material information necessary for Monte Carlo simulation is based on Schneider's approach. In order to validate this simulation system, inter-comparison of depth dose distributions among those obtained from the MCNPX, GEANT4 and FLUKA codes for a 160 MeV monoenergetic proton beam incident normally on the surface of a homogeneous water phantom was performed. For dose validation within the CT phantom, direct comparison with measurement is infeasible. Instead, this study took the approach to indirectly compare the 50% ranges (R 50% ) along the central axis by our system to the NIST CSDA ranges for beams with 160 and 115 MeV energies. Comparison result within the homogeneous phantom shows good agreement. Differences of simulated R 50% among the three codes are less than 1 mm. For results within the CT phantom, the MCNPX simulated water equivalent R eq,50% are compatible with the CSDA water equivalent ranges from the NIST database with differences of 0.7 and 4.1 mm for 160 and 115 MeV beams, respectively. - Highlights: ► Proton dose simulation based on the MCNPX 2.6.0 in homogeneous and CT phantoms. ► CT number (HU) conversion to electron density based on Schneider's approach. ► Good agreement among MCNPX, GEANT4 and FLUKA codes in a homogeneous water phantom. ► Water equivalent R 50 in CT phantoms are compatible to those of NIST database

  20. New MCNPX developments

    Energy Technology Data Exchange (ETDEWEB)

    Hendricks, J. S. (John S.); McKinney, G. W. (Gregg W.); Waters, L. S. (Laurie S.); Hughes, H. G. (Henry Grady); Snow, E. C. (Edward Clark)

    2002-01-01

    The Los Alamos National Laboratory Monte Carlo N-Particle extended (MCNPX) radiation transport code has been upgraded significantly to Version MCNPX2.4.0. It is now based on the latest MCNP4C3 and MCNPX2.3.0 releases to the Radiation Safety Information Computational Center (RSICC). In addition to all of the advances from earlier versions of MCNP and MCNPX, important new capabilities have been developed. The Monte Carlo method was developed at Los Alamos National Laboratory during the Manhattan Project in the early 1940s. MCNP and MCNPX are heirs to those early efforts. Over 400 person-years have been invested in the research, development, programming, documentation, and databases for these codes. MCNP is a general-purpose neutron (0-MeV to 20-MeV), photon (1-keV to 1-GeV), and electron (1-keV to 1-GeV) transport code for calculating *MCNPX, MCNP, LAHET, and LCS are trademarks of the Regents of the University of California, Los Alamos National Laboratory. the time-dependent, continuous-energy transport of these particles in three-dimensional geometries. MCNP is perhaps the most widely used and well-known physics simulation code in the world today. MCNPX extends MCNP to track nearly all particles at all energies. MCNPX combined MCNP and the LAHET Code System (LCS). LCS is based on the Oak Ridge High Energy Transport Code. LCS uses models for particles in physics regimes where there are no tabulated data, including the Bertini and ISABEL models. MCNPX has additional models to LCS, such as the CEM model. MCNPX2.3.0 was released to RSICC in December 2001 and is based on MCNP4B. The principal features of MCNPX2.3.0 are (1) Physics for 34 particle types; (2) High-energy physics above the giga-electron volt range; (3) Neutron, proton, and photonuclear 150-MeV libraries: (4) Photonuclear physics; (5) Mesh tallies; (6) Radiography tallies; (7) Secondary particle production biasing; (8) VAVILOV energy straggling for charged particles; and (9) Automatic configuration for

  1. Recent developments in MCNPX trademark

    International Nuclear Information System (INIS)

    Hughes, H.G.; Adams, K.J.; Chadwick, M.B.

    1998-01-01

    The MCNPX Monte Carlo particle transport code is rapidly developing into a significant computational tool for high-energy transport applications. In this paper, the authors will discuss three recent enhancements to MCNPX: a new charged-particle collisional energy-loss model, a geometry-independent mesh-based tally system, and a radiography simulation capability

  2. Prostate dose calculations for permanent implants using the MCNPX code and the Voxels phantom MAX

    Energy Technology Data Exchange (ETDEWEB)

    Reis Junior, Juraci Passos dos; Silva, Ademir Xavier da, E-mail: jjunior@con.ufrj.b, E-mail: Ademir@con.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Facure, Alessandro N.S., E-mail: facure@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2010-07-01

    This paper presents the modeling of 80, 88 and 100 of {sup 125}I seeds, punctual and volumetric inserted into the phantom spherical volume representing the prostate and prostate phantom voxels MAX. Starting values of minimum and maximum activity, 0.27 mCi and 0.38 mCi, respectively, were simulated in the Monte Carlo code MCNPX in order to determine whether the final dose, according to the integration of the equation of decay at time t = 0 to t = {infinity} corresponds to the default value set by the AAPM 64 which is 144 Gy. The results showed that consider sources results in doses exceeding the percentage discrepancy of the default value of 200%, while volumetric consider sources result in doses close to 144 Gy. (author)

  3. Benchmarking Heavy Ion Transport Codes FLUKA, HETC-HEDS MARS15, MCNPX, and PHITS

    Energy Technology Data Exchange (ETDEWEB)

    Ronningen, Reginald Martin [Michigan State University; Remec, Igor [Oak Ridge National Laboratory; Heilbronn, Lawrence H. [University of Tennessee-Knoxville

    2013-06-07

    Powerful accelerators such as spallation neutron sources, muon-collider/neutrino facilities, and rare isotope beam facilities must be designed with the consideration that they handle the beam power reliably and safely, and they must be optimized to yield maximum performance relative to their design requirements. The simulation codes used for design purposes must produce reliable results. If not, component and facility designs can become costly, have limited lifetime and usefulness, and could even be unsafe. The objective of this proposal is to assess the performance of the currently available codes PHITS, FLUKA, MARS15, MCNPX, and HETC-HEDS that could be used for design simulations involving heavy ion transport. We plan to access their performance by performing simulations and comparing results against experimental data of benchmark quality. Quantitative knowledge of the biases and the uncertainties of the simulations is essential as this potentially impacts the safe, reliable and cost effective design of any future radioactive ion beam facility. Further benchmarking of heavy-ion transport codes was one of the actions recommended in the Report of the 2003 RIA R&D Workshop".

  4. Microdosimetry of alpha particles for simple and 3D voxelised geometries using MCNPX and Geant4 Monte Carlo codes

    International Nuclear Information System (INIS)

    Elbast, M.; Saudo, A.; Franck, D.; Petitot, F.; Desbree, A.

    2008-01-01

    Microdosimetry using Monte Carlo simulation is a suitable technique to describe the stochastic nature of energy deposition by alpha particle at cellular level. Because of its short range, the energy imparted by this particle to the targets is highly non-uniform. Thus, to achieve accurate dosimetric results, the modelling of the geometry should be as realistic as possible. The objectives of the present study were to validate the use of the MCNPX and Geant4 Monte Carlo codes for microdosimetric studies using simple and three-dimensional voxelised geometry and to study their limit of validity in this last case. To that aim, the specific energy (z) deposited in the cell nucleus, the single-hit density of specific energy f 1 (z) and the mean-specific energy 1 > were calculated. Results show a good agreement when compared with the literature using simple geometry. The maximum percentage difference found 1 (z) obtained with MCNPX for <1 μm voxel size presents a significant difference with the shape of non-voxelised geometry. When using Geant4, little differences are observed whatever the voxel size is. Below 1 μm, the use of Geant4 is required. However, the calculation time is 10 times higher with Geant4 than MCNPX code in the same conditions. (authors)

  5. Geant4 and MCNPX simulations of thermal neutron detection with planar silicon detectors

    Energy Technology Data Exchange (ETDEWEB)

    Guardiola, C; Fleta, C; Quirion, D; Lozano, M [Instituto de Microelectronica de Barcelona, (IMB-CNM), CSIC, 08193 Bellaterra, Barcelona (Spain); Amgarou, K [Departamento de FIsica, Universidad Autonoma de Barcelona, 08193 Bellaterra, Barcelona (Spain); GarcIa, F, E-mail: Consuelo.Guardiola@imb-cnm.csic.es [Helsinki Institute of Physics, University of Helsinki, 00014 Helsinki (Finland)

    2011-09-15

    We used Geant4 and MCNPX codes to evaluate the detection efficiency of planar silicon detectors coupled to different Boron-based converters with varied compositions and thicknesses that detect thermal neutrons via the {sup 10}B(n,{alpha}){sup 7}Li nuclear reaction. Few studies about the thermal neutron transport in Geant4 have been reported so far and it is becoming increasingly difficult to ignore its discrepancies with MCNPX in this neutron energy range. In the thermal energy range, Geant4 shows high discrepancies with MCNPX giving a maximum efficiency of about 3.3% in the {sup 10}B case whereas that obtained with MCNPX was 5%. Disagreements obtained between both codes in this energy range are analyzed and discussed.

  6. Benchmark on traveling wave fast reactor with negative reactivity feedback obtained with MCNPX code

    International Nuclear Information System (INIS)

    Gann, V.V.; Gann, A.V.

    2012-01-01

    This paper presents results of computer simulations of traveling wave fast reactor with negative reactivity feedback. The results were obtained using MCNPX code combined with CINDER90 subroutine for depletion calculations. We considered 1-D model of TWR containing 4 m long core made of mixture of 66 at. % 238 U and 34 at. % 10 B. Ignitor made of 235 U was located in the center of the core. Boron was included as imitator of structural in-core materials and coolant. Negative reactivity feedback was adjusted to reactor power of 500 MW. In this case two burning waves originated from the igniter and travel to the ends of the core during the following 40 years; coefficient of utilization of 238 U reached 80 %. Distribution of specific power in traveling wave, isotope concentration of fission products and actinides, neutron flux, fast neutron spectrum, specific activity were calculated. Data of the computer simulation is in qualitative agreement with theoretical results obtained in slow burning wave approximation

  7. Study of the source term of radiation of the CDTN GE-PET trace 8 cyclotron with the MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Benavente C, J. A.; Lacerda, M. A. S.; Fonseca, T. C. F.; Da Silva, T. A. [Centro de Desenvolvimento da Tecnologia Nuclear / CNEN, Av. Pte. Antonio Carlos 6627, 31270-901 Belo Horizonte, Minas Gerais (Brazil); Vega C, H. R., E-mail: jhonnybenavente@gmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

    2015-10-15

    Full text: The knowledge of the neutron spectra in a PET cyclotron is important for the optimization of radiation protection of the workers and individuals of the public. The main objective of this work is to study the source term of radiation of the GE-PET trace 8 cyclotron of the Development Center of Nuclear Technology (CDTN/CNEN) using computer simulation by the Monte Carlo method. The MCNPX version 2.7 code was used to calculate the flux of neutrons produced from the interaction of the primary proton beam with the target body and other cyclotron components, during 18F production. The estimate of the source term and the corresponding radiation field was performed from the bombardment of a H{sub 2}{sup 18}O target with protons of 75 μA current and 16.5 MeV of energy. The values of the simulated fluxes were compared with those reported by the accelerator manufacturer (GE Health care Company). Results showed that the fluxes estimated with the MCNPX codes were about 70% lower than the reported by the manufacturer. The mean energies of the neutrons were also different of that reported by GE Health Care. It is recommended to investigate other cross sections data and the use of physical models of the code itself for a complete characterization of the source term of radiation. (Author)

  8. Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes

    Science.gov (United States)

    Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.

    2017-02-01

    International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.

  9. Dose Distribution Calculation Using MCNPX Code in the Gamma-ray Irradiation Cell

    International Nuclear Information System (INIS)

    Kim, Yong Ho

    1991-02-01

    60 Co-gamma irradiators have long been used for foods sterilization, plant mutation and development of radio-protective agents, radio-sensitizers and other purposes. The Applied Radiological Science Research Institute of Cheju National University has a multipurpose gamma irradiation facility loaded with a MDS Nordin standard 60 Co source (C188), of which the initial activity was 400 TBq (10,800 Ci) on February 19, 2004. This panoramic gamma irradiator is designed to irradiate in all directions various samples such as plants, cultured cells and mice to administer given radiation doses. In order to give accurate doses to irradiation samples, appropriate methods of evaluating, both by calculation and measurement, the radiation doses delivered to the samples should be set up. Computational models have been developed to evaluate the radiation dose distributions inside the irradiation chamber and the radiation doses delivered to typical biolological samples which are frequently irradiated in the facility. The computational models are based on using the MCNPX code. The horizontal and vertical dose distributions has been calculated inside the irradiation chamber and compared the calculated results with measured data obtained with radiation dosimeters to verify the computational models. The radiation dosimeters employed are a Famer's type ion chamber and MOSFET dosimeters. Radiation doses were calculated by computational models, which were delivered to cultured cell samples contained in test tubes and to a mouse fixed in a irradiation cage, and compared the calculated results with the measured data. The computation models are also tested to see if they can accurately simulate the case where a thick lead shield is placed between the source and detector. Three tally options of the MCNPX code, F4, F5 and F6, are alternately used to see which option produces optimum results. The computation models are also used to calculate gamma ray energy spectra of a BGO scintillator at

  10. Assessment of patient dose reduction by bismuth shielding in CT using measurements, GEANT4 and MCNPX simulations.

    Science.gov (United States)

    Mendes, M; Costa, F; Figueira, C; Madeira, P; Teles, P; Vaz, P

    2015-07-01

    This work reports on the use of two different Monte Carlo codes (GEANT4 and MCNPX) for assessing the dose reduction using bismuth shields in computer tomography (CT) procedures in order to protect radiosensitive organs such as eye lens, thyroid and breast. Measurements were performed using head and body PMMA phantoms and an ionisation chamber placed in five different positions of the phantom. Simulations were performed to estimate Computed Tomography Dose Index values using GEANT4 and MCNPX. The relative differences between measurements and simulations were bismuth shielding ranges from 2 to 45 %, depending on the position of the bismuth shield. The percentage of dose reduction was more significant for the area covered by the bismuth shielding (36 % for eye lens, 39 % for thyroid and 45 % for breast shields). © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  11. MCNPX-CINDER'90 Simulation of Photonuclear Mo-99 Production Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kelsey, Charles T. IV [Los Alamos National Laboratory; Chemerizov, Sergey D. [Argonne National Laboratory; Dale, Gregory E. [Los Alamos National Laboratory; Harvey, James T. [NorthStar Medical Radioisotopes; Tkac, Peter [Argonne National Laboratory; Vandegrift, George R III [Argonne National Laboratory

    2011-01-01

    The MCNPX and CINDER'90 codes were used to support design of experiments investigating Mo-99 production with a 20-MeV electron beam. Bremsstrahlung photons produced by the electron beam interacting with the target drive the desired Mo-100({gamma},n)Mo-99 reaction, as well as many undesired reactions important to accurate prediction of radiation hazards. MCNPX is a radiation transport code and CINDER'90 is a transmutation code. They are routinely used together for accelerator activation calculations. Low energy neutron fluxes and production rates for nonneutron and high energy neutron induced reactions computed using MCNPX are inputs to CINDER'90. CINDER'90 presently has only a neutron reaction cross section library up to 25 MeV and normally the other reaction rates come from MCNPX physics models. For this work MCNPX photon flux tallies modified by energy response functions prepared from evaluated photonuclear cross section data were used to tally the reaction rates for CINDER'90 input. The cross section evaluations do not provide isomer to ground state yield ratios so a spin based approximation was used. Post irradiation dose rates were calculated using MCNPX with CINDER'90 produced decay photon spectra. The sensitivity of radionuclide activities and dose rates to beam parameters including energy, position, and profile, as well as underlying isomer assumptions, was investigated. Three experimental production targets were irradiated, two natural Mo and one Mo-100 enriched. Natural Mo foils upstream of the targets were used to analyze beam position and profile by exposing Gafchromic film to the foils after each irradiation. Activation and dose rate calculations were rerun after the experiments using measured beam parameters for comparison with measured Mo-99 activities and dose rates.

  12. Modeling of neutron diffractometry facility of Tehran Research Reactor using Vitess 3.3a and MCNPX codes

    Directory of Open Access Journals (Sweden)

    Z. Gholamzadeh

    2018-02-01

    Full Text Available The neutron powder diffractometer (NPD is used to study a variety of technologically important and scientifically driven materials such as superconductors, multiferroics, catalysts, alloys, ceramics, cements, colossal magnetoresistance perovskites, magnets, thermoelectrics, zeolites, pharmaceuticals, etc. Monte Carlo–based codes are powerful tools to evaluate the neutronic behavior of the NPD. In the present study, MCNPX 2.6.0 and Vitess 3.3a codes were applied to simulate NPD facilities, which could be equipped with different optic devices such as pyrolytic graphite or neutron chopper. So, the Monte Carlo–based codes were used to simulate the NPD facility of the 5 MW Tehran Research Reactor. The simulation results were compared to the experimental data. The theoretical results showed good conformity to experimental data, which indicates acceptable performance of the Vitess 3.3a code in the neutron optic section of calculations. Another extracted result of this work shows that application of neutron chopper instead of monochromator could be efficient to keep neutron flux intensity higher than 106 n/s/cm2 at sample position.

  13. Soft-Rt: software for IMRT simulations based on MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira F, T. C. [Centro de Desenvolvimento da Tecnologia Nuclear / CNEN, Av. Pte. Antonio Carlos 6627, 31270-901 Belo Horizonte, Minas Gerais (Brazil); Campos, T., E-mail: tcff01@gmail.com [Universidade Federal de Minas Gerais, Departamento de Engenharia Nuclear, Programa de Pos Graduacao em Ciencias e Tecnicas Nucleares, Av. Pte. Antonio Carlos 6627, 31270-901 Belo Horizonte, Minas Gerais (Brazil)

    2015-10-15

    Intensity Modulated Radiation Therapy (IMRT) is an advanced treatment technique, widely used in external radiotherapy. This paper presents the Soft-Rt which allows the simulation of an entire IMRT treatment protocol. The Soft-Rt performs a full three-dimensional rendering of a set of patient images, including the definitions of region of interest with organs in risk, and the target tumor volume and margins (PTV). Thus, a more accurate analysis and planning can be performed, taking into account the features and orientation of the radiation beams. The exposed tissues as well as the amount of absorbed dose is depicted in healthy and/or cancerous tissues. As conclusion, Soft-Rt can predict dose on the PTV accurately, preserving the surrounding healthy tissues. Soft-Rt is coupled with SISCODES code. The SISCODES code is firstly applied to segment the set of CT or MRI patient images in distinct tissues pointing out its respective density and chemical compositions. Later, the voxel model is export to the Soft-Rt IMRT planning module in which a full treatment planning is created. All geometrical parameters are sent to the general purpose Monte Carlo transport code - MCNP - to simulate the interaction of each incident beam towards to the PTV avoiding organs in risk. The normalized dose results are exported to the Soft-Rt out-module, in which the three-dimensional model visualization is shown in a transparent glass procedure adopting gray scale for the dependence on the mass density of the correlated tissue; while, a color scale to depict dose values in a superimpose protocol. (Author)

  14. Soft-Rt: software for IMRT simulations based on MCNPX

    International Nuclear Information System (INIS)

    Ferreira F, T. C.; Campos, T.

    2015-10-01

    Intensity Modulated Radiation Therapy (IMRT) is an advanced treatment technique, widely used in external radiotherapy. This paper presents the Soft-Rt which allows the simulation of an entire IMRT treatment protocol. The Soft-Rt performs a full three-dimensional rendering of a set of patient images, including the definitions of region of interest with organs in risk, and the target tumor volume and margins (PTV). Thus, a more accurate analysis and planning can be performed, taking into account the features and orientation of the radiation beams. The exposed tissues as well as the amount of absorbed dose is depicted in healthy and/or cancerous tissues. As conclusion, Soft-Rt can predict dose on the PTV accurately, preserving the surrounding healthy tissues. Soft-Rt is coupled with SISCODES code. The SISCODES code is firstly applied to segment the set of CT or MRI patient images in distinct tissues pointing out its respective density and chemical compositions. Later, the voxel model is export to the Soft-Rt IMRT planning module in which a full treatment planning is created. All geometrical parameters are sent to the general purpose Monte Carlo transport code - MCNP - to simulate the interaction of each incident beam towards to the PTV avoiding organs in risk. The normalized dose results are exported to the Soft-Rt out-module, in which the three-dimensional model visualization is shown in a transparent glass procedure adopting gray scale for the dependence on the mass density of the correlated tissue; while, a color scale to depict dose values in a superimpose protocol. (Author)

  15. Mass Attenuation Coefficients of Human Body Organs using MCNPX Monte Carlo Code

    Directory of Open Access Journals (Sweden)

    Huseyin Tekin

    2017-12-01

    Full Text Available Introduction: Investigation of radiation interaction with living organs has always been a thrust area in medical and radiation physics. The investigated results are being used in medical physics for developing improved and sensitive techniques and minimizing radiation exposure. In this study, mass attenuation coefficients of different human organs and biological materials such as adipose, blood, bone, brain, eye lens, lung, muscle, skin, and tissue have been calculated. Materials and Methods: In the present study, Monte Carlo N-Particle eXtended (MCNP-X version 2.4.0 was used for determining mass attenuation coefficients, and the obtained results were compared with earlier investigations (using GEometry ANd Tracking [GEANT4] and FLUKA computer simulation packages for blood, bone, lung, eye lens, adipose, tissue, muscle, brain, and skin materials at different energies. Results: The results of this study showed that the obtained results from MCNP-X were in high accordance with the National Institute of Standards and Technology data. Conclusion: Our findings would be beneficial for use of present simulation technique and mass attenuation coefficients for medical and radiation physics applications.

  16. Assessment of ocular beta radiation dose distribution due to 106Ru/106Rh brachytherapy applicators using MCNPX Monte Carlo code

    Directory of Open Access Journals (Sweden)

    Nilseia Aparecida Barbosa

    2014-08-01

    Full Text Available Purpose: Melanoma at the choroid region is the most common primary cancer that affects the eye in adult patients. Concave ophthalmic applicators with 106Ru/106Rh beta sources are the more used for treatment of these eye lesions, mainly lesions with small and medium dimensions. The available treatment planning system for 106Ru applicators is based on dose distributions on a homogeneous water sphere eye model, resulting in a lack of data in the literature of dose distributions in the eye radiosensitive structures, information that may be crucial to improve the treatment planning process, aiming the maintenance of visual acuity. Methods: The Monte Carlo code MCNPX was used to calculate the dose distribution in a complete mathematical model of the human eye containing a choroid melanoma; considering the eye actual dimensions and its various component structures, due to an ophthalmic brachytherapy treatment, using 106Ru/106Rh beta-ray sources. Two possibilities were analyzed; a simple water eye and a heterogeneous eye considering all its structures. Two concave applicators, CCA and CCB manufactured by BEBIG and a complete mathematical model of the human eye were modeled using the MCNPX code. Results and Conclusion: For both eye models, namely water model and heterogeneous model, mean dose values simulated for the same eye regions are, in general, very similar, excepting for regions very distant from the applicator, where mean dose values are very low, uncertainties are higher and relative differences may reach 20.4%. For the tumor base and the eye structures closest to the applicator, such as sclera, choroid and retina, the maximum difference observed was 4%, presenting the heterogeneous model higher mean dose values. For the other eye regions, the higher doses were obtained when the homogeneous water eye model is taken into consideration. Mean dose distributions determined for the homogeneous water eye model are similar to those obtained for the

  17. Dose rate distribution of the GammaBeam: 127 irradiator using MCNPX code

    International Nuclear Information System (INIS)

    Gual, Maritza Rodriguez; Batista, Adriana de Souza Medeiros; Pereira, Claubia; Faria, Luiz O. de; Grossi, Pablo Andrade

    2013-01-01

    The GammaBeam - 127 Irradiator is widely used for biological, chemical and medical applications of the gamma irradiation technology using Cobalt 60 radioactive at the Centro de Desenvolvimento da Tecnologia Nuclear CDTN/CNEN, Belo Horizonte, Brazil. The source has maximum activity of 60.000Ci, which is composed by 16 double encapsulated radioactive pencils placed in a rack. The facility is classified by the IAEA as Category II (dry storage facility). The aim of this work is to present a modelling developed to evaluate the dose rates at the irradiation room and the dose distribution at the irradiated products. In addition, the simulations could be used as a predictive tool of dose evaluation in the irradiation facility helping benchmark experiments in new similar facilities. The MCNPX simulated results were compared and validated with radiometric measurements using Fricke and TLDs dosimeters along several positions inside the irradiation room. (author)

  18. Evaluation of equivalent doses in 18F PET/CT using the Monte Carlo method with MCNPX code

    International Nuclear Information System (INIS)

    Belinato, Walmir; Santos, William Souza; Perini, Ana Paula; Neves, Lucio Pereira; Souza, Divanizia N.

    2017-01-01

    The present work used the Monte Carlo method (MMC), specifically the Monte Carlo NParticle - MCNPX, to simulate the interaction of radiation involving photons and particles, such as positrons and electrons, with virtual adult anthropomorphic simulators on PET / CT scans and to determine absorbed and equivalent doses in adult male and female patients

  19. Application of the MCNPX-McStas interface for shielding calculations and guide design at ESS

    DEFF Research Database (Denmark)

    Klinkby, Esben Bryndt; Bergbäck Knudsen, Erik; Willendrup, Peter Kjær

    2013-01-01

    . The generation and moderation of neutrons is simulated using a full scale MCNPX model of the ESS target monolith. Upon entering the beam extraction region, the individual neutron states are handed to McStas via the MCNPX-McStas interface. McStas transports the neutrons through the beam guide and by using newly......Recently, an interface between the Monte Carlo code MCNPX and the neutron ray-tracing code MCNPX was developed[1]. Based on the expected neutronic performance and guide geometries relevant for the ESS, the combined MCNPX-McStas code is used to calculate dose rates along neutron beam guides...... developed event logging capability, the neutron state parameters corresponding to un-reflected neutrons are recorded at each scattering. This information is handed back to MCNPX where it serves as neutron source input for a second MCNPX simulation. This simulation enables calculation of dose rates...

  20. Study of salinity in aqueous medium using X-Ray beam with MCNP-X code

    International Nuclear Information System (INIS)

    Barbosa, Caroline M.; Braz, Delson

    2017-01-01

    In offshore production, it is possible that the produced water presents geochemical characteristics that correspond to the mixture of formation water (connate water) and the sea water (injection water), and the physical-chemical behavior of the injected water allows a considerable variation in the index of salinity altering the water/oil ratio during transportation and/or extraction. Injection water is generally used to raise the reservoir pressure, increasing the percentage of extracted oil. This water has a significant amount of salts that generate some difficulties, such as measuring fractions of volume in multiphase systems. One way to check the effects of salinity would be to regularly measure the amount of salt present in the water. In this way, this work presents a methodology to measure the concentration and the types of salts using nuclear techniques through the MCNP-X computational code. The measurement geometry uses an X-ray beam (40-100 keV) and NaI(Tl) scintillation detector positioned diametrically opposed to the source. The studied samples were the NaCl, KCl and MgCl 2 salts in aqueous solution. The results present the possibility of differentiating the formation and injection waters due to differences in the salt concentrations. (author)

  1. Study of salinity in aqueous medium using X-Ray beam with MCNP-X code

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Caroline M.; Braz, Delson [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Salgado, César M., E-mail: cbarbosa@nuclear.ufrj.br, E-mail: delson@nuclear.ufrj.br, E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    In offshore production, it is possible that the produced water presents geochemical characteristics that correspond to the mixture of formation water (connate water) and the sea water (injection water), and the physical-chemical behavior of the injected water allows a considerable variation in the index of salinity altering the water/oil ratio during transportation and/or extraction. Injection water is generally used to raise the reservoir pressure, increasing the percentage of extracted oil. This water has a significant amount of salts that generate some difficulties, such as measuring fractions of volume in multiphase systems. One way to check the effects of salinity would be to regularly measure the amount of salt present in the water. In this way, this work presents a methodology to measure the concentration and the types of salts using nuclear techniques through the MCNP-X computational code. The measurement geometry uses an X-ray beam (40-100 keV) and NaI(Tl) scintillation detector positioned diametrically opposed to the source. The studied samples were the NaCl, KCl and MgCl{sub 2} salts in aqueous solution. The results present the possibility of differentiating the formation and injection waters due to differences in the salt concentrations. (author)

  2. Study of geometry to obtain the volume fraction of multiphase flows using the MCNP-X code

    International Nuclear Information System (INIS)

    Peixoto, Philippe N.B.; Salgado, Cesar M.

    2015-01-01

    The gamma ray attenuation technique is used in many works to obtaining volume fraction of multiphase flows in the oil industry, because it is a noninvasive technique with good precision. In these studies are simulated various geometries with different flow regime, compositions of materials, source-detector positions and types of collimation for sources. This work aim evaluate the interference in the results of the geometry changes and obtaining the best measuring geometry to provide the volume fractions accurately by evaluating different geometries simulations (ranging the source-detector position, flow schemes and homogeneity Makeup) in the MCNP-X code. The study was performed for two types of biphasic compositions of materials (oil-water and oil-air), two flow regimes (annular and smooth stratified) and was varied the position of each material in relative to source and detector positions. Another study to evaluate the interference of homogeneity of the compositions in the results was also conducted in order to verify the possibility of removing part of the composition and make a homogeneous blend using a mixer equipment. All these variations were simulated with two different types of beam, divergent beam and pencil beam. From the simulated geometries, it was possible to compare the differences between the areas of the spectra generated for each model. The results indicate that the flow regime and the differences in the material's densities interfere in the results being necessary to establish a specific simulation geometry for each flows regime. However, the simulations indicate that changing the type of collimation of sources do not affect the results, but improving the counts statistics, increasing the accurate. (author)

  3. Coupling of MCNPX with a sub-channel code for analysis of a HPLWR fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Waata, C.; Schulenberg, T.; Cheng Xu [Forschungszentrum Karlsruhe, Institute for Nuclear and Energy Technologies, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, (Germany)

    2005-07-01

    Full text of publication follows: The High Performance Light Water Reactor (HPLWR) project was launched in 2000 under the 5. Framework Program of the European Commission. The main objective of this project was to study the technical and economic feasibility of light water reactors operating at supercritical pressure conditions. This study aims to achieve high thermal efficiency of the nuclear power plant with operating conditions of pressure at 25 MPa, coolant temperature of about 510?C and an efficiency of up to 45%. The utilization of supercritical water as coolant and moderator in the HPLWR core introduces some challenges in the design of the HPLWR core due to the special behavior of the thermal-physical properties of water under super-critical pressure conditions. At supercritical pressure conditions, water does not exhibit a phase change. Therefore no boiling phenomenon occurs in the reactor core. However, there exist a strong variation in the water density in the core as the temperature changes across the pseudo-critical value. The strong variation in the water density affects strongly to the neutron-physical behavior in the core. Therefore, for an accurate and detailed design analysis of a HPLWR core, coupled analysis of neutron-physics with thermal-hydraulics is mandatory. Although extensive activities have been carried worldwide on the design of super-critical pressure light water reactors, accurate design analysis with neutron-physical/thermal-hydraulic coupling is still very limited. In the present study, the Monte-Carlo code, MCNPX, is coupled with the sub-channel analysis code, STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions), which was developed specifically for fuel assemblies of supercritical water cooled reactors and is also flexible for complex fuel assembly designs. In this paper, a short description about both codes is given. The coupling methodology and procedure is presented and assessed. A

  4. Geometric optimization of spallation targets for the MYRRHA reactor using MCNPX simulations

    International Nuclear Information System (INIS)

    Rebello Junior, Andre Luiz P.; Martinez, Aquilino S.; Golcalves, Alessandro C.

    2013-01-01

    The present work aims to evaluate the behavior of neutron multiplicity in a spallation target using MCNPX simulations, focusing on its application in the MYRRHA reactor. It was studied the two types of spallation target proposed for the MYRRHA project, windowless and windows target, in order to compare them and nd saturation boundaries. Some saturation boundaries were found and the windowless target proved to be as viable as the windows one. Each one produced nearly the same number of neutrons per incident proton. Using the concept of neutron cost, it was also observed that the optimum conditions on neutron production occur at about 1GeV, for both target designs. (author)

  5. Geometric optimization of spallation targets for the MYRRHA reactor using MCNPX simulations

    Energy Technology Data Exchange (ETDEWEB)

    Rebello Junior, Andre Luiz P.; Martinez, Aquilino S.; Golcalves, Alessandro C., E-mail: junior.rebello@poli.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil). Dept. de Engenharia Nuclear

    2013-07-01

    The present work aims to evaluate the behavior of neutron multiplicity in a spallation target using MCNPX simulations, focusing on its application in the MYRRHA reactor. It was studied the two types of spallation target proposed for the MYRRHA project, windowless and windows target, in order to compare them and nd saturation boundaries. Some saturation boundaries were found and the windowless target proved to be as viable as the windows one. Each one produced nearly the same number of neutrons per incident proton. Using the concept of neutron cost, it was also observed that the optimum conditions on neutron production occur at about 1GeV, for both target designs. (author)

  6. Improvement of the WBC calibration of the Internal Dosimetry Laboratory of the CDTN/CNEN using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Guerra P, F.; Heeren de O, A. [Universidade Federal de Minas Gerais, Departamento de Engenharia Nuclear, Programa de Pos Graduacao em Ciencias e Tecnicas Nucleares, Av. Pte. Antonio Carlos 6627, 31270-901 Belo Horizonte, Minas Gerais (Brazil); Melo, B. M.; Lacerda, M. A. S.; Da Silva, T. A.; Ferreira F, T. C., E-mail: tcff01@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear, Programa de Pos Graduacao / CNEN, Av. Pte. Antonio Carlos 6627, 31270-901 Belo Horizonte, Minas Gerais (Brazil)

    2015-10-15

    The Plan of Radiological Protection licensed by the National Nuclear Energy Commission - CNEN in Brazil includes the risks of assessment of internal and external exposure by implementing a program of individual monitoring which is responsible of controlling exposures and ensuring the maintenance of radiation safety. The Laboratory of Internal Dosimetry of the Center for Development of Nuclear Technology - LID/CDTN is responsible for routine monitoring of internal contamination of the Individuals Occupationally Exposed (IOEs). These are, the IOEs involved in handling {sup 18}F produced by the Unit for Research and Production of Radiopharmaceuticals sources; as well a monitoring of the entire body of workers from the Research Reactor TRIGA IPR-R1/CDTN or whenever there is any risk of accidental incorporation. The determination of photon emitting radionuclides from the human body requires calibration techniques of the counting geometries, in order to obtain a curve of efficiency. The calibration process normally makes use of physical phantoms containing certified activities of the radionuclides of interest. The objective of this project is the calibration of the WBC facility of the LID/CDTN using the BOMAB physical phantom and Monte Carlo simulations. Three steps were needed to complete the calibration process. First, the BOMAB was filled with a KCl solution and several measurements of the gamma ray energy (1.46 MeV) emitted by {sup 40}K were done. Second, simulations using MCNPX code were performed to calculate the counting efficiency (Ce) for the BOMAB model phantom and compared with the measurements Ce results. Third and last step, the modeled BOMAB phantom was used to calculate the Ce covering the energy range of interest. The results showed a good agreement and are within the expected ratio between the measured and simulated results. (Author)

  7. Development of internal dosimetry protocols using the code MCNPx and voxelized phantoms of Reference of ICRP 110; Desenvolvimento de protocolos de dosimetria interna empregando o codigo MCNPx e fantomas voxelizados de referencia da ICRP 110

    Energy Technology Data Exchange (ETDEWEB)

    Mendes, B.M.; Fonseca, T.C.F., E-mail: bmm@cdtn.br [Centro de esenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Trindade, B.M.; Campos, T.P.R. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-04-01

    The objective of this work was to perform internal dosimetry calculations for {sup 18}F-FDG employing the MCNPx code and ICRP 110 voxelized reference phantoms (RCP{sub A}F and RCP{sub A}M). The methodologies developed and validated here represent protocols of internal dosimetry holding a better anthropomorphic and anthropometric representation of the human model in which heterogeneous distributions of the emissions can be adopted, useful in the study of new radiopharmaceuticals and internal contamination cases. The reference phantoms were implemented to run on MCNPx. Biodistribution data of {sup 18}F-FDG radiopharmaceutical provided in ICRP 128 were used in the simulations. The organs average absorbed doses and the effective doses were calculated for each model. The values obtained were compared with two reference works available in the literature for validation purposes. The means of the difference of our values and Zankl et al., 2012 reference values were -0.3% for RCP{sub A}M and -0.4% for RCP{sub A}F. Considering Hadid et al., 2013 reference values, the means of the deviation were -2.9% and -2.2% for RCP{sub A}M and RCP{sub A}F respectively. No statistically significant differences were observed (p <0.01) between the reference values and the values calculated by the internal dosimetry protocols developed by our group. Considering the {sup 18}F-FDG validation study performed in this work, the internal dosimetry protocols developed by our group have produced suitable dosimetry data. (author)

  8. Calculation of dose rate in escape channel of Research Irradiating Facility Army Technology Center using code MCNPX

    International Nuclear Information System (INIS)

    Gomes, Renato G.; Rebello, Wilson F.; Vellozo, Sergio O.; Moreira Junior, Luis; Vital, Helio C.; Rusin, Tiago; Silva, Ademir X.

    2013-01-01

    In order to evaluate new lines of research in the area of irradiation of materials external to the research irradiating facility Army Technology Center (CTEx), it is necessary to study security parameters and magnitude of the dose rates from their channels of escape. The objective was to calculate, with the code MCNPX, dose rates (Gy / min) on the interior and exterior of the four-channel leakage gamma irradiator. The channels were designed to leak radiation on materials properly disposed in the area outside the irradiator larger than the expected volume of irradiation chambers (50 liters). This study aims to assess the magnitude of dose rates within the channels, as well as calculate the angle of beam output range outside the channel for analysis as to its spread, and evaluation of safe conditions of their operators (protection radiological). The computer simulation was performed by distributing virtual dosimeter ferrous sulfate (Fricke) in the longitudinal axis of the vertical drain channels (anterior and posterior) and horizontal (top and bottom). The results showed a collimating the beams irradiated on each of the channels to the outside, with values of the order of tenths of Gy / min as compared to the maximum amount of operation of the irradiator chamber (33 Gy / min). The external beam irradiation in two vertical channels showed a distribution shaped 'trunk pyramid', not collimated, so scattered, opening angle 83 ° in the longitudinal direction and 88 in the transverse direction. Thus, the cases allowed the evaluation of materials for irradiation outside the radiator in terms of the magnitude of the dose rates and positioning of materials, and still be able to take the necessary care in mounting shield for radiation protection by operators, avoiding exposure to ionizing radiation. (author)

  9. Computational Model for the Neutronic Simulation of Pebble Bed Reactor’s Core Using MCNPX

    Directory of Open Access Journals (Sweden)

    J. Rosales

    2014-01-01

    Full Text Available Very high temperature reactor (VHTR designs offer promising performance characteristics; they can provide sustainable energy, improved proliferation resistance, inherent safety, and high temperature heat supply. These designs also promise operation to high burnup and large margins to fuel failure with excellent fission product retention via the TRISO fuel design. The pebble bed reactor (PBR is a design of gas cooled high temperature reactor, candidate for Generation IV of Nuclear Energy Systems. This paper describes the features of a detailed geometric computational model for PBR whole core analysis using the MCNPX code. The validation of the model was carried out using the HTR-10 benchmark. Results were compared with experimental data and calculations of other authors. In addition, sensitivity analysis of several parameters that could have influenced the results and the accuracy of model was made.

  10. Computational intercomparison of the mathematical model of a clinical accelerator LINAC 6MV using two different Monte Carlo codes: MCNPx and EGSnrc; Intercomparacao computacional do modelo matematico de um acelerador clinico LINAC 6MV utilizando dois codigos de Monte Carlo diferentes: MCNPx e EGSnrc

    Energy Technology Data Exchange (ETDEWEB)

    Castelo e Silva, L.A., E-mail: castelo@ifsp.edu.br [Instituto Federal de Sao Paulo (IFSP), SP (Brazil); Mendes, M.B.; Goncalves, B.R.; Santos, D.M.M.; Vieira, M.V.; Fonseca, R.L.M.; Zenobio, M.A.F.; Fonseca, T.C.F. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Paixao, L. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2016-07-01

    The main goal of this work is to publish the results of an inter-comparison simulation exercise of a clinical 10 x 10 cm{sup 2} beam model of a 6 MV LINAC using two different Monte Carlo codes: the MCNPX and EGSnrc. Results obtained for the dosimetric parameters PDD{sub 20,10} and TPR{sub 20,10} were compared with experimental data obtained in Radiotherapy and Megavoltage Institute of Minas Gerais. The main challenges on the computational modeling of this system are reported and discussed for didactic purposes in the area of modeling and simulation. (author)

  11. Parallel of semi-empirical results simulated by MCNP of X-ray spectra with a semiconductor; Paralelo de resultado semi- empiricos simulados por MCNPX de espectros de raios-X com um semicondutor

    Energy Technology Data Exchange (ETDEWEB)

    Santos, L.R.; Vivolo, V.; Potiens, M.P.A., E-mail: dossantos.lucasrodrigues@gmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Navarro, M.V.T.; Santos, W.S. [Universidade Federal de Uberlandia (INFIS/UFU), MG (Brazil). Instituto de Fisica

    2016-07-01

    The aim of this study was to use the MCNPX radiation transport code to simulate X-ray spectra generated by a constant voltage system in a CdTe semiconductor detector. As part of the validation process, we obtained a series of experimental spectra. Comparatively, in all cases there is a good correlation between the two spectra. There were no statistically significant differences between the experimental results with the simulated. (author)

  12. Comparison of electron dose-point kernels in water generated by the Monte Carlo codes, PENELOPE, GEANT4, MCNPX, and ETRAN.

    Science.gov (United States)

    Uusijärvi, Helena; Chouin, Nicolas; Bernhardt, Peter; Ferrer, Ludovic; Bardiès, Manuel; Forssell-Aronsson, Eva

    2009-08-01

    Point kernels describe the energy deposited at a certain distance from an isotropic point source and are useful for nuclear medicine dosimetry. They can be used for absorbed-dose calculations for sources of various shapes and are also a useful tool when comparing different Monte Carlo (MC) codes. The aim of this study was to compare point kernels calculated by using the mixed MC code, PENELOPE (v. 2006), with point kernels calculated by using the condensed-history MC codes, ETRAN, GEANT4 (v. 8.2), and MCNPX (v. 2.5.0). Point kernels for electrons with initial energies of 10, 100, 500, and 1 MeV were simulated with PENELOPE. Spherical shells were placed around an isotropic point source at distances from 0 to 1.2 times the continuous-slowing-down-approximation range (R(CSDA)). Detailed (event-by-event) simulations were performed for electrons with initial energies of less than 1 MeV. For 1-MeV electrons, multiple scattering was included for energy losses less than 10 keV. Energy losses greater than 10 keV were simulated in a detailed way. The point kernels generated were used to calculate cellular S-values for monoenergetic electron sources. The point kernels obtained by using PENELOPE and ETRAN were also used to calculate cellular S-values for the high-energy beta-emitter, 90Y, the medium-energy beta-emitter, 177Lu, and the low-energy electron emitter, 103mRh. These S-values were also compared with the Medical Internal Radiation Dose (MIRD) cellular S-values. The greatest differences between the point kernels (mean difference calculated for distances, electrons was 1.4%, 2.5%, and 6.9% for ETRAN, GEANT4, and MCNPX, respectively, compared to PENELOPE, if omitting the S-values when the activity was distributed on the cell surface for 10-keV electrons. The largest difference between the cellular S-values for the radionuclides, between PENELOPE and ETRAN, was seen for 177Lu (1.2%). There were large differences between the MIRD cellular S-values and those obtained from

  13. Monte Carlo simulation of the electron transport through thin slabs: A comparative study of PENELOPE, GEANT3, GEANT4, EGSnrc and MCNPX

    International Nuclear Information System (INIS)

    Vilches, M.; Garcia-Pareja, S.; Guerrero, R.; Anguiano, M.; Lallena, A.M.

    2007-01-01

    The Monte Carlo simulation of the electron transport through thin slabs is studied with five general purpose codes: PENELOPE, GEANT3, GEANT4, EGSnrc and MCNPX. The different material foils analyzed in the old experiments of Kulchitsky and Latyshev [L.A. Kulchitsky, G.D. Latyshev, Phys. Rev. 61 (1942) 254] and Hanson et al. [A.O. Hanson, L.H. Lanzl, E.M. Lyman, M.B. Scott, Phys. Rev. 84 (1951) 634] are used to perform the comparison between the Monte Carlo codes. Non-negligible differences are observed in the angular distributions of the transmitted electrons obtained with the some of the codes. The experimental data are reasonably well described by EGSnrc, PENELOPE (v.2005) and GEANT4. A general good agreement is found for EGSnrc and PENELOPE (v.2005) in all the cases analyzed

  14. Monte Carlo simulation of the electron transport through thin slabs: A comparative study of PENELOPE, GEANT3, GEANT4, EGSnrc and MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Vilches, M. [Servicio de Fisica y Proteccion Radiologica, Hospital Regional Universitario ' Virgen de las Nieves' , Avda. de las Fuerzas Armadas, 2, E-18014 Granada (Spain)]. E-mail: mvilches@ugr.es; Garcia-Pareja, S. [Servicio de Radiofisica Hospitalaria, Hospital Regional Universitario ' Carlos Haya' , Avda. Carlos Haya, s/n, E-29010 Malaga (Spain)]. E-mail: garciapareja@gmail.com; Guerrero, R. [Servicio de Radiofisica, Hospital Universitario ' San Cecilio' , Avda. Dr. Oloriz, 16, E-18012 Granada (Spain)]. E-mail: rafael.guerrero.alcalde.sspa@juntadeandalucia.es; Anguiano, M. [Departamento de Fisica Atomica, Molecular y Nuclear, Universidad de Granada, E-18071 Granada (Spain)]. E-mail: mangui@ugr.es; Lallena, A.M. [Departamento de Fisica Atomica, Molecular y Nuclear, Universidad de Granada, E-18071 Granada (Spain)]. E-mail: lallena@ugr.es

    2007-01-15

    The Monte Carlo simulation of the electron transport through thin slabs is studied with five general purpose codes: PENELOPE, GEANT3, GEANT4, EGSnrc and MCNPX. The different material foils analyzed in the old experiments of Kulchitsky and Latyshev [L.A. Kulchitsky, G.D. Latyshev, Phys. Rev. 61 (1942) 254] and Hanson et al. [A.O. Hanson, L.H. Lanzl, E.M. Lyman, M.B. Scott, Phys. Rev. 84 (1951) 634] are used to perform the comparison between the Monte Carlo codes. Non-negligible differences are observed in the angular distributions of the transmitted electrons obtained with the some of the codes. The experimental data are reasonably well described by EGSnrc, PENELOPE (v.2005) and GEANT4. A general good agreement is found for EGSnrc and PENELOPE (v.2005) in all the cases analyzed.

  15. Calculation of dose rate in escape channel of Research Irradiating Facility Army Technology Center using code MCNPX; Calculo das taxas de dose no canal de fuga do irradiador gama de pesquisa do Centro Tecnologico do Exercito utilizando o codigo MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Renato G.; Rebello, Wilson F.; Vellozo, Sergio O.; Moreira Junior, Luis, E-mail: renatoguedes@ime.eb.br, E-mail: rebello@ime.eb.br, E-mail: eng.cavaliere@gmail.com, E-mail: vellozo@cbpf.br, E-mail: luisjrmoreira@hotmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil); Vital, Helio C., E-mail: vital@ctex.eb.br [Centro Tecnologico do Exercito (CTEX), Barra de Guaratiba, RJ (Brazil); Rusin, Tiago, E-mail: tiago.rusin@mma.gov.br [Ministerio do Meio Ambiente (MMA), Brasilia, DF (Brazil); Silva, Ademir X., E-mail: ademir@con.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    In order to evaluate new lines of research in the area of irradiation of materials external to the research irradiating facility Army Technology Center (CTEx), it is necessary to study security parameters and magnitude of the dose rates from their channels of escape. The objective was to calculate, with the code MCNPX, dose rates (Gy / min) on the interior and exterior of the four-channel leakage gamma irradiator. The channels were designed to leak radiation on materials properly disposed in the area outside the irradiator larger than the expected volume of irradiation chambers (50 liters). This study aims to assess the magnitude of dose rates within the channels, as well as calculate the angle of beam output range outside the channel for analysis as to its spread, and evaluation of safe conditions of their operators (protection radiological). The computer simulation was performed by distributing virtual dosimeter ferrous sulfate (Fricke) in the longitudinal axis of the vertical drain channels (anterior and posterior) and horizontal (top and bottom). The results showed a collimating the beams irradiated on each of the channels to the outside, with values of the order of tenths of Gy / min as compared to the maximum amount of operation of the irradiator chamber (33 Gy / min). The external beam irradiation in two vertical channels showed a distribution shaped 'trunk pyramid', not collimated, so scattered, opening angle 83 ° in the longitudinal direction and 88 in the transverse direction. Thus, the cases allowed the evaluation of materials for irradiation outside the radiator in terms of the magnitude of the dose rates and positioning of materials, and still be able to take the necessary care in mounting shield for radiation protection by operators, avoiding exposure to ionizing radiation. (author)

  16. MCNPX and GEANT4 simulation of γ-ray polymeric shields

    Indian Academy of Sciences (India)

    safety to evaluate the reliability of MCNPX and GEANT4 results for the new materials, the same calculations were conducted for lead and tungsten which are .... medicine and nanotechnology edited by Brahim Attaf (Shanghai: InTech ISBN: 978-953-307-. 235-7, 2011) pp. 565–592. [5] Y Haruvy, Int. J. Radiat. Appl. Instrum.

  17. Spectral range calculation inside the Research Irradiating Facility Army Technology Center using code MCNPX and comparison with the spectra of energy Caesium 137 raised in laboratory

    International Nuclear Information System (INIS)

    Gomes, Renato G.; Rebello, Wilson F.; Cavaliere, Marcos Paulo; Vellozo, Sergio O.; Moreira Junior, Luis; Vital, Helio C.; Silva, Ademir X.

    2013-01-01

    Using the MCNPX code, the objective was to calculate by means of computer simulation spectroscopy range inside the irradiation chamber upper radiator gamma research irradiating facility Army Technology Center (CTEx). The calculations were performed in the spectral range usual 2 points for research purposes irradiating the energy spectra of gamma rays from the source of Cesium chloride 137. Sought the discretization of the spectrum in 100 channels at points of upper bound of 1cm higher and lower dose rates previously known. It was also conducted in the laboratory lifting the spectrum of Cesium-137 source using NaI scintillator detector and multichannel analyzer. With the source spectrum Cesium-137 contained in the literature and raised in the laboratory, both used as reference for comparison and analysis in terms of probability of emission maximum of 0.661 MeV The spectra were quite consistent in terms of the behavior of the energy distributions with scores. The position of maximum dose rate showed absorption detection almost maximum energy of 0.661 MeV photopeak In the spectrum of the position of minimum dosage rate, it was found that due to the removal of the source point of interest, some loss detection were caused by Compton scattering. (author)

  18. Obtaining of the cellular S values by means of Monte Carlo simulation with Penelope and MCNPX; Obtencion de los valores S celulares mediante simulacion Monte Carlo con PENELOPE y MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Rojas C, E. L.; Avila, O., E-mail: leticia.rojas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    In this work the simulation codes Monte Carlo, Penelope and MCNPX were used to calculate the doses by unit of accumulated activity S(N-N) in water spherical cells models of different radius exposed to mono-energetics electrons coming from punctual sources located in the center of the cellular nucleus. The studied cellular radii were: r{sub n}1=3 r{sub c}1=6; r{sub n}2=5 and r{sub c}2=10; r{sub n}3=9 and r{sub c}3=10 {mu}m; being r{sub n} and r{sub c} the nuclear and cellular radius, respectively. The following initial energies of the electrons were considered: 1, 5, 10, 50, 100, 500, 700 and 1000 keV. Additionally values S(N-N) were calculated for spherical cells of r= 3 {mu}m r{sub c}= 6 {mu}m due to the electrons coming from sources of {sup 111}In, {sup 177}Lu, {sup 99m}Tc, {sup 188}Re and {sup 186}Re. The obtained values are compared with those calculated by the MIRD Committee internationally accepted. The percentage differences between the values reported by this Committee and those calculated by Monte Carlo simulation are inside the interval that is considered valid for this dosimetry type. A major concordance was found among the values calculated by Monte Carlo simulation that among those calculated by MIRD and those obtained by simulation. Considering validated the use of both codes for similar applications, the values S(N-N) and S(N-C y) were obtained of prostate cancer real cells models of the PC3 line. The results were compared among them. The values of S(N-N) obtained with Penelope for the PC3 cells for the electron emissions of {sup 111}In, {sup 177}Lu, {sup 99m}Tc, {sup 188}Re and {sup 186}Re are: 3.19e{sup {sub {sup 4}}}, 3.24e{sup -4}, 1.37e{sup -4}, 1.11e{sup -4} and 1.91e{sup -4} Gy/Bq-s, respectively. Also the obtained results for S(N-C y) are: 2.95e{sup -6}, 3.17e{sup -5}, 2.09e{sup -6}, 1.41e{sup -5}, 1.86e{sup -5} Gy/Bq-s. (Author)

  19. Nuclear densimeter of soil simulated in MCNP-4C code

    Energy Technology Data Exchange (ETDEWEB)

    Braga, Mario R.M.S.S.; Penna, Rodrigo; Vasconcelos, Danilo C.; Pereira, Claubia; Guerra, Bruno T., E-mail: mario@nuclear.ufmg.b [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear; Silva, Clemente J.G.C., E-mail: clementecarneito@yahoo.com.b [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Dept. de Ciencias Exatas e Tecnologicas

    2009-07-01

    The Monte Carlo code (MCNPX) was used to simulate a nuclear densimeter for measuring soil density. An Americium source (E = 60 keV) and a NaI (Tl) detector were placed on soil surface. Results from MCNP shown that scattered photon fluxes may be used to determining soil density. Linear regressions between scattered photons fluxes and soil density were calculated and shown correlation coefficients near unity. (author)

  20. Nuclear densimeter of soil simulated in MCNP-4C code

    International Nuclear Information System (INIS)

    Braga, Mario R.M.S.S.; Penna, Rodrigo; Vasconcelos, Danilo C.; Pereira, Claubia; Guerra, Bruno T.; Silva, Clemente J.G.C.

    2009-01-01

    The Monte Carlo code (MCNPX) was used to simulate a nuclear densimeter for measuring soil density. An Americium source (E = 60 keV) and a NaI (Tl) detector were placed on soil surface. Results from MCNP shown that scattered photon fluxes may be used to determining soil density. Linear regressions between scattered photons fluxes and soil density were calculated and shown correlation coefficients near unity. (author)

  1. MCNPX Model/Table Comparison

    CERN Document Server

    Hendricks, J S

    2003-01-01

    MCNPX is a Monte Carlo N-Particle radiation transport code extending the capabilities of MCNP4C. As with MCNP, MCNPX uses nuclear data tables to transport neutrons, photons, and electrons. Unlike MCNP, MCNPX also uses (1) nuclear data tables to transport protons; (2) physics models to transport 30 additional particle types (deuterons, tritons, alphas, pions, muons, etc.); and (3) physics models to transport neutrons and protons when no tabular data are available or when the data are above the energy range (20 to 150 MeV) where the data tables end. MCNPX can mix and match data tables and physics models throughout a problem. For example, MCNPX can model neutron transport in a bismuth germinate (BGO) particle detector by using data tables for bismuth and oxygen and using physics models for germanium. Also, MCNPX can model neutron transport in UO sub 2 , making the best use of physics models and data tables: below 20 MeV, data tables are used; above 150 MeV, physics models are used; between 20 and 150 MeV, data t...

  2. MCNPX Model/Table Comparison

    International Nuclear Information System (INIS)

    Hendricks, J.S.

    2003-01-01

    MCNPX is a Monte Carlo N-Particle radiation transport code extending the capabilities of MCNP4C. As with MCNP, MCNPX uses nuclear data tables to transport neutrons, photons, and electrons. Unlike MCNP, MCNPX also uses (1) nuclear data tables to transport protons; (2) physics models to transport 30 additional particle types (deuterons, tritons, alphas, pions, muons, etc.); and (3) physics models to transport neutrons and protons when no tabular data are available or when the data are above the energy range (20 to 150 MeV) where the data tables end. MCNPX can mix and match data tables and physics models throughout a problem. For example, MCNPX can model neutron transport in a bismuth germinate (BGO) particle detector by using data tables for bismuth and oxygen and using physics models for germanium. Also, MCNPX can model neutron transport in UO 2 , making the best use of physics models and data tables: below 20 MeV, data tables are used; above 150 MeV, physics models are used; between 20 and 150 MeV, data tables are used for oxygen and models are used for uranium. The mix-and-match capability became available with MCNPX2.5.b (November 2002). For the first time, we present here comparisons that calculate radiation transport in materials with various combinations of data charts and model physics. The physics models are poor at low energies (<150 MeV); thus, data tables should be used when available. Our comparisons demonstrate the importance of the mix-and-match capability and indicate how well physics models work in the absence of data tables

  3. Towards advanced code simulators

    International Nuclear Information System (INIS)

    Scriven, A.H.

    1990-01-01

    The Central Electricity Generating Board (CEGB) uses advanced thermohydraulic codes extensively to support PWR safety analyses. A system has been developed to allow fully interactive execution of any code with graphical simulation of the operator desk and mimic display. The system operates in a virtual machine environment, with the thermohydraulic code executing in one virtual machine, communicating via interrupts with any number of other virtual machines each running other programs and graphics drivers. The driver code itself does not have to be modified from its normal batch form. Shortly following the release of RELAP5 MOD1 in IBM compatible form in 1983, this code was used as the driver for this system. When RELAP5 MOD2 became available, it was adopted with no changes needed in the basic system. Overall the system has been used for some 5 years for the analysis of LOBI tests, full scale plant studies and for simple what-if studies. For gaining rapid understanding of system dependencies it has proved invaluable. The graphical mimic system, being independent of the driver code, has also been used with other codes to study core rewetting, to replay results obtained from batch jobs on a CRAY2 computer system and to display suitably processed experimental results from the LOBI facility to aid interpretation. For the above work real-time execution was not necessary. Current work now centers on implementing the RELAP 5 code on a true parallel architecture machine. Marconi Simulation have been contracted to investigate the feasibility of using upwards of 100 processors, each capable of a peak of 30 MIPS to run a highly detailed RELAP5 model in real time, complete with specially written 3D core neutronics and balance of plant models. This paper describes the experience of using RELAP5 as an analyzer/simulator, and outlines the proposed methods and problems associated with parallel execution of RELAP5

  4. Tokamak simulation code manual

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-01-01

    The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs.

  5. Tokamak simulation code manual

    International Nuclear Information System (INIS)

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won

    1995-01-01

    The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs

  6. Full core analysis of IRIS reactor by using MCNPX.

    Science.gov (United States)

    Amin, E A; Bashter, I I; Hassan, Nabil M; Mustafa, S S

    2016-07-01

    This paper describes neutronic analysis for fresh fuelled IRIS (International Reactor Innovative and Secure) reactor by MCNPX code. The analysis included criticality calculations, radial power and axial power distribution, nuclear peaking factor and axial offset percent at the beginning of fuel cycle. The effective multiplication factor obtained by MCNPX code is compared with previous calculations by HELIOS/NESTLE, CASMO/SIMULATE, modified CORD-2 nodal calculations and SAS2H/KENO-V code systems. It is found that k-eff value obtained by MCNPX is closer to CORD-2 value. The radial and axial powers are compared with other published results carried out using SAS2H/KENO-V code. Moreover, the WIMS-D5 code is used for studying the effect of enriched boron in form of ZrB2 on the effective multiplication factor (K-eff) of the fuel pin. In this part of calculation, K-eff is calculated at different concentrations of Boron-10 in mg/cm at different stages of burnup of unit cell. The results of this part are compared with published results performed by HELIOS code. Copyright © 2016 Elsevier Ltd. All rights reserved.

  7. Determination of absorbed dose distribution in water for COC ophthalmic applicator of {sup 106}Ru/{sup 106}Rh using Monte Carlo code-MCNPX; Determinacao da distribuicao de dose absorvida na agua para o aplicador oftalmico COC de {sup 106}Ru/{sup 106}Rh utilizando o codigo de Monte Carlo - MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Barbosa, Nilseia A.; Rosa, Luiz A. Ribeiro da, E-mail: nilseia@ird.gov.br, E-mail: lrosa@ird.gov.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ),Rio de Janeiro, RJ (Brazil); Braz, Delson, E-mail: delson@nuclear.ufrj.br [Coordenacao dos programas de Pos-Graduacao em Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2014-07-01

    The COC ophthalmic applicators using beta radiation source of {sup 106}Ru/{sup 106}Rh are used in the treatment of intraocular tumors near the optic nerve. In this type of treatment is very important to know the dose distribution in order to provide the best possible delivery of prescribed dose to the tumor, preserves the optic nerve region extremely critical, that if damaged, can compromise the patient's visual acuity, and cause brain sequelae. These dose distributions are complex and doctors, who will have the responsibility on the therapy, only have the source calibration certificate provided by the manufacturer Eckert and Ziegler BEBIG GmbH. These certificates provide 10 absorbed dose values at water depth along the central axis applicator with the uncertainties of the order of 20% isodose and in a plane located 1 mm from the applicator surface. Thus, it is important to know with more detail and precision the dose distributions in water generated by such applicators. To this end, the Monte Carlo simulation was used using MCNPX code. Initially, was validated the simulation by comparing the obtained results to the central axis of the applicator with those provided by the certificate. The different percentages were lower than 5%, validating the used method. Lateral dose profile was calculated for 6 different depths in intervals of 1 mm and the dose rates in mGy.min{sup -1} for the same depths.

  8. Absorbed dose calculation from beta and gamma rays of 131I in ellipsoidal thyroid and other organs of neck with MCNPX code

    Directory of Open Access Journals (Sweden)

    Mohammad Mirzaie

    2012-09-01

    Full Text Available Background: The 131I radioisotope is used for diagnosis and treatment of hyperthyroidism and thyroid cancer. In optimized Iodine therapy, a specific dose must be reached to the thyroid gland with minimum radiation to the cervical spine, cervical vertebrae, neck tissue, subcutaneous fat and skin. Dose measurement inside the alive organ is difficult therefore the aim of this research was dose calculation in the organs by MCNPX code. Materials and Methods: First of all, the input file for MCNPX code has been prepared to calculate F6 and F8 tallies for ellipsoidal thyroid lobes with long axes is tow times of short axes which the 131I is distributed uniformly inside the lobes. Then the code has been run for F6 and F8 tallies for variation of lobe volume from 1 to 25 milliliters. From the output file of tally F6, the gamma absorbed dose in ellipsoidal thyroid, spinal neck, neck bone, neck tissue, subcutaneous fat layer and skin for the volume lobe variation from 1 ml to 25 ml have been derived and the graphs are drew. As well as, form the output of F8 tally the absorbed energy of beta in thyroid and soft tissue of neck is obtained and listed in the table and then absorbed dose of bate has been calculated. Results: The results of this research show that for constant activity in thyroid, the absorbed dose of gamma decreases about 88.3% in thyroid, 6.9% at soft tissue, 19.3% in adipose layer and 17.4% in skin, but it increases 32.1% in spinal of neck and 32.3% in neck bone when the lobe volume varied from 1 to 25 milliliters. For the same situation, the beta absorbed dose decreases 95.9% in thyroid and 64.2% in soft tissue. Conclusion: For the constant activity in thyroid by increasing the thyroid volume, absorbed dose of gamma in thyroid and soft tissue of neck, adipose layer under the skin and skin of neck decreased, but it increased at spinal of neck and neck bone. Also, by increasing of the lobe volume in constant activity, the beta absorbed dose

  9. Study of the radioactive particle tracking technique using gamma-ray attenuation and MCNP-X code to evaluate industrial agitators

    Energy Technology Data Exchange (ETDEWEB)

    Dam, Roos Sophia de F.; Salgado, César M., E-mail: rsophia.dam@gmail.com, E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    Agitators or mixers are highly used in the chemical, food, pharmaceutical and cosmetic industries. During the fabrication process, the equipment may fail and compromise the appropriate stirring or mixing procedure. Besides that, it is also important to determine the right point of homogeneity of the mixture. Thus, it is very important to have a diagnosis tool for these industrial units to assure the quality of the product and to keep the market competitiveness. The radioactive particle tracking (RPT) technique is widely used in the nuclear field. In this paper, a method based on the principles of the RPT technique is presented. Counts obtained by an array of detectors properly positioned around the unit will be correlated to predict the instantaneous positions occupied by the radioactive particle by means of an appropriate mathematical search location algorithm. Detection geometry developed employs eight NaI(Tl) scintillator detectors and a Cs-137 (662 keV) source with isotropic emission of gamma-rays. The modeling of the detection system is performed using the Monte Carlo Method, by means of the MCNP-X code. In this work a methodology is presented to predict the position of a radioactive particle to evaluate the performance of agitators in industrial units by means of an Artificial Neural Network (ANN). (author)

  10. Burn-up calculation of different thorium-based fuel matrixes in a thermal research reactor using MCNPX 2.6 code

    Directory of Open Access Journals (Sweden)

    Gholamzadeh Zohreh

    2014-12-01

    Full Text Available Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view

  11. Importance of All-in-one (MCNPX2.7.0+CINDER2008) Code for Rigorous Transmutation Study

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Oyeon [Institute for Modeling and Simulation Convergence, Daegu (Korea, Republic of); Kim, Kwanghyun [RadTek Co. Ltd., Daejeon (Korea, Republic of)

    2015-10-15

    It can be utilized as a possible mechanism for reducing the volume and hazard of radioactive waste by transforming hazardous radioactive elements with long half-life into less hazardous elements with short halflife. Thus, the understanding of the transmutation mechanism and beneficial machinery design technologies are important and useful. Although the terminology transmutation was rooted back to alchemy which transforms the base metals into gold in the middle ages, Rutherford and Soddy were the first observers by discovering the natural transmutation as a part of radioactive decay of the alpha decay type in early 20th century. Along with the development of computing technology, analysis software, for example, CINDER was developed for rigorous atomic transmutation study. The code has a long history of development from the original work of T. England at Bettis Atomic Power Laboratory (BAPL) in the early 1960s. It has been used to calculate the inventory of nuclides in an irradiated material. CINDER'90 which is recently released involved an upgrade of the code to allow the spontaneous tracking of chains based upon the significant density or pass-by of a nuclide, where pass-by represents the density of a nuclide transforming to other nuclides. Nuclear transmutation process is governed by highly non-linear differential equation. Chaotic nature of the non-linear equation bespeaks the importance of the accurate input data (i.e. number of significant digits). Thus, reducing the human interrogation is very important for the rigorous transmutation study and 'allin- one' code structure is desired. Note that non-linear characteristic of the transmutation equation caused by the flux changes due to the number density change during a given time interval (intrinsic physical phenomena) is not considered in this study. In this study, we only emphasized the effects of human interrogation in the computing process solving nonlinear differential equations, as shown in

  12. Development of the MCNPX depletion capability: A Monte Carlo linked depletion method that automates the coupling between MCNPX and CINDER90 for high fidelity burnup calculations

    Science.gov (United States)

    Fensin, Michael Lorne

    Monte Carlo-linked depletion methods have gained recent interest due to the ability to more accurately model complex 3-dimesional geometries and better track the evolution of temporal nuclide inventory by simulating the actual physical process utilizing continuous energy coefficients. The integration of CINDER90 into the MCNPX Monte Carlo radiation transport code provides a high-fidelity completely self-contained Monte-Carlo-linked depletion capability in a well established, widely accepted Monte Carlo radiation transport code that is compatible with most nuclear criticality (KCODE) particle tracking features in MCNPX. MCNPX depletion tracks all necessary reaction rates and follows as many isotopes as cross section data permits in order to achieve a highly accurate temporal nuclide inventory solution. This work chronicles relevant nuclear history, surveys current methodologies of depletion theory, details the methodology in applied MCNPX and provides benchmark results for three independent OECD/NEA benchmarks. Relevant nuclear history, from the Oklo reactor two billion years ago to the current major United States nuclear fuel cycle development programs, is addressed in order to supply the motivation for the development of this technology. A survey of current reaction rate and temporal nuclide inventory techniques is then provided to offer justification for the depletion strategy applied within MCNPX. The MCNPX depletion strategy is then dissected and each code feature is detailed chronicling the methodology development from the original linking of MONTEBURNS and MCNP to the most recent public release of the integrated capability (MCNPX 2.6.F). Calculation results of the OECD/NEA Phase IB benchmark, H. B. Robinson benchmark and OECD/NEA Phase IVB are then provided. The acceptable results of these calculations offer sufficient confidence in the predictive capability of the MCNPX depletion method. This capability sets up a significant foundation, in a well established

  13. OPTIMIZATION OF A NEUTRON BEAM SHAPING ASSEMBLY DESIGN FOR BNCT AND ITS DOSIMETRY SIMULATION BASED ON MCNPX

    Directory of Open Access Journals (Sweden)

    I Made Ardana

    2017-10-01

    OPTIMASI DESAIN KOLIMATOR NEUTRON UNTUK SISTEM BNCT DAN UJI DOSIMETRINYA MENGGUNAKAN PROGRAM MCNPX. Telah dilakukan penelitian tentang sistem BNCT yang meliputi dua tahapan simulasi dengan menggunakan program MCNPX yaitu uji simulasi untuk optimasi desain kolimator neutron untuk sistem BNCT berbasis Siklotron 30 MeV dan uji simulasi untuk menghitung fluks neutron dan dosimetri radiasi pada kanker sarkoma jaringan lunak pada leher dan kepala. Tujuan simulasi untuk mendapatkan desain kolimator yang paling optimal dalam memoderasi fluks neutron cepat yang dihasilkan dari sistem target berilium sehingga dapat dihasilkan fluks neutron yang sesuai untuk sistem BNCT. Uji optimasi dilakukan dengan cara memvariasikan bahan dan ketebalan masing-masing komponen dalam kolimator seperi reflektor, moderator, filter neutron cepat, filter neutron thermal, filter radiasi gamma dan lubang keluaran. Desain kolimator yang diperoleh dari hasil optimasi tersusun atas moderator berbahan Al dengan ketebalan 39 cm, filter neutron cepat berbahan LiF2 setebal 8,2 cm, dan filter neutron thermal berbahan B4C setebal 0,5 cm. Untuk reflektor, filter radiasi gamma dan lubang keluaran masing-masing menggunakan bahan PbF2, Pb dan Bi. Fluks neutron epithermal yang dihasilkan dari kolimator yang didesain adalah sebesar 2,83 x 109 n/s cm-2 dan telah memenuhi seluruh parameter fluks neutron yang sesuai untuk sistem BNCT. Selanjutnya uji simulasi dosimetri pada kanker sarkoma jaringan lunak pada leher dan kepala dilakukan dengan cara memvariasikan konsentrasi senyawa boron pada model phantom leher manusia (ORNL. Selanjutnya model phantom tersebut diiradiasi dengan fluks neutron yang berasal dari kolimator yang telah didesain sebelumnya. Hasilnya, fluks neutron thermal mencapai nilai tertinggi pada kedalaman 4,8 cm di dalam model phantom leher ORNL dengan laju dosis tertinggi terletak pada area jaringan kanker. Untuk masing-masing variasi konsentrasi senyawa boron pada model phantom leher ORNL supaya

  14. LFSC - Linac Feedback Simulation Code

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, Valentin; /Fermilab

    2008-05-01

    The computer program LFSC (Simulation Code>) is a numerical tool for simulation beam based feedback in high performance linacs. The code LFSC is based on the earlier version developed by a collective of authors at SLAC (L.Hendrickson, R. McEwen, T. Himel, H. Shoaee, S. Shah, P. Emma, P. Schultz) during 1990-2005. That code was successively used in simulation of SLC, TESLA, CLIC and NLC projects. It can simulate as pulse-to-pulse feedback on timescale corresponding to 5-100 Hz, as slower feedbacks, operating in the 0.1-1 Hz range in the Main Linac and Beam Delivery System. The code LFSC is running under Matlab for MS Windows operating system. It contains about 30,000 lines of source code in more than 260 subroutines. The code uses the LIAR ('Linear Accelerator Research code') for particle tracking under ground motion and technical noise perturbations. It uses the Guinea Pig code to simulate the luminosity performance. A set of input files includes the lattice description (XSIF format), and plane text files with numerical parameters, wake fields, ground motion data etc. The Matlab environment provides a flexible system for graphical output.

  15. LFSC - Linac Feedback Simulation Code

    International Nuclear Information System (INIS)

    Ivanov, Valentin; Fermilab

    2008-01-01

    The computer program LFSC ( ) is a numerical tool for simulation beam based feedback in high performance linacs. The code LFSC is based on the earlier version developed by a collective of authors at SLAC (L.Hendrickson, R. McEwen, T. Himel, H. Shoaee, S. Shah, P. Emma, P. Schultz) during 1990-2005. That code was successively used in simulation of SLC, TESLA, CLIC and NLC projects. It can simulate as pulse-to-pulse feedback on timescale corresponding to 5-100 Hz, as slower feedbacks, operating in the 0.1-1 Hz range in the Main Linac and Beam Delivery System. The code LFSC is running under Matlab for MS Windows operating system. It contains about 30,000 lines of source code in more than 260 subroutines. The code uses the LIAR ('Linear Accelerator Research code') for particle tracking under ground motion and technical noise perturbations. It uses the Guinea Pig code to simulate the luminosity performance. A set of input files includes the lattice description (XSIF format), and plane text files with numerical parameters, wake fields, ground motion data etc. The Matlab environment provides a flexible system for graphical output

  16. Neutronic computational modeling of the ASTRA critical facility using MCNPX

    International Nuclear Information System (INIS)

    Rodriguez, L. P.; Garcia, C. R.; Milian, D.; Milian, E. E.; Brayner, C.

    2015-01-01

    The Pebble Bed Very High Temperature Reactor is considered as a prominent candidate among Generation IV nuclear energy systems. Nevertheless the Pebble Bed Very High Temperature Reactor faces an important challenge due to the insufficient validation of computer codes currently available for use in its design and safety analysis. In this paper a detailed IAEA computational benchmark announced by IAEA-TECDOC-1694 in the framework of the Coordinated Research Project 'Evaluation of High Temperature Gas Cooled Reactor (HTGR) Performance' was solved in support of the Generation IV computer codes validation effort using MCNPX ver. 2.6e computational code. In the IAEA-TECDOC-1694 were summarized a set of four calculational benchmark problems performed at the ASTRA critical facility. Benchmark problems include criticality experiments, control rod worth measurements and reactivity measurements. The ASTRA Critical Facility at the Kurchatov Institute in Moscow was used to simulate the neutronic behavior of nuclear pebble bed reactors. (Author)

  17. Modeling and analysis of anti-scatter grids for analogical and digital systems using MCNPX

    International Nuclear Information System (INIS)

    Correa, Samanda C.A.; Souza, Edmilson M.; Silva, Ademir X.; Lopes, Ricardo T.

    2007-01-01

    Monte Carlo code MCNPX was used for modeling of the antiscatter grids. The performance analysis of grid was investigated for analogical and digital systems by calculation of contrast improvement factor (CIF), bucky factor (BF) and signal improvement factor (SIF). The accuracy of simulation of the grids was validated through comparison with measured results demonstrating good agreement, thus increasing the reliability of the results presented in this paper. (author)

  18. Two dimensional plasma simulation code

    International Nuclear Information System (INIS)

    Hazak, G.; Boneh, Y.; Goshen, Sh.; Oreg, J.

    1977-03-01

    An electrostatic two-dimensional particle code for plasma simulation is described. Boundary conditions which take into account the finiteness of the system are presented. An analytic solution for the case of crossed fields plasma acceleration is derived. This solution serves as a check on a computer test run

  19. Proton beam simulation with MCNPX/CINDER'90: Germanium metal activation estimates below 30MeV relevant to the bulk production of arsenic radioisotopes.

    Science.gov (United States)

    Fassbender, M; Taylor, W; Vieira, D; Nortier, M; Bach, H; John, K

    2012-01-01

    Germanium metal targets encapsulated in Nb shells were irradiated in a proton beam. Proton and secondary neutron beam fluences as well as radionuclide activity formation were modeled using MCNPX in combination with CINDER90. Targets were chemically processed using distillation and anion exchange. Good agreement between the measured radiochemical yields and MCNPX/CINDER90 estimates was observed. A target of pentavalent (73,74)As radioarsenic for neutron activation studies was prepared. Copyright © 2011 Elsevier Ltd. All rights reserved.

  20. Monte Carlo simulation of a coded-aperture thermal neutron camera

    International Nuclear Information System (INIS)

    Dioszegi, I.; Salwen, C.; Forman, L.

    2011-01-01

    We employed the MCNPX Monte Carlo code to simulate image formation in a coded-aperture thermal-neutron camera. The camera, developed at Brookhaven National Laboratory (BNL), consists of a 20 x 17 cm 2 active area 3 He-filled position-sensitive wire chamber in a cadmium enclosure box. The front of the box is a coded-aperture cadmium mask (at present with three different resolutions). We tested the detector experimentally with various arrangements of moderated point-neutron sources. The purpose of using the Monte Carlo modeling was to develop an easily modifiable model of the device to predict the detector's behavior using different mask patterns, and also to generate images of extended-area sources or large numbers (up to ten) of them, that is important for nonproliferation and arms-control verification, but difficult to achieve experimentally. In the model, we utilized the advanced geometry capabilities of the MCNPX code to simulate the coded aperture mask. Furthermore, the code simulated the production of thermal neutrons from fission sources surrounded by a thermalizer. With this code we also determined the thermal-neutron shadow cast by the cadmium mask; the calculations encompassed fast- and epithermal-neutrons penetrating into the detector through the mask. Since the process of signal production in 3 He-filled position-sensitive wire chambers is well known, we omitted this part from our modeling. Simplified efficiency values were used for the three (thermal, epithermal, and fast) neutron-energy regions. Electronic noise and the room's background were included as a uniform irradiation component. We processed the experimental- and simulated-images using identical LabVIEW virtual instruments. (author)

  1. Computer Code for Nanostructure Simulation

    Science.gov (United States)

    Filikhin, Igor; Vlahovic, Branislav

    2009-01-01

    Due to their small size, nanostructures can have stress and thermal gradients that are larger than any macroscopic analogue. These gradients can lead to specific regions that are susceptible to failure via processes such as plastic deformation by dislocation emission, chemical debonding, and interfacial alloying. A program has been developed that rigorously simulates and predicts optoelectronic properties of nanostructures of virtually any geometrical complexity and material composition. It can be used in simulations of energy level structure, wave functions, density of states of spatially configured phonon-coupled electrons, excitons in quantum dots, quantum rings, quantum ring complexes, and more. The code can be used to calculate stress distributions and thermal transport properties for a variety of nanostructures and interfaces, transport and scattering at nanoscale interfaces and surfaces under various stress states, and alloy compositional gradients. The code allows users to perform modeling of charge transport processes through quantum-dot (QD) arrays as functions of inter-dot distance, array order versus disorder, QD orientation, shape, size, and chemical composition for applications in photovoltaics and physical properties of QD-based biochemical sensors. The code can be used to study the hot exciton formation/relation dynamics in arrays of QDs of different shapes and sizes at different temperatures. It also can be used to understand the relation among the deposition parameters and inherent stresses, strain deformation, heat flow, and failure of nanostructures.

  2. Monte Carlo simulation code modernization

    CERN Multimedia

    CERN. Geneva

    2015-01-01

    The continual development of sophisticated transport simulation algorithms allows increasingly accurate description of the effect of the passage of particles through matter. This modelling capability finds applications in a large spectrum of fields from medicine to astrophysics, and of course HEP. These new capabilities however come at the cost of a greater computational intensity of the new models, which has the effect of increasing the demands of computing resources. This is particularly true for HEP, where the demand for more simulation are driven by the need of both more accuracy and more precision, i.e. better models and more events. Usually HEP has relied on the "Moore's law" evolution, but since almost ten years the increase in clock speed has withered and computing capacity comes in the form of hardware architectures of many-core or accelerated processors. To harness these opportunities we need to adapt our code to concurrent programming models taking advantages of both SIMD and SIMT architectures. Th...

  3. Analysis of different Monte Carlo simulation codes for its use in radiotherapy

    International Nuclear Information System (INIS)

    Azorin V, C.G.; Rivera M, T.

    2007-01-01

    Full text: At the present time many computer programs that simulate the radiation interaction with the matter using the Monte Carlo method. Presently work is carried out the comparative analysis of four of these codes (MCNPX, EGS4, GEANT, PENELOPE) for their later one use in the development of a simple algorithm that simulates the energy deposit when passing through the matter in patients subjected to radiotherapy. The results of the analysis show that the studied simulators model the interaction of almost all type of particles with the matter, although they have their variations among those the energy intervals that manage, the programming language in which are programmed, as well as the platform under which they are executed can be mentioned. (Author)

  4. Modeling of a double fission chamber using MCNPX for power calibration at the zero-power teaching reactor CROCUS

    International Nuclear Information System (INIS)

    Girardin, Gaetan; Epiney, Aaron; Joneja, Om Parkash

    2010-01-01

    MCNPX-2.5 simulations and experiments were performed to improve the power prediction of the zero-power teaching reactor CROCUS at the Ecole Polytechnique Federale de Lausanne (EPFL) using a calibrated double fission chamber (DFC). The CROCUS facility is a zero-power critical reactor used for educational purposes. Traditionally, the core power is determined by irradiating thin gold foils placed along the core centre and by measuring the 411 keV γ-rays on HPGe detectors. The average 197 Au(n,γ) self-shielded macroscopic cross-section obtained with the deterministic BOXER code (1σ - 10%) is employed to determine the flux and the reactor power. To benchmark the BOXER calculations, a DFC containing known amounts of enriched 235 U and 239 Pu deposits was installed within the reflector core and simulated with MCNPX-2.5/JEF-2.2. Particular care was taken to model the fissile deposits allowing to reduce the power uncertainty to 2% compared to the gold foil technique. A code-to-code comparison (BOXER vs. MCNPX) was performed and the results have shown a good agreement (2 to 5%) for most of the quantities calculated (flux, reaction rates). However, the normalization factor differed by 17% (flux-to-power ratio). Consequently, the core power was overestimated by 17% until now. Finally, the current investigations lead to an improved fission power determination and contribute to better core safety standard. (author)

  5. Reactive transport codes for subsurface environmental simulation

    NARCIS (Netherlands)

    Steefel, C.I.; Appelo, C.A.J.; Arora, B.; Kalbacher, D.; Kolditz, O.; Lagneau, V.; Lichtner, P.C.; Mayer, K.U.; Meeussen, J.C.L.; Molins, S.; Moulton, D.; Shao, D.; Simunek, J.; Spycher, N.; Yabusaki, S.B.; Yeh, G.T.

    2015-01-01

    A general description of the mathematical and numerical formulations used in modern numerical reactive transport codes relevant for subsurface environmental simulations is presented. The formulations are followed by short descriptions of commonly used and available subsurface simulators that

  6. Dose measurement aboard biosatellite BION-M no. 1 (AutoCAD implementation into MCNPX)

    International Nuclear Information System (INIS)

    Dolloso, P.; Miller, A.; Machrafi, R.; Shurshakov, V.; Khulapko, S.; Ivanova, O.

    2015-01-01

    This work outlines the first AutoCAD implementation of the BION-M no. 1 space craft in MCNPX using a simplified model. The MCNPX data obtained from the simulation can be compared with measurements taken from different radiation detectors immediately after the vehicle landing. (author)

  7. Dose measurement aboard biosatellite BION-M no. 1 (AutoCAD implementation into MCNPX)

    Energy Technology Data Exchange (ETDEWEB)

    Dolloso, P.; Miller, A.; Machrafi, R. [University of Ontario Institute of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, ON (Canada); Shurshakov, V.; Khulapko, S.; Ivanova, O. [Institute of Biomedical Problems, Moscow (Russian Federation)

    2015-07-01

    This work outlines the first AutoCAD implementation of the BION-M no. 1 space craft in MCNPX using a simplified model. The MCNPX data obtained from the simulation can be compared with measurements taken from different radiation detectors immediately after the vehicle landing. (author)

  8. Application of the MCNPX-McStas interface for shielding calculations and guide design at ESS

    DEFF Research Database (Denmark)

    Klinkby, Esben Bryndt; Bergbäck Knudsen, Erik; Willendrup, Peter Kjær

    2014-01-01

    . The generation and moderation of neutrons is simulated using a full scale MCNPX model of the ESS target monolith. Upon entering the neutron beam extraction region, the individual neutron states are handed to McStas via the MCNPX-McStas interface. McStas transports the neutrons through the beam guide...

  9. User's manual of Tokamak Simulation Code

    International Nuclear Information System (INIS)

    Nakamura, Yukiharu; Nishino, Tooru; Tsunematsu, Toshihide; Sugihara, Masayoshi.

    1992-12-01

    User's manual for use of Tokamak Simulation Code (TSC), which simulates the time-evolutional process of deformable motion of axisymmetric toroidal plasma, is summarized. For the use at JAERI computer system, the TSC is linked with the data management system GAEA. This manual is forcused on the procedure for the input and output by using the GAEA system. Model equations to give axisymmetric motion, outline of code system, optimal method to get the well converged solution are also described. (author)

  10. Dosimetric reconstruction of radiological accident by numerical simulations by means associating an anthropomorphic model and a Monte Carlo computation code

    International Nuclear Information System (INIS)

    Courageot, Estelle

    2010-01-01

    After a description of the context of radiological accidents (definition, history, context, exposure types, associated clinic symptoms of irradiation and contamination, medical treatment, return on experience) and a presentation of dose assessment in the case of external exposure (clinic, biological and physical dosimetry), this research thesis describes the principles of numerical reconstruction of a radiological accident, presents some computation codes (Monte Carlo code, MCNPX code) and the SESAME tool, and reports an application to an actual case (an accident which occurred in Equator in April 2009). The next part reports the developments performed to modify the posture of voxelized phantoms and the experimental and numerical validations. The last part reports a feasibility study for the reconstruction of radiological accidents occurring in external radiotherapy. This work is based on a Monte Carlo simulation of a linear accelerator, with the aim of identifying the most relevant parameters to be implemented in SESAME in the case of external radiotherapy

  11. Comparison of MCNPX and Albedo method in criticality calculation

    International Nuclear Information System (INIS)

    Cunha, Victor L. Lassance; Rebello, Wilson F.; Cabral, Ronaldo G.; Melo, Fernando da S.; Silva, Ademir X. da

    2009-01-01

    This study aims to conduct a computer simulation that will calculate the reactivity of a homogeneous reactor and compare the results with the calculations made by the albedo method. The simulation will be developed using the MCNPX. The study compared the results calculated for a hypothetical reactor by the albedo method for four groups of energy with those obtained by the MCNPX simulation. The design of the reactor is spherical and homogeneous with a reflector of finite thickness. The value obtained for the neutron effective multiplication factor - k eff will be compared. Different situations were simulated in order to obtain results closer to the compared method and reality. The was Good consistency could be noticed between the calculated results. (author)

  12. Coded aperture optimization using Monte Carlo simulations

    International Nuclear Information System (INIS)

    Martineau, A.; Rocchisani, J.M.; Moretti, J.L.

    2010-01-01

    Coded apertures using Uniformly Redundant Arrays (URA) have been unsuccessfully evaluated for two-dimensional and three-dimensional imaging in Nuclear Medicine. The images reconstructed from coded projections contain artifacts and suffer from poor spatial resolution in the longitudinal direction. We introduce a Maximum-Likelihood Expectation-Maximization (MLEM) algorithm for three-dimensional coded aperture imaging which uses a projection matrix calculated by Monte Carlo simulations. The aim of the algorithm is to reduce artifacts and improve the three-dimensional spatial resolution in the reconstructed images. Firstly, we present the validation of GATE (Geant4 Application for Emission Tomography) for Monte Carlo simulations of a coded mask installed on a clinical gamma camera. The coded mask modelling was validated by comparison between experimental and simulated data in terms of energy spectra, sensitivity and spatial resolution. In the second part of the study, we use the validated model to calculate the projection matrix with Monte Carlo simulations. A three-dimensional thyroid phantom study was performed to compare the performance of the three-dimensional MLEM reconstruction with conventional correlation method. The results indicate that the artifacts are reduced and three-dimensional spatial resolution is improved with the Monte Carlo-based MLEM reconstruction.

  13. General purpose code for Monte Carlo simulations

    International Nuclear Information System (INIS)

    Wilcke, W.W.

    1983-01-01

    A general-purpose computer called MONTHY has been written to perform Monte Carlo simulations of physical systems. To achieve a high degree of flexibility the code is organized like a general purpose computer, operating on a vector describing the time dependent state of the system under simulation. The instruction set of the computer is defined by the user and is therefore adaptable to the particular problem studied. The organization of MONTHY allows iterative and conditional execution of operations

  14. PC-Reactor-core transient simulation code

    International Nuclear Information System (INIS)

    Nakata, H.

    1989-10-01

    PC-REATOR, a reactor core transient simulation code has been developed for the real-time operator training on a IBM-PC microcomputer. The program presents capabilities for on-line exchange of the operating parameters during the transient simulation, by friendly keyboard instructions. The model is based on the point-kinetics approximation, with 2 delayed neutron percursors and up to 11 decay power generating groups. (author) [pt

  15. Development of code PRETOR for stellarator simulation

    International Nuclear Information System (INIS)

    Dies, J.; Fontanet, J.; Fontdecaba, J.M.; Castejon, F.; Alejandre, C.

    1998-01-01

    The Department de Fisica i Enginyeria Nuclear (DFEN) of the UPC has some experience in the development of the transport code PRETOR. This code has been validated with shots of DIII-D, JET and TFTR, it has also been used in the simulation of operational scenarios of ITER fast burnt termination. Recently, the association EURATOM-CIEMAT has started the operation of the TJ-II stellarator. Due to the need of validating the results given by others transport codes applied to stellarators and because all of them made some approximations, as a averaging magnitudes in each magnetic surface, it was thought suitable to adapt the PRETOR code to devices without axial symmetry, like stellarators, which is very suitable for the specific needs of the study of TJ-II. Several modifications are required in PRETOR; the main concerns to the models of: magnetic equilibrium, geometry and transport of energy and particles. In order to solve the complex magnetic equilibrium geometry the powerful numerical code VMEC has been used. This code gives the magnetic surface shape as a Fourier series in terms of the harmonics (m,n). Most of the geometric magnitudes are also obtained from the VMEC results file. The energy and particle transport models will be replaced by other phenomenological models that are better adapted to stellarator simulation. Using the proposed models, it is pretended to reproduce experimental data available from present stellarators, given especial attention to the TJ-II of the association EURATOM-CIEMAT. (Author)

  16. Nuclide Inventory Calculation Using MCNPX for Wolsung Unit 1 Reactor Decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie; Noh, Kyoung Ho; Hah, Chang Joo [KEPCO International Nuclear Graduate School, Daejeon (Korea, Republic of)

    2014-05-15

    The CINDER90 computation process involves utilizing linear Markovian chains to determine the time dependent nuclide densities. The CINDER90 depletion algorithm is implemented the MCNPX code package. The coupled depletion process involves a Monte-Carlo steady-state reaction rate calculation linked to a deterministic depletion calculation. The process is shown in Fig.1. MCNPX runs a steady state calculation to determine the system eigenvalue collision densities, recoverable energies from fission and neutrons per fission events. In order to generate number densities for the next time step, the CINDER90 code takes the MCNPX generated values and performs a depletion calculation. MCNPX then takes the new number densities and caries out a new steady-stated calculation. The process repeats itself until the final time step. This paper describe the preliminary source term and nuclide inventory calculation of Candu single fuel channel using MCNPX, as a part of the activities to support the equilibrium core model development and decommissioning evaluation process of a Candu reactor. The aim of this study was to apply the MCNPX code for source term and nuclide inventory calculation of Candu single fuel channel. Nuclide inventories as a function of burnup will be used to model an equilibrium core for Candu reactor. The core lifetime neutron fluence obtained from the model is used to estimate radioactivity at the stage of decommisioning. In general, as expected, the actinides and fission products build up increase with increasing burnup. Despite the fact that the MCNPX code is still in development we can conclude that the code is capable of obtaining relevant results in burnup and source term calculation. It is recommended that in the future work, the calculation has to be verified on the basis of experimental data or comparison with other codes.

  17. Photopeak efficiency response function of an underwater gamma-ray NaI(Tl) detector using MCNP-X

    Energy Technology Data Exchange (ETDEWEB)

    Salgado, William L., E-mail: william.otero@hotmail.com [Instituto Federal do Rio de Janeiro (IFRJ), RJ (Brazil); Silva, Ademir X., E-mail: ademir@con.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (PEN/COPPE-DNC/UFRJ/EE/CT), Rio de Janeiro, RJ (Brazil); Salgado, Cesar M., E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    This work presents a study to calculate the response function of a 1.5″ x 1″ NaI(Tl) scintillation detector when it is used in the marine environment in the energy range from 20 keV to 662 keV. The method takes into account both the scattering of photons in the water and the detection mechanism of the detector. In addition, the calculation of the response function of the whole system is essential for suppressing the background of the measurement and for estimating the concentration of the involved radionuclides, especially given the greater probability of primary gamma photons undergoing multiple scattering events before they interact with the detector. The experimental photopeak efficiency measurements for point sources were compared with the simulated results under the same conditions of the experimental setup to validate the simulation of the detector. Monte Carlo simulations were performed using the MCNP-X code for the investigation of gamma-ray absorption in water in different brines. The energy resolution curve was used to improve the response of the mathematical simulation of the detector. The detector’s simulation was based on information obtained from the gammagraphy technique. Both dimensions and materials were used for the calculation with the MCNP-X code. The photopeak efficiency of a NaI(Tl) detector for different radionuclides in the aquatic environment with different salinities was calculated. (author)

  18. A Study on MCNPX-CINDER90 System for Activation Analysis

    International Nuclear Information System (INIS)

    Kim, Sung-Min; Kim, Myung Hyun

    2014-01-01

    Neutron spectrum at every cell is different depending on the geometrical characteristics. For the problem of neutron activation to the reactor containment wall, neutron spectra are varying from the center of core to the containment wall. Therefore, process of one group cross section library for all relevant isotopes need an extensive works for all locations. On the other hand, CINDER can concerns 3 dimensional geometry effects and handles up to 3,400 nuclides. It is believed that CINDER is more reliable and accurate compared to ORIGEN because it treats 63-group cross section. In this paper, a new coupling of MCNP-CINDER was tested and compared with MCNP-ORIGEN and MCNPX 2.6.0. MCNPX is a coupled code of MCNP with CINDER90 for fuel depletion chain only. The simple UO2 single pin was modelled in order to compare and evaluate the fission product densities for fuel depletion chains. The simple reactor pressure vessel (RPV) and concrete wall were modelled for the comparison of isotopic inventory chains for activation products simulating the RPV boundaries. The UO2 single pin, simple RPV, and concrete wall model were modeled in order to compare inventory change and radioactivity with MCNPX 2.6..0, ORIGEN 2.1, and CINDER90. It is regarded that CINDER90 is more reliable and accurate compared to ORIGEN 2.1 because it has 63-group multi cross section library. In addition, several error by the approximation of model description and the difference of the cross section library, fission yield data, and et cetera in each code result in the relative error between each code. Also as the decay chain is the simple nuclide, the difference of the result is little between the code and it is the complicated nuclide, the difference of the result is large between the codes. So the error by the difference of the decay chain has to be considered between each code. As a result of this, the practicality of MCNP-CINDER system was verified and it is expected to be used for the study on the

  19. Propagation of Uncertainty in System Parameters of a LWR Model by Sampling MCNPX Calculations - Burnup Analysis

    Science.gov (United States)

    Campolina, Daniel de A. M.; Lima, Claubia P. B.; Veloso, Maria Auxiliadora F.

    2014-06-01

    For all the physical components that comprise a nuclear system there is an uncertainty. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a best estimate calculation that has been replacing the conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. In this work a sample space of MCNPX calculations was used to propagate the uncertainty. The sample size was optimized using the Wilks formula for a 95th percentile and a two-sided statistical tolerance interval of 95%. Uncertainties in input parameters of the reactor considered included geometry dimensions and densities. It was showed the capacity of the sampling-based method for burnup when the calculations sample size is optimized and many parameter uncertainties are investigated together, in the same input.

  20. MCNPX vs. DORT for SNS shielding design studies.

    Science.gov (United States)

    Popova, Irina I

    2005-01-01

    Radiation transport occurs through the 18 m long access way adjacent to the Spallation Neutron Source accelerator tunnel and the 2.2 m thick massive shielding door which closes the access way. A variety of typical materials for accelerator shielding, such as concrete and steel, were used for construction of the door to study radiation penetration. A comparison was carried out using both Monte Carlo (code MCNPX) and discrete ordinates (code DORT) methods. The beam losses during the accelerator operation are the sources for the radiation calculations. Analyses show that the results from the two methods are in good agreement.

  1. Use of GEANT4 vs. MCNPX for the characterization of a boron-lined neutron detector

    Energy Technology Data Exchange (ETDEWEB)

    Ende, B.M. van der; Atanackovic, J.; Erlandson, A.; Bentoumi, G.

    2016-06-01

    This work compares GEANT4 with MCNPX in the characterization of a boron-lined neutron detector. The neutron energy ranges simulated in this work (0.025 eV to 20 MeV) are the traditional domain of MCNP simulations. This paper addresses the question, how well can GEANT4 and MCNPX be employed for detailed thermal neutron detector characterization? To answer this, GEANT4 and MCNPX have been employed to simulate detector response to a {sup 252}Cf energy spectrum point source, as well as to simulate mono-energetic parallel beam source geometries. The {sup 252}Cf energy spectrum simulation results demonstrate agreement in detector count rate within 3% between the two packages, with the MCNPX results being generally closer to experiment than are those from GEANT4. The mono-energetic source simulations demonstrate agreement in detector response within 5% between the two packages for all neutron energies, and within 1% for neutron energies between 100 eV and 5 MeV. Cross-checks between the two types of simulations using ISO-8529 {sup 252}Cf energy bins demonstrates that MCNPX results are more self-consistent than are GEANT4 results, by 3–4%.

  2. Neutronic evaluation of transuranics in a GFR model using MCNPX and scale 6.0

    Energy Technology Data Exchange (ETDEWEB)

    Macedo, Anderson A.P.; Castro, Victor F.; Silva, Clarysson A.M. da; Velasquez, Carlos E.; Pereira, Claubia, E-mail: macedo@nuclear.ufmg.br, E-mail: victorfariacastro@gmail.com, E-mail: clarysson@nuclear.ufmg.br, E-mail: carlosvelcab@ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    In this study, a GFR core model with 100 MWt was evaluated using three different fuel compositions: a conventional (U, Pu)C and two reprocessed fuels reprocessed by UREX+ technique one spiked with depleted uranium, (U,TRU)C, and the other one reprocessed spiked with thorium, (Th,TRU)C. The reprocessed fuel came from a PWR standard fuel (33,000 MWd/T burned) with 3.1% of initial enrichment and left in the pool by 5 years. Some important nuclides were followed during burnup and k{sub inf} was evaluated for 1400 days. The results also include analysis of the B4C insertion and the temperature coefficient. The simulations were performed comparing results between MCNPX and SCALE 6.0 codes. The main goal is to validate the model and evaluate the possibility to use TRU spiked with Th in a GFR. (author)

  3. Study of neutronic flux in IPR-R1 reactor with MCNPX; Estudo do fluxo neutronico no reator IPR-R1 com o MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Melo, J.A.S.; Castrillo, L.S., E-mail: julio.angelo@poli.br, E-mail: lazara@poli.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica; Oliveira, R.M.B.M., E-mail: romero.matias@educacao.pe.gov.br [Secretaria Executiva de Educacao do Estado de Pernambuco (SEE), Recife, PE (Brazil)

    2016-11-01

    MCNPX computer code, one of the latest versions of code MCNP transport were used to study the flux distribution and its neutronic fluence as a function of energy in two research reactor irradiation IPR-R1. The model developed was validated with research conducted by Dalle (2005). Initially, in the simulation is considered fresh fuel whose core configuration contained three neutron rods control, being two of them 100% ejected while the other inserted 3,1 x 10{sup -1} m deep, as adopted in the literature situation. The neutron source used was the critical type, through KSRC card. The results of the neutron flow and neutronic fluence were obtained in the central tube and the turntable on a range of energy spectrum that ranged from 1.0 x 10{sup -9} MeV to 10 MeV, showing good correlations with the model used in validation. Finally, a hypothetical situation wherein the three reactor control rods are ejected simultaneously was simulated. The simulation results showed an increase in the neutron flux of 7% in the central tube and 5% on the turntable.

  4. Developing an interface between MCNP and McStas for simulation of neutron moderators

    DEFF Research Database (Denmark)

    Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik

    2012-01-01

    typically consists of providing analytical fits from MCNP/X neutron spectra to McStas. This method is generally successful, but as will be discussed in the this paper, there are limitations and a more direct coupling between MCNP/X andMcStas could allow for more accurate simulations of e.g. complex......Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using MCNP/X whereas simulations of neutron transport and instrument performance are carried out by neutron ray tracing codes such as McStas. The coupling between the two simulations suites...... moderator geometries, interference between beamlines as well as shielding requirements along the neutron guides. In this paper different possible interfaces between McStas and MCNP/X are discussed and first preliminary performance results are shown....

  5. User's manual for a measurement simulation code

    International Nuclear Information System (INIS)

    Kern, E.A.

    1982-07-01

    The MEASIM code has been developed primarily for modeling process measurements in materials processing facilities associated with the nuclear fuel cycle. In addition, the code computes materials balances and the summation of materials balances along with associated variances. The code has been used primarily in performance assessment of materials' accounting systems. This report provides the necessary information for a potential user to employ the code in these applications. A number of examples that demonstrate most of the capabilities of the code are provided

  6. MCMEG: Simulations of both PDD and TPR for 6 MV LINAC photon beam using different MC codes

    Science.gov (United States)

    Fonseca, T. C. F.; Mendes, B. M.; Lacerda, M. A. S.; Silva, L. A. C.; Paixão, L.; Bastos, F. M.; Ramirez, J. V.; Junior, J. P. R.

    2017-11-01

    The Monte Carlo Modelling Expert Group (MCMEG) is an expert network specializing in Monte Carlo radiation transport and the modelling and simulation applied to the radiation protection and dosimetry research field. For the first inter-comparison task the group launched an exercise to model and simulate a 6 MV LINAC photon beam using the Monte Carlo codes available within their laboratories and validate their simulated results by comparing them with experimental measurements carried out in the National Cancer Institute (INCA) in Rio de Janeiro, Brazil. The experimental measurements were performed using an ionization chamber with calibration traceable to a Secondary Standard Dosimetry Laboratory (SSDL). The detector was immersed in a water phantom at different depths and was irradiated with a radiation field size of 10×10 cm2. This exposure setup was used to determine the dosimetric parameters Percentage Depth Dose (PDD) and Tissue Phantom Ratio (TPR). The validation process compares the MC calculated results to the experimental measured PDD20,10 and TPR20,10. Simulations were performed reproducing the experimental TPR20,10 quality index which provides a satisfactory description of both the PDD curve and the transverse profiles at the two depths measured. This paper reports in detail the modelling process using MCNPx, MCNP6, EGSnrc and Penelope Monte Carlo codes, the source and tally descriptions, the validation processes and the results.

  7. Fast code for Monte Carlo simulations

    International Nuclear Information System (INIS)

    Oliveira, P.M.C. de; Penna, T.J.P.

    1988-01-01

    A computer code to generate the dynamic evolution of the Ising model on a square lattice, following the Metropolis algorithm is presented. The computer time consumption is reduced by a factor of 8 when one compares our code with traditional multiple spin codes. The memory allocation size is also reduced by a factor of 4. The code is easily generalizable for other lattices and models. (author) [pt

  8. Communication Systems Simulator with Error Correcting Codes Using MATLAB

    Science.gov (United States)

    Gomez, C.; Gonzalez, J. E.; Pardo, J. M.

    2003-01-01

    In this work, the characteristics of a simulator for channel coding techniques used in communication systems, are described. This software has been designed for engineering students in order to facilitate the understanding of how the error correcting codes work. To help students understand easily the concepts related to these kinds of codes, a…

  9. Software quality and process improvement in scientific simulation codes

    Energy Technology Data Exchange (ETDEWEB)

    Ambrosiano, J.; Webster, R. [Los Alamos National Lab., NM (United States)

    1997-11-01

    This report contains viewgraphs on the quest to develope better simulation code quality through process modeling and improvement. This study is based on the experience of the authors and interviews with ten subjects chosen from simulation code development teams at LANL. This study is descriptive rather than scientific.

  10. MED101: a laser-plasma simulation code. User guide

    International Nuclear Information System (INIS)

    Rodgers, P.A.; Rose, S.J.; Rogoyski, A.M.

    1989-12-01

    Complete details for running the 1-D laser-plasma simulation code MED101 are given including: an explanation of the input parameters, instructions for running on the Rutherford Appleton Laboratory IBM, Atlas Centre Cray X-MP and DEC VAX, and information on three new graphics packages. The code, based on the existing MEDUSA code, is capable of simulating a wide range of laser-produced plasma experiments including the calculation of X-ray laser gain. (author)

  11. Voxel2MCNP: a framework for modeling, simulation and evaluation of radiation transport scenarios for Monte Carlo codes

    International Nuclear Information System (INIS)

    Pölz, Stefan; Laubersheimer, Sven; Eberhardt, Jakob S; Harrendorf, Marco A; Keck, Thomas; Benzler, Andreas; Breustedt, Bastian

    2013-01-01

    The basic idea of Voxel2MCNP is to provide a framework supporting users in modeling radiation transport scenarios using voxel phantoms and other geometric models, generating corresponding input for the Monte Carlo code MCNPX, and evaluating simulation output. Applications at Karlsruhe Institute of Technology are primarily whole and partial body counter calibration and calculation of dose conversion coefficients. A new generic data model describing data related to radiation transport, including phantom and detector geometries and their properties, sources, tallies and materials, has been developed. It is modular and generally independent of the targeted Monte Carlo code. The data model has been implemented as an XML-based file format to facilitate data exchange, and integrated with Voxel2MCNP to provide a common interface for modeling, visualization, and evaluation of data. Also, extensions to allow compatibility with several file formats, such as ENSDF for nuclear structure properties and radioactive decay data, SimpleGeo for solid geometry modeling, ImageJ for voxel lattices, and MCNPX’s MCTAL for simulation results have been added. The framework is presented and discussed in this paper and example workflows for body counter calibration and calculation of dose conversion coefficients is given to illustrate its application. (paper)

  12. APR1400 Containment Simulation with CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Moon Kyu; Chung, Bub Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    The more realistic containment pressure variation predicted by the CONTAIN code through the coupled analysis during a large break loss of coolant accident in the nuclear power plant is expected to provide more accurate prediction for the plant behavior than a standalone MARS-KS calculation. The input deck has been generated based on the already available ARP- 1400 input for CONTEMPT code. Similarly to the CONTEMPT input deck, a simple two-cell model was adopted to model the containment behavior, one cell for the containment inner volume and another cell for the environment condition. The developed input for the CONTAIN code is to be eventually applied for the coupled code calculation of MARS-KS/CONTAIN

  13. APR1400 Containment Simulation with CONTAIN code

    International Nuclear Information System (INIS)

    Hwang, Moon Kyu; Chung, Bub Dong

    2010-01-01

    The more realistic containment pressure variation predicted by the CONTAIN code through the coupled analysis during a large break loss of coolant accident in the nuclear power plant is expected to provide more accurate prediction for the plant behavior than a standalone MARS-KS calculation. The input deck has been generated based on the already available ARP- 1400 input for CONTEMPT code. Similarly to the CONTEMPT input deck, a simple two-cell model was adopted to model the containment behavior, one cell for the containment inner volume and another cell for the environment condition. The developed input for the CONTAIN code is to be eventually applied for the coupled code calculation of MARS-KS/CONTAIN

  14. Prediction of neutron induced radioactivity in the concrete walls of a PET cyclotron vault room with MCNPX.

    Science.gov (United States)

    Martínez-Serrano, J Javier; Díez de los Ríos, Antonio

    2010-11-01

    The authors want to assess the relevance of the neutron activation of the concrete vault of the PET cyclotron at CIMES (Universidad de Malaga) by predicting specific activities of the main activation products in the vault and their variation profiles as a function of penetration depth into concrete at present and after 10 yr of cyclotron operation. The dual proton cyclotron is used for PET isotopes production, mainly 18F. During the years 2006 and 2008, the using rate has been 1 h/day at single beam (40 microA). From January 2008, using rate is 4 h/day at dual beam (80 microA). The energy of the cyclotron proton beam is 18 MeV. Four point locations were chosen on the walls of the cyclotron room to assess neutron induced activity concentrations. In each wall point location, neutron induced radionuclide specific activity was assessed from the wall surface to a depth of 120 cm within concrete. Simulations were carried out with the Monte Carlo based radiation transport code MCNPX (v2.6.0). According to MCNPX calculations, activity depth profiles of activation products studied, except 54Mn, have a maximum at variable depths from the wall surface never beyond 12 cm. 54Mn activity decreases exponentially in all the studied depth ranges within wall concrete. The activity of 152Eu, 154Eu, 60CO, 134Cs, 46Sc, and 65Zn decreases exponentially beyond a 30 cm depth into concrete. 54Mn activity presents the faster decrease within a concrete vault with an attenuation length of 21 cm. According to MCNPX estimations, present activity in the cyclotron vault is mostly due to 46Sc and 60Co, with highest specific activity near the vault surface of 146 +/- 16 and 50 +/- 4.6 Bq/kg, respectively. 46Sc and 60Co activity measurements near the surface wall present an acceptable match with the estimation within the uncertainties, but measured activities of the other radionuclides are quite over the MCNPX estimations. The calculations after 10 yr of cyclotron operation predict a slight increase

  15. The TESS [Tandem Experiment Simulation Studies] computer code user's manual

    International Nuclear Information System (INIS)

    Procassini, R.J.

    1990-01-01

    TESS (Tandem Experiment Simulation Studies) is a one-dimensional, bounded particle-in-cell (PIC) simulation code designed to investigate the confinement and transport of plasma in a magnetic mirror device, including tandem mirror configurations. Mirror plasmas may be modeled in a system which includes an applied magnetic field and/or a self-consistent or applied electrostatic potential. The PIC code TESS is similar to the PIC code DIPSI (Direct Implicit Plasma Surface Interactions) which is designed to study plasma transport to and interaction with a solid surface. The codes TESS and DIPSI are direct descendants of the PIC code ES1 that was created by A. B. Langdon. This document provides the user with a brief description of the methods used in the code and a tutorial on the use of the code. 10 refs., 2 tabs

  16. Benchmark study of TRIPOLI-4 through experiment and MCNP codes

    Energy Technology Data Exchange (ETDEWEB)

    Michel, M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Coulon, R. [Canberra France, F-78182 Saint Quentin en Yvelines (France); Normand, S. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Huot, N.; Petit, O. [CEA, DEN DANS, SERMA, F-91191 Gif-sur-Yvette (France)

    2011-07-01

    Reliability on simulation results is essential in nuclear physics. Although MCNP5 and MCNPX are the world widely used 3D Monte Carlo radiation transport codes, alternative Monte Carlo simulation tools exist to simulate neutral and charged particles' interactions with matter. Therefore, benchmark are required in order to validate these simulation codes. For instance, TRIPOLI-4.7, developed at the French Alternative Energies and Atomic Energy Commission for neutron and photon transport, now also provides the user with a full feature electron-photon electromagnetic shower. Whereas the reliability of TRIPOLI-4.7 for neutron and photon transport has been validated yet, the new development regarding electron-photon matter interaction needs additional validation benchmarks. We will thus demonstrate how accurately TRIPOLI-4's 'deposited spectrum' tally can simulate gamma spectrometry problems, compared to MCNP's 'F8' tally. The experimental setup is based on an HPGe detector measuring the decay spectrum of an {sup 152}Eu source. These results are then compared with those given by MCNPX 2.6d and TRIPOLI-4 codes. This paper deals with both the experimental aspect and simulation. We will demonstrate that TRIPOLI-4 is a potential alternative to both MCNPX and MCNP5 for gamma-electron interaction simulation. (authors)

  17. OpenQ∗D simulation code for QCD+QED

    DEFF Research Database (Denmark)

    Campos, Isabel; Fritzsch, Patrick; Hansen, Martin

    2018-01-01

    The openQ∗D code for the simulation of QCD+QED with C∗ boundary conditions is presented. This code is based on openQCD-1.6, from which it inherits the core features that ensure its efficiency: the locally-deflated SAP-preconditioned GCR solver, the twisted-mass frequency splitting of the fermion...

  18. Propagation of uncertainty in system parameters of a LWR model by sampling MCNPX calculations - Burnup analysis

    International Nuclear Information System (INIS)

    Campolina, D. de A. M.; Lima, C.P.B.; Veloso, M.A.F.

    2013-01-01

    For all the physical components that comprise a nuclear system there is an uncertainty. Assessing the impact of uncertainties in the simulation of fissionable material systems is essential for a best estimate calculation that has been replacing the conservative model calculations as the computational power increases. The propagation of uncertainty in a simulation using a Monte Carlo code by sampling the input parameters is recent because of the huge computational effort required. In this work a sample space of MCNPX calculations was used to propagate the uncertainty. The sample size was optimized using the Wilks formula for a 95. percentile and a two-sided statistical tolerance interval of 95%. Uncertainties in input parameters of the reactor considered included geometry dimensions and densities. It was showed the capacity of the sampling-based method for burnup when the calculations sample size is optimized and many parameter uncertainties are investigated together, in the same input. Particularly it was shown that during the burnup, the variances when considering all the parameters uncertainties is equivalent to the sum of variances if the parameter uncertainties are sampled separately

  19. Classical diffusion: theory and simulation codes

    International Nuclear Information System (INIS)

    Grad, H.; Hu, P.N.

    1978-03-01

    A survey is given of the development of classical diffusion theory which arose from the observation of Grad and Hogan that the Pfirsch-Schluter and Neoclassical theories are very special and frequently inapplicable because they require that plasma mass flow be treated as transport rather than as a state variable of the plasma. The subsequent theory, efficient numerical algorithms, and results of various operating codes are described

  20. Simulation of Water Chemistry using and Geochemistry Code, PHREEQE

    Energy Technology Data Exchange (ETDEWEB)

    Chi, J.H. [Korea Electric Power Research Institute, Taejeon (Korea)

    2001-07-01

    This report introduces principles and procedures of simulation for water chemistry using a geochemistry code, PHREEQE. As and example of the application of this code, we described the simulation procedure for titration of an aquatic sample with strong acid to investigate the state of Carbonates in aquatic solution. Major contents of this report are as follows; Concepts and principles of PHREEQE, Kinds of chemical reactions which may be properly simulated by PHREEQE, The definition and meaning of each input data, An example of simulation using PHREEQE. (author). 2 figs., 1 tab.

  1. A general purpose code for Monte Carlo simulations

    International Nuclear Information System (INIS)

    Wilcke, W.W.; Rochester Univ., NY

    1984-01-01

    A general-purpose computer code MONTHY has been written to perform Monte Carlo simulations of physical systems. To achieve a high degree of flexibility the code is organized like a general purpose computer, operating on a vector describing the time dependent state of the system under simulation. The instruction set of the 'computer' is defined by the user and is therefore adaptable to the particular problem studied. The organization of MONTHY allows iterative and conditional execution of operations. (orig.)

  2. Scalable Simulation of Electromagnetic Hybrid Codes

    International Nuclear Information System (INIS)

    Perumalla, Kalyan S.; Fujimoto, Richard; Karimabadi, Dr. Homa

    2006-01-01

    New discrete-event formulations of physics simulation models are emerging that can outperform models based on traditional time-stepped techniques. Detailed simulation of the Earth's magnetosphere, for example, requires execution of sub-models that are at widely differing timescales. In contrast to time-stepped simulation which requires tightly coupled updates to entire system state at regular time intervals, the new discrete event simulation (DES) approaches help evolve the states of sub-models on relatively independent timescales. However, parallel execution of DES-based models raises challenges with respect to their scalability and performance. One of the key challenges is to improve the computation granularity to offset synchronization and communication overheads within and across processors. Our previous work was limited in scalability and runtime performance due to the parallelization challenges. Here we report on optimizations we performed on DES-based plasma simulation models to improve parallel performance. The net result is the capability to simulate hybrid particle-in-cell (PIC) models with over 2 billion ion particles using 512 processors on supercomputing platforms

  3. MUSIC: a mesh-unrestricted simulation code

    International Nuclear Information System (INIS)

    Bonalumi, R.A.; Rouben, B.; Dastur, A.R.; Dondale, C.S.; Li, H.Y.H.

    1978-01-01

    A general formalism to solve the G-group neutron diffusion equation is described. The G-group flux is represented by complementing an ''asymptotic'' mode with (G-1) ''transient'' modes. A particular reduction-to-one-group technique gives a high computational efficiency. MUSIC, a 2-group code using the above formalism, is presented. MUSIC is demonstrated on a fine-mesh calculation and on 2 coarse-mesh core calculations: a heavy-water reactor (HWR) problem and the 2-D lightwater reactor (LWR) IAEA benchmark. Comparison is made to finite-difference results

  4. Towards a realistic plasma simulation code

    International Nuclear Information System (INIS)

    Anderson, D.V.

    1991-06-01

    Several new developments in the technology of simulating plasmas, both in particle and fluid models, now allow a stage of synthesis in which many of these advances can be combined into one simulation model. Accuracy and efficiency are the criteria to be satisfied in this quest. We want to build on the following research: 1. the development of the δf method of Barnes. 2. The moving node Galerkin model of Glasser, Miller and Carlson. 3. Particle moving schemes on unstructured grids by Ambrosiano and Bradon. 4. Particle simulations using sorted particles Anderson and Shumaker. Rather than being competing developments,these presumably can be combined into one computational model. We begin by summarizing the physics model for the plasma. The Vlasov equation can be solved as an initial value problem by integrating the plasma distribution function forward in time. 5 refs

  5. NE213/BC501A scintillator−lightguide assembly response to 241Am−Be neutrons: An MCNPX−PHOTRACK hybrid code simulation

    International Nuclear Information System (INIS)

    Tajik, M.; Ghal-Eh, N.; Etaati, G.R.; Afarideh, H.

    2014-01-01

    The response of an NE213 (or its BICRON equivalent, BC501A) scintillator attached to different sizes of polished/painted lightguides when exposed to 241 Am–Be neutrons has been simulated. This kind of simulation basically needs both particle and light transports: the transport of neutrons and neutron-induced charged particles such as alphas, protons, carbon nuclei and so on has been undertaken using MCNPX whilst the scintillation light transport has been performed with PHOTRACK codes. The comparison between simulated and experimental response functions of NE213 attached to different sizes of polished/painted lightguides and also the influence of length/covering of lightguide on the detection efficiency and uniformity of the scintillator–lightguide assembly response have been studied. - Highlights: • The response of NE213 scintillator with/without lightguides to Am–Be neutrons has been simulated. • The MCNPX–PHOTRACK code has been used for simulation studies in order to model radio-optical properties. • The measured and simulated spectra for an NE213 scintillator exposed to Am–Be source represent a good agreement

  6. Code development for nuclear reactor simulation

    International Nuclear Information System (INIS)

    Chauliac, C.; Verwaerde, D.; Pavageau, O.

    2006-01-01

    Full text of publication follows: Since several years, CEA, EDF and FANP have developed several numerical codes which are currently used for nuclear industry applications and will be remain in use for the coming years. Complementary to this set of codes and in order to better meet the present and future needs, a new system is being developed through a joint venture between CEA, EDF and FANP, with a ten year prospect and strong intermediate milestones. The focus is put on a multi-scale and multi-physics approach enabling to take into account phenomena from microscopic to macroscopic scale, and to describe interactions between various physical fields such as neutronics (DESCARTES), thermal-hydraulics (NEPTUNE) and fuel behaviour (PLEIADES). This approach is based on a more rational design of the softwares and uses a common integration platform providing pre-processing, supervision of computation and post-processing. This paper will describe the overall system under development and present the first results obtained. (authors)

  7. Monte Carlo codes and Monte Carlo simulator program

    International Nuclear Information System (INIS)

    Higuchi, Kenji; Asai, Kiyoshi; Suganuma, Masayuki.

    1990-03-01

    Four typical Monte Carlo codes KENO-IV, MORSE, MCNP and VIM have been vectorized on VP-100 at Computing Center, JAERI. The problems in vector processing of Monte Carlo codes on vector processors have become clear through the work. As the result, it is recognized that these are difficulties to obtain good performance in vector processing of Monte Carlo codes. A Monte Carlo computing machine, which processes the Monte Carlo codes with high performances is being developed at our Computing Center since 1987. The concept of Monte Carlo computing machine and its performance have been investigated and estimated by using a software simulator. In this report the problems in vectorization of Monte Carlo codes, Monte Carlo pipelines proposed to mitigate these difficulties and the results of the performance estimation of the Monte Carlo computing machine by the simulator are described. (author)

  8. GEANT4 benchmark with MCNPX and PHITS for activation of concrete

    Science.gov (United States)

    Tesse, Robin; Stichelbaut, Frédéric; Pauly, Nicolas; Dubus, Alain; Derrien, Jonathan

    2018-02-01

    The activation of concrete is a real problem from the point of view of waste management. Because of the complexity of the issue, Monte Carlo (MC) codes have become an essential tool to its study. But various codes or even nuclear models exist in MC. MCNPX and PHITS have already been validated for shielding studies but GEANT4 is also a suitable solution. In these codes, different models can be considered for a concrete activation study. The Bertini model is not the best model for spallation while BIC and INCL model agrees well with previous results in literature.

  9. Development and tests of a mouse voxel model dor MCNPX based on Digimouse images

    International Nuclear Information System (INIS)

    Melo M, B.; Ferreira F, C.; Garcia de A, I.; Machado T, B.; Passos Ribeiro de C, T.

    2015-10-01

    Mice have been widely used in experimental protocols involving ionizing radiation. Biological effects (Be) induced by radiation can compromise studies results. Good estimates of mouse whole body and organs absorbed dose could provide valuable information to researchers. The aim of this study was to create and test a new voxel phantom for mice dosimetry from -Digimouse- project images. Micro CT images from Digimouse project were used in this work. Corel PHOTOPAINT software was utilized in segmentation process. The three-dimensional (3-D) model assembly and its voxel size manipulation were performed by Image J. SISCODES was used to adapt the model to run in MCNPX Monte Carlo code. The resulting model was called DM B RA. The volume and mass of segmented organs were compared with data available in literature. For the preliminary tests the heart was considered the source organ. Photons of diverse energies were simulated and Saf values obtained through F6:p and + F6 MCNPX tallies. The results were compared with reference data. 3-D picturing of absorbed doses patterns and relative errors distribution were generated by a C++ -in house- made program and visualized through Amide software. The organ masses of DM B RA correlated well with two models that were based on same set of images. However some organs, like eyes and adrenals, skeleton and brain showed large discrepancies. Segmentation of an identical image set by different persons and/or methods can result significant organ masses variations. We believe that the main causes of these differences were: i) operator dependent subjectivity in the definition of organ limits during the segmentation processes; and i i) distinct voxel dimensions between evaluated models. Lack of reference data for mice models construction and dosimetry was detected. Comparison with other models originated from different mice strains also demonstrated that the anatomical and size variability can be significant. Use of + F6 tally for mouse phantoms

  10. Development and tests of a mouse voxel model dor MCNPX based on Digimouse images

    Energy Technology Data Exchange (ETDEWEB)

    Melo M, B.; Ferreira F, C. [Centro de Desenvolvimento da Tecnologia Nuclear / CNEN, Pte. Antonio Carlos No. 6627, Belo Horizonte 31270-901, Minas Gerais (Brazil); Garcia de A, I.; Machado T, B.; Passos Ribeiro de C, T., E-mail: bmm@cdtn.br [Universidade Federal de Minas Gerais, Departamento de Engenharia Nuclear, Pte. Antonio Carlos 6627, Belo Horizonte 31270-901, Minas Gerais (Brazil)

    2015-10-15

    Mice have been widely used in experimental protocols involving ionizing radiation. Biological effects (Be) induced by radiation can compromise studies results. Good estimates of mouse whole body and organs absorbed dose could provide valuable information to researchers. The aim of this study was to create and test a new voxel phantom for mice dosimetry from -Digimouse- project images. Micro CT images from Digimouse project were used in this work. Corel PHOTOPAINT software was utilized in segmentation process. The three-dimensional (3-D) model assembly and its voxel size manipulation were performed by Image J. SISCODES was used to adapt the model to run in MCNPX Monte Carlo code. The resulting model was called DM{sub B}RA. The volume and mass of segmented organs were compared with data available in literature. For the preliminary tests the heart was considered the source organ. Photons of diverse energies were simulated and Saf values obtained through F6:p and + F6 MCNPX tallies. The results were compared with reference data. 3-D picturing of absorbed doses patterns and relative errors distribution were generated by a C++ -in house- made program and visualized through Amide software. The organ masses of DM{sub B}RA correlated well with two models that were based on same set of images. However some organs, like eyes and adrenals, skeleton and brain showed large discrepancies. Segmentation of an identical image set by different persons and/or methods can result significant organ masses variations. We believe that the main causes of these differences were: i) operator dependent subjectivity in the definition of organ limits during the segmentation processes; and i i) distinct voxel dimensions between evaluated models. Lack of reference data for mice models construction and dosimetry was detected. Comparison with other models originated from different mice strains also demonstrated that the anatomical and size variability can be significant. Use of + F6 tally for mouse

  11. Effect of the electron transport through thin slabs on the simulation of linear electron accelerators of use in therapy: A comparative study of various Monte Carlo codes

    Energy Technology Data Exchange (ETDEWEB)

    Vilches, M. [Servicio de Fisica y Proteccion Radiologica, Hospital Regional Universitario ' Virgen de las Nieves' , Avda. de las Fuerzas Armadas, 2, E-18014 Granada (Spain)], E-mail: mvilches@ugr.es; Garcia-Pareja, S. [Servicio de Radiofisica Hospitalaria, Hospital Regional Universitario ' Carlos Haya' , Avda. Carlos Haya, s/n, E-29010 Malaga (Spain); Guerrero, R. [Servicio de Radiofisica, Hospital Universitario ' San Cecilio' , Avda. Dr. Oloriz, 16, E-18012 Granada (Spain); Anguiano, M.; Lallena, A.M. [Departamento de Fisica Atomica, Molecular y Nuclear, Universidad de Granada, E-18071 Granada (Spain)

    2007-09-21

    When a therapeutic electron linear accelerator is simulated using a Monte Carlo (MC) code, the tuning of the initial spectra and the renormalization of dose (e.g., to maximum axial dose) constitute a common practice. As a result, very similar depth dose curves are obtained for different MC codes. However, if renormalization is turned off, the results obtained with the various codes disagree noticeably. The aim of this work is to investigate in detail the reasons of this disagreement. We have found that the observed differences are due to non-negligible differences in the angular scattering of the electron beam in very thin slabs of dense material (primary foil) and thick slabs of very low density material (air). To gain insight, the effects of the angular scattering models considered in various MC codes on the dose distribution in a water phantom are discussed using very simple geometrical configurations for the LINAC. The MC codes PENELOPE 2003, PENELOPE 2005, GEANT4, GEANT3, EGSnrc and MCNPX have been used.

  12. Simulation and interpretation codes for the JET ECE diagnostic. Part 1: physics of the codes' operation

    International Nuclear Information System (INIS)

    Bartlett, D.V.

    1983-06-01

    The codes which have been developed for the analysis of electron cyclotron emission measurements in JET are described. Their principal function is to interpret the spectra measured by the diagnostic so as to give the spatial distribution of the electron temperature in the poloidal cross-section. Various systematic effects in the data are corrected using look-up tables generated by an elaborate simulation code. The part of this code responsible for the accurate calculation of single-pass emission and refraction has been written at CNR-Milan and is described in a separate report. The present report is divided into two parts. This first part describes the methods used for the simulation and interpretation of spectra, the physical/mathematical basis of the codes written at CEA-Fontenay and presents some illustrative results

  13. Nexus: A modular workflow management system for quantum simulation codes

    Science.gov (United States)

    Krogel, Jaron T.

    2016-01-01

    The management of simulation workflows represents a significant task for the individual computational researcher. Automation of the required tasks involved in simulation work can decrease the overall time to solution and reduce sources of human error. A new simulation workflow management system, Nexus, is presented to address these issues. Nexus is capable of automated job management on workstations and resources at several major supercomputing centers. Its modular design allows many quantum simulation codes to be supported within the same framework. Current support includes quantum Monte Carlo calculations with QMCPACK, density functional theory calculations with Quantum Espresso or VASP, and quantum chemical calculations with GAMESS. Users can compose workflows through a transparent, text-based interface, resembling the input file of a typical simulation code. A usage example is provided to illustrate the process.

  14. Edge-relevant plasma simulations with the continuum code COGENT

    Science.gov (United States)

    Dorf, M.; Dorr, M.; Ghosh, D.; Hittinger, J.; Rognlien, T.; Cohen, R.; Lee, W.; Schwartz, P.

    2016-10-01

    We describe recent advances in cross-separatrix and other edge-relevant plasma simulations with COGENT, a continuum gyro-kinetic code being developed by the Edge Simulation Laboratory (ESL) collaboration. The distinguishing feature of the COGENT code is its high-order finite-volume discretization methods, which employ arbitrary mapped multiblock grid technology (nearly field-aligned on blocks) to handle the complexity of tokamak divertor geometry with high accuracy. This paper discusses the 4D (axisymmetric) electrostatic version of the code, and the presented topics include: (a) initial simulations with kinetic electrons and development of reduced fluid models; (b) development and application of implicit-explicit (IMEX) time integration schemes; and (c) conservative modeling of drift-waves and the universal instability. Work performed for USDOE, at LLNL under contract DE-AC52-07NA27344 and at LBNL under contract DE-AC02-05CH11231.

  15. Building a dynamic code to simulate new reactor concepts

    International Nuclear Information System (INIS)

    Catsaros, N.; Gaveau, B.; Jaekel, M.-T.; Maillard, J.; Maurel, G.; Savva, P.; Silva, J.; Varvayanni, M.

    2012-01-01

    Highlights: ► We develop a stochastic neutronic code based on an existing High Energy Physics code. ► The code simulates innovative reactor designs including Accelerator Driven Systems. ► Core materials evolution will be dynamically simulated, including fuel burnup. ► Continuous feedback between the main inter-related parameters will be established. ► A description of the current research development and achievements is also given. - Abstract: Innovative nuclear reactor designs have been proposed, such as the Accelerator Driven Systems (ADSs), the “candle” reactors, etc. These reactor designs introduce computational nuclear technology problems the solution of which necessitates a new, global and dynamic computational approach of the system. A continuous feedback procedure must be established between the main inter-related parameters of the system such as the chemical, physical and isotopic composition of the core, the neutron flux distribution and the temperature field. Furthermore, as far as ADSs are concerned, the ability of the computational tool to simulate the nuclear cascade created from the interaction of accelerated protons with the spallation target as well as the produced neutrons, is also required. The new Monte Carlo code ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is being developed based on the GEANT3 High Energy Physics code, aiming to progressively satisfy all the above requirements. A description of the capabilities and methodologies implemented in the present version of ANET is given here, together with some illustrative applications of the code.

  16. The CombLayer build of the MCNPX models for the design of the fast neutron ports in the European Spallation Source

    International Nuclear Information System (INIS)

    Milocco, A.; Gorini, G.; Zanini, L.; Ansell, S.

    2013-01-01

    Building MCNPX models is a time consuming process. At ISIS, a modeling architecture called 'CombLayer' has been developed, which allows MCNPX models to be produced rapidly, and in a highly parametric manner. In this work, CombLayer has been used for neutronic studies for the European Spallation Source (ESS). Initially, MCNPX models of the ESS were modified to include the irradiation ports and material test volumes. The computational time required to run each of these models was prohibitive, which precludes running multiple configurations for optimizations. To help mitigate this performance problem, we built the model using the CombLayer coding model. CombLayer is a C++ tool-kit for building geometric models. These models can then be rewritten in various formats, including MCNPX. As the CombLayer program has a working model, it is able to perform simple geometric optimizations (e.g. minimizing the cell literals, removing complementary objects) and some computed variance reduction

  17. Performance of MPI parallel processing implemented by MCNP5/ MCNPX for criticality benchmark problems

    International Nuclear Information System (INIS)

    Mark Dennis Usang; Mohd Hairie Rabir; Mohd Amin Sharifuldin Salleh; Mohamad Puad Abu

    2012-01-01

    MPI parallelism are implemented on a SUN Workstation for running MCNPX and on the High Performance Computing Facility (HPC) for running MCNP5. 23 input less obtained from MCNP Criticality Validation Suite are utilized for the purpose of evaluating the amount of speed up achievable by using the parallel capabilities of MPI. More importantly, we will study the economics of using more processors and the type of problem where the performance gain are obvious. This is important to enable better practices of resource sharing especially for the HPC facilities processing time. Future endeavours in this direction might even reveal clues for best MCNP5/ MCNPX coding practices for optimum performance of MPI parallelisms. (author)

  18. DART: a simulation code for charged particle beams

    International Nuclear Information System (INIS)

    White, R.C.; Barr, W.L.; Moir, R.W.

    1988-01-01

    This paper presents a recently modified verion of the 2-D DART code designed to simulate the behavior of a beam of charged particles whose paths are affected by electric and magnetic fields. This code was originally used to design laboratory-scale and full-scale beam direct converters. Since then, its utility has been expanded to allow more general applications. The simulation technique includes space charge, secondary electron effects, and neutral gas ionization. Calculations of electrode placement and energy conversion efficiency are described. Basic operation procedures are given including sample input files and output. 7 refs., 18 figs

  19. Simulation of the turbine discharge transient with the code Trace

    International Nuclear Information System (INIS)

    Mejia S, D. M.; Filio L, C.

    2014-10-01

    In this paper the results of the simulation of the turbine discharge transient are shown, occurred in Unit 1 of nuclear power plant of Laguna Verde (NPP-L V), carried out with the model of this unit for the best estimate code Trace. The results obtained by the code Trace are compared with those obtained from the Process Information Integral System (PIIS) of the NPP-L V. The reactor pressure, level behavior in the down-comer, steam flow and flow rate through the recirculation circuits are compared. The results of the simulation for the operation power of 2027 MWt, show concordance with the system PIIS. (Author)

  20. Benchmark measurements and simulations of dose perturbations due to metallic spheres in proton beams

    International Nuclear Information System (INIS)

    Newhauser, Wayne D.; Rechner, Laura; Mirkovic, Dragan; Yepes, Pablo; Koch, Nicholas C.; Titt, Uwe; Fontenot, Jonas D.; Zhang, Rui

    2013-01-01

    Monte Carlo simulations are increasingly used for dose calculations in proton therapy due to its inherent accuracy. However, dosimetric deviations have been found using Monte Carlo code when high density materials are present in the proton beamline. The purpose of this work was to quantify the magnitude of dose perturbation caused by metal objects. We did this by comparing measurements and Monte Carlo predictions of dose perturbations caused by the presence of small metal spheres in several clinical proton therapy beams as functions of proton beam range and drift space. Monte Carlo codes MCNPX, GEANT4 and Fast Dose Calculator (FDC) were used. Generally good agreement was found between measurements and Monte Carlo predictions, with the average difference within 5% and maximum difference within 17%. The modification of multiple Coulomb scattering model in MCNPX code yielded improvement in accuracy and provided the best overall agreement with measurements. Our results confirmed that Monte Carlo codes are well suited for predicting multiple Coulomb scattering in proton therapy beams when short drift spaces are involved. - Highlights: • We compared measurements and Monte Carlo predictions of dose perturbations caused by the metal objects in proton beams. • Different Monte Carlo codes were used, including MCNPX, GEANT4 and Fast Dose Calculator. • Good agreement was found between measurements and Monte Carlo simulations. • The modification of multiple Coulomb scattering model in MCNPX code yielded improved accuracy. • Our results confirmed that Monte Carlo codes are well suited for predicting multiple Coulomb scattering in proton therapy

  1. Developments of HTGR thermofluid dynamic analysis codes and HTGR plant dynamic simulation code

    International Nuclear Information System (INIS)

    Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi

    1983-01-01

    In nuclear power plants as well as high temperature gas-cooled reactor plants, the design is mostly performed on the basis of the results after their characteristics have been grasped by carrying out the numerical simulation using the analysis code. Also in Kawasaki Heavy Industries Ltd., on the basis of the system engineering accumulated with gas-cooled reactors since several years ago, the preparation and systematization of analysis codes have been advanced, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In this report, a part of the results is described. The example of the analysis applying the two-dimensional compressible flow analysis codes SOLA-VOF and SALE-2D, which were developed by Los Alamos National Laboratory in USA and modified for use in Kawasaki, to HTGR system is reported. Besides, Kawasaki has developed the control characteristics analyzing code DYSCO by which the change of system composition is easy and high versatility is available. The outline, fundamental equations, fundamental algorithms and examples of application of the SOLA-VOF and SALE-2D, the present status of system characteristic simulation codes and the outline of the DYSCO are described. (Kako, I.)

  2. Simulations of linear and Hamming codes using SageMath

    Science.gov (United States)

    Timur, Tahta D.; Adzkiya, Dieky; Soleha

    2018-03-01

    Digital data transmission over a noisy channel could distort the message being transmitted. The goal of coding theory is to ensure data integrity, that is, to find out if and where this noise has distorted the message and what the original message was. Data transmission consists of three stages: encoding, transmission, and decoding. Linear and Hamming codes are codes that we discussed in this work, where encoding algorithms are parity check and generator matrix, and decoding algorithms are nearest neighbor and syndrome. We aim to show that we can simulate these processes using SageMath software, which has built-in class of coding theory in general and linear codes in particular. First we consider the message as a binary vector of size k. This message then will be encoded to a vector with size n using given algorithms. And then a noisy channel with particular value of error probability will be created where the transmission will took place. The last task would be decoding, which will correct and revert the received message back to the original message whenever possible, that is, if the number of error occurred is smaller or equal to the correcting radius of the code. In this paper we will use two types of data for simulations, namely vector and text data.

  3. DART: A simulation code for charged particle beams

    International Nuclear Information System (INIS)

    White, R.C.; Barr, W.L.; Moir, R.W.

    1989-01-01

    This paper presents a recently modified version of the 2-D code, DART, which can simulate the behavior of a beam of charged particles whose trajectories are determined by electric and magnetic fields. This code was originally used to design laboratory-scale and full-scale beam direct converters. Since then, its utility has been expanded to allow more general applications. The simulation includes space charge, secondary electrons, and the ionization of neutral gas. A beam can contain up to nine superimposed beamlets of different energy and species. The calculation of energy conversion efficiency and the method of specifying the electrode geometry are described. Basic procedures for using the code are given, and sample input and output fields are shown. 7 refs., 18 figs

  4. Coding considerations for standalone molecular dynamics simulations of atomistic structures

    Science.gov (United States)

    Ocaya, R. O.; Terblans, J. J.

    2017-10-01

    The laws of Newtonian mechanics allow ab-initio molecular dynamics to model and simulate particle trajectories in material science by defining a differentiable potential function. This paper discusses some considerations for the coding of ab-initio programs for simulation on a standalone computer and illustrates the approach by C language codes in the context of embedded metallic atoms in the face-centred cubic structure. The algorithms use velocity-time integration to determine particle parameter evolution for up to several thousands of particles in a thermodynamical ensemble. Such functions are reusable and can be placed in a redistributable header library file. While there are both commercial and free packages available, their heuristic nature prevents dissection. In addition, developing own codes has the obvious advantage of teaching techniques applicable to new problems.

  5. Simulations of the Ondine experiment with the solitude code

    International Nuclear Information System (INIS)

    Gouard, P.; Gardelle, J.

    1992-11-01

    A new version of the SOLITUDE code, including an axial magnetic field and a cylindrical waveguide, is presented. It allows to simulate the ONDINE experiment at CESTA and to study the effects and behaviour of an actual electron beam in a Free Electron Laser amplifier experiment

  6. Simulating magnetised plasma with the versatile advection code

    NARCIS (Netherlands)

    Keppens, R.; Toth, G.; Palma, J. M. L.; Dongarra, J.; Hernandez, V.

    1999-01-01

    Matter in the universe mainly consists of plasma. The dynamics of plasmas is controlled by magnetic fields. To simulate the evolution of magnetised plasma, we solve the equations of magnetohydrodynamics using the Versatile Advection Code (VAC). To demonstrate the versatility of VAC, we present

  7. Contribution to study and design of PWR plant simulation code

    International Nuclear Information System (INIS)

    Delourme, Didier.

    1980-11-01

    This paper presents an improvement of PICOLO, a package for PWR plants simulation. Its describes principally the integration to the code of a primary loop and pressurizer model and the corresponding control loops. Fast transients are tested on the packages and results are compared with real transients obtained on plants [fr

  8. Evolution calculations of fuel for a GFR using MCNPX-C90 and Tripoli-4-D; Calculos de evolucion de combustible para un GFR usando MCNPX-C90 y TRIPOLI-4-D

    Energy Technology Data Exchange (ETDEWEB)

    Reyes R, R.; Martin del Campo M, C.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico); Brun, E.; Dumonteil, E.; Malvagi, F., E-mail: emeric.brun@cea.fr [Commissariat a l' Energie Atomique et aux Energies Alternative, Service d' Etude des Reacteurs et de Mathematiques Appliquees, Saclay, DEN/DM2S/SERMA/LTSD, Bat 470, 91191 Gif-sur-Yvette Cedex (France)

    2011-11-15

    Burnt calculations were realized for a fuel model based on the technology of the Gas-cooled Fast Reactor, GFR. The fuel design is based on bars. The code MCNPX-CINDER90 and the CSADA method for the burnt calculations were used. Models of homogeneous and heterogeneous fuel assembly were studied; for the burnt calculations of the fuel homogeneous model was considered the tracking of three series (Tiers) of evolution of the fission products. The Tier 1 tracks a reduced group of fission products, the Tier 2 tracks to the arrangement of fission products that are contained in the library of cross sections XSDIR of MCNPX; and the Tier 3 tracks 1325 fission products. The results were compared with those obtained with Tripoli-4-D in function of the calculation methods: 1) Explicit Euler, as method of first order; and 2) CSADA, as method of second order. According to the results was observed that the infinite multiplication factor varies in function of the fission products quantity that are tracked. The calculation time used by MCNPX-C90 with the series Tier 3 is more than double than the used by Tripoli-4-D, therefore this last code has advantage over MCNPX-C90 in the case of neutrons analysis of fast reactors. (Author)

  9. A Pass Band Performance Simulation Code of Coupled Cavities

    CERN Document Server

    Tao, X

    2004-01-01

    A simulation code of accelerating cavities named PPSC is developed by the solutions of the microwave equivalent circuit equations. PPSC can give the pass band performance of periodic or non-periodic accelerating structures, such as the dispersion frequency and the reflection factor of the cavity, the field distribution of each mode and so on. The natural parameters of the structure, such as the number of the cavities, the resonant frequencies and Q-factors of each cavity, the coupling factor between two cavities, and the locations of the couplers, can be changed easily to see the different results of the simulation. The code is written based on MS Visual Basic under MS windows. With these, a user-friendly interface is made. Some simple examples was simulated and gave reliable results.

  10. Enhanced Verification Test Suite for Physics Simulation Codes

    Energy Technology Data Exchange (ETDEWEB)

    Kamm, J R; Brock, J S; Brandon, S T; Cotrell, D L; Johnson, B; Knupp, P; Rider, W; Trucano, T; Weirs, V G

    2008-10-10

    This document discusses problems with which to augment, in quantity and in quality, the existing tri-laboratory suite of verification problems used by Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and Sandia National Laboratories (SNL). The purpose of verification analysis is demonstrate whether the numerical results of the discretization algorithms in physics and engineering simulation codes provide correct solutions of the corresponding continuum equations. The key points of this document are: (1) Verification deals with mathematical correctness of the numerical algorithms in a code, while validation deals with physical correctness of a simulation in a regime of interest. This document is about verification. (2) The current seven-problem Tri-Laboratory Verification Test Suite, which has been used for approximately five years at the DOE WP laboratories, is limited. (3) Both the methodology for and technology used in verification analysis have evolved and been improved since the original test suite was proposed. (4) The proposed test problems are in three basic areas: (a) Hydrodynamics; (b) Transport processes; and (c) Dynamic strength-of-materials. (5) For several of the proposed problems we provide a 'strong sense verification benchmark', consisting of (i) a clear mathematical statement of the problem with sufficient information to run a computer simulation, (ii) an explanation of how the code result and benchmark solution are to be evaluated, and (iii) a description of the acceptance criterion for simulation code results. (6) It is proposed that the set of verification test problems with which any particular code be evaluated include some of the problems described in this document. Analysis of the proposed verification test problems constitutes part of a necessary--but not sufficient--step that builds confidence in physics and engineering simulation codes. More complicated test cases, including physics models of

  11. Towards simulating a collisionless shock wave using the FLASH code

    International Nuclear Information System (INIS)

    Farley, D.R.; Shigemori K; Azechi, H.

    2005-01-01

    Full text: In recent years, there has been considerable interest in conducting laser-plasma experiments of relevance to astrophysical research. Results from these experiments can be used to study fundamental physics, as well as benchmark astrophysical codes. The FLASH code, a robust hydrodynamics code produced by the University of Chicago under the Accelerated Strategic Computing Initiative (ASCI), is Bed to simulate a blast wave experiment conducted at the University of Osaka Institute for Laser Engineering (ILE). The code was run in one and two dimensions (axis-symmetric) using a perfect gas equation of state approximation. The shock wave experiment involved irradiating a 2 micron thick piece of plastic (CH) with a high-power, short-pulse laser. The ablated material and resulting shock front from the plastic target Propagated a blast wave into ambient argon at 1.0 Torr. Evolution of the blast wave differs slightly between the cases of Spitzer-Harm conductivity on and off, and neither case matches well with experiments. Due to the high temperatures involved, a thermal wave should be expected such that the Spitzer-Harm conductivity on vase is more likely. Density an velocity profiles from experiment and numerical simulation are compared. The experiment and simulation results compare well blast wave theory

  12. High performance computer code for molecular dynamics simulations

    International Nuclear Information System (INIS)

    Levay, I.; Toekesi, K.

    2007-01-01

    Complete text of publication follows. Molecular Dynamics (MD) simulation is a widely used technique for modeling complicated physical phenomena. Since 2005 we are developing a MD simulations code for PC computers. The computer code is written in C++ object oriented programming language. The aim of our work is twofold: a) to develop a fast computer code for the study of random walk of guest atoms in Be crystal, b) 3 dimensional (3D) visualization of the particles motion. In this case we mimic the motion of the guest atoms in the crystal (diffusion-type motion), and the motion of atoms in the crystallattice (crystal deformation). Nowadays, it is common to use Graphics Devices in intensive computational problems. There are several ways to use this extreme processing performance, but never before was so easy to programming these devices as now. The CUDA (Compute Unified Device) Architecture introduced by nVidia Corporation in 2007 is a very useful for every processor hungry application. A Unified-architecture GPU include 96-128, or more stream processors, so the raw calculation performance is 576(!) GFLOPS. It is ten times faster, than the fastest dual Core CPU [Fig.1]. Our improved MD simulation software uses this new technology, which speed up our software and the code run 10 times faster in the critical calculation code segment. Although the GPU is a very powerful tool, it has a strongly paralleled structure. It means, that we have to create an algorithm, which works on several processors without deadlock. Our code currently uses 256 threads, shared and constant on-chip memory, instead of global memory, which is 100 times slower than others. It is possible to implement the total algorithm on GPU, therefore we do not need to download and upload the data in every iteration. On behalf of maximal throughput, every thread run with the same instructions

  13. Neutronic simulation of a research reactor core of (232 Th, 235 U ...

    Indian Academy of Sciences (India)

    Home; Journals; Pramana – Journal of Physics; Volume 80; Issue 1. Neutronic simulation of a research reactor core of (232Th, 235U)O2 fuel using MCNPX2.6 code. Seyed Amir ... Many researchers consider long core life with no on-site refuelling activity as a primary feature for the small reactor design. Long core life can be ...

  14. ZENO: N-body and SPH Simulation Codes

    Science.gov (United States)

    Barnes, Joshua E.

    2011-02-01

    The ZENO software package integrates N-body and SPH simulation codes with a large array of programs to generate initial conditions and analyze numerical simulations. Written in C, the ZENO system is portable between Mac, Linux, and Unix platforms. It is in active use at the Institute for Astronomy (IfA), at NRAO, and possibly elsewhere. Zeno programs can perform a wide range of simulation and analysis tasks. While many of these programs were first created for specific projects, they embody algorithms of general applicability and embrace a modular design strategy, so existing code is easily applied to new tasks. Major elements of the system include: Structured data file utilities facilitate basic operations on binary data, including import/export of ZENO data to other systems.Snapshot generation routines create particle distributions with various properties. Systems with user-specified density profiles can be realized in collisionless or gaseous form; multiple spherical and disk components may be set up in mutual equilibrium.Snapshot manipulation routines permit the user to sift, sort, and combine particle arrays, translate and rotate particle configurations, and assign new values to data fields associated with each particle.Simulation codes include both pure N-body and combined N-body/SPH programs: Pure N-body codes are available in both uniprocessor and parallel versions.SPH codes offer a wide range of options for gas physics, including isothermal, adiabatic, and radiating models. Snapshot analysis programs calculate temporal averages, evaluate particle statistics, measure shapes and density profiles, compute kinematic properties, and identify and track objects in particle distributions.Visualization programs generate interactive displays and produce still images and videos of particle distributions; the user may specify arbitrary color schemes and viewing transformations.

  15. Simulation of water hammer phenomena using the system code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Bratfisch, Christoph; Koch, Marco K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group

    2017-07-15

    Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.

  16. Simulation of water hammer phenomena using the system code ATHLET

    International Nuclear Information System (INIS)

    Bratfisch, Christoph; Koch, Marco K.

    2017-01-01

    Water Hammer Phenomena can endanger the integrity of structures leading to a possible failure of pipes in nuclear power plants as well as in many industrial applications. These phenomena can arise in nuclear power plants in the course of transients and accidents induced by the start-up of auxiliary feed water systems or emergency core cooling systems in combination with rapid acting valves and pumps. To contribute to further development and validation of the code ATHLET (Analysis of Thermalhydraulics of Leaks and Transients), an experiment performed in the test facility Pilot Plant Pipework (PPP) at Fraunhofer UMSICHT is simulated using the code version ATHLET 3.0A.

  17. Scientific codes developed and used at GRS. Nuclear simulation chain

    Energy Technology Data Exchange (ETDEWEB)

    Schaffrath, Andreas; Sonnenkalb, Martin; Sievers, Juergen; Luther, Wolfgang; Velkov, Kiril [Gesellschaft fuer Anlagen und Reaktorsicherheit (GRS) gGmbH, Garching/Muenchen (Germany). Forschungszentrum

    2016-05-15

    Over 60 technical experts of the reactor safety research division of the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH are developing and validating reliable methods and computer codes - summarized under the term nuclear simulation chain - for the safety-related assessment for all types of nuclear power plants (NPP) and other nuclear facilities considering the current state of science and technology. This nuclear simulation chain has to be able to simulate and assess all relevant physical processes and phenomena for all operating states and (severe) accidents. In the present contribution, the nuclear simulation chain developed and applied by GRS as well as selected examples of its application are presented. The latter demonstrate impressively the width of its scope and its performance. The GRS codes can be passed on request to other (national as well as international) organizations. This contributes to a worldwide increase of the nuclear safety standards. The code transfer is especially important for developing and emerging countries lacking the financial means and/or the necessary know-how for this purpose. At the end of this contribution, the respective course of action is described.

  18. Simulations of Laboratory Astrophysics Experiments using the CRASH code

    Science.gov (United States)

    Trantham, Matthew; Kuranz, Carolyn; Fein, Jeff; Wan, Willow; Young, Rachel; Keiter, Paul; Drake, R. Paul

    2015-11-01

    Computer simulations can assist in the design and analysis of laboratory astrophysics experiments. The Center for Radiative Shock Hydrodynamics (CRASH) at the University of Michigan developed a code that has been used to design and analyze high-energy-density experiments on OMEGA, NIF, and other large laser facilities. This Eulerian code uses block-adaptive mesh refinement (AMR) with implicit multigroup radiation transport, electron heat conduction and laser ray tracing. This poster will demonstrate some of the experiments the CRASH code has helped design or analyze including: Kelvin-Helmholtz, Rayleigh-Taylor, magnetized flows, jets, and laser-produced plasmas. This work is funded by the following grants: DEFC52-08NA28616, DE-NA0001840, and DE-NA0002032.

  19. Steam explosion simulation code JASMINE v.3 user's guide

    International Nuclear Information System (INIS)

    Moriyama, Kiyofumi; Maruyama, Yu; Nakamura, Hideo

    2008-07-01

    A steam explosion occurs when hot liquid contacts with cold volatile liquid. In this phenomenon, fine fragmentation of the hot liquid causes extremely rapid heat transfer from the hot liquid to the cold volatile liquid, and explosive vaporization, bringing shock waves and destructive forces. The steam explosion due to the contact of the molten core material and coolant water during severe accidents of light water reactors has been regarded as a potential threat to the integrity of the containment vessel. We developed a mechanistic steam explosion simulation code, JASMINE, that is applicable to plant scale assessment of the steam explosion loads. This document, as a manual for users of JASMINE code, describes the models, numerical solution methods, and also some verification and example calculations, as well as practical instructions for input preparation and usage of the code. (author)

  20. openQ*D simulation code for QCD+QED

    Science.gov (United States)

    Campos, Isabel; Fritzsch, Patrick; Hansen, Martin; Krstić Marinković, Marina; Patella, Agostino; Ramos, Alberto; Tantalo, Nazario

    2018-03-01

    The openQ*D code for the simulation of QCD+QED with C* boundary conditions is presented. This code is based on openQCD-1.6, from which it inherits the core features that ensure its efficiency: the locally-deflated SAP-preconditioned GCR solver, the twisted-mass frequency splitting of the fermion action, the multilevel integrator, the 4th order OMF integrator, the SSE/AVX intrinsics, etc. The photon field is treated as fully dynamical and C* boundary conditions can be chosen in the spatial directions. We discuss the main features of openQ*D, and we show basic test results and performance analysis. An alpha version of this code is publicly available and can be downloaded from http://rcstar.web.cern.ch/.

  1. HYDRASTAR - a code for stochastic simulation of groundwater flow

    International Nuclear Information System (INIS)

    Norman, S.

    1992-05-01

    The computer code HYDRASTAR was developed as a tool for groundwater flow and transport simulations in the SKB 91 safety analysis project. Its conceptual ideas can be traced back to a report by Shlomo Neuman in 1988, see the reference section. The main idea of the code is the treatment of the rock as a stochastic continuum which separates it from the deterministic methods previously employed by SKB and also from the discrete fracture models. The current report is a comprehensive description of HYDRASTAR including such topics as regularization or upscaling of a hydraulic conductivity field, unconditional and conditional simulation of stochastic processes, numerical solvers for the hydrology and streamline equations and finally some proposals for future developments

  2. Parallelization of a Monte Carlo particle transport simulation code

    Science.gov (United States)

    Hadjidoukas, P.; Bousis, C.; Emfietzoglou, D.

    2010-05-01

    We have developed a high performance version of the Monte Carlo particle transport simulation code MC4. The original application code, developed in Visual Basic for Applications (VBA) for Microsoft Excel, was first rewritten in the C programming language for improving code portability. Several pseudo-random number generators have been also integrated and studied. The new MC4 version was then parallelized for shared and distributed-memory multiprocessor systems using the Message Passing Interface. Two parallel pseudo-random number generator libraries (SPRNG and DCMT) have been seamlessly integrated. The performance speedup of parallel MC4 has been studied on a variety of parallel computing architectures including an Intel Xeon server with 4 dual-core processors, a Sun cluster consisting of 16 nodes of 2 dual-core AMD Opteron processors and a 200 dual-processor HP cluster. For large problem size, which is limited only by the physical memory of the multiprocessor server, the speedup results are almost linear on all systems. We have validated the parallel implementation against the serial VBA and C implementations using the same random number generator. Our experimental results on the transport and energy loss of electrons in a water medium show that the serial and parallel codes are equivalent in accuracy. The present improvements allow for studying of higher particle energies with the use of more accurate physical models, and improve statistics as more particles tracks can be simulated in low response time.

  3. KULL: LLNL's ASCI Inertial Confinement Fusion Simulation Code

    International Nuclear Information System (INIS)

    Rathkopf, J. A.; Miller, D. S.; Owen, J. M.; Zike, M. R.; Eltgroth, P. G.; Madsen, N. K.; McCandless, K. P.; Nowak, P. F.; Nemanic, M. K.; Gentile, N. A.; Stuart, L. M.; Keen, N. D.; Palmer, T. S.

    2000-01-01

    KULL is a three dimensional, time dependent radiation hydrodynamics simulation code under development at Lawrence Livermore National Laboratory. A part of the U.S. Department of Energy's Accelerated Strategic Computing Initiative (ASCI), KULL's purpose is to simulate the physical processes in Inertial Confinement Fusion (ICF) targets. The National Ignition Facility, where ICF experiments will be conducted, and ASCI are part of the experimental and computational components of DOE's Stockpile Stewardship Program. This paper provides an overview of ASCI and describes KULL, its hydrodynamic simulation capability and its three methods of simulating radiative transfer. Particular emphasis is given to the parallelization techniques essential to obtain the performance required of the Stockpile Stewardship Program and to exploit the massively parallel processor machines that ASCI is procuring

  4. Electron cloud effects: codes and simulations at KEK

    CERN Document Server

    Ohmi, K.

    2013-04-22

    Electron cloud effects had been studied at KEK-Photon Factory since 1995. e-p instability had been studied in proton rings since 1965 in BINP, ISR and PSR. Study of electron cloud effects with the present style, which was based on numerical simulations, started at 1995 in positron storage rings. The instability observed in KEKPF gave a strong impact to B factories, KEKB and PEPII, which were final stage of their design in those days. History of cure for electron cloud instability overlapped the progress of luminosity performance in KEKB. The studies on electron cloud codes and simulations in KEK are presented.

  5. Systematic effects in CALOR simulation code to model experimental configurations

    International Nuclear Information System (INIS)

    Job, P.K.; Proudfoot, J.; Handler, T.

    1991-01-01

    CALOR89 code system is being used to simulate test beam results and the design parameters of several calorimeter configurations. It has been bench-marked against the ZEUS, Dθ and HELIOS data. This study identifies the systematic effects in CALOR simulation to model the experimental configurations. Five major systematic effects are identified. These are the choice of high energy nuclear collision model, material composition, scintillator saturation, shower integration time, and the shower containment. Quantitative estimates of these systematic effects are presented. 23 refs., 6 figs., 7 tabs

  6. Enhanced verification test suite for physics simulation codes

    Energy Technology Data Exchange (ETDEWEB)

    Kamm, James R.; Brock, Jerry S.; Brandon, Scott T.; Cotrell, David L.; Johnson, Bryan; Knupp, Patrick; Rider, William J.; Trucano, Timothy G.; Weirs, V. Gregory

    2008-09-01

    This document discusses problems with which to augment, in quantity and in quality, the existing tri-laboratory suite of verification problems used by Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and Sandia National Laboratories (SNL). The purpose of verification analysis is demonstrate whether the numerical results of the discretization algorithms in physics and engineering simulation codes provide correct solutions of the corresponding continuum equations.

  7. Use of advanced simulations in fuel performance codes

    International Nuclear Information System (INIS)

    Van Uffelen, P.

    2015-01-01

    The simulation of the cylindrical fuel rod behaviour in a reactor or a storage pool for spent fuel requires a fuel performance code. Such tool solves the equations for the heat transfer, the stresses and strains in fuel and cladding, the evolution of several isotopes and the behaviour of various fission products in the fuel rod. The main equations along with their limitations are briefly described. The current approaches adopted for overcoming these limitations and the perspectives are also outlined. (author)

  8. CHOLLA: A NEW MASSIVELY PARALLEL HYDRODYNAMICS CODE FOR ASTROPHYSICAL SIMULATION

    International Nuclear Information System (INIS)

    Schneider, Evan E.; Robertson, Brant E.

    2015-01-01

    We present Computational Hydrodynamics On ParaLLel Architectures (Cholla ), a new three-dimensional hydrodynamics code that harnesses the power of graphics processing units (GPUs) to accelerate astrophysical simulations. Cholla models the Euler equations on a static mesh using state-of-the-art techniques, including the unsplit Corner Transport Upwind algorithm, a variety of exact and approximate Riemann solvers, and multiple spatial reconstruction techniques including the piecewise parabolic method (PPM). Using GPUs, Cholla evolves the fluid properties of thousands of cells simultaneously and can update over 10 million cells per GPU-second while using an exact Riemann solver and PPM reconstruction. Owing to the massively parallel architecture of GPUs and the design of the Cholla code, astrophysical simulations with physically interesting grid resolutions (≳256 3 ) can easily be computed on a single device. We use the Message Passing Interface library to extend calculations onto multiple devices and demonstrate nearly ideal scaling beyond 64 GPUs. A suite of test problems highlights the physical accuracy of our modeling and provides a useful comparison to other codes. We then use Cholla to simulate the interaction of a shock wave with a gas cloud in the interstellar medium, showing that the evolution of the cloud is highly dependent on its density structure. We reconcile the computed mixing time of a turbulent cloud with a realistic density distribution destroyed by a strong shock with the existing analytic theory for spherical cloud destruction by describing the system in terms of its median gas density

  9. 3D code for simulations of fluid flows

    International Nuclear Information System (INIS)

    Skandera, D.

    2004-01-01

    In this paper, a present status in the development of the new numerical code is reported. The code is considered for simulations of fluid flows. The finite volume approach is adopted for solving standard fluid equations. They are treated in a conservative form to ensure a correct conservation of fluid quantities. Thus, a nonlinear hyperbolic system of conservation laws is numerically solved. The code uses the Eulerian description of the fluid and is designed as a high order central numerical scheme. The central approach employs no (approximate) Riemann solver and is less computational expensive. The high order WENO strategy is adopted in the reconstruction step to achieve results comparable with more accurate Riemann solvers. A combination of the central approach with an iterative solving of a local Riemann problem is tested and behaviour of such numerical flux is reported. An extension to three dimensions is implemented using a dimension by dimension approach, hence, no complicated dimensional splitting need to be introduced. The code is fully parallelized with the MPI library. Several standard hydrodynamic tests in one, two and three dimensions were performed and their results are presented. (author)

  10. Comparing DINA code simulations with TCV experimental plasma equilibrium responses

    International Nuclear Information System (INIS)

    Khayrutdinov, R.R.; Lister, J.B.; Lukash, V.E.; Wainwright, J.P.

    2000-08-01

    The DINA non-linear time dependent simulation code has been validated against an extensive set of plasma equilibrium response experiments carried out on the TCV tokamak. Limited and diverted plasmas are found to be well modelled during the plasma current flat top. In some simulations the application of the PF coil voltage stimulation pulse sufficiently changed the plasma equilibrium that the vertical position feedback control loop became unstable. This behaviour was also found in the experimental work, and cannot be reproduced using linear time-independent models. A single null diverted plasma discharge was also simulated from start-up to shut-down and the results were found to accurately reproduce their experimental equivalents. The most significant difference noted was the penetration time of the poloidal flux, leading to a delayed onset of sawtoothing in the DINA simulation. The complete set of frequency stimulation experiments used to measure the open loop tokamak plasma equilibrium response was also simulated using DINA and the results were analysed in an identical fashion to the experimental data. The frequency response of the DINA simulations agrees with the experimental results. Comparisons with linear models are also discussed to identify areas of good and only occasionally less good agreement. (author)

  11. A computer code to simulate X-ray imaging techniques

    Energy Technology Data Exchange (ETDEWEB)

    Duvauchelle, Philippe E-mail: philippe.duvauchelle@insa-lyon.fr; Freud, Nicolas; Kaftandjian, Valerie; Babot, Daniel

    2000-09-01

    A computer code was developed to simulate the operation of radiographic, radioscopic or tomographic devices. The simulation is based on ray-tracing techniques and on the X-ray attenuation law. The use of computer-aided drawing (CAD) models enables simulations to be carried out with complex three-dimensional (3D) objects and the geometry of every component of the imaging chain, from the source to the detector, can be defined. Geometric unsharpness, for example, can be easily taken into account, even in complex configurations. Automatic translations or rotations of the object can be performed to simulate radioscopic or tomographic image acquisition. Simulations can be carried out with monochromatic or polychromatic beam spectra. This feature enables, for example, the beam hardening phenomenon to be dealt with or dual energy imaging techniques to be studied. The simulation principle is completely deterministic and consequently the computed images present no photon noise. Nevertheless, the variance of the signal associated with each pixel of the detector can be determined, which enables contrast-to-noise ratio (CNR) maps to be computed, in order to predict quantitatively the detectability of defects in the inspected object. The CNR is a relevant indicator for optimizing the experimental parameters. This paper provides several examples of simulated images that illustrate some of the rich possibilities offered by our software. Depending on the simulation type, the computation time order of magnitude can vary from 0.1 s (simple radiographic projection) up to several hours (3D tomography) on a PC, with a 400 MHz microprocessor. Our simulation tool proves to be useful in developing new specific applications, in choosing the most suitable components when designing a new testing chain, and in saving time by reducing the number of experimental tests.

  12. A computer code to simulate X-ray imaging techniques

    International Nuclear Information System (INIS)

    Duvauchelle, Philippe; Freud, Nicolas; Kaftandjian, Valerie; Babot, Daniel

    2000-01-01

    A computer code was developed to simulate the operation of radiographic, radioscopic or tomographic devices. The simulation is based on ray-tracing techniques and on the X-ray attenuation law. The use of computer-aided drawing (CAD) models enables simulations to be carried out with complex three-dimensional (3D) objects and the geometry of every component of the imaging chain, from the source to the detector, can be defined. Geometric unsharpness, for example, can be easily taken into account, even in complex configurations. Automatic translations or rotations of the object can be performed to simulate radioscopic or tomographic image acquisition. Simulations can be carried out with monochromatic or polychromatic beam spectra. This feature enables, for example, the beam hardening phenomenon to be dealt with or dual energy imaging techniques to be studied. The simulation principle is completely deterministic and consequently the computed images present no photon noise. Nevertheless, the variance of the signal associated with each pixel of the detector can be determined, which enables contrast-to-noise ratio (CNR) maps to be computed, in order to predict quantitatively the detectability of defects in the inspected object. The CNR is a relevant indicator for optimizing the experimental parameters. This paper provides several examples of simulated images that illustrate some of the rich possibilities offered by our software. Depending on the simulation type, the computation time order of magnitude can vary from 0.1 s (simple radiographic projection) up to several hours (3D tomography) on a PC, with a 400 MHz microprocessor. Our simulation tool proves to be useful in developing new specific applications, in choosing the most suitable components when designing a new testing chain, and in saving time by reducing the number of experimental tests

  13. Calibration of the TIME2 environmental simulation code

    International Nuclear Information System (INIS)

    Wilmot, R.D.; Hiscock, K.; Lloyd, J.

    1991-04-01

    The TARGET finite-difference groundwater modelling code has been used to reconstruct the hydrogeological environment of the area around Killingholme, Humberside, UK. Reconstructions have been made for the present day and for three periods during the past 120,000 years. Permeability development in the Chalk and the stratified nature of the current groundwater system act as boundary conditions for these reconstructions. The results from these reconstructions have been compared with values used by the environmental simulation code TIME2. With optimisation of partition coefficients within the water budget sub-model, values for recharge from TIME2 accord closely with those from this study for temperate and boreal conditions. TIME2 over-estimates recharge during tundra climate states because it does not account for permafrost. (author)

  14. ELEGANT: A flexible SDDS-compliant code for accelerator simulation

    International Nuclear Information System (INIS)

    Borland, M.

    2000-01-01

    ELEGANT (ELEctron Generation ANd Tracking) is the principle accelerator simulation code used at the Advanced Photon Source (APS) for circular and one-pass machines. Capabilities include 6-D tracking using matrices up to third order, canonical integration, and numerical integration. Standard beamline elements are supported, as well as coherent synchrotron radiation, wakefields, rf elements, kickers, apertures, scattering, and more. In addition to tracking with and without errors, ELEGANT performs optimization of tracked properties, as well as computation and optimization of Twiss parameters, radiation integrals, matrices, and floor coordinates. Orbit/trajectory, tune, and chromaticity correction are supported. ELEGANT is fully compliant with the Self Describing Data Sets (SDDS) file protocol, and hence uses the SDDS Toolkit for pre- and post-processing. This permits users to prepare scripts to run the code in a flexible and automated fashion. It is particularly well suited to multistage simulation and concurrent simulation on many workstations. Several examples of complex projects performed with ELEGANT are given, including top-up safety analysis of the APS and design of the APS bunch compressor

  15. Galerkin algorithm for multidimensional plasma simulation codes. Informal report

    International Nuclear Information System (INIS)

    Godfrey, B.B.

    1979-03-01

    A Galerkin finite element differencing scheme has been developed for a computer simulation of plasmas. The new difference equations identically satisfy an equation of continuity. Thus, the usual current correction procedure, involving inversion of Poisson's equation, is unnecessary. The algorithm is free of many numerical Cherenkov instabilities. This differencing scheme has been implemented in CCUBE, an already existing relativistic, electromagnetic, two-dimensional PIC code in arbitrary separable, orthogonal coordinates. The separability constraint is eliminated by the new algorithm. The new version of CCUBE exhibits good stability and accuracy with reduced computer memory and time requirements. Details of the algorithm and its implementation are presented

  16. GOTHIC code simulation of thermal stratification in POOLEX facility

    Energy Technology Data Exchange (ETDEWEB)

    Li, H.; Kudinov, P. (Royal Institute of Technology (KTH) (Sweden))

    2009-07-15

    Pressure suppression pool is an important element of BWR containment. It serves as a heat sink and steam condenser to prevent containment pressure buildup during loss of coolant accident or safety relief valve opening during normal operations of a BWR. Insufficient mixing in the pool, in case of low mass flow rate of steam, can cause development of thermal stratification and reduction of pressure suppression pool capacity. For reliable prediction of mixing and stratification phenomena validation of simulation tools has to be performed. Data produced in POOLEX/PPOOLEX facility at Lappeenranta University of Technology about development of thermal stratification in a large scale model of a pressure suppression pool is used for GOTHIC lumped and distributed parameter validation. Sensitivity of GOTHIC solution to different boundary conditions and grid convergence study for 2D simulations of POOLEX STB-20 experiment are performed in the present study. CFD simulation was carried out with FLUENT code in order to get additional insights into physics of stratification phenomena. In order to support development of experimental procedures for new tests in the PPOOLEX facility lumped parameter pre-test GOTHIC simulations were performed. Simulations show that drywell and wetwell pressures can be kept within safety margins during a long transient necessary for development of thermal stratification. (au)

  17. GOTHIC code simulation of thermal stratification in POOLEX facility

    International Nuclear Information System (INIS)

    Li, H.; Kudinov, P.

    2009-07-01

    Pressure suppression pool is an important element of BWR containment. It serves as a heat sink and steam condenser to prevent containment pressure buildup during loss of coolant accident or safety relief valve opening during normal operations of a BWR. Insufficient mixing in the pool, in case of low mass flow rate of steam, can cause development of thermal stratification and reduction of pressure suppression pool capacity. For reliable prediction of mixing and stratification phenomena validation of simulation tools has to be performed. Data produced in POOLEX/PPOOLEX facility at Lappeenranta University of Technology about development of thermal stratification in a large scale model of a pressure suppression pool is used for GOTHIC lumped and distributed parameter validation. Sensitivity of GOTHIC solution to different boundary conditions and grid convergence study for 2D simulations of POOLEX STB-20 experiment are performed in the present study. CFD simulation was carried out with FLUENT code in order to get additional insights into physics of stratification phenomena. In order to support development of experimental procedures for new tests in the PPOOLEX facility lumped parameter pre-test GOTHIC simulations were performed. Simulations show that drywell and wetwell pressures can be kept within safety margins during a long transient necessary for development of thermal stratification. (au)

  18. Computer code for the atomistic simulation of lattice defects and dynamics. [COMENT code

    Energy Technology Data Exchange (ETDEWEB)

    Schiffgens, J.O.; Graves, N.J.; Oster, C.A.

    1980-04-01

    This document has been prepared to satisfy the need for a detailed, up-to-date description of a computer code that can be used to simulate phenomena on an atomistic level. COMENT was written in FORTRAN IV and COMPASS (CDC assembly language) to solve the classical equations of motion for a large number of atoms interacting according to a given force law, and to perform the desired ancillary analysis of the resulting data. COMENT is a dual-purpose intended to describe static defect configurations as well as the detailed motion of atoms in a crystal lattice. It can be used to simulate the effect of temperature, impurities, and pre-existing defects on radiation-induced defect production mechanisms, defect migration, and defect stability.

  19. Computer code for simulating pressurized water reactor core

    International Nuclear Information System (INIS)

    Serrano, A.M.B.

    1978-01-01

    A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numerically. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistance added to the film coefficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (author)

  20. Code for the core simulation in pressurized water reactors

    International Nuclear Information System (INIS)

    Serrano, M.A.B.

    1978-08-01

    A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numericaly. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistence added to the film coeficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (Author) [pt

  1. Simulation of Code Spectrum and Code Flow of Cultured Neuronal Networks.

    Science.gov (United States)

    Tamura, Shinichi; Nishitani, Yoshi; Hosokawa, Chie; Miyoshi, Tomomitsu; Sawai, Hajime

    2016-01-01

    It has been shown that, in cultured neuronal networks on a multielectrode, pseudorandom-like sequences (codes) are detected, and they flow with some spatial decay constant. Each cultured neuronal network is characterized by a specific spectrum curve. That is, we may consider the spectrum curve as a "signature" of its associated neuronal network that is dependent on the characteristics of neurons and network configuration, including the weight distribution. In the present study, we used an integrate-and-fire model of neurons with intrinsic and instantaneous fluctuations of characteristics for performing a simulation of a code spectrum from multielectrodes on a 2D mesh neural network. We showed that it is possible to estimate the characteristics of neurons such as the distribution of number of neurons around each electrode and their refractory periods. Although this process is a reverse problem and theoretically the solutions are not sufficiently guaranteed, the parameters seem to be consistent with those of neurons. That is, the proposed neural network model may adequately reflect the behavior of a cultured neuronal network. Furthermore, such prospect is discussed that code analysis will provide a base of communication within a neural network that will also create a base of natural intelligence.

  2. Large interface simulation in an averaged two-fluid code

    International Nuclear Information System (INIS)

    Henriques, A.

    2006-01-01

    Different ranges of size of interfaces and eddies are involved in multiphase flow phenomena. Classical formalisms focus on a specific range of size. This study presents a Large Interface Simulation (LIS) two-fluid compressible formalism taking into account different sizes of interfaces. As in the single-phase Large Eddy Simulation, a filtering process is used to point out Large Interface (LI) simulation and Small interface (SI) modelization. The LI surface tension force is modelled adapting the well-known CSF method. The modelling of SI transfer terms is done calling for classical closure laws of the averaged approach. To simulate accurately LI transfer terms, we develop a LI recognition algorithm based on a dimensionless criterion. The LIS model is applied in a classical averaged two-fluid code. The LI transfer terms modelling and the LI recognition are validated on analytical and experimental tests. A square base basin excited by a horizontal periodic movement is studied with the LIS model. The capability of the model is also shown on the case of the break-up of a bubble in a turbulent liquid flow. The break-up of a large bubble at a grid impact performed regime transition between two different scales of interface from LI to SI and from PI to LI. (author) [fr

  3. Overview of the Tusas Code for Simulation of Dendritic Solidification

    Energy Technology Data Exchange (ETDEWEB)

    Trainer, Amelia J. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Newman, Christopher Kyle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Francois, Marianne M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-01-07

    The aim of this project is to conduct a parametric investigation into the modeling of two dimensional dendrite solidification, using the phase field model. Specifically, we use the Tusas code, which is for coupled heat and phase-field simulation of dendritic solidification. Dendritic solidification, which may occur in the presence of an unstable solidification interface, results in treelike microstructures that often grow perpendicular to the rest of the growth front. The interface may become unstable if the enthalpy of the solid material is less than that of the liquid material, or if the solute is less soluble in solid than it is in liquid, potentially causing a partition [1]. A key motivation behind this research is that a broadened understanding of phase-field formulation and microstructural developments can be utilized for macroscopic simulations of phase change. This may be directly implemented as a part of the Telluride project at Los Alamos National Laboratory (LANL), through which a computational additive manufacturing simulation tool is being developed, ultimately to become part of the Advanced Simulation and Computing Program within the U.S. Department of Energy [2].

  4. Development and assessment of ASTEC code for severe accident simulation

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.P.; Pignet, S.; Seropian, C.; Montanelli, T.; Giordano, P.; Jacq, F.; Schwinges, B.

    2005-01-01

    Full text of publication follows: The ASTEC integral code, jointly developed by IRSN and GRS since several years for evaluation of source term during a severe accident (SA) in a Light Water Reactor, will play a central role in the SARNET network of excellence of the 6. Framework Programme (FwP) of the European Commission which started in spring 2004. It should become the reference European SA integral code in the next years. The version V1.1, released in June 2004, allows to model most of the main physical phenomena (except steam explosion) near or at the state of the art. 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 usually takes less than one day of real time to be simulated on a PC computer. Important efforts are being made on validation by covering more than 30 reference experiments, often International Standard Problems from OECD (CORA, LOFT, PACTEL, BETA, VANAM, ACE-RTF, Phebus.FPT1...). The code is also used for the detailed interpretation of all the integral Phebus.FP experiments. Eighteen European partners performed a first independent evaluation of the code capabilities in 2000-03 within the frame of the EVITA 5. FwP project on one hand by comparison to experiments and on another hand by benchmarking with MAAP4 and MELCOR integral codes on plant applications on PWR and VVER. Their main conclusions were the needs of improvement of code robustness (especially the 2 new modules CESAR and DIVA simulating respectively circuit thermal hydraulics and core degradation) and of post-processing tools. Some improvements have already been achieved in the latest version V 1.1 on these two aspects. A new module MEDICIS devoted to Molten Core Concrete Interaction (MCCI) is implemented in this version, with a tight coupling to the containment thermal hydraulics module CPA. The paper presents a detailed analysis of a TMLB sequence on a French 900 MWe PWR, from

  5. The MCUCN simulation code for ultracold neutron physics

    Science.gov (United States)

    Zsigmond, G.

    2018-02-01

    Ultracold neutrons (UCN) have very low kinetic energies 0-300 neV, thereby can be stored in specific material or magnetic confinements for many hundreds of seconds. This makes them a very useful tool in probing fundamental symmetries of nature (for instance charge-parity violation by neutron electric dipole moment experiments) and contributing important parameters for the Big Bang nucleosynthesis (neutron lifetime measurements). Improved precision experiments are in construction at new and planned UCN sources around the world. MC simulations play an important role in the optimization of such systems with a large number of parameters, but also in the estimation of systematic effects, in benchmarking of analysis codes, or as part of the analysis. The MCUCN code written at PSI has been extensively used for the optimization of the UCN source optics and in the optimization and analysis of (test) experiments within the nEDM project based at PSI. In this paper we present the main features of MCUCN and interesting benchmark and application examples.

  6. Neoclassical Simulation of Tokamak Plasmas using Continuum Gyrokinetc Code TEMPEST

    International Nuclear Information System (INIS)

    Xu, X Q

    2007-01-01

    We present gyrokinetic neoclassical simulations of tokamak plasmas with self-consistent electric field for the first time using a fully nonlinear (full-f) continuum code TEMPEST in a circular geometry. A set of gyrokinetic equations are discretized on a five dimensional computational grid in phase space. The present implementation is a Method of Lines approach where the phase-space derivatives are discretized with finite differences and implicit backwards differencing formulas are used to advance the system in time. The fully nonlinear Boltzmann model is used for electrons. The neoclassical electric field is obtained by solving gyrokinetic Poisson equation with self-consistent poloidal variation. With our 4D (ψ, θ, ε, μ) version of the TEMPEST code we compute radial particle and heat flux, the Geodesic-Acoustic Mode (GAM), and the development of neoclassical electric field, which we compare with neoclassical theory with a Lorentz collision model. The present work provides a numerical scheme and a new capability for self-consistently studying important aspects of neoclassical transport and rotations in toroidal magnetic fusion devices

  7. Development of a methodology for simulation of gas cooled reactors with purpose of transmutation

    International Nuclear Information System (INIS)

    Silva, Clarysson Alberto da

    2009-01-01

    This work proposes a methodology of MHR (Modular Helium Reactor) simulation using the WIMSD-5B (Winfrith Improved Multi/group Scheme) nuclear code which is validated by MCNPX 2.6.0 (Monte Carlo N-Particle transport eXtend) nuclear code. The goal is verify the capability of WIMSD-5B to simulate a reactor type GT-MHR (Gas Turbine Modular Helium Reactor), considering all the fuel recharges possibilities. Also is evaluated the possibility of WIMSD-5B to represent adequately the fuel evolution during the fuel recharge. Initially was verified the WIMSD-5B capability to simulate the recharge specificities of this model by analysis of neutronic parameters and isotopic composition during the burnup. After the model was simulated using both WIMSD-5B and MCNPX 2.6.0 codes and the results of k eff , neutronic flux and isotopic composition were compared. The results show that the deterministic WIMSD-5B code can be applied to a qualitative evaluation, representing adequately the core behavior during the fuel recharges being possible in a short period of time to inquire about the burned core that, once optimized, can be quantitatively evaluated by a code type MCNPX 2.6.0. (author)

  8. A new MCNPX PTRAC coincidence capture file capability: a tool for neutron detector design

    International Nuclear Information System (INIS)

    Evans, Louise G.; Schear, Melissa A.; Hendricks, John S.; Swinhoe, Martyn T.; Tobin, Stephen J.; Croft, Stephen

    2010-01-01

    The existing MCNPX(trademark) PTRAC coincidence capture file allows a full list of neutron capture events to be recorded in any simulated detection medium. The originating event history number (e.g. spontaneous fission events), capture time, location and source particle number are tracked and output to file for post-processing. We have developed a new MCNPX PTRAC coincidence capture file capability to aid detector design studies. New features include the ability to track the isotopes that emitted the detected neutrons as well as induced fission chains in mixed samples before detection (both generation number and isotope). Here, the power of this tool is demonstrated using a detector design that has been developed for the non-destructive assay (NDA) of spent nuclear fuel. Individual capture time distributions have been generated for neutrons originating from Curium-244 source spontaneous fission events and induced fission events in fissile isotopes of interest: namely Plutonium-239, Plutonium-241, and Uranium-235. Through this capability, a full picture for the attribution of neutron capture events in the detector can be simulated.

  9. Particle tracking code of simulating global RF feedback

    International Nuclear Information System (INIS)

    Mestha, L.K.

    1991-09-01

    It is well known in the ''control community'' that a good feedback controller design is deeply rooted in the physics of the system. For example, when accelerating the beam we must keep several parameters under control so that the beam travels within the confined space. Important parameters include the frequency and phase of the rf signal, the dipole field, and the cavity voltage. Because errors in these parameters will progressively mislead the beam from its projected path in the tube, feedback loops are used to correct the behavior. Since the feedback loop feeds energy to the system, it changes the overall behavior of the system and may drive it to instability. Various types of controllers are used to stabilize the feedback loop. Integrating the beam physics with the feedback controllers allows us to carefully analyze the beam behavior. This will not only guarantee optimal performance but will also significantly enhance the ability of the beam control engineer to deal effectively with the interaction of various feedback loops. Motivated by this theme, we developed a simple one-particle tracking code to simulate particle behavior with feedback controllers. In order to achieve our fundamental objective, we can ask some key questions: What are the input and output parameters? How can they be applied to the practical machine? How can one interface the rf system dynamics such as the transfer characteristics of the rf cavities and phasing between the cavities? Answers to these questions can be found by considering a simple case of a single cavity with one particle, tracking it turn-by-turn with appropriate initial conditions, then introducing constraints on crucial parameters. Critical parameters are rf frequency, phase, and amplitude once the dipole field has been given. These are arranged in the tracking code so that we can interface the feedback system controlling them

  10. Monte-Carlo simulation of proton radiotherapy for human eye

    International Nuclear Information System (INIS)

    Liu Yunpeng; Tang Xiaobin; Xie Qin; Chen Feida; Geng Changran; Chen Da

    2010-01-01

    The 62 MeV proton beam was selected to develop a MCNPX model of the human eye to approximate dose delivered from proton therapy by. In the course of proton therapy, two treatment simulations were considered. The first simulation was an ideal treatment scenario. In this case, the dose of tumor was 50.03 Gy, which was at the level of effective treatment, while other organizations were in the range of acceptable dose. The second case was a worst case scenario to simulate a patient gazing directly into the treatment beam during therapy. The bulk of dose deposited in the cornea, lens, and anterior chamber region. However, the dose of tumor area was zero. The calculated results show an agreement accordance with the relative reference, which confirmed that the MCNPX code can simulate proton radiotherapy perfectly, and is a capable platform for patient planning. The data from the worst case can be used for dose reconstruction of the clinical accident. (authors)

  11. Simulation of water hammer experiments using RELAP5 code

    International Nuclear Information System (INIS)

    Kaliatka, A.; Vaisnoras, M.

    2005-01-01

    The rapid closing or opening of a valve causes pressure transients in pipelines. The fast deceleration of the liquid results in high pressure surges upstream the valve, thus the kinetic energy is transformed into the potential energy, which leads to the temporary pressure increases. This phenomenon is called water hammer. The intensity of water hammer effects will depend upon the rate of change in the velocity or momentum. Generally water hammer can occur in any thermal-hydraulic systems and it is extremely dangerous for the thermal-hydraulic system since, if the pressure induced exceeds the pressure range of a pipe given by the manufacturer, it can lead to the failure of the pipeline integrity. Due to its potential for damage of pipes, water hammer has been a subject of study since the middle of the nineteenth century. Many theoretical and experimental investigations were performed. The experimental investigation of the water hammer tests performed at Fraunhofer Institute for Environmental, Safety and Energy Technology (UMSICHT) [1] and Cold Water Hammer experiment performed by Forschungszentrum Rossendorf (CWHTF) [2] should be mentioned. The UMSICHT facility in Oberhausen was modified in order to simulate a piping system and associated supports that are typical for a nuclear power plant [3]. The Cold water hammer experiment is interesting and instructive because it covers a wide spectrum of particularities. One of them is sub-cooled water interaction with condensing steam at the closed end of the vertical pipe at room temperature and corresponding saturation pressure [4]. In the paper, the capabilities of RELAP5 code to correctly represent the water hammer phenomenon are presented. Paper presents the comparison of RELAP5 calculated and measured at UMSICHT and CWHTF test facilities pressure transient values after the fast closure (opening) of valves. The analyses of rarefaction wave travels inside the pipe and condensation of vapour bubbles in the liquid column

  12. Numerical simulations of hydrodynamic instabilities: perturbation codes Pansy, Perle, and 2D code Chic applied to a realistic LIL target

    Energy Technology Data Exchange (ETDEWEB)

    Hallo, L.; Olazabal-Loume, M.; Maire, P.H.; Breil, J.; Schurtz, G. [CELIA, 33 - Talence (France); Morse, R.L. [Arizona Univ., Dept. of Nuclear Engineering, Tucson (United States)

    2006-06-15

    This paper deals with ablation front instabilities simulations in the context of direct drive inertial confinement fusion. A simplified deuterium-tritium target, representative of realistic target on LIL (laser integration line at Megajoule laser facility) is considered. We describe here two numerical approaches: the linear perturbation method using the perturbation codes Perle (planar) and Pansy (spherical) and the direct simulation method using our bi-dimensional hydrodynamic code Chic. Our work shows a good behaviour of all methods even for large wavenumbers during the acceleration phase of the ablation front. We also point out a good agreement between model and numerical predictions at ablation front during the shock wave transit.

  13. Comparison of physics model for 600 MeV protons 290 MeV·{sup n-}1 oxygen ions on carbon in MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Arim; Kim, Dong Hyun; Jung, Nam Suk; Oh, Joo Hee [Pohang Accelerator Laboratory, POSTECH, Pohang (Korea, Republic of); Oranj, Leila Mokhtari [Pohang University of Science and Technology, Pohang (Korea, Republic of)

    2016-06-15

    With the increase in the number of particle accelerator facilities under either operation or construction, the accurate calculation using Monte Carlo codes become more important in the shielding design and radiation safety evaluation of accelerator facilities. The calculations with different physics models were applied in both of cases: using only physics model and using the mix and match method of MCNPX code. The issued conditions were the interactions of 600 MeV proton and 290 MeV·{sup n-}1 oxygen with a carbon target. Both of cross-section libraries, JENDL High Energy File 2007 (JENDL/HE-2007) and LA150, were tested in this calculation. In the case of oxygen ion interactions, the calculation results using LAQGSM physics model and JENDL/HE-2007 library were compared with D. Satoh's experimental data. Other Monte Carlo calculations using PHITS and FLUKA codes were also carried out for further benchmarking study. It was clearly found that the physics models, especially intra-nuclear cascade model, gave a great effect to determine proton-induced secondary neutron spectrum in MCNPX code. The variety of physics models related to heavy ion interactions did not make big difference on the secondary particle productions. The variations of secondary neutron spectra and particle transports depending on various physics models in MCNPX code were studied and the result of this study can be used for the shielding design and radiation safety evaluation.

  14. Computer codes for simulating atomic-displacement cascades in solids subject to irradiation

    International Nuclear Information System (INIS)

    Asaoka, Takumi; Taji, Yukichi; Tsutsui, Tsuneo; Nakagawa, Masayuki; Nishida, Takahiko

    1979-03-01

    In order to study atomic displacement cascades originating from primary knock-on atoms in solids subject to incident radiation, the simulation code CASCADE/CLUSTER is adapted for use on FACOM/230-75 computer system. In addition, the code is modified so as to plot the defect patterns in crystalline solids. As other simulation code of the cascade process, MARLOWE is also available for use on the FACOM system. To deal with the thermal annealing of point defects produced in the cascade process, the code DAIQUIRI developed originally for body-centered cubic crystals is modified to be applicable also for face-centered cubic lattices. By combining CASCADE/CLUSTER and DAIQUIRI, we then prepared a computer code system CASCSRB to deal with heavy irradiation or saturation damage state of solids at normal temperature. Furthermore, a code system for the simulation of heavy irradiations CASCMARL is available, in which MARLOWE code is substituted for CASCADE in the CASCSRB system. (author)

  15. Adaptive mesh simulations of astrophysical detonations using the ASCI flash code

    Science.gov (United States)

    Fryxell, B.; Calder, A. C.; Dursi, L. J.; Lamb, D. Q.; MacNeice, P.; Olson, K.; Ricker, P.; Rosner, R.; Timmes, F. X.; Truran, J. W.; Tufo, H. M.; Zingale, M.

    2001-08-01

    The Flash code was developed at the University of Chicago as part of the Department of Energy's Accelerated Strategic Computing Initiative (ASCI). The code was designed specifically to simulate thermonuclear flashes in compact stars (white dwarfs and neutron stars). This paper will give a brief introduction to the astrophysics problems we wish to address, followed by a description of the current version of the Flash code. Finally, we discuss two simulations of astrophysical detonations that we have carried out with the code. The first is of a helium detonation in an X-ray burst. The other simulation models a carbon detonation in a Type Ia supernova explosion. .

  16. SITA version 0. A simulation and code testing assistant for TOUGH2 and MARNIE

    Energy Technology Data Exchange (ETDEWEB)

    Seher, Holger; Navarro, Martin

    2016-06-15

    High quality standards have to be met by those numerical codes that are applied in long-term safety assessments for deep geological repositories for radioactive waste. The software environment SITA (''a simulation and code testing assistant for TOUGH2 and MARNIE'') has been developed by GRS in order to perform automated regression testing for the flow and transport simulators TOUGH2 and MARNIE. GRS uses the codes TOUGH2 and MARNIE in order to assess the performance of deep geological repositories for radioactive waste. With SITA, simulation results of TOUGH2 and MARNIE can be compared to analytical solutions and simulations results of other code versions. SITA uses data interfaces to operate with codes whose input and output depends on the code version. The present report is part of a wider GRS programme to assure and improve the quality of TOUGH2 and MARNIE. It addresses users as well as administrators of SITA.

  17. Development of 2D particle-in-cell code to simulate high current, low ...

    Indian Academy of Sciences (India)

    Abstract. A code for 2D space-charge dominated beam dynamics study in beam trans- port lines is developed. The code is used for particle-in-cell (PIC) simulation of z-uniform beam in a channel containing solenoids and drift space. It can also simulate a transport line where quadrupoles are used for focusing the beam.

  18. TreePM: A Code for Cosmological N-Body Simulations

    Indian Academy of Sciences (India)

    We describe the TreePM method for carrying out large N-Body simulations to study formation and evolution of the large scale structure in the Universe. This method is a combination of Barnes and Hut tree code and Particle-Mesh code. It combines the automatic inclusion of periodic boundary conditions of PM simulations ...

  19. Parallel of semi-empirical results simulated by MCNP of X-ray spectra with a semiconductor

    International Nuclear Information System (INIS)

    Santos, L.R.; Vivolo, V.; Potiens, M.P.A.; Navarro, M.V.T.; Santos, W.S.

    2016-01-01

    The aim of this study was to use the MCNPX radiation transport code to simulate X-ray spectra generated by a constant voltage system in a CdTe semiconductor detector. As part of the validation process, we obtained a series of experimental spectra. Comparatively, in all cases there is a good correlation between the two spectra. There were no statistically significant differences between the experimental results with the simulated. (author)

  20. A Framework for Retargetable Code Generation using Simulated Annealing

    NARCIS (Netherlands)

    Visser, B.S.

    2000-01-01

    embedded systems. Retargetable code generation is a co-designing method to map a high-level software description onto a variety of hardware architectures without the need to rewrite a compiler. Highly efficient code generation is required to meet, for example, timing, area and low-power constraints.

  1. Computer codes in particle transport physics

    International Nuclear Information System (INIS)

    Pesic, M.

    2004-01-01

    Simulation of transport and interaction of various particles in complex media and wide energy range (from 1 MeV up to 1 TeV) is very complicated problem that requires valid model of a real process in nature and appropriate solving tool - computer code and data library. A brief overview of computer codes based on Monte Carlo techniques for simulation of transport and interaction of hadrons and ions in wide energy range in three dimensional (3D) geometry is shown. Firstly, a short attention is paid to underline the approach to the solution of the problem - process in nature - by selection of the appropriate 3D model and corresponding tools - computer codes and cross sections data libraries. Process of data collection and evaluation from experimental measurements and theoretical approach to establishing reliable libraries of evaluated cross sections data is Ion g, difficult and not straightforward activity. For this reason, world reference data centers and specialized ones are acknowledged, together with the currently available, state of art evaluated nuclear data libraries, as the ENDF/B-VI, JEF, JENDL, CENDL, BROND, etc. Codes for experimental and theoretical data evaluations (e.g., SAMMY and GNASH) together with the codes for data processing (e.g., NJOY, PREPRO and GRUCON) are briefly described. Examples of data evaluation and data processing to generate computer usable data libraries are shown. Among numerous and various computer codes developed in transport physics of particles, the most general ones are described only: MCNPX, FLUKA and SHIELD. A short overview of basic application of these codes, physical models implemented with their limitations, energy ranges of particles and types of interactions, is given. General information about the codes covers also programming language, operation system, calculation speed and the code availability. An example of increasing computation speed of running MCNPX code using a MPI cluster compared to the code sequential option

  2. Sensibility analysis of fuel depletion using different nuclear fuel depletion codes

    Energy Technology Data Exchange (ETDEWEB)

    Martins, F.; Velasquez, C.E.; Castro, V.F.; Pereira, C.; Silva, C. A. Mello da, E-mail: felipmartins94@gmail.com, E-mail: carlosvelcab@hotmail.com, E-mail: victorfariascastro@gmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: clarysson@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    Nowadays, the utilization of different nuclear codes to perform the depletion and criticality calculations has been used to simulated nuclear reactors problems. Therefore, the goal is to analyze the sensibility of the fuel depletion of a PWR assembly using three different nuclear fuel depletion codes. The burnup calculations are performed using the codes MCNP5/ORIGEN2.1 (MONTEBURNS), KENO-VI/ORIGEN-S (TRITONSCALE6.0) and MCNPX (MCNPX/CINDER90). Each nuclear code performs the burnup using different depletion codes. Each depletion code works with collapsed energies from a master library in 1, 3 and 63 groups, respectively. Besides, each code uses different ways to obtain neutron flux that influences the depletions calculation. The results present a comparison of the neutronic parameters and isotopes composition such as criticality and nuclides build-up, the deviation in results are going to be assigned to features of the depletion code in use, such as the different radioactive decay internal libraries and the numerical method involved in solving the coupled differential depletion equations. It is also seen that the longer the period is and the more time steps are chosen, the larger the deviation become. (author)

  3. SPIDERMAN: Fast code to simulate secondary transits and phase curves

    Science.gov (United States)

    Louden, Tom; Kreidberg, Laura

    2017-11-01

    SPIDERMAN calculates exoplanet phase curves and secondary eclipses with arbitrary surface brightness distributions in two dimensions. The code uses a geometrical algorithm to solve exactly the area of sections of the disc of the planet that are occulted by the star. Approximately 1000 models can be generated per second in typical use, which makes making Markov Chain Monte Carlo analyses practicable. The code is modular and allows comparison of the effect of multiple different brightness distributions for a dataset.

  4. ARC Code TI: Multi-Fidelity Simulator (MFSim)

    Data.gov (United States)

    National Aeronautics and Space Administration — Multi-Fidelity Simulator, MFSim is a pluggable framework for creating an air traffic flow simulator at multiple levels of fidelity. The framework is designed to...

  5. Sensitivity analysis of the titan hybrid deterministic transport code for SPECT simulation

    International Nuclear Information System (INIS)

    Royston, Katherine K.; Haghighat, Alireza

    2011-01-01

    Single photon emission computed tomography (SPECT) has been traditionally simulated using Monte Carlo methods. The TITAN code is a hybrid deterministic transport code that has recently been applied to the simulation of a SPECT myocardial perfusion study. For modeling SPECT, the TITAN code uses a discrete ordinates method in the phantom region and a combined simplified ray-tracing algorithm with a fictitious angular quadrature technique to simulate the collimator and generate projection images. In this paper, we compare the results of an experiment with a physical phantom with predictions from the MCNP5 and TITAN codes. While the results of the two codes are in good agreement, they differ from the experimental data by ∼ 21%. In order to understand these large differences, we conduct a sensitivity study by examining the effect of different parameters including heart size, collimator position, collimator simulation parameter, and number of energy groups. (author)

  6. A new Scheme for ATLAS Trigger Simulation using Legacy Code

    CERN Document Server

    Galster, G; The ATLAS collaboration; Wiedenmann, W

    2014-01-01

    An accurate simulation of the trigger response is necessary for high quality data analyses. This poses a challenge. For event generation and simulated data reconstruction the latest software is used to be in best agreement with the reconstructed data. Contrary the trigger response simulation needs to be in agreement with when the data was taken. The approach we follow is to use trigger software and conditions data that matches the simulated data-taking period potentially dating many years back. Having a strategy for running old software in a modern environment thus becomes essential when data simulated for past years start to present a sizable fraction of the total.\

  7. Program Code Generator for Cardiac Electrophysiology Simulation with Automatic PDE Boundary Condition Handling.

    Directory of Open Access Journals (Sweden)

    Florencio Rusty Punzalan

    Full Text Available Clinical and experimental studies involving human hearts can have certain limitations. Methods such as computer simulations can be an important alternative or supplemental tool. Physiological simulation at the tissue or organ level typically involves the handling of partial differential equations (PDEs. Boundary conditions and distributed parameters, such as those used in pharmacokinetics simulation, add to the complexity of the PDE solution. These factors can tailor PDE solutions and their corresponding program code to specific problems. Boundary condition and parameter changes in the customized code are usually prone to errors and time-consuming. We propose a general approach for handling PDEs and boundary conditions in computational models using a replacement scheme for discretization. This study is an extension of a program generator that we introduced in a previous publication. The program generator can generate code for multi-cell simulations of cardiac electrophysiology. Improvements to the system allow it to handle simultaneous equations in the biological function model as well as implicit PDE numerical schemes. The replacement scheme involves substituting all partial differential terms with numerical solution equations. Once the model and boundary equations are discretized with the numerical solution scheme, instances of the equations are generated to undergo dependency analysis. The result of the dependency analysis is then used to generate the program code. The resulting program code are in Java or C programming language. To validate the automatic handling of boundary conditions in the program code generator, we generated simulation code using the FHN, Luo-Rudy 1, and Hund-Rudy cell models and run cell-to-cell coupling and action potential propagation simulations. One of the simulations is based on a published experiment and simulation results are compared with the experimental data. We conclude that the proposed program code

  8. Tech-X Corporation releases simulation code for solving complex problems in plasma physics : VORPAL code provides a robust environment for simulating plasma processes in high-energy physics, IC fabrications and material processing applications

    CERN Multimedia

    2005-01-01

    Tech-X Corporation releases simulation code for solving complex problems in plasma physics : VORPAL code provides a robust environment for simulating plasma processes in high-energy physics, IC fabrications and material processing applications

  9. The FLUKA code: An accurate simulation tool for particle therapy

    CERN Document Server

    Battistoni, Giuseppe; Böhlen, Till T; Cerutti, Francesco; Chin, Mary Pik Wai; Dos Santos Augusto, Ricardo M; Ferrari, Alfredo; Garcia Ortega, Pablo; Kozlowska, Wioletta S; Magro, Giuseppe; Mairani, Andrea; Parodi, Katia; Sala, Paola R; Schoofs, Philippe; Tessonnier, Thomas; Vlachoudis, Vasilis

    2016-01-01

    Monte Carlo (MC) codes are increasingly spreading in the hadrontherapy community due to their detailed description of radiation transport and interaction with matter. The suitability of a MC code for application to hadrontherapy demands accurate and reliable physical models capable of handling all components of the expected radiation field. This becomes extremely important for correctly performing not only physical but also biologically-based dose calculations, especially in cases where ions heavier than protons are involved. In addition, accurate prediction of emerging secondary radiation is of utmost importance in innovative areas of research aiming at in-vivo treatment verification. This contribution will address the recent developments of the FLUKA MC code and its practical applications in this field. Refinements of the FLUKA nuclear models in the therapeutic energy interval lead to an improved description of the mixed radiation field as shown in the presented benchmarks against experimental data with bot...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-15

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

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

    International Nuclear Information System (INIS)

    Li, Yankai; Lin, Meng; Yang, Yanhua

    2016-01-01

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

  12. FARO and KROTOS code simulation and analysis at JRC Ispra

    Energy Technology Data Exchange (ETDEWEB)

    Annunziato, A.; Yerkess, A.; Addabbo, C. [European Commission-Joint Research Centre, Inst. for Systems, Informatics and Safety, 21020 Ispra (Italy)

    1998-01-01

    The paper summarizes relevant results from the pre and post test calculations of fuel coolant interaction and quenching tests performed in the FARO and KROTOS test facilities. The main analytical tools adopted at JRC Ispra are the COMETA and the TEXAS codes. COMETA pre and post test calculations of FARO Test L-20 as well as an application of the code to KROTOS test facility are presented. The analysis provides the need to account for H{sub 2} generation models into the pre-mixing calculations. In addition salient results from the application of TEXAS to FARO and KROTOS tests are shown. (author)

  13. Effect of the multiple scattering of electrons in Monte Carlo simulation of LINACS

    International Nuclear Information System (INIS)

    Vilches, Manuel; Garcia-Pareja, Salvador; Guerrero, Rafael; Anguiano, Marta; Lallena, Antonio M.

    2008-01-01

    Results obtained from Monte Carlo simulations of the transport of electrons in thin slabs of dense material media and air slabs with different widths are analyzed. Various general purpose Monte Carlo codes have been used: PENELOPE, GEANT3, GEANT4, EGSnrc, MCNPX. Non-negligible differences between the angular and radial distributions after the slabs have been found. The effects of these differences on the depth doses measured in water are also discussed

  14. LOLA SYSTEM: A code block for nodal PWR simulation. Part. I - Simula-3 Code

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J. M.; Ahnert, C.; Gomez Santamaria, J.; Rodriguez Olabarria, I.

    1985-07-01

    Description of the theory and users manual of the SIMULA-3 code, which is part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. SIMULA-3 is the main module of the system, it uses a modified nodal theory, with interface leakages equivalent to the diffusion theory. (Author) 4 refs.

  15. LOLA SYSTEM: A code block for nodal PWR simulation. Part. I - Simula-3 Code

    International Nuclear Information System (INIS)

    Aragones, J. M.; Ahnert, C.; Gomez Santamaria, J.; Rodriguez Olabarria, I.

    1985-01-01

    Description of the theory and users manual of the SIMULA-3 code, which is part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. SIMULA-3 is the main module of the system, it uses a modified nodal theory, with interface leakages equivalent to the diffusion theory. (Author) 4 refs

  16. Simulations of X-ray synchrotron beams using the EGS4 code system in medical applications

    International Nuclear Information System (INIS)

    Orion, I.; Henn, A.; Sagi, I.; Dilmanian, F.A.; Pena, L.; Rosenfeld, A.B.

    2001-01-01

    X-ray synchrotron beams are commonly used in biological and medical research. The availability of intense, polarized low-energy photons from the synchrotron beams provides a high dose transfer to biological materials. The EGS4 code system, which includes the photoelectron angular distribution, electron motion inside a magnetic field, and the LSCAT package, found to be the appropriate Monte Carlo code for synchrotron-produced X-ray simulations. The LSCAT package was developed in 1995 for the EGS4 code to contain the routines to simulate the linear polarization, the bound Compton, and the incoherent scattering functions. Three medical applications were demonstrated using the EGS4 Monte Carlo code as a proficient simulation code system for the synchrotron low-energy X-ray source. (orig.)

  17. HELIOS/DRAGON/NESTLE codes' simulation of void reactivity in a CANDU core

    International Nuclear Information System (INIS)

    Sarsour, H.N.; Rahnema, F.; Mosher, S.; Turinsky, P.J.; Serghiuta, D.; Marleau, G.; Courau, T.

    2002-01-01

    This paper presents results of simulation of void reactivity in a CANDU core using the NESTLE core simulator, cross sections from the HELIOS lattice physics code in conjunction with incremental cross sections from the DRAGON lattice physics code. First, a sub-region of a CANDU6 core is modeled using the NESTLE core simulator and predictions are contrasted with predictions by the MCNP Monte Carlo simulation code utilizing a continuous energy model. In addition, whole core modeling results are presented using the NESTLE finite difference method (FDM), NESTLE nodal method (NM) without assembly discontinuity factors (ADF), and NESTLE NM with ADF. The work presented in this paper has been performed as part of a project sponsored by the Canadian Nuclear Safety Commission (CNSC). The purpose of the project was to gather information and assess the accuracy of best estimate methods using calculational methods and codes developed independently from the CANDU industry. (author)

  18. SUPERBOX: Particle-multi-mesh code to simulate galaxies

    Science.gov (United States)

    Fellhauer, M.; Kroupa, P.; Baumgardt, H.; Bien, R.; Boily, C. M.; Spurzem, R.; Wassmer, N.

    2015-07-01

    SUPERBOX is a particle-mesh code that uses moving sub-grids to track and resolve high-density peaks in the particle distribution and a nearest grid point force-calculation scheme based on the second derivatives of the potential. The code implements a fast low-storage FFT-algorithm and allows a highly resolved treatment of interactions in clusters of galaxies, such as high-velocity encounters between elliptical galaxies and the tidal disruption of dwarf galaxies, as sub-grids follow the trajectories of individual galaxies. SUPERBOX is efficient in that the computational overhead is kept as slim as possible and is also memory efficient since it uses only one set of grids to treat galaxies in succession.

  19. pTSC: Data file editing for the Tokamak Simulation Code

    International Nuclear Information System (INIS)

    Meiss, J.D.

    1987-09-01

    The code pTSC is an editor for the data files needed to run the Princeton Tokamak Simulation Code (TSC). pTSC utilizes the Macintosh interface to create a graphical environment for entering the data. As most of the data to run TSC consists of conductor positions, the graphical interface is especially appropriate

  20. pTSC: Data file editing for the Tokamak Simulation Code

    Energy Technology Data Exchange (ETDEWEB)

    Meiss, J.D.

    1987-09-01

    The code pTSC is an editor for the data files needed to run the Princeton Tokamak Simulation Code (TSC). pTSC utilizes the Macintosh interface to create a graphical environment for entering the data. As most of the data to run TSC consists of conductor positions, the graphical interface is especially appropriate.

  1. Premar-2: a Monte Carlo code for radiative transport simulation in atmospheric environments

    International Nuclear Information System (INIS)

    Cupini, E.

    1999-01-01

    The peculiarities of the PREMAR-2 code, aimed at radiation transport Monte Carlo simulation in atmospheric environments in the infrared-ultraviolet frequency range, are described. With respect to the previously developed PREMAR code, besides plane multilayers, spherical multilayers and finite sequences of vertical layers, each one with its own atmospheric behaviour, are foreseen in the new code, together with the refraction phenomenon, so that long range, highly slanted paths can now be more faithfully taken into account. A zenithal angular dependence of the albedo coefficient has moreover been introduced. Lidar systems, with spatially independent source and telescope, are allowed again to be simulated, and, in this latest version of the code, sensitivity analyses to be performed. According to this last feasibility, consequences on radiation transport of small perturbations in physical components of the atmospheric environment may be analyze and the related effects on searched results estimated. The availability of a library of physical data (reaction coefficients, phase functions and refraction indexes) is required by the code, providing the essential features of the environment of interest needed of the Monte Carlo simulation. Variance reducing techniques have been enhanced in the Premar-2 code, by introducing, for instance, a local forced collision technique, especially apt to be used in Lidar system simulations. Encouraging comparisons between code and experimental results carried out at the Brasimone Centre of ENEA, have so far been obtained, even if further checks of the code are to be performed [it

  2. ICOOL: A Simulation Code for Ionization Cooling of Muon Beams

    International Nuclear Information System (INIS)

    Fernow, R. C.

    1999-01-01

    Current ideas [1,2] for designing a high luminosity muon collider require significant cooling of the phase space of the muon beams. The only known method that can cool the beams in a time comparable to the muon lifetime is ionization cooling [3,4]. This method requires directing the particles in the beam at a large angle through a low Z absorber material in a strong focusing magnetic channel and then restoring the longitudinal momentum with an rf cavity. We have developed a new 3-D tracking code ICOOL for examining possible configurations for muon cooling. A cooling system is described in terms of a series of longitudinal regions with associated material and field properties. The tracking takes place in a coordinate system that follows a reference orbit through the system. The code takes into account decays and interactions of ∼50-500 MeV/c muons in matter. Material geometry regions include cylinders and wedges. A number of analytic models are provided for describing the field configurations. Simple diagnostics are built into the code, including calculation of emittances and correlations, longitudinal traces, histograms and scatter plots. A number of auxiliary files can be generated for post-processing analysis by the user

  3. Simulations of the Ondine experiment with the solitude code; Simulations de l`experience Ondine a l`aide du code solitude

    Energy Technology Data Exchange (ETDEWEB)

    Gouard, P.; Gardelle, J.

    1992-11-01

    A new version of the SOLITUDE code, including an axial magnetic field and a cylindrical waveguide, is presented. It allows to simulate the ONDINE experiment at CESTA and to study the effects and behaviour of an actual electron beam in a Free Electron Laser amplifier experiment.

  4. Benchmarking and scaling studies of pseudospectral code Tarang for turbulence simulations

    KAUST Repository

    VERMA, MAHENDRA K

    2013-09-21

    Tarang is a general-purpose pseudospectral parallel code for simulating flows involving fluids, magnetohydrodynamics, and Rayleigh–Bénard convection in turbulence and instability regimes. In this paper we present code validation and benchmarking results of Tarang. We performed our simulations on 10243, 20483, and 40963 grids using the HPC system of IIT Kanpur and Shaheen of KAUST. We observe good ‘weak’ and ‘strong’ scaling for Tarang on these systems.

  5. ARC Code TI: Mission Simulation ToolKit (MST)

    Data.gov (United States)

    National Aeronautics and Space Administration — The MST is a simulation framework, supporting the development of autonomy technology for planetary exploration vehicles. The MST provides a software test bed which...

  6. ATLAS trigger simulation with legacy code using virtualization techniques

    CERN Document Server

    Galster, G; The ATLAS collaboration; Wiedenmann, W

    2014-01-01

    Abstract. Several scenarios, both present and future, requires re-simulation of the trigger response in ATLAS. While software for the detector response simulation and event reconstruction is allowed to change and improve, the trigger response simulation has to reflect the conditions at which data was taken. This poses a massive maintenance and data preservation problem. Several strategies have been considered and a proof-of-concept model using CernVM has been developed. While the virtualization with CernVM elegantly solves several aspects of the data preservation problem, the low maturity for contextualization as well as data format incompatibilities in the currently used data format introduces new challenges. In this proceeding these challenges, their current solutions and the proof of concept model for precise trigger simulation are discussed.

  7. On the adequacy of numerical codes for the simulation of vapour cloud explosions

    International Nuclear Information System (INIS)

    Wingerden, G.J.M.v.; Berg, A.C.v.d.

    1984-01-01

    Three spherically symmetric blast simulation codes have been evaluated: a low-flame-speed model (Piston model) and two gasdynamic blast simulation codes (BLAST and CLOUD). Self-similar flow fields in front of constant velocity flames and large- and small-scale spherically symmetric explosions experiments were simulated. The Piston model can be used for the simulation of spherically symmetric explosions at flame speeds -1 whereas BLAST and CLOUD are adequate for flame speeds exceeding 100 ms -1 . An adapted Piston code has been investigated with respect to the capability of simulating blast due to explosions of pancake-shaped clouds. Comparison to an acoustic approach showed that the Piston model can be regarded as an acoustic model with the possibility of handling every imaginable flame path. The research was part of the indirect action research programme on LWR Safety of the Commission of the European Communities. (project 12B, contract 008 SRN)

  8. Lacan - a global simulation code for laser isotope separation

    International Nuclear Information System (INIS)

    Goldstein, S.; Quaegebeur, J.P.

    1990-01-01

    Dimensioning a Laser Isotope Separation (LIS) plant means calculating the values of a large number of parameters in order to optimize some objective function. In such algorithms the calculation of the objective function must be repeated thousands of times, therefore each elementary calculation must consume little time. LACAN uses simple models to describe the elementary physical processes: evaporation, vapour expansion, interaction between photons and atoms, ion extraction etc ... These simple models are derived from refined modeling codes or are empirical. As an example the optimization of the separative work of an uranium facility is discussed

  9. The status of simulation codes for extraction process using mixer-settler

    International Nuclear Information System (INIS)

    Byeon, Kee Hoh; Lee, Eil Hee; Kwon, Seong Gil; Kim, Kwang Wook; Yang, Han Beom; Chung, Dong Yong; Lim, Jae Kwan; Shin, Hyun Kyoo; Kim, Soo Ho

    1999-10-01

    We have studied and analyzed the mixer-settler simulation codes such as three kinds of SEPHIS series, PUBG, and EXTRA.M, which is the most recently developed code. All of these are sufficiently satisfactory codes in the fields of process/device modeling, but it is necessary to formulate the accurate distribution data and chemical reaction mechanism for the aspect of accuracy and reliability. In the aspect of application to be the group separation process, the mixer-settler model of these codes have no problems, but the accumulation and formulation of partitioning and reaction equilibrium data of chemical elements used in group separation process is very important. (author)

  10. The TOUGH codes - a family of simulation tools for multiphase flowand transport processes in permeable media

    Energy Technology Data Exchange (ETDEWEB)

    Pruess, Karsten

    2003-08-08

    Numerical simulation has become a widely practiced andaccepted technique for studying flow and transport processes in thevadose zone and other subsurface flow systems. This article discusses asuite of codes, developed primarily at Lawrence Berkeley NationalLaboratory (LBNL), with the capability to model multiphase flows withphase change. We summarize history and goals in the development of theTOUGH codes, and present the governing equations for multiphase,multicomponent flow. Special emphasis is given to space discretization bymeans of integral finite differences (IFD). Issues of code implementationand architecture are addressed, as well as code applications,maintenance, and future developments.

  11. ASAS: Computational code for Analysis and Simulation of Atomic Spectra

    Directory of Open Access Journals (Sweden)

    Jhonatha R. dos Santos

    2017-01-01

    Full Text Available The laser isotopic separation process is based on the selective photoionization principle and, because of this, it is necessary to know the absorption spectrum of the desired atom. Computational resource has become indispensable for the planning of experiments and analysis of the acquired data. The ASAS (Analysis and Simulation of Atomic Spectra software presented here is a helpful tool to be used in studies involving atomic spectroscopy. The input for the simulations is friendly and essentially needs a database containing the energy levels and spectral lines of the atoms subjected to be studied.

  12. Monte Carlo simulation in UWB1 depletion code

    International Nuclear Information System (INIS)

    Lovecky, M.; Prehradny, J.; Jirickova, J.; Skoda, R.

    2015-01-01

    U W B 1 depletion code is being developed as a fast computational tool for the study of burnable absorbers in the University of West Bohemia in Pilsen, Czech Republic. In order to achieve higher precision, the newly developed code was extended by adding a Monte Carlo solver. Research of fuel depletion aims at development and introduction of advanced types of burnable absorbers in nuclear fuel. Burnable absorbers (BA) allow the compensation of the initial reactivity excess of nuclear fuel and result in an increase of fuel cycles lengths with higher enriched fuels. The paper describes the depletion calculations of VVER nuclear fuel doped with rare earth oxides as burnable absorber based on performed depletion calculations, rare earth oxides are divided into two equally numerous groups, suitable burnable absorbers and poisoning absorbers. According to residual poisoning and BA reactivity worth, rare earth oxides marked as suitable burnable absorbers are Nd, Sm, Eu, Gd, Dy, Ho and Er, while poisoning absorbers include Sc, La, Lu, Y, Ce, Pr and Tb. The presentation slides have been added to the article

  13. Evaluation of the Aleph PIC Code on Benchmark Simulations

    Science.gov (United States)

    Boerner, Jeremiah; Pacheco, Jose; Grillet, Anne

    2016-09-01

    Aleph is a massively parallel, 3D unstructured mesh, Particle-in-Cell (PIC) code, developed to model low temperature plasma applications. In order to verify and validate performance, Aleph is benchmarked against a series of canonical problems to demonstrate statistical indistinguishability in the results. Here, a series of four problems is studied: Couette flows over a range of Knudsen number, sheath formation in an undriven plasma, the two-stream instability, and a capacitive discharge. These problems respectively exercise collisional processes, particle motion in electrostatic fields, electrostatic field solves coupled to particle motion, and a fully coupled reacting plasma. Favorable comparison with accepted results establishes confidence in Aleph's capability and accuracy as a general purpose PIC code. Finally, Aleph is used to investigate the sensitivity of a triggered vacuum gap switch to the particle injection conditions associated with arc breakdown at the trigger. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000.

  14. Simulation of loss of feedwater transient of MASLWR test facility by MARS-KS code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Juyeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is being current developed domestically also adopts helical coil steam generator, KINS has joined this ICSP to evaluate performance of domestic regulatory audit thermal-hydraulic code (MARS-KS code) in various respects including wall-to-fluid heat transfer model modification implemented in the code by independent international experiment database. In the ICSP, two types of transient experiments have been focused and they are loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels (SP-3). In the present study, KINS simulation results by the MARS-KS code (KS-002 version) for the SP-2 experiment are presented in detail and conclusions on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the loss of feedwater transient of the MASLWR test facility. Steady state run shows helical coil specific heat transfer models implemented in the code is reasonable. However, through the transient run, it is also found that three-dimensional effect within the HPC and axial conduction effect through the HTP are not well reproduced by the code.

  15. SIMIFR: A code to simulate material movement in the Integral Fast Reactor

    International Nuclear Information System (INIS)

    White, A.M.; Orechwa, Yuri.

    1991-01-01

    The SIMIFR code has been written to simulate the movement of material through a process. This code can be used to investigate inventory differences in material balances, assist in process design, and to produce operational scheduling. The particular process that is of concern to the authors is that centered around Argonne National Laboratory's Integral Fast Reactor. This is a process which involves the irradiation of fissile material for power production, and the recycling of the irradiated reactor fuel pins into fresh fuel elements. To adequately simulate this process it is necessary to allow for locations which can contain either discrete items or homogeneous mixtures. It is also necessary to allow for a very flexible process control algorithm. Further, the code must have the capability of transmuting isotopic compositions and computing internally the fraction of material from a process ending up in a given location. The SIMIFR code has been developed to perform all of these tasks. In addition to simulating the process, the code is capable of generating random measurement values and sampling errors for all locations, and of producing a restart deck so that terminated problems may be continued. In this paper the authors first familiarize the reader with the IFR fuel cycle. The different capabilities of the SIMIFR code are described. Finally, the simulation of the IFR fuel cycle using the SIMIFR code is discussed. 4 figs

  16. Applications of the lahet simulation code to relativistic heavy ion detectors

    Energy Technology Data Exchange (ETDEWEB)

    Waters, L.; Gavron, A. [Los Alamos National Lab., NM (United States)

    1991-12-31

    The Los Alamos High Energy Transport (LAHET) simulation code has been applied to test beam data from the lead/scintillator Participant Calorimeter of BNL AGS experiment E814. The LAHET code treats hadronic interactions with the LANL version of the Oak Ridge code HETC. LAHET has now been expanded to handle hadrons with kinetic energies greater than 5 GeV with the FLUKA code, while HETC is used exclusively below 2.0 GeV. FLUKA is phased in linearly between 2.0 and 5.0 GeV. Transport of electrons and photons is done with EGS4, and an interface to the Los Alamos HMCNP3B library based code is provided to analyze neutrons with kinetic energies less than 20 MeV. Excellent agreement is found between the test data and simulation, and results for 2.46 GeV/c protons and pions are illustrated in this article.

  17. Applications of the LAHET simulation code to relativistic heavy ion detectors

    International Nuclear Information System (INIS)

    Waters, L.S.; Gavron, A.

    1991-01-01

    The Los Alamos High Energy Transport (LAHET) simulation code has been applied to test beam data from the lead/scintillator Participant Calorimeter of BNL AGS experiment E814. The LAHET code treats hadronic interactions with the LANL version of the Oak Ridge code HETC. LAHET has now been expanded to handle hadrons with kinetic energies greater than 5 GeV with the FLUKA code, while HETC is used exclusively below 2.0 GeV. FLUKA is phased in linearly between 2.0 and 5.0 GeV. Transport of electrons and photons is done with EGS4, and an interface to the Los Alamos HMCNP3B library based code is provided to analyze neutrons with kinetic energies less than 20 MeV. Excellent agreement is found between the test data and simulation, and results for 2.46 GeV/c protons and pions are illustrated in this article

  18. The UPSF code: a metaprogramming-based high-performance automatically parallelized plasma simulation framework

    Science.gov (United States)

    Gao, Xiatian; Wang, Xiaogang; Jiang, Binhao

    2017-10-01

    UPSF (Universal Plasma Simulation Framework) is a new plasma simulation code designed for maximum flexibility by using edge-cutting techniques supported by C++17 standard. Through use of metaprogramming technique, UPSF provides arbitrary dimensional data structures and methods to support various kinds of plasma simulation models, like, Vlasov, particle in cell (PIC), fluid, Fokker-Planck, and their variants and hybrid methods. Through C++ metaprogramming technique, a single code can be used to arbitrary dimensional systems with no loss of performance. UPSF can also automatically parallelize the distributed data structure and accelerate matrix and tensor operations by BLAS. A three-dimensional particle in cell code is developed based on UPSF. Two test cases, Landau damping and Weibel instability for electrostatic and electromagnetic situation respectively, are presented to show the validation and performance of the UPSF code.

  19. On the Development of a Gridless Inflation Code for Parachute Simulations

    Energy Technology Data Exchange (ETDEWEB)

    STRICKLAND,JAMES H.; HOMICZ,GREGORY F.; GOSSLER,ALBERT A.; WOLFE,WALTER P.; PORTER,VICKI L.

    2000-08-29

    In this paper the authors present the current status of an unsteady 3D parachute simulation code which is being developed at Sandia National Laboratories under the Department of Energy's Accelerated Strategic Computing Initiative (ASCI). The Vortex Inflation PARachute code (VIPAR) which embodies this effort will eventually be able to perform complete numerical simulations of ribbon parachute deployment, inflation, and steady descent. At the present time they have a working serial version of the uncoupled fluids code which can simulate unsteady 3D incompressible flows around bluff bodies made up of triangular membrane elements. A parallel version of the code has just been completed which will allow one to compute flows over complex geometries utilizing several thousand processors on one of the new DOE teraFLOP computers.

  20. Parallelization of a beam dynamics code and first large scale radio frequency quadrupole simulations

    Directory of Open Access Journals (Sweden)

    J. Xu

    2007-01-01

    Full Text Available The design and operation support of hadron (proton and heavy-ion linear accelerators require substantial use of beam dynamics simulation tools. The beam dynamics code TRACK has been originally developed at Argonne National Laboratory (ANL to fulfill the special requirements of the rare isotope accelerator (RIA accelerator systems. From the beginning, the code has been developed to make it useful in the three stages of a linear accelerator project, namely, the design, commissioning, and operation of the machine. To realize this concept, the code has unique features such as end-to-end simulations from the ion source to the final beam destination and automatic procedures for tuning of a multiple charge state heavy-ion beam. The TRACK code has become a general beam dynamics code for hadron linacs and has found wide applications worldwide. Until recently, the code has remained serial except for a simple parallelization used for the simulation of multiple seeds to study the machine errors. To speed up computation, the TRACK Poisson solver has been parallelized. This paper discusses different parallel models for solving the Poisson equation with the primary goal to extend the scalability of the code onto 1024 and more processors of the new generation of supercomputers known as BlueGene (BG/L. Domain decomposition techniques have been adapted and incorporated into the parallel version of the TRACK code. To demonstrate the new capabilities of the parallelized TRACK code, the dynamics of a 45 mA proton beam represented by 10^{8} particles has been simulated through the 325 MHz radio frequency quadrupole and initial accelerator section of the proposed FNAL proton driver. The results show the benefits and advantages of large-scale parallel computing in beam dynamics simulations.

  1. An Advanced simulation Code for Modeling Inductive Output Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Thuc Bui; R. Lawrence Ives

    2012-04-27

    During the Phase I program, CCR completed several major building blocks for a 3D large signal, inductive output tube (IOT) code using modern computer language and programming techniques. These included a 3D, Helmholtz, time-harmonic, field solver with a fully functional graphical user interface (GUI), automeshing and adaptivity. Other building blocks included the improved electrostatic Poisson solver with temporal boundary conditions to provide temporal fields for the time-stepping particle pusher as well as the self electric field caused by time-varying space charge. The magnetostatic field solver was also updated to solve for the self magnetic field caused by time changing current density in the output cavity gap. The goal function to optimize an IOT cavity was also formulated, and the optimization methodologies were investigated.

  2. Traveling-wave-tube simulation; The IBC code

    Energy Technology Data Exchange (ETDEWEB)

    Morey, I.J.; Birdsall, C.K. (California Univ., Berkeley, CA (USA). Dept. of Electrical Engineering and Computer Sciences)

    1990-06-01

    A beam-circuit code is presented, to run interactively on fast PC's or workstations, for purposes of first-cut design of Traveling-Wave Tubes (TWT's) at small and large amplitudes. The new physics parts are the use of particle-in-cell methods to obtain the space-charge forces, and the following of the electron beam over the full length of the tube. The model is fully nonlinear and one-dimensional, with the transverse space-charge fields approximated by one mode. The slow-wave circuit is modeled by a transmission line. All variables are displayed continuously, such as the velocity displacement of all the particles (phase space), beam charge and current densities, space-charge field, circuit field, voltage and current, circuit power, and the location of the added loss. Some initial runs are presented.

  3. VERB code simulations using realistic whistler wave parameters

    Science.gov (United States)

    Drozdov, Alexander; Orlova, Ksenia; Shprits, Yuri

    The dynamics of the outer electron radiation belt is dramatically variable and can be described in terms of electron phase space density (PSD). The PSD variation can be modeled using the Fokker-Planck diffusion equation in terms of the radial distance, energy, and equatorial pitch angle. We present the results of the modelling performed using the Versatile Electron Radiation Belt (VERB) code. The pitch-angle, mixed, and energy diffusion coefficients are calculated using recent statistical studies of chorus and hiss waves. Wave amplitudes, wave normal angle distribution, and wave spectral density are parameterized as a functions of magnetic latitude, radial distance, and Kp-index. For our calculations we use model of plasma density that depends on the latitude. The obtained results are compared with observations from the MagEIS and REPT instruments on the Van Allen Probes spacecraft.

  4. Simulation of Two-group IATE models with EAGLE code

    International Nuclear Information System (INIS)

    Nguyen, V. T.; Bae, B. U.; Song, C. H.

    2011-01-01

    The two-group transport equation should be employed in order to describe correctly the interfacial area transport in various two phase flow regimes, especially at the bubbly-to-slug flow transition. This is because the differences in bubble sizes or shapes cause substantial differences in their transport mechanisms and interaction phenomena. The basic concept of two group interfacial area transport equations have been formulated and demonstrated for vertical gas-liquid bubbly-to-slug flow transition by Hibiki and his coworkers. More than twelve adjustable parameters need to be determined based on extensive experimental database. It should be noted that these parameters were adjusted only in one-dimensional approach by area averaged flow parameters in a vertical pipe under adiabatic and steady conditions. This obviously brings up the following experimental issue: how to adjust all these parameters as independently as possible by considering experiments where a single physical phenomenon is of importance. The vertical air-water loop (VAWL) has been used for investigating the transport phenomena of two-phase flow at Korea Atomic Energy Research Institute (KAERI). The data for local void fraction and interfacial area concentration are measured by using five-sensor conductivity probe method and classified into two groups, the small spherical bubble group and the cap/slug one. The initial bubble size, which has a big influence on the interaction mechanism between phases, was controlled. In the present work, two-group interfacial area transport equation (IATE) was implemented in the EAGLE code and assessed against VAWL data. The purpose of this study is to investigate the capability of coefficients derived by Hibiki in the two-group interfacial area transport equations with CFD code

  5. Code modernization and modularization of APEX and SWAT watershed simulation models

    Science.gov (United States)

    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...

  6. SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients

    International Nuclear Information System (INIS)

    Dias, A.F.V.

    1980-10-01

    Fuel rod behavior during a stationary and a transitory operation, is studied. A computer code aiming at simulating PWR type rods, was developed; however, it can be adapted for simulating other type of rods. A finite difference method was used. (E.G.) [pt

  7. Coding Instructions, Worksheets, and Keypunch Sheets for M.E.T.R.O.-APEX Simulation.

    Science.gov (United States)

    Michigan Univ., Ann Arbor. Environmental Simulation Lab.

    Compiled in this resource are coding instructions, worksheets, and keypunch sheets for use in the M.E.T.R.O.-APEX simulation, described in detail in documents ED 064 530 through ED 064 550. Air Pollution Exercise (APEX) is a computerized college and professional level "real world" simulation of a community with urban and rural problems, industrial…

  8. Numerical analysis of reflood simulation based on a mechanistic, best-estimate, approach by KREWET code

    International Nuclear Information System (INIS)

    Chun, Moon-Hyun; Jeong, Eun-Soo

    1983-01-01

    A new computer code entitled KREWET has been developed in an effort to improve the accuracy and applicability of the existing reflood heat transfer simulation computer code. Sample calculations for temperature histories and heat transfer coefficient are made using KREWET code and the results are compared with the predictions of REFLUX, QUEN1D, and the PWR-FLECHT data for various conditions. These show favourable agreement in terms of clad temperature versus time. For high flooding rates (5-15cm/sec) and high pressure (∼413 Kpa), reflood predictions are reasonably well predicted by KREWET code as well as with other codes. For low flooding rates (less than ∼4cm/sec) and low pressure (∼138Kpa), predictions show considerable error in evaluating the rewet position versus time. This observation is common to all the codes examined in the present work

  9. Simulation according to the New Romanian Civil Code

    Directory of Open Access Journals (Sweden)

    G. TIŢA-NICOLESCU

    2012-01-01

    Full Text Available Simulation arises when for two parties and for the same legal relationship there are two legal acts (more precisely two variants of the same legal act that have different contents, especially essential provisions. One of the two juridical acts is referred to as a public act or the apparent act (but, in fact, simulated, being stated as such by the parties and reflecting the ”official”( but false variant of the agreement between the parties, as it is, reached in front of the Notary, of the lawyer or a private signature act. The other mentioned act is the secret act (referred to as the secret agreement, which represents, in fact, the true agreement between the parties, but it is not included in an official act, being known only by the parties.

  10. Process monitoring and simulation code verification using interactive computer animation

    International Nuclear Information System (INIS)

    Curtis, J.N.; Beelman, R.J.; Schwieder, D.H.; Stewart, H.D.

    1984-01-01

    At the Idaho National Engineering Laboratory (INEL), EG and G Idaho, Inc., has developed techniques by which schematics, created for and displayed at color graphics terminals, can be driven by actual or calculated data. These input data cause changes to occur within the displayed schematic. This research is presently being done to develop a prototype to be used in nuclear power plant control rooms. Work stations have already been developed to analyze data that are produced during actual and simulated nuclear reactor experiments

  11. GAPD: a GPU-accelerated atom-based polychromatic diffraction simulation code.

    Science.gov (United States)

    E, J C; Wang, L; Chen, S; Zhang, Y Y; Luo, S N

    2018-03-01

    GAPD, a graphics-processing-unit (GPU)-accelerated atom-based polychromatic diffraction simulation code for direct, kinematics-based, simulations of X-ray/electron diffraction of large-scale atomic systems with mono-/polychromatic beams and arbitrary plane detector geometries, is presented. This code implements GPU parallel computation via both real- and reciprocal-space decompositions. With GAPD, direct simulations are performed of the reciprocal lattice node of ultralarge systems (∼5 billion atoms) and diffraction patterns of single-crystal and polycrystalline configurations with mono- and polychromatic X-ray beams (including synchrotron undulator sources), and validation, benchmark and application cases are presented.

  12. Experience gained in running the EPRI MMS code with an in-house simulation language

    International Nuclear Information System (INIS)

    Weber, D.S.

    1987-01-01

    The EPRI Modular Modeling System (MMS) code represents a collection of component models and a steam/water properties package. This code has undergone extensive verification and validation testing. Currently, the code requires a commercially available simulation language to run. The Philadelphia Electric Company (PECO) has been modeling power plant systems for over the past sixteen years. As a result, an extensive number of models have been developed. In addition, an extensive amount of experience has been developed and gained using an in-house simulation language. The objective of this study was to explore the possibility of developing an MMS pre-processor which would allow the use of the MMS package with other simulation languages such as the PECO in-house simulation language

  13. A Novel Technique for Running the NASA Legacy Code LAPIN Synchronously With Simulations Developed Using Simulink

    Science.gov (United States)

    Vrnak, Daniel R.; Stueber, Thomas J.; Le, Dzu K.

    2012-01-01

    This report presents a method for running a dynamic legacy inlet simulation in concert with another dynamic simulation that uses a graphical interface. The legacy code, NASA's LArge Perturbation INlet (LAPIN) model, was coded using the FORTRAN 77 (The Portland Group, Lake Oswego, OR) programming language to run in a command shell similar to other applications that used the Microsoft Disk Operating System (MS-DOS) (Microsoft Corporation, Redmond, WA). Simulink (MathWorks, Natick, MA) is a dynamic simulation that runs on a modern graphical operating system. The product of this work has both simulations, LAPIN and Simulink, running synchronously on the same computer with periodic data exchanges. Implementing the method described in this paper avoided extensive changes to the legacy code and preserved its basic operating procedure. This paper presents a novel method that promotes inter-task data communication between the synchronously running processes.

  14. MOCCA code for star cluster simulation: comparison with optical observations using COCOA

    Science.gov (United States)

    Askar, Abbas; Giersz, Mirek; Pych, Wojciech; Olech, Arkadiusz; Hypki, Arkadiusz

    2016-02-01

    We introduce and present preliminary results from COCOA (Cluster simulatiOn Comparison with ObservAtions) code for a star cluster after 12 Gyr of evolution simulated using the MOCCA code. The COCOA code is being developed to quickly compare results of numerical simulations of star clusters with observational data. We use COCOA to obtain parameters of the projected cluster model. For comparison, a FITS file of the projected cluster was provided to observers so that they could use their observational methods and techniques to obtain cluster parameters. The results show that the similarity of cluster parameters obtained through numerical simulations and observations depends significantly on the quality of observational data and photometric accuracy.

  15. SPACE CHARGE SIMULATION METHODS INCORPORATED IN SOME MULTI - PARTICLE TRACKING CODES AND THEIR RESULTS COMPARISON

    International Nuclear Information System (INIS)

    BEEBE - WANG, J.; LUCCIO, A.U.; D IMPERIO, N.; MACHIDA, S.

    2002-01-01

    Space charge in high intensity beams is an important issue in accelerator physics. Due to the complicity of the problems, the most effective way of investigating its effect is by computer simulations. In the resent years, many space charge simulation methods have been developed and incorporated in various 2D or 3D multi-particle-tracking codes. It has becoming necessary to benchmark these methods against each other, and against experimental results. As a part of global effort, we present our initial comparison of the space charge methods incorporated in simulation codes ORBIT++, ORBIT and SIMPSONS. In this paper, the methods included in these codes are overviewed. The simulation results are presented and compared. Finally, from this study, the advantages and disadvantages of each method are discussed

  16. Srna-Monte Carlo codes for proton transport simulation in combined and voxelized geometries

    International Nuclear Information System (INIS)

    Ilic, R.D.; Lalic, D.; Stankovic, S.J.

    2002-01-01

    This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D) dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtained through the PETRA and GEANT programs. The simulation of the proton beam characterization by means of the Multi-Layer Faraday Cup and spatial distribution of positron emitters obtained by our program indicate the imminent application of Monte Carlo techniques in clinical practice. (author)

  17. Benchmark test of drift-kinetic and gyrokinetic codes through neoclassical transport simulations

    International Nuclear Information System (INIS)

    Satake, S.; Sugama, H.; Watanabe, T.-H.; Idomura, Yasuhiro

    2009-09-01

    Two simulation codes that solve the drift-kinetic or gyrokinetic equation in toroidal plasmas are benchmarked by comparing the simulation results of neoclassical transport. The two codes are the drift-kinetic δf Monte Carlo code (FORTEC-3D) and the gyrokinetic full- f Vlasov code (GT5D), both of which solve radially-global, five-dimensional kinetic equation with including the linear Fokker-Planck collision operator. In a tokamak configuration, neoclassical radial heat flux and the force balance relation, which relates the parallel mean flow with radial electric field and temperature gradient, are compared between these two codes, and their results are also compared with the local neoclassical transport theory. It is found that the simulation results of the two codes coincide very well in a wide rage of plasma collisionality parameter ν * = 0.01 - 10 and also agree with the theoretical estimations. The time evolution of radial electric field and particle flux, and the radial profile of the geodesic acoustic mode frequency also coincide very well. These facts guarantee the capability of GT5D to simulate plasma turbulence transport with including proper neoclassical effects of collisional diffusion and equilibrium radial electric field. (author)

  18. Srna - Monte Carlo codes for proton transport simulation in combined and voxelized geometries

    Directory of Open Access Journals (Sweden)

    Ilić Radovan D.

    2002-01-01

    Full Text Available This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtained through the PETRA and GEANT programs. The simulation of the proton beam characterization by means of the Multi-Layer Faraday Cup and spatial distribution of positron emitters obtained by our program indicate the imminent application of Monte Carlo techniques in clinical practice.

  19. Simulations of Edge Current Driven Kink Modes with BOUT + + code

    Science.gov (United States)

    Li, G. Q.; Xu, X. Q.; Snyder, P. B.; Turnbull, A. D.; Xia, T. Y.; Ma, C. H.; Xi, P. W.

    2013-10-01

    Edge kink modes (or peeling modes) play a key role in the ELMs. The edge kink modes are driven by peak edge current, which comes from the bootstrap current. We calculated sequences of equilibria with different edge current using CORSICA by keeping total current and pressure profile fixed. Based on these equilibria, with the 3-field BOUT + + code, we calculated the MHD instabilities driven by edge current. For linear low-n ideal MHD modes, BOUT + + results agree with GATO results. With the edge current increasing, the dominant modes are changed from high-n ballooning modes to low-n kink modes. The edge current provides also stabilizing effects on high-n ballooning modes. Furthermore, for edge current scan without keeping total current fixed, the increasing edge current can stabilize the high-n ballooning modes and cannot drive kink modes. The diamagnetic effect can stabilize the high-n ballooning modes, but has no effect on the low-n kink modes. Also, the nonlinear behavior of kink modes is analyzed. Work supported by China MOST grant 2013GB111000 and by China NSF grant 10975161. Also performed for USDOE by LLNL under DE-AC52-07NA27344.

  20. Simulated evolution applied to study the genetic code optimality using a model of codon reassignments.

    Science.gov (United States)

    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

  1. Simulated evolution applied to study the genetic code optimality using a model of codon reassignments

    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

  2. Monte Carlo Simulation of a Segmented Detector for Low-Energy Electron Antineutrinos

    Science.gov (United States)

    Qomi, H. Akhtari; Safari, M. J.; Davani, F. Abbasi

    2017-11-01

    Detection of low-energy electron antineutrinos is of importance for several purposes, such as ex-vessel reactor monitoring, neutrino oscillation studies, etc. The inverse beta decay (IBD) is the interaction that is responsible for detection mechanism in (organic) plastic scintillation detectors. Here, a detailed study will be presented dealing with the radiation and optical transport simulation of a typical segmented antineutrino detector withMonte Carlo method using MCNPX and FLUKA codes. This study shows different aspects of the detector, benefiting from inherent capabilities of the Monte Carlo simulation codes.

  3. The use of best estimate codes to improve the simulation in real time

    International Nuclear Information System (INIS)

    Rivero, N.; Esteban, J. A.; Lenhardt, G.

    2007-01-01

    Best estimate codes are assumed to be the technology solution providing the most realistic and accurate response. Best estimate technology provides a complementary solution to the conservative simulation technology usually applied to determine plant safety margins and perform security related studies. Tecnatom in the early 90's, within the MAS project, pioneered the initiative to implement best estimate code in its training simulators. Result of this project was the implementation of the first six-equations thermal hydraulic code worldwide (TRAC R T), running in a training environment. To meet real time and other specific training requirements, it was necessary to overcome important difficulties. Tecnatom has just adapted the Global Nuclear Fuel core Design code: PANAC 11, and is about to complete the General Electric TRACG04 thermal hydraulic code adaptation. This technology features a unique solution for nuclear plants aiming at providing the highest fidelity in simulation, enabling to consider the simulator as a multipurpose: engineering and training, simulation platform. Besides, a visual environment designed to optimize the models life cycle, covering both pre and post-processing activities, is in its late development phase. (Author)

  4. Divergence-free MHD Simulations with the HERACLES Code

    Directory of Open Access Journals (Sweden)

    Vides J.

    2013-12-01

    Full Text Available Numerical simulations of the magnetohydrodynamics (MHD equations have played a significant role in plasma research over the years. The need of obtaining physical and stable solutions to these equations has led to the development of several schemes, all requiring to satisfy and preserve the divergence constraint of the magnetic field numerically. In this paper, we aim to show the importance of maintaining this constraint numerically. We investigate in particular the hyperbolic divergence cleaning technique applied to the ideal MHD equations on a collocated grid and compare it to the constrained transport technique that uses a staggered grid to maintain the property. The methods are implemented in the software HERACLES and several numerical tests are presented, where the robustness and accuracy of the different schemes can be directly compared.

  5. A multiscale numerical algorithm for heat transfer simulation between multidimensional CFD and monodimensional system codes

    Science.gov (United States)

    Chierici, A.; Chirco, L.; Da Vià, R.; Manservisi, S.; Scardovelli, R.

    2017-11-01

    Nowadays the rapidly-increasing computational power allows scientists and engineers to perform numerical simulations of complex systems that can involve many scales and several different physical phenomena. In order to perform such simulations, two main strategies can be adopted: one may develop a new numerical code where all the physical phenomena of interest are modelled or one may couple existing validated codes. With the latter option, the creation of a huge and complex numerical code is avoided but efficient methods for data exchange are required since the performance of the simulation is highly influenced by its coupling techniques. In this work we propose a new algorithm that can be used for volume and/or boundary coupling purposes for both multiscale and multiphysics numerical simulations. The proposed algorithm is used for a multiscale simulation involving several CFD domains and monodimensional loops. We adopt the overlapping domain strategy, so the entire flow domain is simulated with the system code. We correct the system code solution by matching averaged inlet and outlet fields located at the boundaries of the CFD domains that overlap parts of the monodimensional loop. In particular we correct pressure losses and enthalpy values with source-sink terms that are imposed in the system code equations. The 1D-CFD coupling is a defective one since the CFD code requires point-wise values on the coupling interfaces and the system code provides only averaged quantities. In particular we impose, as inlet boundary conditions for the CFD domains, the mass flux and the mean enthalpy that are calculated by the system code. With this method the mass balance is preserved at every time step of the simulation. The coupling between consecutive CFD domains is not a defective one since with the proposed algorithm we can interpolate the field solutions on the boundary interfaces. We use the MED data structure as the base structure where all the field operations are

  6. Modification of PRETOR Code to Be Applied to Transport Simulation in Stellarators

    Energy Technology Data Exchange (ETDEWEB)

    Fontanet, J.; Castejon, F.; Dies, J.; Fontdecaba, J.; Alejaldre, C.

    2001-07-01

    The 1.5 D transport code PRETOR, that has been previously used to simulate tokamak plasmas, has been modified to perform transport analysis in stellarator geometry. The main modifications that have been introduced in the code are related with the magnetic equilibrium and with the modelling of energy and particle transport. Therefore, PRETOR- Stellarator version has been achieved and the code is suitable to perform simulations on stellarator plasmas. As an example, PRETOR- Stellarator has been used in the transport analysis of several Heliac Flexible TJ-II shots, and the results are compared with those obtained using PROCTR code. These results are also compared with the obtained using the tokamak version of PRETOR to show the importance of the introduced changes. (Author) 18 refs.

  7. Electromagnetic behaviour of a plasma in fluid and relativistic regimes: simulation code R H E A

    International Nuclear Information System (INIS)

    Bonnaud, G.; Dussy, S.; Lefebvre, E.; Bouchut, F.

    1998-01-01

    This report presents a numerical model to simulate the electromagnetic processes involved by electrically-charged relativistic fluids. The physical model is first given. Second, the numerical methods are explained with the various packages of the code RHEA, with indication methods are explained with the various packages of the code RHEA, with indication of its performances, within a 1.5.- dimensional framework. Results from test-simulations are shown to validate the use of the code, for both academic situations and realistic context of laser-plasma interaction, for which the code has been designed: the non-linear phenomena in the context of inertial confinement fusion and the ultra-intense laser pulses. (author)

  8. Modification of PRETOR Code to Be Applied to Transport Simulation in Stellarators

    International Nuclear Information System (INIS)

    Fontanet, J.; Castejon, F.; Dies, J.; Fontdecaba, J.; Alejaldre, C.

    2001-01-01

    The 1.5 D transport code PRETOR, that has been previously used to simulate tokamak plasmas, has been modified to perform transport analysis in stellarator geometry. The main modifications that have been introduced in the code are related with the magnetic equilibrium and with the modelling of energy and particle transport. Therefore, PRETOR- Stellarator version has been achieved and the code is suitable to perform simulations on stellarator plasmas. As an example, PRETOR- Stellarator has been used in the transport analysis of several Heliac Flexible TJ-II shots, and the results are compared with those obtained using PROCTR code. These results are also compared with the obtained using the tokamak version of PRETOR to show the importance of the introduced changes. (Author) 18 refs

  9. A simulation of driven reconnection by a high precision MHD code

    International Nuclear Information System (INIS)

    Kusano, Kanya; Ouchi, Yasuo; Hayashi, Takaya; Horiuchi, Ritoku; Watanabe, Kunihiko; Sato, Tetsuya.

    1988-01-01

    A high precision MHD code, which has the fourth-order accuracy for both the spatial and time steps, is developed, and is applied to the simulation studies of two dimensional driven reconnection. It is confirm that the numerical dissipation of this new scheme is much less than that of two-step Lax-Wendroff scheme. The effect of the plasma compressibility on the reconnection dynamics is investigated by means of this high precision code. (author)

  10. FAST: a three-dimensional time-dependent FEL simulation code

    International Nuclear Information System (INIS)

    Saldin, E.L.; Schneidmiller, E.A.; Yurkov, M.V.

    1999-01-01

    In this report we briefly describe the three-dimensional, time-dependent FEL simulation code FAST. The equations of motion of the particles and Maxwell's equations are solved simultaneously taking into account the slippage effect. Radiation fields are calculated using an integral solution of Maxwell's equations. A special technique has been developed for fast calculations of the radiation field, drastically reducing the required CPU time. As a result, the developed code allows one to use a personal computer for time-dependent simulations. The code allows one to simulate the radiation from the electron bunch of any transverse and longitudinal bunch shape; to simulate simultaneously an external seed with superimposed noise in the electron beam; to take into account energy spread in the electron beam and the space charge fields; and to simulate a high-gain, high-efficiency FEL amplifier with a tapered undulator. It is important to note that there are no significant memory limitations in the developed code and an electron bunch of any length can be simulated

  11. Tree Particle-Mesh: An Adaptive, Efficient, and Parallel Code for Collisionless Cosmological Simulation

    Science.gov (United States)

    Bode, Paul; Ostriker, Jeremiah P.

    2003-03-01

    An improved implementation of an N-body code for simulating collisionless cosmological dynamics is presented. TPM (tree particle-mesh) combines the PM method on large scales with a tree code to handle particle-particle interactions at small separations. After the global PM forces are calculated, spatially distinct regions above a given density contrast are located; the tree code calculates the gravitational interactions inside these denser objects at higher spatial and temporal resolution. The new implementation includes individual particle time steps within trees, an improved treatment of tidal forces on trees, new criteria for higher force resolution and choice of time step, and parallel treatment of large trees. TPM is compared to P3M and a tree code (GADGET) and is found to give equivalent results in significantly less time. The implementation is highly portable (requiring a FORTRAN compiler and MPI) and efficient on parallel machines. The source code can be found on the World Wide Web.

  12. Simulation of disc-bulge-halo galaxies using parallel GPU based codes

    Science.gov (United States)

    Veles, O.; Berczik, P.; Just, A.

    2016-02-01

    We compare the performance of the very popular Tree-GPU code BONSAI with the older Particle-(Multi)Mesh code SUPERBOX. Both code we run on a same hardware using the GPU acceleration for the force calculation. SUPERBOX is a particle-mesh code with high resolution sub-grid and a higher order NGP (nearest grid point) force-calculation scheme. In our research, we are aiming to demonstrate that the new parallel version of SUPERBOX is capable to do the high resolution simulations of the interaction of the system of disc-bulge-halo composed galaxy. We describe the improvement of performance and scalability of SUPERBOX particularly for the Kepler cluster (NVIDIA K20 GPU). A comparison was made with the very popular and publicly available Tree-GPU code BONSAI†.

  13. Simulation of MASPn experiments in MISTRA test facility with COCOSYS code

    International Nuclear Information System (INIS)

    Povilaitis, M.; Urbonavicius, E.

    2007-01-01

    Paper describes simulation of MASPn experiments, which were performed in the MISTRA test facility, with lumped-parameter code COCOSYS. MASPn experiments belong to the SARNET spray benchmark, which was initiated in the Containment Atmosphere Mixing work package. The objective of this benchmark is to evaluate the spray modelling in the containment codes. The paper presents developed MISTRA nodalisation scheme for COCOSYS code, and the results of performed analysis. It is shown that a clear specification of experiments initial conditions is needed to perform the simulation of the experiments. The performed parametric analysis shows that in the simulation the heat losses through the external walls behind the lower condenser installed in the MISTRA facility determines the long-term depressurisation rate. (author)

  14. Development and Test of 2.5-Dimensional Electromagnetic PIC Simulation Code

    Directory of Open Access Journals (Sweden)

    Sang-Yun Lee

    2015-03-01

    Full Text Available We have developed a 2.5-dimensional electromagnetic particle simulation code using the particle-in-cell (PIC method to investigate electromagnetic phenomena that occur in space plasmas. Our code is based on the leap-frog method and the centered difference method for integration and differentiation of the governing equations. We adopted the relativistic Buneman-Boris method to solve the Lorentz force equation and the Esirkepov method to calculate the current density while maintaining charge conservation. Using the developed code, we performed test simulations for electron two-stream instability and electron temperature anisotropy induced instability with the same initial parameters as used in previously reported studies. The test simulation results are almost identical with those of the previous papers.

  15. Srna-Monte Carlo codes for proton transport simulation in combined and voxelized geometries

    CERN Document Server

    Ilic, R D; Stankovic, S J

    2002-01-01

    This paper describes new Monte Carlo codes for proton transport simulations in complex geometrical forms and in materials of different composition. The SRNA codes were developed for three dimensional (3D) dose distribution calculation in proton therapy and dosimetry. The model of these codes is based on the theory of proton multiple scattering and a simple model of compound nucleus decay. The developed package consists of two codes: SRNA-2KG and SRNA-VOX. The first code simulates proton transport in combined geometry that can be described by planes and second order surfaces. The second one uses the voxelized geometry of material zones and is specifically adopted for the application of patient computer tomography data. Transition probabilities for both codes are given by the SRNADAT program. In this paper, we will present the models and algorithms of our programs, as well as the results of the numerical experiments we have carried out applying them, along with the results of proton transport simulation obtaine...

  16. Validation of thermohydraulic codes by comparison of experimental results with computer simulations

    International Nuclear Information System (INIS)

    Madeira, A.A.; Galetti, M.R.S.; Pontedeiro, A.C.

    1989-01-01

    The results obtained by simulation of three cases from CANON depressurization experience, using the TRAC-PF1 computer code, version 7.6, implanted in the VAX-11/750 computer of Brazilian CNEN, are presented. The CANON experience was chosen as first standard problem in thermo-hydraulic to be discussed at ENFIR for comparing results from different computer codes with results obtained experimentally. The ability of TRAC-PF1 code to prevent the depressurization phase of a loss of primary collant accident in pressurized water reactors is evaluated. (M.C.K.) [pt

  17. Sensitivity Analysis and Uncertainty Quantification for the LAMMPS Molecular Dynamics Simulation Code

    Energy Technology Data Exchange (ETDEWEB)

    Picard, Richard Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bhat, Kabekode Ghanasham [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-07-18

    We examine sensitivity analysis and uncertainty quantification for molecular dynamics simulation. Extreme (large or small) output values for the LAMMPS code often occur at the boundaries of input regions, and uncertainties in those boundary values are overlooked by common SA methods. Similarly, input values for which code outputs are consistent with calibration data can also occur near boundaries. Upon applying approaches in the literature for imprecise probabilities (IPs), much more realistic results are obtained than for the complacent application of standard SA and code calibration.

  18. Calibration of Monte Carlo simulation code to low voltage electron beams through radiachromic dosimetry

    International Nuclear Information System (INIS)

    Weiss, D.E.; Kalweit, H.W.; Kensek, R.P.

    1994-01-01

    A simple multilayer slab model of an electron beam using the ITS/TIGER code can consistently account for about 80% of the actual dose delivered by a low voltage electron beam. The difference in calculated values is principally due to the 3D hibachi structure which blocks 22% of the beam. A 3D model was constructed using the ITS/ACCEPT code to improve upon the TIGER simulations. A rectangular source description update to the code and reproduction of all key geometric elements involved, including the hibachi, accounted for 90-95% of the dose received by routine dosimetry

  19. The assessment of containment codes by experiments simulating severe accident scenarios

    International Nuclear Information System (INIS)

    Karwat, H.

    1992-01-01

    Hitherto, a generally applicable validation matrix for codes simulating the containment behaviour under severe accident conditions did not exist. Past code applications have shown that most problems may be traced back to inaccurate thermalhydraulic parameters governing gas- or aerosol-distribution events. A provisional code-validation matrix is proposed, based on a careful selection of containment experiments performed during recent years in relevant test facilities under various operating conditions. The matrix focuses on the thermalhydraulic aspects of the containment behaviour after severe accidents as a first important step. It may be supplemented in the future by additional suitable tests

  20. Comparison of TITAN hybrid deterministic transport code and MCNP5 for simulation of SPECT

    International Nuclear Information System (INIS)

    Royston, K.; Haghighat, A.; Yi, C.

    2010-01-01

    Traditionally, Single Photon Emission Computed Tomography (SPECT) simulations use Monte Carlo methods. The hybrid deterministic transport code TITAN has recently been applied to the simulation of a SPECT myocardial perfusion study. The TITAN SPECT simulation uses the discrete ordinates formulation in the phantom region and a simplified ray-tracing formulation outside of the phantom. A SPECT model has been created in the Monte Carlo Neutral particle (MCNP)5 Monte Carlo code for comparison. In MCNP5 the collimator is directly modeled, but TITAN instead simulates the effect of collimator blur using a circular ordinate splitting technique. Projection images created using the TITAN code are compared to results using MCNP5 for three collimator acceptance angles. Normalized projection images for 2.97 deg, 1.42 deg and 0.98 deg collimator acceptance angles had maximum relative differences of 21.3%, 11.9% and 8.3%, respectively. Visually the images are in good agreement. Profiles through the projection images were plotted to find that the TITAN results followed the shape of the MCNP5 results with some differences in magnitude. A timing comparison on 16 processors found that the TITAN code completed the calculation 382 to 2787 times faster than MCNP5. Both codes exhibit good parallel performance. (author)

  1. Development of JURA -A Simulation Code for Electrochemical Processes in Molten Salt-

    OpenAIRE

    小林 嗣幸

    2003-01-01

    A simulation code for electrochemical processes in molten salt named JURA has been developed. This code can simulate the time history of the processes such as the metal pyro-process and the oxide pyro-process by the diffusion layer theory at various temperature and melts. This report describes the specific formulations of the theory for various electrodes such as the solid cathode, the Cd cathode, and the Cd pool anode. Explanation of the input data and sample calculations of the solid cathod...

  2. Evaluation of SNS Beamline Shielding Configurations using MCNPX Accelerated by ADVANTG

    International Nuclear Information System (INIS)

    Risner, Joel M; Johnson, Seth R.; Remec, Igor; Bekar, Kursat B.

    2015-01-01

    Shielding analyses for the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory pose significant computational challenges, including highly anisotropic high-energy sources, a combination of deep penetration shielding and an unshielded beamline, and a desire to obtain well-converged nearly global solutions for mapping of predicted radiation fields. The majority of these analyses have been performed using MCNPX with manually generated variance reduction parameters (source biasing and cell-based splitting and Russian roulette) that were largely based on the analyst's insight into the problem specifics. Development of the variance reduction parameters required extensive analyst time, and was often tailored to specific portions of the model phase space. We previously applied a developmental version of the ADVANTG code to an SNS beamline study to perform a hybrid deterministic/Monte Carlo analysis and showed that we could obtain nearly global Monte Carlo solutions with essentially uniform relative errors for mesh tallies that cover extensive portions of the model with typical voxel spacing of a few centimeters. The use of weight window maps and consistent biased sources produced using the FW-CADIS methodology in ADVANTG allowed us to obtain these solutions using substantially less computer time than the previous cell-based splitting approach. While those results were promising, the process of using the developmental version of ADVANTG was somewhat laborious, requiring user-developed Python scripts to drive much of the analysis sequence. In addition, limitations imposed by the size of weight-window files in MCNPX necessitated the use of relatively coarse spatial and energy discretization for the deterministic Denovo calculations that we used to generate the variance reduction parameters. We recently applied the production version of ADVANTG to this beamline analysis, which substantially streamlined the analysis process. We also tested importance function

  3. Interfacing VPSC with finite element codes. Demonstration of irradiation growth simulation in a cladding tube

    Energy Technology Data Exchange (ETDEWEB)

    Patra, Anirban [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tome, Carlos [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-23

    This Milestone report shows good progress in interfacing VPSC with the FE codes ABAQUS and MOOSE, to perform component-level simulations of irradiation-induced deformation in Zirconium alloys. In this preliminary application, we have performed an irradiation growth simulation in the quarter geometry of a cladding tube. We have benchmarked VPSC-ABAQUS and VPSC-MOOSE predictions with VPSC-SA predictions to verify the accuracy of the VPSCFE interface. Predictions from the FE simulations are in general agreement with VPSC-SA simulations and also with experimental trends.

  4. Interfacing VPSC with finite element codes. Demonstration of irradiation growth simulation in a cladding tube

    International Nuclear Information System (INIS)

    Patra, Anirban; Tome, Carlos

    2016-01-01

    This Milestone report shows good progress in interfacing VPSC with the FE codes ABAQUS and MOOSE, to perform component-level simulations of irradiation-induced deformation in Zirconium alloys. In this preliminary application, we have performed an irradiation growth simulation in the quarter geometry of a cladding tube. We have benchmarked VPSC-ABAQUS and VPSC-MOOSE predictions with VPSC-SA predictions to verify the accuracy of the VPSCFE interface. Predictions from the FE simulations are in general agreement with VPSC-SA simulations and also with experimental trends.

  5. Research progress on the numerical simulation of Z-pinch implosion using mared code

    International Nuclear Information System (INIS)

    Ding Ning; Wu Jiming; Yang Zhenhua; Fu Shangwu; Ning Cheng; Shu Xiaojian; Zhang Yang; Dai Zihuan; Yao Yanzhong; Yin Li; Sun Shunkai

    2010-01-01

    The physical scheme of the MARED code, a two-dimensional three-temperature radiation magneto-hydrodynamics code for Z-pinch implosion simulation, is described. Results from the one- and two-dimensional calculation tests demonstrate the MARED code is able to simulate Z-pinch implosions of a wide range of accelerator and load parameters. It is able to present the primary dynamic characteristics of Z-pinch implosions, and the calculated images and rules qualitatively agree with the theoretical analyses and experimental observations. Compared with the experimental data, simulation results show that, under the same condition, the tungsten wire-array implosion has higher X-ray radiation power output than aluminum wire arrays. With same load parameters, the X-ray radiation power increases with the load current. Under the certain drive condition, the X-ray output decreases with the load mass. The MARED code is also used to simulate the radiation field formation of the wire-array filled with foam. The preliminary results on the Z machine are qualitative consistent with the simulation results from the Sandia laboratory. (authors)

  6. A generic method for automatic translation between input models for different versions of simulation codes

    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

  7. 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.

  8. Development of a computational framework on fluid-solid mixture flow simulations for the COMPASS code

    International Nuclear Information System (INIS)

    Zhang, Shuai; Morita, Koji; Shirakawa, Noriyuki; Yamamoto, Yuichi

    2010-01-01

    The COMPASS code is designed based on the moving particle semi-implicit method to simulate various complex mesoscale phenomena relevant to core disruptive accidents of sodium-cooled fast reactors. In this study, a computational framework for fluid-solid mixture flow simulations was developed for the COMPASS code. The passively moving solid model was used to simulate hydrodynamic interactions between fluid and solids. Mechanical interactions between solids were modeled by the distinct element method. A multi-time-step algorithm was introduced to couple these two calculations. The proposed computational framework for fluid-solid mixture flow simulations was verified by the comparison between experimental and numerical studies on the water-dam break with multiple solid rods. (author)

  9. A FEW ASPECTS REGARDING THE SIMULATION OF CONTRACT IN THE ROMANIAN CIVIL CODE

    Directory of Open Access Journals (Sweden)

    Tudor Vlad RĂDULESCU

    2017-05-01

    Full Text Available The article aims to analyze some key aspects of simulation in contracts, as regulated by the Romanian Civil Code . The process of simulation will be explained, based on the provisions of the previous Civil Code, but also with reference to the relevant provisions of the legislation of some European countries. The analyse will focus on the apparent act, and also on the secret one and a special emphasis on intention to simulate, animo simulandi, the key aspect of the matter. Also the effects of the simulation will be reviewed, both from the point of view of the parties and that of third parties, the concept of third parties having another meaning in this procedure.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-15

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

  11. Simulation facility for development of high level control code at the SSC

    International Nuclear Information System (INIS)

    Bourianoff, G.; Cole, B.; Botlo, M.; Hunt, S.; Romero, A.; Malitsky, N.; Reshetov, A.

    1994-01-01

    An interactive simulator has been built to serve as a development platform for High Level Application Code (HLAC). It is built around a powerful parallel computer that serves as a particle tracking engine. An intermediate hardware and software layer incorporates fully the EPICS control system. Details of the control system are hidden from the HLAC by an isolation layer (AIC) which consists of C++ class libraries. A closed orbit correction module developed on the simulator is described. ((orig.))

  12. Simulation of hydrogen deflagration experiment – Benchmark exercise with lumped-parameter codes

    International Nuclear Information System (INIS)

    Kljenak, Ivo; Kuznetsov, Mikhail; Kostka, Pal; Kubišova, Lubica; Maltsev, Mikhail; Manzini, Giovanni; Povilaitis, Mantas

    2015-01-01

    Highlights: • Blind and open simulations of hydrogen combustion experiment in large-scale containment-like facility with different lumped-parameter codes. • Simulation of axial as well as radial flame propagation. • Confirmation of adequacy of lumped-parameter codes for safety analyses of actual nuclear power plants. - Abstract: An experiment on hydrogen deflagration (Upward Flame Propagation Experiment – UFPE) was proposed by the Jozef Stefan Institute (Slovenia) and performed in the HYKA A2 facility at the Karlsruhe Institute of Technology (Germany). The experimental results were used to organize a benchmark exercise for lumped-parameter codes. Six organizations (JSI, AEP, LEI, NUBIKI, RSE and UJD SR) participated in the benchmark exercise, using altogether four different computer codes: ANGAR, ASTEC, COCOSYS and ECART. Both blind and open simulations were performed. In general, all the codes provided satisfactory results of the pressure increase, whereas the results of the temperature show a wider dispersal. Concerning the flame axial and radial velocities, the results may be considered satisfactory, given the inherent simplification of the lumped-parameter description compared to the local instantaneous description

  13. Fire simulation in nuclear facilities--the FIRAC code and supporting experiments

    International Nuclear Information System (INIS)

    Burkett, M.W.; Martin, R.A.; Fenton, D.L.; Gunaji, M.V.

    1985-01-01

    The fire accident analysis computer code FIRAC was designed to estimate radioactive and nonradioactive source terms and predict fire-induced flows and thermal and material transport within the ventilation systems of nuclear fuel cycle facilities. FIRAC maintains its basic structure and features and has been expanded and modified to include the capabilities of the zone-type compartment fire model computer code FIRIN developed by Battelle Pacific Northwest Laboratory. The two codes have been coupled to provide an improved simulation of a fire-induced transient within a facility. The basic material transport capability of FIRAC has been retained and includes estimates of entrainment, convection, deposition, and filtration of material. The interrelated effects of filter plugging, heat transfer, gas dynamics, material transport, and fire and radioactive source terms also can be simulated. Also, a sample calculation has been performed to illustrate some of the capabilities of the code and how a typical facility is modeled with FIRAC. In addition to the analytical work being performed at Los Alamos, experiments are being conducted at the New Mexico State University to support the FIRAC computer code development and verification. This paper summarizes two areas of the experimental work that support the material transport capabilities of the code: the plugging of high-efficiency particulate air (HEPA) filters by combustion aerosols and the transport and deposition of smoke in ventilation system ductwork

  14. Testing the new stochastic neutronic code ANET in simulating safety important parameters

    International Nuclear Information System (INIS)

    Xenofontos, T.; Delipei, G.-K.; Savva, P.; Varvayanni, M.; Maillard, J.; Silva, J.; Catsaros, N.

    2017-01-01

    Highlights: • ANET is a new neutronics stochastic code. • Criticality calculations in both subcritical and critical nuclear systems of conventional design were conducted. • Simulations of thermal, lower epithermal and fast neutron fluence rates were performed. • Axial fission rate distributions in standard and MOX fuel pins were computed. - Abstract: ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is an under development Monte Carlo code for simulating both GEN II/III reactors as well as innovative nuclear reactor designs, based on the high energy physics code GEANT3.21 of CERN. ANET is built through continuous GEANT3.21 applicability amplifications, comprising the simulation of particles’ transport and interaction in low energy along with the accessibility of user-provided libraries and tracking algorithms for energies below 20 MeV, as well as the simulation of elastic and inelastic collision, capture and fission. Successive testing applications performed throughout the ANET development have been utilized to verify the new code capabilities. In this context the ANET reliability in simulating certain reactor parameters important to safety is here examined. More specifically the reactor criticality as well as the neutron fluence and fission rates are benchmarked and validated. The Portuguese Research Reactor (RPI) after its conversion to low enrichment in U-235 and the OECD/NEA VENUS-2 MOX international benchmark were considered appropriate for the present study, the former providing criticality and neutron flux data and the latter reaction rates. Concerning criticality benchmarking, the subcritical, Training Nuclear Reactor of the Aristotle University of Thessaloniki (TNR-AUTh) was also analyzed. The obtained results are compared with experimental data from the critical infrastructures and with computations performed by two different, well established stochastic neutronics codes, i.e. TRIPOLI-4.8 and MCNP5. Satisfactory agreement

  15. Parallelization issues of a code for physically-based simulation of fabrics

    Science.gov (United States)

    Romero, Sergio; Gutiérrez, Eladio; Romero, Luis F.; Plata, Oscar; Zapata, Emilio L.

    2004-10-01

    The simulation of fabrics, clothes, and flexible materials is an essential topic in computer animation of realistic virtual humans and dynamic sceneries. New emerging technologies, as interactive digital TV and multimedia products, make necessary the development of powerful tools to perform real-time simulations. Parallelism is one of such tools. When analyzing computationally fabric simulations we found these codes belonging to the complex class of irregular applications. Frequently this kind of codes includes reduction operations in their core, so that an important fraction of the computational time is spent on such operations. In fabric simulators these operations appear when evaluating forces, giving rise to the equation system to be solved. For this reason, this paper discusses only this phase of the simulation. This paper analyzes and evaluates different irregular reduction parallelization techniques on ccNUMA shared memory machines, applied to a real, physically-based, fabric simulator we have developed. Several issues are taken into account in order to achieve high code performance, as exploitation of data access locality and parallelism, as well as careful use of memory resources (memory overhead). In this paper we use the concept of data affinity to develop various efficient algorithms for reduction parallelization exploiting data locality.

  16. The NEST Dry-Run Mode: Efficient Dynamic Analysis of Neuronal Network Simulation Code

    Directory of Open Access Journals (Sweden)

    Susanne Kunkel

    2017-06-01

    Full Text Available NEST is a simulator for spiking neuronal networks that commits to a general purpose approach: It allows for high flexibility in the design of network models, and its applications range from small-scale simulations on laptops to brain-scale simulations on supercomputers. Hence, developers need to test their code for various use cases and ensure that changes to code do not impair scalability. However, running a full set of benchmarks on a supercomputer takes up precious compute-time resources and can entail long queuing times. Here, we present the NEST dry-run mode, which enables comprehensive dynamic code analysis without requiring access to high-performance computing facilities. A dry-run simulation is carried out by a single process, which performs all simulation steps except communication as if it was part of a parallel environment with many processes. We show that measurements of memory usage and runtime of neuronal network simulations closely match the corresponding dry-run data. Furthermore, we demonstrate the successful application of the dry-run mode in the areas of profiling and performance modeling.

  17. SCAR - Post-Accident Simulator SIPA with safety analysis code CATHARE-2 and PWR cold shutdown state simulation

    International Nuclear Information System (INIS)

    Farvacque, M.; Faydide, B.; Dufeil, Ph.; Raimond, E.

    2003-01-01

    The use of Cathare in the simulators of pressurized water reactors has been effective since the beginning of the nineties. Scar project is the second stage of the Cathare strategy for the simulators, its main objective is the extension of the field of simulation to the accident situations in cold shutdown states. Work was carried out in 3 major areas: modelling, optimization and integration in the simulator. Throughout the project, the developments were part of a 3 stages validation strategy: -) elementary tests of the developments of new model on the N4 (1450 MW PWR); -) analytical tests and systems to ensure non regression of the validation of the physical laws of the Cathare code during the modifications carried out within the optimization stage; and -) overall tests of the SIPA-CP1 (900 MW PWR) simulator, controlled automatically by programmed scenarios including the transients which are carried out in PWR, the transients of the Regulatory Guides and the accident transients

  18. TDA - a three-dimensional axisymmetric code for free-electron-laser (FEL) simulation

    International Nuclear Information System (INIS)

    Tran, T.M.; Wurtele, J.S.

    1988-09-01

    A particle simulation code, TDA, which models the single-pass amplification process in a free-electron-laser (FEL) is developed and tested. The code allows for the treatment of the fully three-dimensional electron dynamics, thus taking into account the transverse betatron motion as well as the longitudinal bunching of the electrons. The paraxial wave equation that governs the growth and the diffraction of the selfconsistent radiation field (assumed to be axisymmetrtic), is discretized in the radial direction by the finite difference method. The benchmark study indicates that the single-pass gain, as well as the optical guiding phenomena can be well described by the code with a reasonable number of simulation particles (N ∼ 1000) and a radial mesh number not exceeding 64. A detailed discussion of the numerical method is presented. (author) 6 figs., 1 tab., 17 refs

  19. Computational simulation of natural circulation and rewetting experiments using the TRAC/PF1 code

    International Nuclear Information System (INIS)

    Silva, J.D. da.

    1994-05-01

    In this work the TRAC code was used to simulate experiments of natural circulation performed in the first Brazilian integral test facility at (COPESP), Sao Paulo and a rewetting experiment in a single tube test section carried out at CDTN, Belo Horizonte, Brazil. In the first simulation the loop behavior in two transient conditions with different thermal power, namely 20 k W and 120 k W, was verified in the second one the quench front propagation, the liquid mass collected in the carry over measuring tube and the wall temperature at different elevations during the flooding experiment was measured. A comparative analysis, for code consistency, shows a good agreement between the code results and experimental data, except for the quench from velocity. (author). 15 refs, 19 figs, 12 tabs

  20. Development of 2D particle-in-cell code to simulate high current, low ...

    Indian Academy of Sciences (India)

    Home; Journals; Pramana – Journal of Physics; Volume 69; Issue 4. Development of 2D particle-in-cell code to simulate high current, low energy beam in a beam transport system. S C L Srivastava S V L S Rao ... Keywords. Space-charge; particle-in-cell; beam dynamics; Poisson's equation; solenoids; quadrupole magnets.

  1. Programming Video Games and Simulations in Science Education: Exploring Computational Thinking through Code Analysis

    Science.gov (United States)

    Garneli, Varvara; Chorianopoulos, Konstantinos

    2018-01-01

    Various aspects of computational thinking (CT) could be supported by educational contexts such as simulations and video-games construction. In this field study, potential differences in student motivation and learning were empirically examined through students' code. For this purpose, we performed a teaching intervention that took place over five…

  2. A PIC-MCC code for simulation of streamer propagation in air

    DEFF Research Database (Denmark)

    Chanrion, Olivier Arnaud; Neubert, Torsten

    2008-01-01

    particles are followed in a Cartesian mesh and the electric field is updated with Poisson's equation from the charged particle densities. Collisional processes between electrons and air molecules are simulated with a Monte Carlo technique, according to cross section probabilities. The code also includes...

  3. Simulation of multibunch motion with the Headtail code and application to the CERN SPS and LHC

    CERN Document Server

    Mounet, N; Rumolo, G

    2011-01-01

    Multibunch instabilities due to beam-coupling impedance can be a critical limitation for synchrotrons operating with many bunches. It is particularly true for the LHC under nominal conditions, where according to theoretical predictions the 2808 bunches rely entirely on the performance of the transverse feedback system to remain stable. To study these instabilities, the HEADTAIL code has been extended to simulate the motion of many bunches under the action of wake fields. All the features already present in the single-bunch version of the code, such as synchrotron motion, chromaticity, amplitude detuning due to octupoles and the ability to load any kind of wake fields through tables, have remained available. This new code has been then parallelized in order to track thousands of bunches in a reasonable amount of time. The code was benchmarked against theory and exhibited a good agreement. We also show results for bunch trains in the LHC and compare them with beam-based measurements.

  4. A New Code SORD for Simulation of Polarized Light Scattering in the Earth Atmosphere

    Science.gov (United States)

    Korkin, Sergey; Lyapustin, Alexei; Sinyuk, Aliaksandr; Holben, Brent

    2016-01-01

    We report a new publicly available radiative transfer (RT) code for numerical simulation of polarized light scattering in plane-parallel atmosphere of the Earth. Using 44 benchmark tests, we prove high accuracy of the new RT code, SORD (Successive ORDers of scattering). We describe capabilities of SORD and show run time for each test on two different machines. At present, SORD is supposed to work as part of the Aerosol Robotic NETwork (AERONET) inversion algorithm. For natural integration with the AERONET software, SORD is coded in Fortran 90/95. The code is available by email request from the corresponding (first) author or from ftp://climate1.gsfc.nasa.gov/skorkin/SORD/.

  5. Uncertainty and sensitivity analysis in the scenario simulation with RELAP/SCDAP and MELCOR codes

    International Nuclear Information System (INIS)

    Garcia J, T.; Cardenas V, J.

    2015-09-01

    A methodology was implemented for analysis of uncertainty in simulations of scenarios with RELAP/SCDAP V- 3.4 bi-7 and MELCOR V-2.1 codes, same that are used to perform safety analysis in the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS). The uncertainty analysis methodology chosen is a probabilistic method of type Propagation of uncertainty of the input parameters to the departure parameters. Therefore, it began with the selection of the input parameters considered uncertain and are considered of high importance in the scenario for its direct effect on the output interest variable. These parameters were randomly sampled according to intervals of variation or probability distribution functions assigned by expert judgment to generate a set of input files that were run through the simulation code to propagate the uncertainty to the output parameters. Then, through the use or ordered statistical and formula Wilks, was determined that the minimum number of executions required to obtain the uncertainty bands that include a population of 95% at a confidence level of 95% in the results is 93, is important to mention that in this method that number of executions does not depend on the number of selected input parameters. In the implementation routines in Fortran 90 that allowed automate the process to make the uncertainty analysis in transients for RELAP/SCDAP code were generated. In the case of MELCOR code for severe accident analysis, automation was carried out through complement Dakota Uncertainty incorporated into the Snap platform. To test the practical application of this methodology, two analyzes were performed: the first with the simulation of closing transient of the main steam isolation valves using the RELAP/SCDAP code obtaining the uncertainty band of the dome pressure of the vessel; while in the second analysis, the accident simulation of the power total loss (Sbo) was carried out with the Macarol code obtaining the uncertainty band for the

  6. Monte Carlo method studies and a comparative between GEANT4 tool kit and MCNPX to depth dose in medical physics; Estudos do metodo Monte Carlo e um comparativo entre a ferramenta GEANT4 e MCNPX para doses profundas em fisica medica

    Energy Technology Data Exchange (ETDEWEB)

    Magalhaes, Antonio H.M.; Lemke, Ney; Hormaza, Joel M.; Silva, Danilo A. da; Inocente, Guilherme F.; Pazianotto, Mauricio T., E-mail: ahmmagalhaes@gmail.co [UNESP, Botucatu, SP (Brazil). Inst. de Biociencias. Dept. de Fisica e Biofisica

    2009-07-01

    Knowing the depth dose at the central axis is fundamental for the accurate planning of medical treatment systems involving ionizing radiation. With the evolution of the informatics it is possible the utilization of various computational tools such as GEANT4 and the MCNPX, which use the Monte Carlo Method for simulation of such situations, This paper makes a comparative between the two tools for the this type of application

  7. Implementation of Hydrodynamic Simulation Code in Shock Experiment Design for Alkali Metals

    Science.gov (United States)

    Coleman, A. L.; Briggs, R.; Gorman, M. G.; Ali, S.; Lazicki, A.; Swift, D. C.; Stubley, P. G.; McBride, E. E.; Collins, G.; Wark, J. S.; McMahon, M. I.

    2017-10-01

    Shock compression techniques enable the investigation of extreme P-T states. In order to probe off-Hugoniot regions of P-T space, target makeup and laser pulse parameters must be carefully designed. HYADES is a hydrodynamic simulation code which has been successfully utilised to simulate shock compression events and refine the experimental parameters required in order to explore new P-T states in alkali metals. Here we describe simulations and experiments on potassium, along with the techniques required to access off-Hugoniot states.

  8. Simulation of Aircraft Landing Gears with a Nonlinear Dynamic Finite Element Code

    Science.gov (United States)

    Lyle, Karen H.; Jackson, Karen E.; Fasanella, Edwin L.

    2000-01-01

    Recent advances in computational speed have made aircraft and spacecraft crash simulations using an explicit, nonlinear, transient-dynamic, finite element analysis code more feasible. This paper describes the development of a simple landing gear model, which accurately simulates the energy absorbed by the gear without adding substantial complexity to the model. For a crash model, the landing gear response is approximated with a spring where the force applied to the fuselage is computed in a user-written subroutine. Helicopter crash simulations using this approach are compared with previously acquired experimental data from a full-scale crash test of a composite helicopter.

  9. Simulation of containment phenomena during the Phebus FPT1 test with the CONTAIN code

    International Nuclear Information System (INIS)

    Kljenak, I.; Mavko, B.

    2002-01-01

    Thermal-hydraulic and aerosol phenomena which occurred in the containment vessel of the Phebus integral experimental facility during the first 30000 s of the Phebus FPT1 test were simulated with the CONTAIN thermal-hydraulic computer code. A single-cell input model of the vessel was developed, and boundary and initial conditions that were determined during the experiment were applied. The comparison of experimental and calculated results shows that, although the atmosphere temperature was well simulated, the calculated condensation rate was apparently too high, resulting in a lower pressure of the containment atmosphere. The aerosol deposition process was well simulated.(author)

  10. 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

  11. L-PICOLA: A parallel code for fast dark matter simulation

    Science.gov (United States)

    Howlett, C.; Manera, M.; Percival, W. J.

    2015-09-01

    Robust measurements based on current large-scale structure surveys require precise knowledge of statistical and systematic errors. This can be obtained from large numbers of realistic mock galaxy catalogues that mimic the observed distribution of galaxies within the survey volume. To this end we present a fast, distributed-memory, planar-parallel code, L-PICOLA, which can be used to generate and evolve a set of initial conditions into a dark matter field much faster than a full non-linear N-Body simulation. Additionally, L-PICOLA has the ability to include primordial non-Gaussianity in the simulation and simulate the past lightcone at run-time, with optional replication of the simulation volume. Through comparisons to fully non-linear N-Body simulations we find that our code can reproduce the z = 0 power spectrum and reduced bispectrum of dark matter to within 2% and 5% respectively on all scales of interest to measurements of Baryon Acoustic Oscillations and Redshift Space Distortions, but 3 orders of magnitude faster. The accuracy, speed and scalability of this code, alongside the additional features we have implemented, make it extremely useful for both current and next generation large-scale structure surveys. L-PICOLA is publicly available at https://cullanhowlett.github.io/l-picola.

  12. Schnek: A C++ library for the development of parallel simulation codes on regular grids

    Science.gov (United States)

    Schmitz, Holger

    2018-05-01

    A large number of algorithms across the field of computational physics are formulated on grids with a regular topology. We present Schnek, a library that enables fast development of parallel simulations on regular grids. Schnek contains a number of easy-to-use modules that greatly reduce the amount of administrative code for large-scale simulation codes. The library provides an interface for reading simulation setup files with a hierarchical structure. The structure of the setup file is translated into a hierarchy of simulation modules that the developer can specify. The reader parses and evaluates mathematical expressions and initialises variables or grid data. This enables developers to write modular and flexible simulation codes with minimal effort. Regular grids of arbitrary dimension are defined as well as mechanisms for defining physical domain sizes, grid staggering, and ghost cells on these grids. Ghost cells can be exchanged between neighbouring processes using MPI with a simple interface. The grid data can easily be written into HDF5 files using serial or parallel I/O.

  13. Simulating magnetospheres with numerical relativity: The GiRaFFE code

    Science.gov (United States)

    Babiuc-Hamilton, Maria; Etienne, Zach

    2016-01-01

    Numerical Relativity has shown success over the past several years, especially in the simulation of black holes and gravitational waves. In recent years, teams have tackled the problem of the interaction of gravitational and electromagnetic waves. But where there are plasmas, the simulations often have trouble reproducing nature. Neutron stars, black hole accretion disks, astrophysical jets—all of these represent extreme environments both gravitationally and electromagnetically. We are creating the first open-source, dynamical spacetime general relativity force-free electrodynamics code: GiRaFFE.We present here the performance of GiRaFFE in testing. With this code, we will simulate neutron star magnetospheres, collisions between neutron stars and black holes, and particular attention will be paid to the production of jets through the Blandford-Znajek mechanism.GiRaFFE will be made available to the community.

  14. Phase 1 Validation Testing and Simulation for the WEC-Sim Open Source Code

    Science.gov (United States)

    Ruehl, K.; Michelen, C.; Gunawan, B.; Bosma, B.; Simmons, A.; Lomonaco, P.

    2015-12-01

    WEC-Sim is an open source code to model wave energy converters performance in operational waves, developed by Sandia and NREL and funded by the US DOE. The code is a time-domain modeling tool developed in MATLAB/SIMULINK using the multibody dynamics solver SimMechanics, and solves the WEC's governing equations of motion using the Cummins time-domain impulse response formulation in 6 degrees of freedom. The WEC-Sim code has undergone verification through code-to-code comparisons; however validation of the code has been limited to publicly available experimental data sets. While these data sets provide preliminary code validation, the experimental tests were not explicitly designed for code validation, and as a result are limited in their ability to validate the full functionality of the WEC-Sim code. Therefore, dedicated physical model tests for WEC-Sim validation have been performed. This presentation provides an overview of the WEC-Sim validation experimental wave tank tests performed at the Oregon State University's Directional Wave Basin at Hinsdale Wave Research Laboratory. Phase 1 of experimental testing was focused on device characterization and completed in Fall 2015. Phase 2 is focused on WEC performance and scheduled for Winter 2015/2016. These experimental tests were designed explicitly to validate the performance of WEC-Sim code, and its new feature additions. Upon completion, the WEC-Sim validation data set will be made publicly available to the wave energy community. For the physical model test, a controllable model of a floating wave energy converter has been designed and constructed. The instrumentation includes state-of-the-art devices to measure pressure fields, motions in 6 DOF, multi-axial load cells, torque transducers, position transducers, and encoders. The model also incorporates a fully programmable Power-Take-Off system which can be used to generate or absorb wave energy. Numerical simulations of the experiments using WEC-Sim will be

  15. MOLOCH computer code for molecular-dynamics simulation of processes in condensed matter

    Directory of Open Access Journals (Sweden)

    Derbenev I.V.

    2011-01-01

    Full Text Available Theoretical and experimental investigation into properties of condensed matter is one of the mainstreams in RFNC-VNIITF scientific activity. The method of molecular dynamics (MD is an innovative method of theoretical materials science. Modern supercomputers allow the direct simulation of collective effects in multibillion atom sample, making it possible to model physical processes on the atomistic level, including material response to dynamic load, radiation damage, influence of defects and alloying additions upon material mechanical properties, or aging of actinides. During past ten years, the computer code MOLOCH has been developed at RFNC-VNIITF. It is a parallel code suitable for massive parallel computing. Modern programming techniques were used to make the code almost 100% efficient. Practically all instruments required for modelling were implemented in the code: a potential builder for different materials, simulation of physical processes in arbitrary 3D geometry, and calculated data processing. A set of tests was developed to analyse algorithms efficiency. It can be used to compare codes with different MD implementation between each other.

  16. Overall simulation of a HTGR plant with the gas adapted MANTA code

    International Nuclear Information System (INIS)

    Emmanuel Jouet; Dominique Petit; Robert Martin

    2005-01-01

    Full text of publication follows: AREVA's subsidiary Framatome ANP is developing a Very High Temperature Reactor nuclear heat source that can be used for electricity generation as well as cogeneration including hydrogen production. The selected product has an indirect cycle architecture which is easily adapted to all possible uses of the nuclear heat source. The coupling to the applications is implemented through an Intermediate Heat exchanger. The system code chosen to calculate the steady-state and transient behaviour of the plant is based on the MANTA code. The flexible and modular MANTA code that is originally a system code for all non LOCA PWR plant transients, has been the subject of new developments to simulate all the forced convection transients of a nuclear plant with a gas cooled High Temperature Reactor including specific core thermal hydraulics and neutronics modelizations, gas and water steam turbomachinery and control structure. The gas adapted MANTA code version is now able to model a total HTGR plant with a direct Brayton cycle as well as indirect cycles. To validate these new developments, a modelization with the MANTA code of a real plant with direct Brayton cycle has been performed and steady-states and transients compared with recorded thermal hydraulic measures. Finally a comparison with the RELAP5 code has been done regarding transient calculations of the AREVA indirect cycle HTR project plant. Moreover to improve the user-friendliness in order to use MANTA as a systems conception, optimization design tool as well as a plant simulation tool, a Man- Machine-Interface is available. Acronyms: MANTA Modular Advanced Neutronic and Thermal hydraulic Analysis; HTGR High Temperature Gas-Cooled Reactor. (authors)

  17. A general concurrent algorithm for plasma particle-in-cell simulation codes

    International Nuclear Information System (INIS)

    Liewer, P.C.; Decyk, V.K.

    1989-01-01

    We have developed a new algorithm for implementing plasma particle-in-cell (PIC) simulation codes on concurrent processors with distributed memory. This algorithm, named the general concurrent PIC algorithm (GCPIC), has been used to implement an electrostatic PIC code on the 33-node JPL Mark III Hypercube parallel computer. To decompose at PIC code using the GCPIC algorithm, the physical domain of the particle simulation is divided into sub-domains, equal in number to the number of processors, such that all sub-domains have roughly equal numbers of particles. For problems with non-uniform particle densities, these sub-domains will be of unequal physical size. Each processor is assigned a sub-domain and is responsible for updating the particles in its sub-domain. This algorithm has led to a a very efficient parallel implementation of a well-benchmarked 1-dimensional PIC code. The dominant portion of the code, updating the particle positions and velocities, is nearly 100% efficient when the number of particles is increased linearly with the number of hypercube processors used so that the number of particles per processor is constant. For example, the increase in time spent updating particles in going from a problem with 11,264 particles run on 1 processor to 360,448 particles on 32 processors was only 3% (parallel efficiency of 97%). Although implemented on a hypercube concurrent computer, this algorithm should also be efficient for PIC codes on other parallel architectures and for large PIC codes on sequential computers where part of the data must reside on external disks. copyright 1989 Academic Press, Inc

  18. Evaluation of the Trac-PF1 code for simulating the Neptun reflooding experiment

    International Nuclear Information System (INIS)

    Pontedeiro, A.C.; Galetti, M.R.S.

    1991-01-01

    The present work presents an assessment of the TRAC-BF1 code using the results of the NEPTUN experiment which simulates the reflooding in a loss-of-coolant accident (LOCA) in a PWR. The NEPTUN experiment is composed of an array of electrically-heated tubes where the reflooding condition can be tested. Two types of tests results are presented and compared with the values obtained with the TRAC-BF1 code. From this comparison it is concluded that TRAC is suitable for verifying accident analysis. (author)

  19. DynaPhoPy: A code for extracting phonon quasiparticles from molecular dynamics simulations

    Science.gov (United States)

    Carreras, Abel; Togo, Atsushi; Tanaka, Isao

    2017-12-01

    We have developed a computational code, DYNAPHOPY, that allows us to extract the microscopic anharmonic phonon properties from molecular dynamics (MD) simulations using the normal-mode-decomposition technique as presented by Sun et al. (2014). Using this code we calculated the quasiparticle phonon frequencies and linewidths of crystalline silicon at different temperatures using both of first-principles and the Tersoff empirical potential approaches. In this work we show the dependence of these properties on the temperature using both approaches and compare them with reported experimental data obtained by Raman spectroscopy (Balkanski et al., 1983; Tsu and Hernandez, 1982).

  20. Chemical Reactivity and Spectroscopy Explored From QM/MM Molecular Dynamics Simulations Using the LIO Code

    Directory of Open Access Journals (Sweden)

    Juan P. Marcolongo

    2018-03-01

    Full Text Available In this work we present the current advances in the development and the applications of LIO, a lab-made code designed for density functional theory calculations in graphical processing units (GPU, that can be coupled with different classical molecular dynamics engines. This code has been thoroughly optimized to perform efficient molecular dynamics simulations at the QM/MM DFT level, allowing for an exhaustive sampling of the configurational space. Selected examples are presented for the description of chemical reactivity in terms of free energy profiles, and also for the computation of optical properties, such as vibrational and electronic spectra in solvent and protein environments.

  1. Development of Monte Carlo input code for proton, alpha and heavy ion microdosimetric trac structure simulations

    International Nuclear Information System (INIS)

    Douglass, M.; Bezak, E.

    2010-01-01

    Full text: Radiobiology science is important for cancer treatment as it improves our understanding of radiation induced cell death. Monte Carlo simulations playa crucial role in developing improved knowledge of cellular processes. By model Ii ng the cell response to radiation damage and verifying with experimental data, understanding of cell death through direct radiation hits and bystander effects can be obtained. A Monte Carlo input code was developed using 'Geant4' to simulate cellular level radiation interactions. A physics list which enables physically accurate interactions of heavy ions to energies below 100 e V was implemented. A simple biological cell model was also implemented. Each cell consists of three concentric spheres representing the nucleus, cytoplasm and the membrane. This will enable all critical cell death channels to be investigated (i.e. membrane damage, nucleus/DNA). The current simulation has the ability to predict the positions of ionization events within the individual cell components on I micron scale. We have developed a Geant4 simulation for investigation of radiation damage to cells on sub-cellular scale (∼I micron). This code currently allows the positions of the ionisation events within the individual components of the cell enabling a more complete picture of cell death to be developed. The next stage will include expansion of the code to utilise non-regular cell lattice. (author)

  2. Portable parallel code for plasma simulations: Development experience and initial results

    International Nuclear Information System (INIS)

    Liewer, P.; Karmesin, S.R.; Brackbill, J.

    1994-01-01

    Plasma physics covers a wide variety of phenomena that are beyond the reach of symbolic mathematics, and for which experiments are often difficult. It is necessary to rely on computer experiments to test theories and understand the data. The authors describe progress in constructing a portable parallel Particle in Cell (PIC) code for three dimensional plasma simulations, and initial physics results using it for the Global Heliosphere problem. The code is designed to scale well to large parallel machines to take advantage of the fastest computers that are likely to be found in the foreseeable future. It is designed to allow the user to do a straight fluid (MHD) simulation, a kinetic PIC simulation to sample the velocity-space behavior of a system or a FLIP PIC simulation to model a low dissipation continuum fluid. It is designed to allow as much as possible of the algorithm to be written in dimension-independent style to allow the code to be used in 1, 2 or 3 dimensional systems. It is designed to be able to handle moderately complex geometries efficiently through the use of multiple patches, each of which is a deformable logically cartesian mesh. It is designed for several types of portability: to different numerics, physics, architectures, and people

  3. Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC).

    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.

  4. Coupled simulations of the NACIE facility using RELAP5 and ANSYS FLUENT codes

    International Nuclear Information System (INIS)

    Martelli, Daniele; Forgione, Nicola; Barone, Gianluca; Di Piazza, Ivan

    2017-01-01

    Highlights: • A preliminary coupling tool assesment is presented. • A STH/CFD coupling tool is developed for thermal-hydraulics analyses. • Explicit numerical coupling scheme is implemented. • Natural, assisted circulation and ULOF tests are simulated. - Abstract: This work deals with the development and preliminarily assessment of a coupling methodology between a modified version of RELAP5/Mod3.3 STH code and FLUENT commercial CFD code, applied to the NACIE (natural circulation experiment) LBE (lead bismuth eutectic) experimental loop (built and located at the ENEA Brasimone research centre). The coupling tool is used to simulate experiments representative of both natural circulation conditions and isothermal gas enhanced (assisted) circulation. Furthermore, an accidental test reproducing an Unprotected Loss of Flow (ULOF) scenario is also simulated and the outcomes are presented. A preliminary sensitivity analysis has shown that, to guarantee a suitable numerical convergence, the assisted circulation tests require a time step one order of magnitude lower compared to natural circulation ones. The comparison between the RELAP5 stand-alone simulations and RELAP5/FLUENT coupled simulations proved the capability to simulate the thermal-hydraulic behaviour of a loop experimental facility for all the examined conditions.

  5. SimCommSys: taking the errors out of error-correcting code simulations

    Directory of Open Access Journals (Sweden)

    Johann A. Briffa

    2014-06-01

    Full Text Available In this study, we present SimCommSys, a simulator of communication systems that we are releasing under an open source license. The core of the project is a set of C + + libraries defining communication system components and a distributed Monte Carlo simulator. Of principal interest is the error-control coding component, where various kinds of binary and non-binary codes are implemented, including turbo, LDPC, repeat-accumulate and Reed–Solomon. The project also contains a number of ready-to-build binaries implementing various stages of the communication system (such as the encoder and decoder, a complete simulator and a system benchmark. Finally, SimCommSys also provides a number of shell and python scripts to encapsulate routine use cases. As long as the required components are already available in SimCommSys, the user may simulate complete communication systems of their own design without any additional programming. The strict separation of development (needed only to implement new components and use (to simulate specific constructions encourages reproducibility of experimental work and reduces the likelihood of error. Following an overview of the framework, we provide some examples of how to use the framework, including the implementation of a simple codec, the specification of communication systems and their simulation.

  6. Development of NSSS Simulation Engine for SMART Simulator Using the Best Estimate Code, MARS3.1

    International Nuclear Information System (INIS)

    Kim, K. D.; Lee, S. W.; Lee, Sung Chul; Suh, Yong Suk; Suh, Jae Seung

    2011-01-01

    Limited computational capability and crude thermalhydraulic modeling in early 1980s forced the use of overly simplified physical models and assumptions for a real-time calculation at the cost of fidelity. Rapid advances in computer technology make it possible to improve the fidelity of the simulator models. These efforts have been made based on RELAP5 in the US, and CATHARE2 in France. The NSSS thermalhydraulic engines adopted in the most domestic fullscope power plant simulators have been replaced with RELAP5 based engines which were provided by US vendors. Since the technology dependency of the NSSS T/H engine by foreign vendors, it may cause difficulties in maintenance and model improvement. KAERI has started to develop a realistic NSSS calculation engine based on the best-estimate code MARS 3.1 for the SMART full-scope simulator. Even though we are developing the NSSS calculation engine for SMART simulator, it can be easily extended to light water reactors and GEN-IV reactors, etc. The verification of the NSSS calculation engine for SMART simulator has been conducted by an integrated test in the simulator environment, Jade 4.0, developed by GSE of Windows 2003. This paper briefly presents our efforts for the NSSS calculation engine for SMART simulator and verification test results of SAT (Site Acceptance Test)

  7. Sampling-based nuclear data uncertainty quantification for continuous energy Monte-Carlo codes

    International Nuclear Information System (INIS)

    Zhu, T.

    2015-01-01

    Research on the uncertainty of nuclear data is motivated by practical necessity. Nuclear data uncertainties can propagate through nuclear system simulations into operation and safety related parameters. The tolerance for uncertainties in nuclear reactor design and operation can affect the economic efficiency of nuclear power, and essentially its sustainability. The goal of the present PhD research is to establish a methodology of nuclear data uncertainty quantification (NDUQ) for MCNPX, the continuous-energy Monte-Carlo (M-C) code. The high fidelity (continuous-energy treatment and flexible geometry modelling) of MCNPX makes it the choice of routine criticality safety calculations at PSI/LRS, but also raises challenges for NDUQ by conventional sensitivity/uncertainty (S/U) methods. For example, only recently in 2011, the capability of calculating continuous energy κ eff sensitivity to nuclear data was demonstrated in certain M-C codes by using the method of iterated fission probability. The methodology developed during this PhD research is fundamentally different from the conventional S/U approach: nuclear data are treated as random variables and sampled in accordance to presumed probability distributions. When sampled nuclear data are used in repeated model calculations, the output variance is attributed to the collective uncertainties of nuclear data. The NUSS (Nuclear data Uncertainty Stochastic Sampling) tool is based on this sampling approach and implemented to work with MCNPX’s ACE format of nuclear data, which also gives NUSS compatibility with MCNP and SERPENT M-C codes. In contrast, multigroup uncertainties are used for the sampling of ACE-formatted pointwise-energy nuclear data in a groupwise manner due to the more limited quantity and quality of nuclear data uncertainties. Conveniently, the usage of multigroup nuclear data uncertainties allows consistent comparison between NUSS and other methods (both S/U and sampling-based) that employ the same

  8. A MCNP simulation study of neutronic calculations of spallation targets

    Directory of Open Access Journals (Sweden)

    Feghhi Seyed Amir Hossein

    2013-01-01

    Full Text Available The accelerator driven system is an innovative reactor which is being considered as a dedicated high-level waste burner. The function of the spallation target in accelerator driven system is to convert the incident high-energy particle beam to low-energy neutrons. One of the quantities of most interest for practical purposes is the number of neutrons produced per proton in a spallation target. However, this vital value depends not only on the material, but on the size of the target as well, due to the internuclear cascade. The MCNPX 2.4 code can be used for spallation target computation. Some benchmark results have been compared with MCNPX 2.4 simulations to verify the code's potential for calculating various parameters of an accelerator driven system target. Using the computation method, neutron interaction processes such as loss, capture and (n, xn into a spallation target have been studied for W, Ta, Pb, Bi, and LBE spallation targets in different target dimensions. With relative errors less than 10%, the numerical simulation provided by the MCNPX code agrees qualitatively with other simulation results previously carried out, qualifying it for spallation calculations. Among the studied targets, W and Ta targets resulted in a higher neutron spallation yield using lesser target dimensions. Pb, Bi, and LBE spallation targets behave similarly regarding the accessible leaked neutron yield on the outer surface of the spallation target. By use of a thicker target, LBE can compete with both W and Ta targets regarding the neutron yield parameter.

  9. A high-resolution code for large eddy simulation of incompressible turbulent boundary layer flows

    KAUST Repository

    Cheng, Wan

    2014-03-01

    We describe a framework for large eddy simulation (LES) of incompressible turbulent boundary layers over a flat plate. This framework uses a fractional-step method with fourth-order finite difference on a staggered mesh. We present several laminar examples to establish the fourth-order accuracy and energy conservation property of the code. Furthermore, we implement a recycling method to generate turbulent inflow. We use the stretched spiral vortex subgrid-scale model and virtual wall model to simulate the turbulent boundary layer flow. We find that the case with Reθ ≈ 2.5 × 105 agrees well with available experimental measurements of wall friction, streamwise velocity profiles and turbulent intensities. We demonstrate that for cases with extremely large Reynolds numbers (Reθ = 1012), the present LES can reasonably predict the flow with a coarse mesh. The parallel implementation of the LES code demonstrates reasonable scaling on O(103) cores. © 2013 Elsevier Ltd.

  10. An Enhanced GINGER Simulation Code with Harmonic Emission and HDF5 IO Capabilities

    International Nuclear Information System (INIS)

    Fawley, William M.

    2006-01-01

    GINGER [1] is an axisymmetric, polychromatic (r-z-t) FEL simulation code originally developed in the mid-1980's to model the performance of single-pass amplifiers. Over the past 15 years GINGER's capabilities have been extended to include more complicated configurations such as undulators with drift spaces, dispersive sections, and vacuum chamber wakefield effects; multi-pass oscillators; and multi-stage harmonic cascades. Its coding base has been tuned to permit running effectively on platforms ranging from desktop PC's to massively parallel processors such as the IBM-SP. Recently, we have made significant changes to GINGER by replacing the original predictor-corrector field solver with a new direct implicit algorithm, adding harmonic emission capability, and switching to the HDF5 IO library [2] for output diagnostics. In this paper, we discuss some details regarding these changes and also present simulation results for LCLS SASE emission at λ = 0.15 nm and higher harmonics

  11. TMCC: a transient three-dimensional neutron transport code by the direct simulation method - 222

    International Nuclear Information System (INIS)

    Shen, H.; Li, Z.; Wang, K.; Yu, G.

    2010-01-01

    A direct simulation method (DSM) is applied to solve the transient three-dimensional neutron transport problems. DSM is based on the Monte Carlo method, and can be considered as an application of the Monte Carlo method in the specific type of problems. In this work, the transient neutronics problem is solved by simulating the dynamic behaviors of neutrons and precursors of delayed neutrons during the transient process. DSM gets rid of various approximations which are always necessary to other methods, so it is precise and flexible in the requirement of geometric configurations, material compositions and energy spectrum. In this paper, the theory of DSM is introduced first, and the numerical results obtained with the new transient analysis code, named TMCC (Transient Monte Carlo Code), are presented. (authors)

  12. Simulating hypervelocity impact effects on structures using the smoothed particle hydrodynamics code MAGI

    Science.gov (United States)

    Libersky, Larry; Allahdadi, Firooz A.; Carney, Theodore C.

    1992-01-01

    Analysis of interaction occurring between space debris and orbiting structures is of great interest to the planning and survivability of space assets. Computer simulation of the impact events using hydrodynamic codes can provide some understanding of the processes but the problems involved with this fundamental approach are formidable. First, any realistic simulation is necessarily three-dimensional, e.g., the impact and breakup of a satellite. Second, the thickness of important components such as satellite skins or bumper shields are small with respect to the dimension of the structure as a whole, presenting severe zoning problems for codes. Thirdly, the debris cloud produced by the primary impact will yield many secondary impacts which will contribute to the damage and possible breakup of the structure. The problem was approached by choosing a relatively new computational technique that has virtues peculiar to space impacts. The method is called Smoothed Particle Hydrodynamics.

  13. Development of computer code SIMPSEX for simulation of FBR fuel reprocessing flowsheets: II. additional benchmarking results

    International Nuclear Information System (INIS)

    Shekhar Kumar; Koganti, S.B.

    2003-07-01

    Benchmarking and application of a computer code SIMPSEX for high plutonium FBR flowsheets was reported recently in an earlier report (IGC-234). Improvements and recompilation of the code (Version 4.01, March 2003) required re-validation with the existing benchmarks as well as additional benchmark flowsheets. Improvements in the high Pu region (Pu Aq >30 g/L) resulted in better results in the 75% Pu flowsheet benchmark. Below 30 g/L Pu Aq concentration, results were identical to those from the earlier version (SIMPSEX Version 3, code compiled in 1999). In addition, 13 published flowsheets were taken as additional benchmarks. Eleven of these flowsheets have a wide range of feed concentrations and few of them are β-γ active runs with FBR fuels having a wide distribution of burnup and Pu ratios. A published total partitioning flowsheet using externally generated U(IV) was also simulated using SIMPSEX. SIMPSEX predictions were compared with listed predictions from conventional SEPHIS, PUMA, PUNE and PUBG. SIMPSEX results were found to be comparable and better than the result from above listed codes. In addition, recently reported UREX demo results along with AMUSE simulations are also compared with SIMPSEX predictions. Results of the benchmarking SIMPSEX with these 14 benchmark flowsheets are discussed in this report. (author)

  14. Development of a space radiation Monte Carlo computer simulation based on the FLUKA and ROOT codes

    CERN Document Server

    Pinsky, L; Ferrari, A; Sala, P; Carminati, F; Brun, R

    2001-01-01

    This NASA funded project is proceeding to develop a Monte Carlo-based computer simulation of the radiation environment in space. With actual funding only initially in place at the end of May 2000, the study is still in the early stage of development. The general tasks have been identified and personnel have been selected. The code to be assembled will be based upon two major existing software packages. The radiation transport simulation will be accomplished by updating the FLUKA Monte Carlo program, and the user interface will employ the ROOT software being developed at CERN. The end-product will be a Monte Carlo-based code which will complement the existing analytic codes such as BRYNTRN/HZETRN presently used by NASA to evaluate the effects of radiation shielding in space. The planned code will possess the ability to evaluate the radiation environment for spacecraft and habitats in Earth orbit, in interplanetary space, on the lunar surface, or on a planetary surface such as Mars. Furthermore, it will be usef...

  15. Simulating Molecular Clouds in Dwarf Spheroidal Galaxies: Simplified Aarseth N body Code with Super Storage

    Science.gov (United States)

    Brecht, J.; Byrd, G.

    1996-12-01

    Variations have been implemented on the standard Aarseth individual time step n body code for future use in simulations of the dynamical effects of molecular clouds in dwarf spheriodal galaxies. The clouds will be many times more massive than a typical star so that various simplifying approximations can be made to speed up the code. One variation has been to assume that large variations from sphericity will not occur so that only the first (m=0) multipole approximation will be needed i.e. particles interior to the clould act as a common mass at the center and particles exterior have no effect. Only the cloud is felt as a single particle by the stars in the galaxy. We will describe various strategies which are used to speed operation of the code under these assumptions in terms of tabulating interior and exterior particles. We also discuss how the individual time step nature of the Aarseth code can be used to greatly save on storage space required to record the positions and velocities of stars and clouds at different times during the simulations. This work was supported by NSF REU grant AST-9424226

  16. Further development of the V-code for recirculating linear accelerator simulations

    Energy Technology Data Exchange (ETDEWEB)

    Franke, Sylvain; Ackermann, Wolfgang; Weiland, Thomas [Institut fuer Theorie Elektromagnetischer Felder, Technische Universitaet Darmstadt (Germany); Eichhorn, Ralf; Hug, Florian; Kleinmann, Michaela; Platz, Markus [Institut fuer Kernphysik, Technische Universitaet Darmstadt (Germany)

    2011-07-01

    The Superconducting Darmstaedter LINear Accelerator (S-DALINAC) installed at the institute of nuclear physics (IKP) at TU Darmstadt is designed as a recirculating linear accelerator. The beam is first accelerated up to 10 MeV in the injector beam line. Then it is deflected by 180 degrees into the main linac. The linac section with eight superconducting cavities is passed up to three times, providing a maximal energy gain of 40 MeV on each passage. Due to this recirculating layout it is complicated to find an accurate setup for the various beam line elements. Fast online beam dynamics simulations can advantageously assist the operators because they provide a more detailed insight into the actual machine status. In this contribution further developments of the moment based simulation tool V-code which enables to simulate recirculating machines are presented together with simulation results.

  17. Comparison of EAS simulation results using CORSIKA code for different high-energy interaction models

    International Nuclear Information System (INIS)

    Thakuria, Chabin; Boruah, K.

    2011-01-01

    Interpretation of EAS characteristics for primary energy above the energy attained by man made accelerator is dominated by model predictions and detailed simulations. Differences and uncertainties in these hadronic interaction models play crucial role in the prediction of EAS characteristics. A comparative study for muon numbers to primary energy ratio and depth of shower maximum for different hadronic interaction models available in monte-carlo simulation code CORSIKA viz. QGSJET01, QGSJET-II, DPMJET, EPOS, SYBILL, and VENUS for two primaries (viz. proton and iron) in the energy range 10 14 eV to 10 17 eV is presented in this work.

  18. Benchmarking of Simulation Codes Based on the Montague Resonance in the CERN Proton Synchrotron

    CERN Document Server

    Hofmann, Ingo; Cousineau, Sarah M; Franchetti, Giuliano; Giovannozzi, Massimo; Holmes, Jeffrey Alan; Jones, Frederick W; Luccio, Alfredo U; Machida, Shinji; Métral, E; Qiang, Ji; Ryne, Robert D; Spentzouris, Panagiotis

    2005-01-01

    Experimental data on emittance exchange by the space charge driven ‘‘Montague resonance'' have been obtained at the CERN Proton Synchrotron in 2002-04 as a function of the working point. These data are used to advance the benchmarking of major simulation codes (ACCSIM, IMPACT, MICROMAP, ORBIT, SIMBAD, SIMPSONS, SYNERGIA) currently employed world-wide in the design or performance improvement of high intensity circular accelerators. In this paper we summarize the experimental findings and compare them with the first three steps of simulation results of the still progressing work.

  19. Particle-in-Cell Code BEAMPATH for Beam Dynamics Simulations in Linear Accelerators and Beamlines

    Energy Technology Data Exchange (ETDEWEB)

    Batygin, Y.

    2004-10-28

    A code library BEAMPATH for 2 - dimensional and 3 - dimensional space charge dominated beam dynamics study in linear particle accelerators and beam transport lines is developed. The program is used for particle-in-cell simulation of axial-symmetric, quadrupole-symmetric and z-uniform beams in a channel containing RF gaps, radio-frequency quadrupoles, multipole lenses, solenoids and bending magnets. The programming method includes hierarchical program design using program-independent modules and a flexible combination of modules to provide the most effective version of the structure for every specific case of simulation. Numerical techniques as well as the results of beam dynamics studies are presented.

  20. Simulation of atmosphere stratification in the HDR test facility with the CONTAIN code

    International Nuclear Information System (INIS)

    Skerlavaj, A.; Mavko, B.; Kljenak, I.

    2001-01-01

    The test E11.2 'Hydrogen distribution in loop flow geometry', which was performed in the Heissdampf Reaktor containment test facility in Germany, was simulated with the CONTAIN computer code. The predicted pressure history and thermal stratification are in relatively good agreement with the measurements. The compositional stratification within the containment was qualitatively well predicted, although the degree of the stratification in the dome area was slightly underestimated. The analysis of simulation results enabled a better understanding of the physical phenomena during the test.(author)

  1. 3D and 4D Simulations of the Dynamics of the Radiation Belts using VERB code

    Science.gov (United States)

    Shprits, Yuri; Kellerman, Adam; Drozdov, Alexander; Orlova, Ksenia

    2015-04-01

    Modeling and understanding of ring current and higher energy radiation belts has been a grand challenge since the beginning of the space age. In this study we show long term simulations with a 3D VERB code of modeling the radiation belts with boundary conditions derived from observations around geosynchronous orbit. We also present 4D VERB simulations that include convective transport, radial diffusion, pitch angle scattering and local acceleration. We show that while lower energy radial transport is dominated by the convection and higher energy transport is dominated by the diffusive radial transport. We also show there exists an intermediate range of energies for electrons for which both processes work simultaneously.

  2. Simulation of nonlinear propagation of biomedical ultrasound using PZFlex and the KZK Texas code

    Energy Technology Data Exchange (ETDEWEB)

    Qiao, Shan, E-mail: shan.qiao@eng.ox.ac.uk; Jackson, Edward; Coussios, Constantin-C; Cleveland, Robin [Institute of Biomedical Engineering, Department of Engineering Science, University of Oxford, Oxford (United Kingdom)

    2015-10-28

    In biomedical ultrasound nonlinear acoustics can be important in both diagnostic and therapeutic applications and robust simulations tools are needed in the design process but also for day-to-day use such as treatment planning. For most biomedical application the ultrasound sources generate focused sound beams of finite amplitude. The KZK equation is a common model as it accounts for nonlinearity, absorption and paraxial diffraction and there are a number of solvers available, primarily developed by research groups. We compare the predictions of the KZK Texas code (a finite-difference time-domain algorithm) to an FEM-based commercial software, PZFlex. PZFlex solves the continuity equation and momentum conservation equation with a correction for nonlinearity in the equation of state incorporated using an incrementally linear, 2nd order accurate, explicit algorithm in time domain. Nonlinear ultrasound beams from two transducers driven at 1 MHz and 3.3 MHz respectively were simulated by both the KZK Texas code and PZFlex, and the pressure field was also measured by a fibre-optic hydrophone to validate the models. Further simulations were carried out a wide range of frequencies. The comparisons showed good agreement for the fundamental frequency for PZFlex, the KZK Texas code and the experiments. For the harmonic components, the KZK Texas code was in good agreement with measurements but PZFlex underestimated the amplitude: 32% for the 2nd harmonic and 66% for the 3rd harmonic. The underestimation of harmonics by PZFlex was more significant when the fundamental frequency increased. Furthermore non-physical oscillations in the axial profile of harmonics occurred in the PZFlex results when the amplitudes were relatively low. These results suggest that careful benchmarking of nonlinear simulations is important.

  3. Implementation of generalized perturbation theory into the 3-D nodal code SIMULATE

    International Nuclear Information System (INIS)

    Bowman, S.M.; Williams, M.L.; Dodds, H.L.

    1980-01-01

    Determining the effects of changes in design and data parameters upon reactor performance can be very expensive when many perturbations must be considered for a complex system. In order to substantially reduce the cost involved in calculations such as these, generalized perturbation theory (GPT) capability has been implemented into the 3-D LWR nodal reactor analysis code SIMULATE. This capability makes 3-D sensitivity analysis of a realistic LWR practical for the first time. Applications to design analysis will be discussed

  4. 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

  5. A quick and easy improvement of Monte Carlo codes for simulation

    Science.gov (United States)

    Lebrere, A.; Talhi, R.; Tripathy, M.; Pyée, M.

    The simulation of trials of independent random variables of given distribution is a critical element of running Monte-Carlo codes. This is usually performed by using pseudo-random number generators (and in most cases linearcongruential ones). We present here an alternative way to generate sequences with given statistical properties. This sequences are purely deterministic and are given by closed formulae, and can give in some cases better results than classical generators.

  6. Penelope-2006: a code system for Monte Carlo simulation of electron and photon transport

    International Nuclear Information System (INIS)

    2006-01-01

    The computer code system PENELOPE (version 2006) performs Monte Carlo simulation of coupled electron-photon transport in arbitrary materials for a wide energy range, from a few hundred eV to about 1 GeV. Photon transport is simulated by means of the standard, detailed simulation scheme. Electron and positron histories are generated on the basis of a mixed procedure, which combines detailed simulation of hard events with condensed simulation of soft interactions. A geometry package called PENGEOM permits the generation of random electron-photon showers in material systems consisting of homogeneous bodies limited by quadric surfaces, i.e. planes, spheres, cylinders, etc. This report is intended not only to serve as a manual of the PENELOPE code system, but also to provide the user with the necessary information to understand the details of the Monte Carlo algorithm. These proceedings contain the corresponding manual and teaching notes of the PENELOPE-2006 workshop and training course, held on 4-7 July 2006 in Barcelona, Spain. (author)

  7. YT: A Multi-Code Analysis Toolkit for Astrophysical Simulation Data

    Energy Technology Data Exchange (ETDEWEB)

    Turk, Matthew J.; /San Diego, CASS; Smith, Britton D.; /Michigan State U.; Oishi, Jeffrey S.; /KIPAC, Menlo Park /Stanford U., Phys. Dept.; Skory, Stephen; Skillman, Samuel W.; /Colorado U., CASA; Abel, Tom; /KIPAC, Menlo Park /Stanford U., Phys. Dept.; Norman, Michael L.; /aff San Diego, CASS

    2011-06-23

    The analysis of complex multiphysics astrophysical simulations presents a unique and rapidly growing set of challenges: reproducibility, parallelization, and vast increases in data size and complexity chief among them. In order to meet these challenges, and in order to open up new avenues for collaboration between users of multiple simulation platforms, we present yt (available at http://yt.enzotools.org/) an open source, community-developed astrophysical analysis and visualization toolkit. Analysis and visualization with yt are oriented around physically relevant quantities rather than quantities native to astrophysical simulation codes. While originally designed for handling Enzo's structure adaptive mesh refinement data, yt has been extended to work with several different simulation methods and simulation codes including Orion, RAMSES, and FLASH. We report on its methods for reading, handling, and visualizing data, including projections, multivariate volume rendering, multi-dimensional histograms, halo finding, light cone generation, and topologically connected isocontour identification. Furthermore, we discuss the underlying algorithms yt uses for processing and visualizing data, and its mechanisms for parallelization of analysis tasks.

  8. HELIOS/DRAGON/NESTLE codes' simulation of the Gentilly-2 loss of class 4 power event

    International Nuclear Information System (INIS)

    Sarsour, H.N.; Turinsky, P.J.; Rahnema, F.; Mosher, S.; Serghiuta, D.; Marleau, G.; Courau, T.

    2002-01-01

    A loss of electrical power occurred at Gentilly-2 in September of 1995 while the station was operating at full power. There was an unexpectedly rapid core power increase initiated by the drainage of the zone controllers and accelerated by coolant boiling. The core transient was terminated by Shutdown System No 1 (SDS1) tripping when the out-of-core ion chambers exceeded the 10%/sec high rate of power increase trip setpoint at 1.29 sec. This resulted in the station automatically shutting down within 2 sec of event initiation. In the first 2 sec, 26 of the 58 SDS1 and SDS2 in-core flux detectors reached there overpower trip (ROPT) setpoints. The peak reactor power reached approximately 110%FP. Reference 1 presented detailed results of the simulations performed with coupled thermalhydraulics and 3D neutron kinetics codes, SOPHT-G2 and the CERBERUS module of RFSP, and the various adjustments of these codes and plant representation that were needed to obtain the neutronic response observed in 1995. The purposes of this paper are to contrast a simulation prediction of the peak prompt core thermal power transient versus experimental estimate, and to note the impact of spatial discretization approach utilized on the prompt core thermal power transient and the channel power distribution as a function of time. In addition, adequacy of the time-step sizes employed and sensitivity to core's transient thermal-hydraulics conditions are studied. The work presented in this paper has been performed as part of a project sponsored by the Canadian Nuclear Safety Commission (CNSC). The purpose of the project was to gather information and assess the accuracy of best estimate methods using calculation methods and codes developed independently from the CANDU industry. The simulation of the accident was completed using the NESTLE core simulator, employing cross sections generated by the HELIOS lattice physics code, and incremental cross sections generated by the DRAGON lattice physics code

  9. TERRA: a computer code for simulating the transport of environmentally released radionuclides through agriculture

    Energy Technology Data Exchange (ETDEWEB)

    Baes, C.F. III; Sharp, R.D.; Sjoreen, A.L.; Hermann, O.W.

    1984-11-01

    TERRA is a computer code which calculates concentrations of radionuclides and ingrowing daughters in surface and root-zone soil, produce and feed, beef, and milk from a given deposition rate at any location in the conterminous United States. The code is fully integrated with seven other computer codes which together comprise a Computerized Radiological Risk Investigation System, CRRIS. Output from either the long range (> 100 km) atmospheric dispersion code RETADD-II or the short range (<80 km) atmospheric dispersion code ANEMOS, in the form of radionuclide air concentrations and ground deposition rates by downwind location, serves as input to TERRA. User-defined deposition rates and air concentrations may also be provided as input to TERRA through use of the PRIMUS computer code. The environmental concentrations of radionuclides predicted by TERRA serve as input to the ANDROS computer code which calculates population and individual intakes, exposures, doses, and risks. TERRA incorporates models to calculate uptake from soil and atmospheric deposition on four groups of produce for human consumption and four groups of livestock feeds. During the environmental transport simulation, intermediate calculations of interception fraction for leafy vegetables, produce directly exposed to atmospherically depositing material, pasture, hay, and silage are made based on location-specific estimates of standing crop biomass. Pasture productivity is estimated by a model which considers the number and types of cattle and sheep, pasture area, and annual production of other forages (hay and silage) at a given location. Calculations are made of the fraction of grain imported from outside the assessment area. TERRA output includes the above calculations and estimated radionuclide concentrations in plant produce, milk, and a beef composite by location.

  10. TERRA: a computer code for simulating the transport of environmentally released radionuclides through agriculture

    International Nuclear Information System (INIS)

    Baes, C.F. III; Sharp, R.D.; Sjoreen, A.L.; Hermann, O.W.

    1984-11-01

    TERRA is a computer code which calculates concentrations of radionuclides and ingrowing daughters in surface and root-zone soil, produce and feed, beef, and milk from a given deposition rate at any location in the conterminous United States. The code is fully integrated with seven other computer codes which together comprise a Computerized Radiological Risk Investigation System, CRRIS. Output from either the long range (> 100 km) atmospheric dispersion code RETADD-II or the short range (<80 km) atmospheric dispersion code ANEMOS, in the form of radionuclide air concentrations and ground deposition rates by downwind location, serves as input to TERRA. User-defined deposition rates and air concentrations may also be provided as input to TERRA through use of the PRIMUS computer code. The environmental concentrations of radionuclides predicted by TERRA serve as input to the ANDROS computer code which calculates population and individual intakes, exposures, doses, and risks. TERRA incorporates models to calculate uptake from soil and atmospheric deposition on four groups of produce for human consumption and four groups of livestock feeds. During the environmental transport simulation, intermediate calculations of interception fraction for leafy vegetables, produce directly exposed to atmospherically depositing material, pasture, hay, and silage are made based on location-specific estimates of standing crop biomass. Pasture productivity is estimated by a model which considers the number and types of cattle and sheep, pasture area, and annual production of other forages (hay and silage) at a given location. Calculations are made of the fraction of grain imported from outside the assessment area. TERRA output includes the above calculations and estimated radionuclide concentrations in plant produce, milk, and a beef composite by location

  11. Development of simulation code for MOX dissolution using silver-mediated electrochemical method (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Kida, Takashi; Umeda, Miki; Sugikawa, Susumu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    MOX dissolution using silver-mediated electrochemical method will be employed for the preparation of plutonium nitrate solution in the criticality safety experiments in the Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF). A simulation code for the MOX dissolution has been developed for the operating support. The present report describes the outline of the simulation code, a comparison with the experimental data and a parameter study on the MOX dissolution. The principle of this code is based on the Zundelevich's model for PuO{sub 2} dissolution using Ag(II). The influence of nitrous acid on the material balance of Ag(II) is taken into consideration and the surface area of MOX powder is evaluated by particle size distribution in this model. The comparison with experimental data was carried out to confirm the validity of this model. It was confirmed that the behavior of MOX dissolution could adequately be simulated using an appropriate MOX dissolution rate constant. It was found from the result of parameter studies that MOX particle size was major governing factor on the dissolution rate. (author)

  12. Design and implementation of a software tool intended for simulation and test of real time codes

    International Nuclear Information System (INIS)

    Le Louarn, C.

    1986-09-01

    The objective of real time software testing is to show off processing errors and unobserved functional requirements or timing constraints in a code. In the perspective of safety analysis of nuclear equipments of power plants testing should be carried independently from the physical process (which is not generally available), and because casual hardware failures must be considered. We propose here a simulation and test tool, integrally software, with large interactive possibilities for testing assembly code running on microprocessor. The OST (outil d'aide a la simulation et au Test de logiciels temps reel) simulates code execution and hardware or software environment behaviour. Test execution is closely monitored and many useful informations are automatically saved. The present thesis work details, after exposing methods and tools dedicated to real time software, the OST system. We show the internal mechanisms and objects of the system: particularly ''events'' (which describe evolutions of the system under test) and mnemonics (which describe the variables). Then, we detail the interactive means available to the user for constructing the test data and the environment of the tested software. Finally, a prototype implementation is presented along with the results of the tests carried out. This demonstrates the many advantages of the use of an automatic tool over a manual investigation. As a conclusion, further developments, nececessary to complete the final tool are rewieved [fr

  13. TEMPEST code modifications and testing for erosion-resisting sludge simulations

    International Nuclear Information System (INIS)

    Onishi, Y.; Trent, D.S.

    1998-01-01

    The TEMPEST computer code has been used to address many waste retrieval operational and safety questions regarding waste mobilization, mixing, and gas retention. Because the amount of sludge retrieved from the tank is directly related to the sludge yield strength and the shear stress acting upon it, it is important to incorporate the sludge yield strength into simulations of erosion-resisting tank waste retrieval operations. This report describes current efforts to modify the TEMPEST code to simulate pump jet mixing of erosion-resisting tank wastes and the models used to test for erosion of waste sludge with yield strength. Test results for solid deposition and diluent/slurry jet injection into sludge layers in simplified tank conditions show that the modified TEMPEST code has a basic ability to simulate both the mobility and immobility of the sludges with yield strength. Further testing, modification, calibration, and verification of the sludge mobilization/immobilization model are planned using erosion data as they apply to waste tank sludges

  14. A Comparison of Nuclear Power Plant Simulator with RELAP5/MOD3 code about Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    Kim, Sung Hyun; Moon, Chan Ki; Park, Sung Baek; Na, Man Gyun

    2013-01-01

    The RELAP5/MOD3 code introduced in cooperation with U. S. NRC has been utilized mainly for validation calculation of accident analysis submitted by licensee in Korea. The Korea Institute of Nuclear Safety has built a verification system of LWR accident analysis with RELAP5/MOD3 code engine. Therefore, the simulator replicates the design basis accident and its results are compared with RELAP5/MOD3 code results that will have important implications in the verification of the simulator in the future. The SGTR simulations were performed by the simulator and its results were compared with ones by RELAP5/MOD3 code in this study. Thus, the results of this study can be used as materials to build the verification system of the nuclear power plant simulator. We tried to compare with RELAP5/MOD3 verification code by replicating major parameters of steam generator tube rupture using the simulator for OPR-1000 in Yonggwang training center. By comparing the changes in temperature, pressure and inventory of the reactor coolant system and main steam system during the SGTR, it was confirmed that the main behaviors of SGTR which the simulator and RELAP5/MOD3 code showed are similar. However, the behavior of SG pressure and level that are important parameters to diagnose the accident were a little different. We estimated that RELAP5/MOD3 code was not reflected the major control systems in detail, such as FWCS, SBCS and PPCS. The different behaviors of SG level and pressure in this study should be needed an additional review. As a result of the comparison, the major simulation parameters behavior by RELAP5/MOD3 code agreed well with the one by the simulator. Therefore, it is thought that RELAP5/MOD3 code is used as a tool for validation of NPP simulator in the near future through this study

  15. Simulation of the loss of residual heat removal system of an integral test facility using French computer code CATHARE

    International Nuclear Information System (INIS)

    Troshko, A.A.; Hassan, Y.A.

    1997-01-01

    Thermal hydraulic CATHARE V1.3U code has been used to simulate an International Standard Problem (ISP38) experiment conducted at BETHSY Integral Test Facility located in Grenoble, France. This experiment dealt with simulation of the loss of Residual Heat Removal System (RHRS) during midloop operation. Overall, the code's prediction and experimental data were found to be in a reasonable qualitative agreement. However, the code underestimated the time of the core uncover and the actuation of the gravity feed injection. It was found that the code's model of the upward tee junction needs to be refined for the low pressure ranges

  16. mocca code for star cluster simulations - VI. Bimodal spatial distribution of blue stragglers

    Science.gov (United States)

    Hypki, Arkadiusz; Giersz, Mirek

    2017-11-01

    The paper presents an analysis of formation mechanism and properties of spatial distributions of blue stragglers in evolving globular clusters, based on numerical simulations done with the mocca code. First, there are presented N-body and mocca simulations which try to reproduce the simulations presented by Ferraro et al. (2012). Then, we show the agreement between N-body and the mocca code. Finally, we discuss the formation process of the bimodal distribution. We report that we could not reproduce simulations from Ferraro et al. (2012). Moreover, we show that the so-called bimodal spatial distribution of blue stragglers is a very transient feature. It is formed for one snapshot in time and it can easily vanish in the next one. Moreover, we show that the radius of avoidance proposed by Ferraro et al. (2012) goes out of sync with the apparent minimum of the bimodal distribution after about two half-mass relaxation times (without finding out what is the reason for that). This finding creates a real challenge for the dynamical clock, which uses this radius to determine the dynamical age of globular clusters. Additionally, the paper discusses a few important problems concerning the apparent visibilities of the bimodal distributions, which have to be taken into account while studying the spatial distributions of blue stragglers.

  17. Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.

    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

  18. Simulations of corrosion product transfer with the OSCAR V1.2 code

    International Nuclear Information System (INIS)

    Dacquait, F.; Francescatto, J.; Broutin, F.; Genin, J.B.; Benier, G.; Girard, M.; You, D.; Ranchoux, G.; Bonnefon, J.; Bachet, M.; Riot, G.

    2012-09-01

    Activated Corrosion Products (ACPs) generate a radiation field in PWRs, which is the major contributor to the dose absorbed by nuclear power plant staff working during shutdown operations and maintenance. Therefore, a thorough understanding of the mechanisms that control the corrosion product transfer is of the highest importance. Since the 1970's, the R and D strategy in France has been based on experiments in test loops representative of PWR conditions, on in-situ gamma spectrometry measurements of the PWR primary system contamination and on simulation code development. The simulation of corrosion product transfers in PWR primary circuits is a major challenge since it involves many physical and chemical phenomena including: corrosion, dissolution, precipitation, erosion, deposition, convection, activation... In addition to the intrinsic difficulty of multi-physics modelling, the primary systems present severe operating conditions (300 deg. C, 150 bar, neutron flux, fluid velocity up to 15 m.s -1 and very low corrosion product concentrations). The purpose of the OSCAR code, developed by the CEA in cooperation with EDF and AREVA NP, is to predict the PWR primary system contamination by corrosion and fission products. The OSCAR code is considered to be not only a tool for numerical simulations and predictions (operational practices improvements and new-built PWRs design) but also one that might combine and organise all new knowledge useful to progress on contamination. The OSCAR code for Products of Corrosion, OSCAR PC, allows researchers to analyse the corrosion product behaviour and to calculate the ACP volume and surface activities of the primary and auxiliary systems. In the new version, OSCAR PC V1.2, the corrosion product transfer in the particulate form is enhanced and a new feature is the possibility to simulate cold shutdowns. In order to validate this version, the contamination transfer has been simulated in 5 French PWRs with different operating and

  19. Comparison of Geant4-DNA simulation of S-values with other Monte Carlo codes

    Energy Technology Data Exchange (ETDEWEB)

    André, T. [Université Bordeaux 1, CENBG, UMR 5797, F-33170 Gradignan (France); CNRS, IN2P3, CENBG, UMR 5797, F-33170 Gradignan (France); Morini, F. [Research Group of Theoretical Chemistry and Molecular Modelling, Hasselt University, Agoralaan Gebouw D, B-3590 Diepenbeek (Belgium); Karamitros, M. [Université Bordeaux 1, CENBG, UMR 5797, F-33170 Gradignan (France); CNRS, IN2P3, CENBG, UMR 5797, F-33170 Gradignan (France); CNRS, INCIA, UMR 5287, F-33400 Talence (France); Delorme, R. [LPSC, Université Joseph Fourier Grenoble 1, CNRS/IN2P3, Grenoble INP, 38026 Grenoble (France); CEA, LIST, F-91191 Gif-sur-Yvette (France); Le Loirec, C. [CEA, LIST, F-91191 Gif-sur-Yvette (France); Campos, L. [Departamento de Física, Universidade Federal de Sergipe, São Cristóvão (Brazil); Champion, C. [Université Bordeaux 1, CENBG, UMR 5797, F-33170 Gradignan (France); CNRS, IN2P3, CENBG, UMR 5797, F-33170 Gradignan (France); Groetz, J.-E.; Fromm, M. [Université de Franche-Comté, Laboratoire Chrono-Environnement, UMR CNRS 6249, Besançon (France); Bordage, M.-C. [Laboratoire Plasmas et Conversion d’Énergie, UMR 5213 CNRS-INPT-UPS, Université Paul Sabatier, Toulouse (France); Perrot, Y. [Laboratoire de Physique Corpusculaire, UMR 6533, Aubière (France); Barberet, Ph. [Université Bordeaux 1, CENBG, UMR 5797, F-33170 Gradignan (France); CNRS, IN2P3, CENBG, UMR 5797, F-33170 Gradignan (France); and others

    2014-01-15

    Monte Carlo simulations of S-values have been carried out with the Geant4-DNA extension of the Geant4 toolkit. The S-values have been simulated for monoenergetic electrons with energies ranging from 0.1 keV up to 20 keV, in liquid water spheres (for four radii, chosen between 10 nm and 1 μm), and for electrons emitted by five isotopes of iodine (131, 132, 133, 134 and 135), in liquid water spheres of varying radius (from 15 μm up to 250 μm). The results have been compared to those obtained from other Monte Carlo codes and from other published data. The use of the Kolmogorov–Smirnov test has allowed confirming the statistical compatibility of all simulation results.

  20. Simulations of the Dynamics of the Coupled Energetic and Relativistic Electrons Using VERB Code

    Science.gov (United States)

    Shprits, Y.; Kellerman, A. C.; Drozdov, A.

    2015-12-01

    Modeling and understanding of ring current and radiation belt coupled system has been a grand challenge since the beginning of the space age. In this study we show long term simulations with a 3D VERB code of modeling the radiation belts with boundary conditions derived from observations around geosynchronous orbit. We also present 4D VERB simulations that include convective transport, radial diffusion, pitch angle scattering and local acceleration. VERB simulations show that the lower energy inward transport is dominated by the convection and higher energy transport is dominated by the diffusive radial transport. We also show that at energies of 100s of keV a number of processes work simultaneously including convective transport, radial diffusion, local acceleration, loss to the loss cone and loss to the magnetopause. The results of the simulaiton of March 2013 storm are compared with Van Allen Probes observations.

  1. Simulation of single-phase rod bundle flow. Comparison between CFD-code ESTET, PWR core code THYC and experimental results

    International Nuclear Information System (INIS)

    Mur, J.; Larrauri, D.

    1998-07-01

    Computer simulation of flow in configurations close to pressurized water reactor (PWR) geometry is of great interest for Electricite de France (EDF). Although simulation of the flow through a whole PWR core with an all purpose CFD-code is not yet achievable, such a tool cna be quite useful to perform numerical experiments in order to try and improve the modeling introduced in computer codes devoted to reactor core thermal-hydraulic analysis. Further to simulation in small bare rod bundle configurations, the present study is focused on the simulation, with CFD-code ESTET and PWR core code THYC, of the flow in the experimental configuration VATICAN-1. ESTET simulation results are compared on the one hand to local velocity and concentration measurements, on the other hand with subchannel averaged values calculated by THYC. As far as the comparison with measurements is concerned, ESTET results are quite satisfactory relatively to available experimental data and their uncertainties. The effect of spacer grids and the prediction of the evolution of an unbalanced velocity profile seem to be correctly treated. As far as the comparison with THYC subchannel averaged values is concerned, the difficulty of a direct comparison between subchannel averaged and local values is pointed out. ESTET calculated local values are close to experimental local values. ESTET subchannel averaged values are also close to THYC calculation results. Thus, THYC results are satisfactory whereas their direct comparison to local measurements could show some disagreement. (author)

  2. Simulation of natural convection cooling phenomena for research reactors using the code PARET

    International Nuclear Information System (INIS)

    Hainoun, A.; Al-Habit, E.

    2006-01-01

    This study deals with testing the capacity of the code PARET to simulate natural circulation phenomena under different boundary conditions in addition to assessment of some new options related to simulation of control rod movement and the reactivity effect of thermal expansion fuel elements. the experiments of the simple thermal hydraulic loop of Missouri University about natural circulation phenomena in narrow parallel channel were used to validate the code. The results indicate good agreements regarding the evolution of coolant velocity and clad temperature. In particular the heat transfer coefficient of natural convection has been calculated in good agreement with the experiment. On the other hand, the core of MNSR reactor has been modelled to stimulate the reactor dynamic behaviour under natural circulation condition for different initial power level. The observed oscillations during the initial phase vanish gradually with passing time. In this context three experiment of step reactivity insertion were calculated using two different options of boundary conditions, either using initial velocity or pressure drop along the core. The results indicate good agreement with the experiments regarding the evolution of relative power. The validations included also sensitivity analysis against some important parameters like initial velocity and radial distance of fuel rod. The new option for simulation of control rod movement was also tested. For this purpose the MNSR experiment of all control rod withdraw was selected. This means control rod velocity was estimated using experimental measurement. The simulation result of relative power evolution shows good agreement with the experiment during the first phase of the transient. However, an increased deviation is observed in the following phase due to the effect of closed hydrodynamics loop, which can be modelled with the code PARET. (Authors)

  3. Enhancing the ABAQUS thermomechanics code to simulate multipellet steady and transient LWR fuel rod behavior

    International Nuclear Information System (INIS)

    Williamson, R.L.

    2011-01-01

    Highlights: → The ABAQUS thermomechanics code is enhanced to enable simulation of nuclear fuel behavior. → Comparisons are made between discrete and smeared fuel pellet analysis. → Multidimensional and multipellet analysis is important for accurate prediction of PCMI. → Fully coupled thermomechanics results in very smooth prediction of fuel-clad gap closure. → A smeared-pellet approximation results in significant underprediction of clad radial displacements and plastic strain. - Abstract: A powerful multidimensional fuels performance analysis capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO 2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth, gap heat transfer, and gap/plenum gas behavior during irradiation. This new capability is demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multipellet fuel rod, during both steady and transient operation. Comparisons are made between discrete and smeared-pellet simulations. Computational results demonstrate the importance of a multidimensional, multipellet, fully-coupled thermomechanical approach. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermomechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.

  4. LOLA SYSTEM: A code block for nodal PWR simulation. Part. II - MELON-3, CONCON and CONAXI Codes

    International Nuclear Information System (INIS)

    Aragones, J. M.; Ahnert, C.; Gomez Santamaria, J.; Rodriguez Olabarria, I.

    1985-01-01

    Description of the theory and users manual of the MELON-3, CONCON and CONAXI codes, which are part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. These auxiliary codes, provide some of the input data for the main module SIMULA-3; these are, the reactivity correlations constants, the albe does and the transport factors. (Author) 7 refs

  5. LOLA SYSTEM: A code block for nodal PWR simulation. Part. II - MELON-3, CONCON and CONAXI Codes

    Energy Technology Data Exchange (ETDEWEB)

    Aragones, J. M.; Ahnert, C.; Gomez Santamaria, J.; Rodriguez Olabarria, I.

    1985-07-01

    Description of the theory and users manual of the MELON-3, CONCON and CONAXI codes, which are part of the core calculation system by nodal theory in one group, called LOLA SYSTEM. These auxiliary codes, provide some of the input data for the main module SIMULA-3; these are, the reactivity correlations constants, the albe does and the transport factors. (Author) 7 refs.

  6. Supercomputing with TOUGH2 family codes for coupled multi-physics simulations of geologic carbon sequestration

    Science.gov (United States)

    Yamamoto, H.; Nakajima, K.; Zhang, K.; Nanai, S.

    2015-12-01

    Powerful numerical codes that are capable of modeling complex coupled processes of physics and chemistry have been developed for predicting the fate of CO2 in reservoirs as well as its potential impacts on groundwater and subsurface environments. However, they are often computationally demanding for solving highly non-linear models in sufficient spatial and temporal resolutions. Geological heterogeneity and uncertainties further increase the challenges in modeling works. Two-phase flow simulations in heterogeneous media usually require much longer computational time than that in homogeneous media. Uncertainties in reservoir properties may necessitate stochastic simulations with multiple realizations. Recently, massively parallel supercomputers with more than thousands of processors become available in scientific and engineering communities. Such supercomputers may attract attentions from geoscientist and reservoir engineers for solving the large and non-linear models in higher resolutions within a reasonable time. However, for making it a useful tool, it is essential to tackle several practical obstacles to utilize large number of processors effectively for general-purpose reservoir simulators. We have implemented massively-parallel versions of two TOUGH2 family codes (a multi-phase flow simulator TOUGH2 and a chemically reactive transport simulator TOUGHREACT) on two different types (vector- and scalar-type) of supercomputers with a thousand to tens of thousands of processors. After completing implementation and extensive tune-up on the supercomputers, the computational performance was measured for three simulations with multi-million grid models, including a simulation of the dissolution-diffusion-convection process that requires high spatial and temporal resolutions to simulate the growth of small convective fingers of CO2-dissolved water to larger ones in a reservoir scale. The performance measurement confirmed that the both simulators exhibit excellent

  7. Look at nuclear artillery yield options using JANUS, a wargame simulation code

    International Nuclear Information System (INIS)

    Andre, C.G.

    1982-01-01

    JANUS, a two-sided, interactive wargame simulation code, was used to explore how using each of several different yield options in a nuclear artillery shell might affect a tactical battlefield simulation. In a general sense, the results or outcomes of these simulations support the results or outcomes of previous studies. In these simulations the Red player knew of the anticipated nuclear capability of the Blue player. Neither side experienced a decisive win over the other, and both continued fighting and experienced losses that, under most historical circumstances, would have been termed unacceptable - that is, something else would have happened (the attack would have been called off). During play, each side had only fragmentary knowledge of the remaining resources on the other side - thus each side desired to continue fighting on the basis of known information. We found that the anticipated use of nuclear weapons by either side affects the character of a game significantly and that, if the employment of nuclear weapons is to have a decided effect on the progress and outcome of a battle, each side will have to have an adequate number of nuclear weapons. In almost all the simulations we ran using JANUS, enhanced radiation (ER) weapons were more effective than 1-kt fission weapons in imposing overall losses on Red. The typical visibility in the JANUS simulation limited each side's ability to acquire units deep into enemy territory and so the 10-kt fission weapon was not useful against enemy tanks that were not engaged in battle

  8. CASINO, a code for simulation of charged particles in an axisymmetric Tokamak

    International Nuclear Information System (INIS)

    Dillner, Oe.

    1992-01-01

    The present report comprises a documentation of CASINO, a simulation code developed as a means for the study of high energy charged particles in an axisymmetric Tokamak. The background of the need for such a numerical tool is presented. In the description of the numerical model used for the orbit integration, the method using constants of motion, the Lao-Hirsman geometry for the flux surfaces and a method for reducing the necessary number of particles is elucidated. A brief outline of the calculational sequence is given as a flow chart. The essential routines and functions as well as the common blocks are briefly described. The input and output routines are shown. Finally the documentation is completed by a short discussion of possible extensions of the code and a test case. (au)

  9. Simulation of the MHD stabilities of the experiment on HL-2A tokamak by GATO code

    International Nuclear Information System (INIS)

    Pan Wei; Chen Liaoyuan; Dong Jiaqi; Shen Yong; Zhang Jinhua

    2009-01-01

    The ideal two-dimensional MHD stabilities code, GATO, has been successfully immigrated to the high-performance computing system of HL-2A and used to the simulation study of the ideal MHD stabilities of the plasmas produced by one of the pellets injection experiments on HL-2A tokamak. The EFIT code was used to reconstruct the equilibrium configures firstly and the GATO was used to compute their MHD stabilities secondly whose source data were obtained by the NO.4050 discharge of the experiments on HL-2A, and finally by analyzing these results the preliminary conclusion was devised that the confinement performance of the plasma was improved because of the stabilization effect of the anti-sheared configures created by the pellets injection. (authors)

  10. Preface: Research advances in vadose zone hydrology through simulations with the TOUGH codes

    International Nuclear Information System (INIS)

    Finsterle, Stefan; Oldenburg, Curtis M.

    2004-01-01

    Numerical simulators are playing an increasingly important role in advancing our fundamental understanding of hydrological systems. They are indispensable tools for managing groundwater resources, analyzing proposed and actual remediation activities at contaminated sites, optimizing recovery of oil, gas, and geothermal energy, evaluating subsurface structures and mining activities, designing monitoring systems, assessing the long-term impacts of chemical and nuclear waste disposal, and devising improved irrigation and drainage practices in agricultural areas, among many other applications. The complexity of subsurface hydrology in the vadose zone calls for sophisticated modeling codes capable of handling the strong nonlinearities involved, the interactions of coupled physical, chemical and biological processes, and the multiscale heterogeneities inherent in such systems. The papers in this special section of ''Vadose Zone Journal'' are illustrative of the enormous potential of such numerical simulators as applied to the vadose zone. The papers describe recent developments and applications of one particular set of codes, the TOUGH family of codes, as applied to nonisothermal flow and transport in heterogeneous porous and fractured media (http://www-esd.lbl.gov/TOUGH2). The contributions were selected from presentations given at the TOUGH Symposium 2003, which brought together developers and users of the TOUGH codes at the Lawrence Berkeley National Laboratory (LBNL) in Berkeley, California, for three days of information exchange in May 2003 (http://www-esd.lbl.gov/TOUGHsymposium). The papers presented at the symposium covered a wide range of topics, including geothermal reservoir engineering, fracture flow and vadose zone hydrology, nuclear waste disposal, mining engineering, reactive chemical transport, environmental remediation, and gas transport. This Special Section of ''Vadose Zone Journal'' contains revised and expanded versions of selected papers from the

  11. Premar-2: a Monte Carlo code for radiative transport simulation in atmospheric environments

    Energy Technology Data Exchange (ETDEWEB)

    Cupini, E. [ENEA, Centro Ricerche Ezio Clementel, Bologna, (Italy). Dipt. Innovazione

    1999-07-01

    The peculiarities of the PREMAR-2 code, aimed at radiation transport Monte Carlo simulation in atmospheric environments in the infrared-ultraviolet frequency range, are described. With respect to the previously developed PREMAR code, besides plane multilayers, spherical multilayers and finite sequences of vertical layers, each one with its own atmospheric behaviour, are foreseen in the new code, together with the refraction phenomenon, so that long range, highly slanted paths can now be more faithfully taken into account. A zenithal angular dependence of the albedo coefficient has moreover been introduced. Lidar systems, with spatially independent source and telescope, are allowed again to be simulated, and, in this latest version of the code, sensitivity analyses to be performed. According to this last feasibility, consequences on radiation transport of small perturbations in physical components of the atmospheric environment may be analyze and the related effects on searched results estimated. The availability of a library of physical data (reaction coefficients, phase functions and refraction indexes) is required by the code, providing the essential features of the environment of interest needed of the Monte Carlo simulation. Variance reducing techniques have been enhanced in the Premar-2 code, by introducing, for instance, a local forced collision technique, especially apt to be used in Lidar system simulations. Encouraging comparisons between code and experimental results carried out at the Brasimone Centre of ENEA, have so far been obtained, even if further checks of the code are to be performed. [Italian] Nel presente rapporto vengono descritte le principali caratteristiche del codice di calcolo PREMAR-2, che esegue la simulazione Montecarlo del trasporto della radiazione elettromagnetica nell'atmosfera, nell'intervallo di frequenza che va dall'infrarosso all'ultravioletto. Rispetto al codice PREMAR precedentemente sviluppato, il codice

  12. TRANPZ - A computer code for the simulation of reactors with axial dependence

    International Nuclear Information System (INIS)

    Sampaio, L.C.M.

    1980-12-01

    A computer code was developed to simulate a PWR reactor in steady state and during transients. The solution of one speed diffusion equation in the axial direction is obtained numerically dividing the core in various axial segments and the axial power distribution is obtained there from. A method was developed to determine the transient solution. The external reactivity effects are caused by the motion of the control rods, starting from the steady condition with the control rods in any position. The heat conduction equation in the fuel is numerically solved in the radial direction. Various tests were performed in steady state and transient conditions and the validity of the present model was verified. Results were compared in steady state condition with the code CITATION and a reasonable agreement was found. (E.G.) [pt

  13. Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRACRT

    International Nuclear Information System (INIS)

    Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto

    2011-01-01

    In February 2010 NA-SA (Nucleoelectrica Argentina S.A.) awarded Tecnatom the Atucha II full scope simulator project. NA-SA is a public company owner of the Argentinean nuclear power plants. Atucha II is due to enter in operation shortly. Atucha II NPP is a PHWR type plant cooled by the water of the Parana River and has the same design as the Atucha I unit, doubling its power capacity. Atucha II will produce 745 MWe utilizing heavy water as coolant and moderator, and natural uranium as fuel. A plant singular feature is the permanent core refueling. TRAC R T is the first real time thermal hydraulic six-equations code used in the training simulation industry for NSSS modeling. It is the result from adapting to real time the best estimate code TRACG. TRAC R T is based on first principle conservation equations for mass, energy and momentum for liquid and steam phases, with two phase flows under non homogeneous and non equilibrium conditions. At present, it has been successfully implemented in twelve full scope replica simulators in different training centers throughout the world. To ease the modeling task, TRAC R T includes a graphical pre-processing tool designed to optimize this process and alleviate the burden of entering alpha numerical data in an input file. (author)

  14. Supersonic propulsion simulation by incorporating component models in the large perturbation inlet (LAPIN) computer code

    Science.gov (United States)

    Cole, Gary L.; Richard, Jacques C.

    1991-01-01

    An approach to simulating the internal flows of supersonic propulsion systems is presented. The approach is based on a fairly simple modification of the Large Perturbation Inlet (LAPIN) computer code. LAPIN uses a quasi-one dimensional, inviscid, unsteady formulation of the continuity, momentum, and energy equations. The equations are solved using a shock capturing, finite difference algorithm. The original code, developed for simulating supersonic inlets, includes engineering models of unstart/restart, bleed, bypass, and variable duct geometry, by means of source terms in the equations. The source terms also provide a mechanism for incorporating, with the inlet, propulsion system components such as compressor stages, combustors, and turbine stages. This requires each component to be distributed axially over a number of grid points. Because of the distributed nature of such components, this representation should be more accurate than a lumped parameter model. Components can be modeled by performance map(s), which in turn are used to compute the source terms. The general approach is described. Then, simulation of a compressor/fan stage is discussed to show the approach in detail.

  15. GeNN: a code generation framework for accelerated brain simulations

    Science.gov (United States)

    Yavuz, Esin; Turner, James; Nowotny, Thomas

    2016-01-01

    Large-scale numerical simulations of detailed brain circuit models are important for identifying hypotheses on brain functions and testing their consistency and plausibility. An ongoing challenge for simulating realistic models is, however, computational speed. In this paper, we present the GeNN (GPU-enhanced Neuronal Networks) framework, which aims to facilitate the use of graphics accelerators for computational models of large-scale neuronal networks to address this challenge. GeNN is an open source library that generates code to accelerate the execution of network simulations on NVIDIA GPUs, through a flexible and extensible interface, which does not require in-depth technical knowledge from the users. We present performance benchmarks showing that 200-fold speedup compared to a single core of a CPU can be achieved for a network of one million conductance based Hodgkin-Huxley neurons but that for other models the speedup can differ. GeNN is available for Linux, Mac OS X and Windows platforms. The source code, user manual, tutorials, Wiki, in-depth example projects and all other related information can be found on the project website http://genn-team.github.io/genn/.

  16. Simulation of image formation in x-ray coded aperture microscopy with polycapillary optics.

    Science.gov (United States)

    Korecki, P; Roszczynialski, T P; Sowa, K M

    2015-04-06

    In x-ray coded aperture microscopy with polycapillary optics (XCAMPO), the microstructure of focusing polycapillary optics is used as a coded aperture and enables depth-resolved x-ray imaging at a resolution better than the focal spot dimensions. Improvements in the resolution and development of 3D encoding procedures require a simulation model that can predict the outcome of XCAMPO experiments. In this work we introduce a model of image formation in XCAMPO which enables calculation of XCAMPO datasets for arbitrary positions of the object relative to the focal plane as well as to incorporate optics imperfections. In the model, the exit surface of the optics is treated as a micro-structured x-ray source that illuminates a periodic object. This makes it possible to express the intensity of XCAMPO images as a convolution series and to perform simulations by means of fast Fourier transforms. For non-periodic objects, the model can be applied by enforcing artificial periodicity and setting the spatial period larger then the field-of-view. Simulations are verified by comparison with experimental data.

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

    Science.gov (United States)

    2015-11-01

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

  18. VINE-A NUMERICAL CODE FOR SIMULATING ASTROPHYSICAL SYSTEMS USING PARTICLES. II. IMPLEMENTATION AND PERFORMANCE CHARACTERISTICS

    International Nuclear Information System (INIS)

    Nelson, Andrew F.; Wetzstein, M.; Naab, T.

    2009-01-01

    We continue our presentation of VINE. In this paper, we begin with a description of relevant architectural properties of the serial and shared memory parallel computers on which VINE is intended to run, and describe their influences on the design of the code itself. We continue with a detailed description of a number of optimizations made to the layout of the particle data in memory and to our implementation of a binary tree used to access that data for use in gravitational force calculations and searches for smoothed particle hydrodynamics (SPH) neighbor particles. We describe the modifications to the code necessary to obtain forces efficiently from special purpose 'GRAPE' hardware, the interfaces required to allow transparent substitution of those forces in the code instead of those obtained from the tree, and the modifications necessary to use both tree and GRAPE together as a fused GRAPE/tree combination. We conclude with an extensive series of performance tests, which demonstrate that the code can be run efficiently and without modification in serial on small workstations or in parallel using the OpenMP compiler directives on large-scale, shared memory parallel machines. We analyze the effects of the code optimizations and estimate that they improve its overall performance by more than an order of magnitude over that obtained by many other tree codes. Scaled parallel performance of the gravity and SPH calculations, together the most costly components of most simulations, is nearly linear up to at least 120 processors on moderate sized test problems using the Origin 3000 architecture, and to the maximum machine sizes available to us on several other architectures. At similar accuracy, performance of VINE, used in GRAPE-tree mode, is approximately a factor 2 slower than that of VINE, used in host-only mode. Further optimizations of the GRAPE/host communications could improve the speed by as much as a factor of 3, but have not yet been implemented in VINE

  19. Vine—A Numerical Code for Simulating Astrophysical Systems Using Particles. II. Implementation and Performance Characteristics

    Science.gov (United States)

    Nelson, Andrew F.; Wetzstein, M.; Naab, T.

    2009-10-01

    We continue our presentation of VINE. In this paper, we begin with a description of relevant architectural properties of the serial and shared memory parallel computers on which VINE is intended to run, and describe their influences on the design of the code itself. We continue with a detailed description of a number of optimizations made to the layout of the particle data in memory and to our implementation of a binary tree used to access that data for use in gravitational force calculations and searches for smoothed particle hydrodynamics (SPH) neighbor particles. We describe the modifications to the code necessary to obtain forces efficiently from special purpose "GRAPE" hardware, the interfaces required to allow transparent substitution of those forces in the code instead of those obtained from the tree, and the modifications necessary to use both tree and GRAPE together as a fused GRAPE/tree combination. We conclude with an extensive series of performance tests, which demonstrate that the code can be run efficiently and without modification in serial on small workstations or in parallel using the OpenMP compiler directives on large-scale, shared memory parallel machines. We analyze the effects of the code optimizations and estimate that they improve its overall performance by more than an order of magnitude over that obtained by many other tree codes. Scaled parallel performance of the gravity and SPH calculations, together the most costly components of most simulations, is nearly linear up to at least 120 processors on moderate sized test problems using the Origin 3000 architecture, and to the maximum machine sizes available to us on several other architectures. At similar accuracy, performance of VINE, used in GRAPE-tree mode, is approximately a factor 2 slower than that of VINE, used in host-only mode. Further optimizations of the GRAPE/host communications could improve the speed by as much as a factor of 3, but have not yet been implemented in VINE

  20. Leveraging Quick Response Code Technology to Facilitate Simulation-Based Leaderboard Competition.

    Science.gov (United States)

    Chang, Todd P; Doughty, Cara B; Mitchell, Diana; Rutledge, Chrystal; Auerbach, Marc A; Frisell, Karin; Jani, Priti; Kessler, David O; Wolfe, Heather; MacKinnon, Ralph J; Dewan, Maya; Pirie, Jonathan; Lemke, Daniel; Khattab, Mona; Tofil, Nancy; Nagamuthu, Chenthila; Walsh, Catharine M

    2018-02-01

    Leaderboards provide feedback on relative performance and a competitive atmosphere for both self-guided improvement and social comparison. Because simulation can provide substantial quantitative participant feedback, leaderboards can be used, not only locally but also in a multidepartment, multicenter fashion. Quick Response (QR) codes can be integrated to allow participants to access and upload data. We present the development, implementation, and initial evaluation of an online leaderboard employing principles of gamification using points, badges, and leaderboards designed to enhance competition among healthcare providers. This article details the fundamentals behind the development and implementation of a user-friendly, online, multinational leaderboard that employs principles of gamification to enhance competition and integrates a QR code system to promote both self-reporting of performance data and data integrity. An open-ended survey was administered to capture perceptions of leaderboard implementation. Conceptual step-by-step instructions detailing how to apply the QR code system to any leaderboard using simulated or real performance metrics are outlined using an illustrative example of a leaderboard that employed simulated cardiopulmonary resuscitation performance scores to compare participants across 17 hospitals in 4 countries for 16 months. The following three major descriptive categories that captured perceptions of leaderboard implementation emerged from initial evaluation data from 10 sites: (1) competition, (2) longevity, and (3) perceived deficits. A well-designed leaderboard should be user-friendly and encompass best practices in gamification principles while collecting and storing data for research analyses. Easy storage and export of data allow for longitudinal record keeping that can be leveraged both to track compliance and to enable social competition.

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

    International Nuclear Information System (INIS)

    Bertolotto, D.

    2011-11-01

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

  2. Prediction of {sup 211}At production using the Monte Carlo code MCNPX

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Gyehong; Chun, Kwonsoo; Kim, Byungil; Yu, Inkong [Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Park, Sungho [Asan Medical Center, Seoul (Korea, Republic of)

    2014-05-15

    {sup 211}At is the most promising radionuclide for targeted cancer therapy due to its decay properties. {sup 211}At has a half-life of 7.214 h, which is sufficient for its production, labeling, dispensing, transportation, quality control and administering the radiolabeled compound. The range of α-particles produced by the decay of {sup 211}At are less than 70 μm in water and soft animal tissues with a LET between 100 and 130 keV/μm, which is about the maximum RBE for heavy ions. The survey carried out by Barbet et al revealed that the most favorable radionuclides for therapeutic applications were {sup 211}At and {sup 67}Cu. The most preferred production route for {sup 211}At production is via cyclotron bombardment of natural bismuth targets with about 29 MeV α-particles. In this study, the production method is based on the nuclear reactions {sup 209}Bi(α,2n){sup 211}At, which has a threshold around 21 MeV and reaches the maximum cross section of about 900 mb at 30 MeV. However, one cannot take advantage of the full range of the beam energies suitable for production of {sup 211}At because of concerns about generating {sup 210}At with half-life of 8.3 h. This radionuclide is problematic because its decay leads to the production of daughter {sup 210}Po, which is an α-particle (5.304 MeV) emitting radionuclide with a physical half-life of 138.4 d and a biological half-lives ranging from 30 to 50 d, unnecessarily giving rise to bone marrow toxicity. Our goal has been to model fluxes from a {sup 209}Bi target and to subsequently calculate the yields of α-emitter {sup 211}At and {sup 210}At using 45 MeV α-beam.

  3. SU-E-T-753: Three-Dimensional Dose Distributions of Incident Proton Particle in the Polymer Gel Dosimeter and the Radiochromic Gel Dosimeter: A Simulation Study with MCNP Code

    International Nuclear Information System (INIS)

    Park, M; Kim, G; Ji, Y; Kim, K; Park, S; Jung, H

    2015-01-01

    Purpose: The purpose of this study is to estimate the three-dimensional dose distributions in the polymer and the radiochromic gel dosimeter, and to identify the detectability of both gel dosimeters by comparing with the water phantom in case of irradiating the proton particles. Methods: The normoxic polymer gel and the LCV micelle radiochromic gel were used in this study. The densities of polymer and the radiochromic gel dosimeter were 1.024 and 1.005 g/cm 3 , respectively. The dose distributions of protons in the polymer and radiochromic gel were simulated using Monte Carlo radiation transport code (MCNPX, Los Alamos National Laboratory). The shape of phantom irradiated by proton particles was a hexahedron with the dimension of 12.4 × 12.4 × 15.0 cm 3 . The energies of proton beam were 50, 80, and 140 MeV energies were directed to top of the surface of phantom. The cross-sectional view of proton dose distribution in both gel dosimeters was estimated with the water phantom and evaluated by the gamma evaluation method. In addition, the absorbed dose(Gy) was also calculated for evaluating the proton detectability. Results: The evaluation results show that dose distributions in both gel dosimeters at intermediated section and Bragg-peak region are similar with that of the water phantom. At entrance section, however, inconsistencies of dose distribution are represented, compared with water. The relative absorbed doses in radiochromic and polymer gel dosimeter were represented to be 0.47 % and 2.26 % difference, respectively. These results show that the radiochromic gel dosimeter was better matched than the water phantom in the absorbed dose evaluation. Conclusion: The polymer and the radiochromic gel dosimeter show similar characteristics in dose distributions for the proton beams at intermediate section and Bragg-peak region. Moreover the calculated absorbed dose in both gel dosimeters represents similar tendency by comparing with that in water phantom

  4. DISCRETE DYNAMIC MODEL OF BEVEL GEAR – VERIFICATION THE PROGRAM SOURCE CODE FOR NUMERICAL SIMULATION

    Directory of Open Access Journals (Sweden)

    Krzysztof TWARDOCH

    2014-06-01

    Full Text Available In the article presented a new model of physical and mathematical bevel gear to study the influence of design parameters and operating factors on the dynamic state of the gear transmission. Discusses the process of verifying proper operation of copyright calculation program used to determine the solutions of the dynamic model of bevel gear. Presents the block diagram of a computing algorithm that was used to create a program for the numerical simulation. The program source code is written in an interactive environment to perform scientific and engineering calculations, MATLAB

  5. Design and construction of a graphical interface for automatic generation of simulation code GEANT4

    International Nuclear Information System (INIS)

    Driss, Mozher; Bouzaine Ismail

    2007-01-01

    This work is set in the context of the engineering studies final project; it is accomplished in the center of nuclear sciences and technologies in Sidi Thabet. This project is about conceiving and developing a system based on graphical user interface which allows an automatic codes generation for simulation under the GEANT4 engine. This system aims to facilitate the use of GEANT4 by scientific not necessary expert in this engine and to be used in different areas: research, industry and education. The implementation of this project uses Root library and several programming languages such as XML and XSL. (Author). 5 refs

  6. BlazeDEM3D-GPU A Large Scale DEM simulation code for GPUs

    Directory of Open Access Journals (Sweden)

    Govender Nicolin

    2017-01-01

    Full Text Available Accurately predicting the dynamics of particulate materials is of importance to numerous scientific and industrial areas with applications ranging across particle scales from powder flow to ore crushing. Computational discrete element simulations is a viable option to aid in the understanding of particulate dynamics and design of devices such as mixers, silos and ball mills, as laboratory scale tests comes at a significant cost. However, the computational time required to simulate an industrial scale simulation which consists of tens of millions of particles can take months to complete on large CPU clusters, making the Discrete Element Method (DEM unfeasible for industrial applications. Simulations are therefore typically restricted to tens of thousands of particles with highly detailed particle shapes or a few million of particles with often oversimplified particle shapes. However, a number of applications require accurate representation of the particle shape to capture the macroscopic behaviour of the particulate system. In this paper we give an overview of the recent extensions to the open source GPU based DEM code, BlazeDEM3D-GPU, that can simulate millions of polyhedra and tens of millions of spheres on a desktop computer with a single or multiple GPUs.

  7. BlazeDEM3D-GPU A Large Scale DEM simulation code for GPUs

    Science.gov (United States)

    Govender, Nicolin; Wilke, Daniel; Pizette, Patrick; Khinast, Johannes

    2017-06-01

    Accurately predicting the dynamics of particulate materials is of importance to numerous scientific and industrial areas with applications ranging across particle scales from powder flow to ore crushing. Computational discrete element simulations is a viable option to aid in the understanding of particulate dynamics and design of devices such as mixers, silos and ball mills, as laboratory scale tests comes at a significant cost. However, the computational time required to simulate an industrial scale simulation which consists of tens of millions of particles can take months to complete on large CPU clusters, making the Discrete Element Method (DEM) unfeasible for industrial applications. Simulations are therefore typically restricted to tens of thousands of particles with highly detailed particle shapes or a few million of particles with often oversimplified particle shapes. However, a number of applications require accurate representation of the particle shape to capture the macroscopic behaviour of the particulate system. In this paper we give an overview of the recent extensions to the open source GPU based DEM code, BlazeDEM3D-GPU, that can simulate millions of polyhedra and tens of millions of spheres on a desktop computer with a single or multiple GPUs.

  8. Numerical simulation of Ge solar cells using D-AMPS-1D code

    Energy Technology Data Exchange (ETDEWEB)

    Barrera, Marcela, E-mail: barrera@tandar.cnea.gov.ar [Comision Nacional de Energia Atomica, Avenida General Paz 1499, San Martin 1650, Buenos Aires (Argentina); Consejo Nacional de Investigaciones Cientificas y Tecnicas (CONICET) (Argentina); Rubinelli, Francisco [Instituto de Desarrollo Tecnologico para la Industria Quimica (INTEC)-CONICET, Gueemes 3450, Santa Fe 3000 (Argentina); Rey-Stolle, Ignacio [Instituto de Energia Solar, Universidad Politecnica de Madrid, Avenida Complutense 30, Madrid 28040 (Spain); Pla, Juan [Comision Nacional de Energia Atomica, Avenida General Paz 1499, San Martin 1650, Buenos Aires (Argentina); Consejo Nacional de Investigaciones Cientificas y Tecnicas (CONICET) (Argentina)

    2012-08-15

    A solar cell is a solid state device that converts the energy of sunlight directly into electricity by the photovoltaic effect. When light with photon energies greater than the band gap is absorbed by a semiconductor material, free electrons and free holes are generated by optical excitation in the material. The main characteristic of a photovoltaic device is the presence of internal electric field able to separate the free electrons and holes so they can pass out of the material to the external circuit before they recombine. Numerical simulation of photovoltaic devices plays a crucial role in their design, performance prediction, and comprehension of the fundamental phenomena ruling their operation. The electrical transport and the optical behavior of the solar cells discussed in this work were studied with the simulation code D-AMPS-1D. This software is an updated version of the one-dimensional (1D) simulation program Analysis of Microelectronic and Photonic Devices (AMPS) that was initially developed at The Penn State University, USA. Structures such as homojunctions, heterojunctions, multijunctions, etc., resulting from stacking layers of different materials can be studied by appropriately selecting characteristic parameters. In this work, examples of cells simulation made with D-AMPS-1D are shown. Particularly, results of Ge photovoltaic devices are presented. The role of the InGaP buffer on the device was studied. Moreover, a comparison of the simulated electrical parameters with experimental results was performed.

  9. Extremely Scalable Spiking Neuronal Network Simulation Code: From Laptops to Exascale Computers.

    Science.gov (United States)

    Jordan, Jakob; Ippen, Tammo; Helias, Moritz; Kitayama, Itaru; Sato, Mitsuhisa; Igarashi, Jun; Diesmann, Markus; Kunkel, Susanne

    2018-01-01

    State-of-the-art software tools for neuronal network simulations scale to the largest computing systems available today and enable investigations of large-scale networks of up to 10 % of the human cortex at a resolution of individual neurons and synapses. Due to an upper limit on the number of incoming connections of a single neuron, network connectivity becomes extremely sparse at this scale. To manage computational costs, simulation software ultimately targeting the brain scale needs to fully exploit this sparsity. Here we present a two-tier connection infrastructure and a framework for directed communication among compute nodes accounting for the sparsity of brain-scale networks. We demonstrate the feasibility of this approach by implementing the technology in the NEST simulation code and we investigate its performance in different scaling scenarios of typical network simulations. Our results show that the new data structures and communication scheme prepare the simulation kernel for post-petascale high-performance computing facilities without sacrificing performance in smaller systems.

  10. Extremely Scalable Spiking Neuronal Network Simulation Code: From Laptops to Exascale Computers

    Science.gov (United States)

    Jordan, Jakob; Ippen, Tammo; Helias, Moritz; Kitayama, Itaru; Sato, Mitsuhisa; Igarashi, Jun; Diesmann, Markus; Kunkel, Susanne

    2018-01-01

    State-of-the-art software tools for neuronal network simulations scale to the largest computing systems available today and enable investigations of large-scale networks of up to 10 % of the human cortex at a resolution of individual neurons and synapses. Due to an upper limit on the number of incoming connections of a single neuron, network connectivity becomes extremely sparse at this scale. To manage computational costs, simulation software ultimately targeting the brain scale needs to fully exploit this sparsity. Here we present a two-tier connection infrastructure and a framework for directed communication among compute nodes accounting for the sparsity of brain-scale networks. We demonstrate the feasibility of this approach by implementing the technology in the NEST simulation code and we investigate its performance in different scaling scenarios of typical network simulations. Our results show that the new data structures and communication scheme prepare the simulation kernel for post-petascale high-performance computing facilities without sacrificing performance in smaller systems. PMID:29503613

  11. Introduction to the simulation with MCNP Monte Carlo code and its applications in Medical Physics

    International Nuclear Information System (INIS)

    Parreno Z, F.; Paucar J, R.; Picon C, C.

    1998-01-01

    The simulation by Monte Carlo is tool which Medical Physics counts with it for the development of its research, the interest by this tool is growing, as we may observe in the main scientific journals for the years 1995-1997 where more than 27 % of the papers treat over Monte Carlo and/or its applications in the radiation transport.In the Peruvian Institute of Nuclear Energy we are implementing and making use of the MCNP4 and EGS4 codes. In this work are presented the general features of the Monte Carlo method and its more useful applications in Medical Physics. Likewise, it is made a simulation of the calculation of isodose curves in an interstitial treatment with Ir-192 wires in a mammary gland carcinoma. (Author)

  12. Simulation of the preliminary General Electric SP-100 space reactor concept using the ATHENA computer code

    International Nuclear Information System (INIS)

    Fletcher, C.D.

    1986-01-01

    The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes. 6 refs., 17 figs., 1 tab

  13. Performance Characteristics of HYDRA - a Multi-Physics simulation code from Lawrence Livermore National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Langer, Steven H. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Karlin, Ian [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marinak, Marty M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-01-09

    HYDRA is used to simulate a variety of experiments carried out at the National Ignition Facility (NIF) [4] and other high energy density physics facilities. HYDRA has packages to simulate radiation transfer, atomic physics, hydrodynamics, laser propagation, and a number of other physics effects. HYDRA has over one million lines of code and includes both MPI and thread-level (OpenMP and pthreads) parallelism. This paper measures the performance characteristics of HYDRA using hardware counters on an IBM BlueGene/Q system. We report key ratios such as bytes/instruction and memory bandwidth for several different physics packages. The total number of bytes read and written per time step is also reported. We show that none of the packages which use significant time are memory bandwidth limited on a Blue Gene/Q. HYDRA currently issues very few SIMD instructions. The pressure on memory bandwidth will increase if high levels of SIMD instructions can be achieved.

  14. Modelling and Simulation of National Electronic Product Code Network Demonstrator Project

    Science.gov (United States)

    Mo, John P. T.

    The National Electronic Product Code (EPC) Network Demonstrator Project (NDP) was the first large scale consumer goods track and trace investigation in the world using full EPC protocol system for applying RFID technology in supply chains. The NDP demonstrated the methods of sharing information securely using EPC Network, providing authentication to interacting parties, and enhancing the ability to track and trace movement of goods within the entire supply chain involving transactions among multiple enterprise. Due to project constraints, the actual run of the NDP was 3 months only and was unable to consolidate with quantitative results. This paper discusses the modelling and simulation of activities in the NDP in a discrete event simulation environment and provides an estimation of the potential benefits that can be derived from the NDP if it was continued for one whole year.

  15. Computer code to simulate transients in a steam generator of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Silva, J.M. da.

    1979-01-01

    A digital computer code KIBE was developed to simulate the transient behavior of a Steam Generator used in Pressurized Water Reactor Power PLants. The equations of Conservation of mass, energy and momentum were numerically integrated by an implicit method progressively in the several axial sections into which the Steam Generator was divided. Forced convection heat transfer was assumed on the primary side, while on the secondary side all the different modes of heat transfer were permitted and deternined from the various correlations. The stability of the stationary state was verified by its reproducibility during the integration of the conservation equation without any pertubation. Transient behavior resulting from pertubations in the flow and the internal energy (temperature) at the inlet of the primary side were simulated. The results obtained exhibited satisfactory behaviour. (author) [pt

  16. Steady-state and transient simulations of gas cooled reactor with the computer code CATHARE

    International Nuclear Information System (INIS)

    Tauveron, N.; Saez, M.; Marchand, M.; Chataing, T.; Geffraye, G.; Cherel, J. M.

    2003-01-01

    This work concerns the design and safety analysis of Gas Cooled Reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbomachine trip, 10 inch cold duct break, 10 inch cold duct break combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation

  17. Density-matrix simulation of small surface codes under current and projected experimental noise

    Science.gov (United States)

    O'Brien, T. E.; Tarasinski, B.; DiCarlo, L.

    2017-09-01

    We present a density-matrix simulation of the quantum memory and computing performance of the distance-3 logical qubit Surface-17, following a recently proposed quantum circuit and using experimental error parameters for transmon qubits in a planar circuit QED architecture. We use this simulation to optimize components of the QEC scheme (e.g., trading off stabilizer measurement infidelity for reduced cycle time) and to investigate the benefits of feedback harnessing the fundamental asymmetry of relaxation-dominated error in the constituent transmons. A lower-order approximate calculation extends these predictions to the distance-5 Surface-49. These results clearly indicate error rates below the fault-tolerance threshold of the surface code, and the potential for Surface-17 to perform beyond the break-even point of quantum memory. However, Surface-49 is required to surpass the break-even point of computation at state-of-the-art qubit relaxation times and readout speeds.

  18. Uranium oxide fuel cycle analysis in VVER-1000 with VISTA simulation code

    Science.gov (United States)

    Mirekhtiary, Seyedeh Fatemeh; Abbasi, Akbar

    2018-02-01

    The VVER-1000 Nuclear power plant generates about 20-25 tons of spent fuel per year. In this research, the fuel transmutation of Uranium Oxide (UOX) fuel was calculated by using of nuclear fuel cycle simulation system (VISTA) code. In this simulation, we evaluated the back end components fuel cycle. The back end component calculations are Spent Fuel (SF), Actinide Inventory (AI) and Fission Product (FP) radioisotopes. The SF, AI and FP values were obtained 23.792178 ton/y, 22.811139 ton/y, 0.981039 ton/y, respectively. The obtained value of spent fuel, major actinide, and minor actinide and fission products were 23.8 ton/year, 22.795 ton/year, 0.024 ton/year and 0.981 ton/year, respectively.

  19. Solar cosmic ray induced ionization in the Earth's atmosphere obtained with CORSIKA code simulations

    International Nuclear Information System (INIS)

    Mishev, A.; Velinov, P.

    2008-01-01

    In the present work ionization profiles due to solar cosmic rays in the Earth's atmosphere are obtained. Recent simulations are carried out with CORSIKA 6.52 code using FLUKA 2006 and QGSJET II hadronic interaction subroutines. The energy deposit of proton induced cascades in the atmosphere is the main result of simulations. We consider vertical events. The atmosphere was divided into 1036 layers per 1 g/cm 2 , which permitted good precision for longitudinal energy deposit estimation. The energy of the incoming protons is accepted 10, 20, 40 and 60 MeV. The ion pair production in the atmosphere is calculated according to the obtained ionization yield function Y , taking into account the steepness of solar cosmic ray spectrum. (authors)

  20. CMS/RPC background particle simulation with the GEANT code preliminary results

    CERN Document Server

    Jamil, M

    2005-01-01

    A method to simulate the background particles of compact muon solenoid (CMS) endcap resistive plate chambers (RPCs) is described using a realistic Monte Carlo simulation based on the geometry and tracking (GEANT) code and analyzed with physics analysis workstation (PAW) interfaces. Sensitivity calculations were performed for particles such as gamma 's, e/sup -/'s and e/sup +/'s in the range 0.1 - 100 MeV for their respective spectra. For the evaluation of the response of detector in the Large Hadron Collider (LHC) background environment, the gamma , e/sup -/ and e/sup +/ energy spectra expected in the CMS muon endcap region were taken into account whereas the RPC sensitivity was evaluated as a function of the detector size.

  1. Simulation of high-energy radiation belt electron fluxes using NARMAX-VERB coupled codes

    Science.gov (United States)

    Pakhotin, I. P.; Drozdov, A. Y.; Shprits, Y. Y.; Boynton, R. J.; Subbotin, D. A.; Balikhin, M. A.

    2014-10-01

    This study presents a fusion of data-driven and physics-driven methodologies of energetic electron flux forecasting in the outer radiation belt. Data-driven NARMAX (Nonlinear AutoRegressive Moving Averages with eXogenous inputs) model predictions for geosynchronous orbit fluxes have been used as an outer boundary condition to drive the physics-based Versatile Electron Radiation Belt (VERB) code, to simulate energetic electron fluxes in the outer radiation belt environment. The coupled system has been tested for three extended time periods totalling several weeks of observations. The time periods involved periods of quiet, moderate, and strong geomagnetic activity and captured a range of dynamics typical of the radiation belts. The model has successfully simulated energetic electron fluxes for various magnetospheric conditions. Physical mechanisms that may be responsible for the discrepancies between the model results and observations are discussed.

  2. Development of an interface between MCNP and ORIGEN codes for calculations of fuel evolution in nuclear systems. Initial project

    International Nuclear Information System (INIS)

    Campolina, Daniel de Almeida Magalhaes

    2009-01-01

    In Many situations of nuclear system study, it is necessary to know the detailed particle flux in a geometry. Deterministic 1-D and 2-D methods aren't suitable to represent some strong 3-D behavior configurations, for example in cores where the neutron flux varies considerably in the space and Monte Carlo analysis are necessary. The majority of Monte Carlo transport calculation codes, performs time static simulations, in terms of fuel isotopic composition. This work is a initial project to incorporate depletion capability to the MCNP code, by means of a connection with ORIGEN2.1 burnup code. The method to develop the program proposed followed the methodology of other programs used to the same purpose. Essentially, MCNP data library are used to generate one group microscopic cross sections that override default ORIGEN libraries. To verify the actual implemented part, comparisons which MCNPX (version 2.6.0) results were made. The neutron flux and criticality value of core agree. The neutron flux and criticality value of the core agree, especially in beginning of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB). Next step of this work is to adapt MCNP version 4C to work with a memory higher than its standard value (4MB), in order to allow a greater number of isotopes in the transport model. (author)

  3. Numerical Simulations of Slow Stick Slip Events with PFC, a DEM Based Code

    Science.gov (United States)

    Ye, S. H.; Young, R. P.

    2017-12-01

    Nonvolcanic tremors around subduction zone have become a fascinating subject in seismology in recent years. Previous studies have shown that the nonvolcanic tremor beneath western Shikoku is composed of low frequency seismic waves overlapping each other. This finding provides direct link between tremor and slow earthquakes. Slow stick slip events are considered to be laboratory scaled slow earthquakes. Slow stick slip events are traditionally studied with direct shear or double direct shear experiment setup, in which the sliding velocity can be controlled to model a range of fast and slow stick slips. In this study, a PFC* model based on double direct shear is presented, with a central block clamped by two side blocks. The gauge layers between the central and side blocks are modelled as discrete fracture networks with smooth joint bonds between pairs of discrete elements. In addition, a second model is presented in this study. This model consists of a cylindrical sample subjected to triaxial stress. Similar to the previous model, a weak gauge layer at a 45 degrees is added into the sample, on which shear slipping is allowed. Several different simulations are conducted on this sample. While the confining stress is maintained at the same level in different simulations, the axial loading rate (displacement rate) varies. By varying the displacement rate, a range of slipping behaviour, from stick slip to slow stick slip are observed based on the stress-strain relationship. Currently, the stick slip and slow stick slip events are strictly observed based on the stress-strain relationship. In the future, we hope to monitor the displacement and velocity of the balls surrounding the gauge layer as a function of time, so as to generate a synthetic seismogram. This will allow us to extract seismic waveforms and potentially simulate the tremor-like waves found around subduction zones. *Particle flow code, a discrete element method based numerical simulation code developed by

  4. Multibunch and multiparticle simulation code with an alternative approach to wakefield effects

    Directory of Open Access Journals (Sweden)

    M. Migliorati

    2015-03-01

    Full Text Available The simulation of beam dynamics in the presence of collective effects requires a strong computational effort to take into account, in a self-consistent way, the wakefield acting on a given charge and produced by all the others. Generally this is done by means of a convolution integral or sum. Moreover, if the electromagnetic fields consist of resonant modes with high quality factors, responsible, for example, for coupled bunch instabilities, a charge is also affected by itself in previous turns, and a very long record of wakefield must be properly taken into account. In this paper we present a new simulation code for the longitudinal beam dynamics in a circular accelerator, which exploits an alternative approach to the currently used convolution sum, reducing the computing time and avoiding the issues related to the length of wakefield for coupled bunch instabilities. With this approach it is possible to simulate, without the need for large computing power, simultaneously, the single and multibunch beam dynamics including intrabunch motion. Moreover, for a given machine, generally both the coupling impedance and the wake potential of a short Gaussian bunch are known. However, a classical simulation code needs in input the so-called “Green” function, that is the wakefield produced by a point charge, making necessary some manipulations to use the wake potential instead of the Green function. The method that we propose does not need the wakefield as input, but a particular fitting of the coupling impedance requiring the use of the resonator impedance model, thus avoiding issues related to the knowledge of the Green function. The same approach can also be applied to the transverse case and to linear accelerators as well.

  5. Computer code for the atomistic simulation of lattice defects and dynamics

    International Nuclear Information System (INIS)

    Schiffgens, J.O.; Graves, N.J.; Oster, C.A.

    1980-04-01

    The computer code COMENT used to simulate the effect of temperature, impurities, and pre-existing defects on radiation-induced defect production mechanisms, defect properties, defect migration, and defect stability. This report documents Version IV of COMENT (models, methods, and implementation) and defines current code options. Version IV of COMENT generates only face-centered-cubic (fcc) crystal lattices. However, an effort was made to structure COMENT to allow addition of new options with a minimum of change in the existing version of the code. This document describes the calling program and thirty-two user-defined subroutines. Fourteen subroutines (ALOYORD, DASPKA, DFCT, DSLOAN, DSLOIN, EXPAND, POT1, POT2, POT3, POT4, POT5, POT6, POT7, and THRMAL) are associated with the selection of program options; only a few of these are used in any given analysis. Seven of the other subroutines (CRYSTL, IEAF, INCBOX, LABLE, MINILAT, SPEFORS, and SQUEZ are used to establish a variety of arrays and conditions required for each analysis; most of them are used once in a given calculation. The remaining eleven subroutines (DAMP, DIRECT, IDDEF, NEAF, INBIN, FILBIN, FTBIN, PAC3, UNPAC3, PACF, and UNPACF) are used many times in each calculation; the last eight of these are used many times in each time step during the integration and, therefore, are written in COMPASS (CDC assembly language). The COMPASS subroutines are described in sufficient detail to permit easy conversion to some other assembly language or to FORTRAN

  6. Software Abstractions and Methodologies for HPC Simulation Codes on Future Architectures

    Directory of Open Access Journals (Sweden)

    Anshu Dubey

    2014-07-01

    Full Text Available Simulations with multi-physics modeling have become crucial to many science and engineering fields, and multi-physics capable scientific software is as important to these fields as instruments and facilities are to experimental sciences. The current generation of mature multi-physics codes would have sustainably served their target communities with modest amount of ongoing investment for enhancing capabilities. However, the revolution occurring in the hardware architecture has made it necessary to tackle the parallelism and performance management in these codes at multiple levels. The requirements of various levels are often at cross-purposes with one another, and therefore hugely complicate the software design. All of these considerations make it essential to approach this challenge cooperatively as a community. We conducted a series of workshops under an NSF-SI2 conceptualization grant to get input from various stakeholders, and to identify broad approaches that might lead to a solution. In this position paper we detail the major concerns articulated by the application code developers, and emerging trends in utilization of programming abstractions that we found through these workshops.

  7. Development of a Beam-Beam Simulation Code for e+e- Colliders

    CERN Document Server

    Zhang, Yuan

    2005-01-01

    BEPC will be upgraded into BEPCII, and the luminosity will be about 100 times higher. We developed a three dimensional strong-strong PIC code to study the beam-beam effects in BEPCII. The transportation through the arc is the same as that in Hirata's weak-strong code. The beam-beam force is computed directly by solving the Poisson equation using the FACR method, and the boundary potential is computed by circular convolution. The finite bunch length effect is included by longitudinal slices. An interpolation scheme is used to reduce the required slice number in simulations. The standard message passing interface (MPI) is used to parallelize the code. The computing time increases linearly with (n+1), where n is the slice number. The calculated luminosity of BEPCII at the design operating point is less than the design value. The best area in the tune space is near (0.505,0.57) according to the survey, where the degradation of luminosity can be improved.

  8. A Mathematical Model and MATLAB Code for Muscle-Fluid-Structure Simulations.

    Science.gov (United States)

    Battista, Nicholas A; Baird, Austin J; Miller, Laura A

    2015-11-01

    This article provides models and code for numerically simulating muscle-fluid-structure interactions (FSIs). This work was presented as part of the symposium on Leading Students and Faculty to Quantitative Biology through Active Learning at the society-wide meeting of the Society for Integrative and Comparative Biology in 2015. Muscle mechanics and simple mathematical models to describe the forces generated by muscular contractions are introduced in most biomechanics and physiology courses. Often, however, the models are derived for simplifying cases such as isometric or isotonic contractions. In this article, we present a simple model of the force generated through active contraction of muscles. The muscles' forces are then used to drive the motion of flexible structures immersed in a viscous fluid. An example of an elastic band immersed in a fluid is first presented to illustrate a fully-coupled FSI in the absence of any external driving forces. In the second example, we present a valveless tube with model muscles that drive the contraction of the tube. We provide a brief overview of the numerical method used to generate these results. We also include as Supplementary Material a MATLAB code to generate these results. The code was written for flexibility so as to be easily modified to many other biological applications for educational purposes. © The Author 2015. Published by Oxford University Press on behalf of the Society for Integrative and Comparative Biology. All rights reserved. For permissions please email: journals.permissions@oup.com.

  9. Traveling-wave-tube simulation: The IBC (Interactive Beam-Circuit) code

    Energy Technology Data Exchange (ETDEWEB)

    Morey, I.J.; Birdsall, C.K.

    1989-09-26

    Interactive Beam-Circuit (IBC) is a one-dimensional many particle simulation code which has been developed to run interactively on a PC or Workstation, and displaying most of the important physics of a traveling-wave-tube. The code is a substantial departure from previous efforts, since it follows all of the particles in the tube, rather than just those in one wavelength, as commonly done. This step allows for nonperiodic inputs in time, a nonuniform line and a large set of spatial diagnostics. The primary aim is to complement a microwave tube lecture course, although past experience has shown that such codes readily become research tools. Simple finite difference methods are used to model the fields of the coupled slow-wave transmission line. The coupling between the beam and the transmission line is based upon the finite difference equations of Brillouin. The space-charge effects are included, in a manner similar to that used by Hess; the original part is use of particle-in-cell techniques to model the space-charge fields. 11 refs., 11 figs.

  10. Implementing Scientific Simulation Codes Highly Tailored for Vector Architectures Using Custom Configurable Computing Machines

    Science.gov (United States)

    Rutishauser, David

    2006-01-01

    The motivation for this work comes from an observation that amidst the push for Massively Parallel (MP) solutions to high-end computing problems such as numerical physical simulations, large amounts of legacy code exist that are highly optimized for vector supercomputers. Because re-hosting legacy code often requires a complete re-write of the original code, which can be a very long and expensive effort, this work examines the potential to exploit reconfigurable computing machines in place of a vector supercomputer to implement an essentially unmodified legacy source code. Custom and reconfigurable computing resources could be used to emulate an original application's target platform to the extent required to achieve high performance. To arrive at an architecture that delivers the desired performance subject to limited resources involves solving a multi-variable optimization problem with constraints. Prior research in the area of reconfigurable computing has demonstrated that designing an optimum hardware implementation of a given application under hardware resource constraints is an NP-complete problem. The premise of the approach is that the general issue of applying reconfigurable computing resources to the implementation of an application, maximizing the performance of the computation subject to physical resource constraints, can be made a tractable problem by assuming a computational paradigm, such as vector processing. This research contributes a formulation of the problem and a methodology to design a reconfigurable vector processing implementation of a given application that satisfies a performance metric. A generic, parametric, architectural framework for vector processing implemented in reconfigurable logic is developed as a target for a scheduling/mapping algorithm that maps an input computation to a given instance of the architecture. This algorithm is integrated with an optimization framework to arrive at a specification of the architecture parameters

  11. A Steam Jet Plume Simulation in a Large Bulk Space with a System Code MARS

    International Nuclear Information System (INIS)

    Bae, Sung Won; Chung, Bub Dong

    2006-01-01

    From May 2002, the OECD-SETH group has launched the PANDA Project in order to provide an experimental data base for a multi-dimensional code assessment. OECD-SETH group expects the PANDA Project will meet the increasing needs for adequate experimental data for a 3D distribution of relevant variables like the temperature, velocity and steam-air concentrations that are measured with a sufficient resolution and accuracy. The scope of the PANDA Project is the mixture stratification and mixing phenomena in a large bulk space. Total of 24 test series are still being performed in PSI, Switzerland. The PANDA facility consists of 2 main large vessels and 1 connection pipe Within the large vessels, a steam injection nozzle and outlet vent are arranged for each test case. These tests are categorized into 3 modes, i.e. the high momentum, near wall plume, and free plume tests. KAERI has also participated in the SETH group since 1997 so that the multi-dimensional capability of the MARS code could be assessed and developed. Test 17, the high steam jet injection test, has already been simulated by MARS and shows promising results. Now, the test 9 and 9bis cases which use a low speed horizontal steam jet flow have been simulated and investigated

  12. Development of compressible density-based steam explosion simulation code ESE-2

    International Nuclear Information System (INIS)

    Leskovar, M.

    2004-01-01

    A steam explosion is a fuel coolant interaction process by which the energy of the corium is transferred to water in a time-scale smaller than the time-scale for system pressure relief and induces dynamic loading of surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. To help finding answers on open questions regarding steam explosion understanding and modelling, the steam explosion simulation code ESE-2 is being developed. In contrast to the developed simulation code ESE-1, where the multiphase flow equations are solved with pressure-based numerical methods (best suited for incompressible flow), in ESE-2 densitybased numerical methods (best suited for compressible flow) are used. Therefore ESE-2 will enable an accurate treatment of the whole steam explosion process, which consists of the premixing, triggering, propagation and expansion phase. In the paper the basic characteristics of the mathematical model and the numerical solution procedure in ESE-2 are described. The essence of the numerical treatment is that the convective terms in the multiphase flow equations are calculated with the AUSM+ scheme, which is very time efficient since no field-by-field wave decomposition is needed, using second order accurate discretization. (author)

  13. Development of a nuclear spallation simulation code and calculations of primary spallation products

    International Nuclear Information System (INIS)

    Nishida, Takahiko; Nakahara, Yasuaki; Tsutsui, Tsuneo

    1986-08-01

    In order to make evaluations of computational models for the nuclear spallation reaction from a nuclear physics point of view, a simulation code NUCLEUS has been developed by modifying and combining the Monte Carlo codes NMTC/JAERI and NMTA/JAERI for calculating only the nuclear spallation reaction (intranuclear cascade + evaporation and/or fast fission) between a nucleus and a projectile without taking into consideration of internuclear transport. New several plotting routines have been provided for the rapid process of much more event data, obtained by using the ARGUS plotting system. The results obtained by our code can be directly compared with the experimental results using by thin foil experiments in which internuclear multiple collisions have little effects, and will serve to upgrade the calculational methods and the values of nuclear parameters currently used in the calculations. Some discussions are done about the preliminary computational results obtained by using NUCLEUS. The mass distribution and charge dispersion of reaction products are examined in some detail for the nuclear spallation reaction between incident protons and target nuclei, such as U, Pb and Ag, in the energy range from 0.5 GeV to 3.0 GeV. These results show that the distribution of reaction products ceases to change its form as the proton energy increases over about 2 GeV. The same tendency is seen in the energy dependence of the number of primary particles emitted from a nucleus. After spallation reactions, a variety of nuclei, especially many neutron deficient nuclides with nuclear charges nearly equal to ones of a target nucleus, are produced. Due to their short lifetime most of them will change to stable nuclides in due time. Finally, some important issues are discussed to improve the present simulation method. (author)

  14. HBT+: an improved code for finding subhaloes and building merger trees in cosmological simulations

    Science.gov (United States)

    Han, Jiaxin; Cole, Shaun; Frenk, Carlos S.; Benitez-Llambay, Alejandro; Helly, John

    2018-02-01

    Dark matter subhalos are the remnants of (incomplete) halo mergers. Identifying them and establishing their evolutionary links in the form of merger trees is one of the most important applications of cosmological simulations. The HBT (Hierachical Bound-Tracing) code identifies haloes as they form and tracks their evolution as they merge, simultaneously detecting subhaloes and building their merger trees. Here we present a new implementation of this approach, HBT+ , that is much faster, more user friendly, and more physically complete than the original code. Applying HBT+ to cosmological simulations, we show that both the subhalo mass function and the peak-mass function are well fitted by similar double-Schechter functions. The ratio between the two is highest at the high-mass end, reflecting the resilience of massive subhaloes that experience substantial dynamical friction but limited tidal stripping. The radial distribution of the most-massive subhaloes is more concentrated than the universal radial distribution of lower mass subhaloes. Subhalo finders that work in configuration space tend to underestimate the masses of massive subhaloes, an effect that is stronger in the host centre. This may explain, at least in part, the excess of massive subhaloes in galaxy cluster centres inferred from recent lensing observations. We demonstrate that the peak-mass function is a powerful diagnostic of merger tree defects, and the merger trees constructed using HBT+ do not suffer from the missing or switched links that tend to afflict merger trees constructed from more conventional halo finders. We make the HBT+ code publicly available.

  15. FENICIA: a generic plasma simulation code using a flux-independent field-aligned coordinate approach

    International Nuclear Information System (INIS)

    Hariri, Farah

    2013-01-01

    The primary thrust of this work is the development and implementation of a new approach to the problem of field-aligned coordinates in magnetized plasma turbulence simulations called the FCI approach (Flux-Coordinate Independent). The method exploits the elongated nature of micro-instability driven turbulence which typically has perpendicular scales on the order of a few ion gyro-radii, and parallel scales on the order of the machine size. Mathematically speaking, it relies on local transformations that align a suitable coordinate to the magnetic field to allow efficient computation of the parallel derivative. However, it does not rely on flux coordinates, which permits discretizing any given field on a regular grid in the natural coordinates such as (x, y, z) in the cylindrical limit. The new method has a number of advantages over methods constructed starting from flux coordinates, allowing for more flexible coding in a variety of situations including X-point configurations. In light of these findings, a plasma simulation code FENICIA has been developed based on the FCI approach with the ability to tackle a wide class of physical models. The code has been verified on several 3D test models. The accuracy of the approach is tested in particular with respect to the question of spurious radial transport. Tests on 3D models of the drift wave propagation and of the Ion Temperature Gradient (ITG) instability in cylindrical geometry in the linear regime demonstrate again the high quality of the numerical method. Finally, the FCI approach is shown to be able to deal with an X-point configuration such as one with a magnetic island with good convergence and conservation properties. (author) [fr

  16. Full modelling of the MOSAIC animal PET system based on the GATE Monte Carlo simulation code

    International Nuclear Information System (INIS)

    Merheb, C; Petegnief, Y; Talbot, J N

    2007-01-01

    Positron emission tomography (PET) systems dedicated to animal imaging are now widely used for biological studies. The scanner performance strongly depends on the design and the characteristics of the system. Many parameters must be optimized like the dimensions and type of crystals, geometry and field-of-view (FOV), sampling, electronics, lightguide, shielding, etc. Monte Carlo modelling is a powerful tool to study the effect of each of these parameters on the basis of realistic simulated data. Performance assessment in terms of spatial resolution, count rates, scatter fraction and sensitivity is an important prerequisite before the model can be used instead of real data for a reliable description of the system response function or for optimization of reconstruction algorithms. The aim of this study is to model the performance of the Philips Mosaic(TM) animal PET system using a comprehensive PET simulation code in order to understand and describe the origin of important factors that influence image quality. We use GATE, a Monte Carlo simulation toolkit for a realistic description of the ring PET model, the detectors, shielding, cap, electronic processing and dead times. We incorporate new features to adjust signal processing to the Anger logic underlying the Mosaic(TM) system. Special attention was paid to dead time and energy spectra descriptions. Sorting of simulated events in a list mode format similar to the system outputs was developed to compare experimental and simulated sensitivity and scatter fractions for different energy thresholds using various models of phantoms describing rat and mouse geometries. Count rates were compared for both cylindrical homogeneous phantoms. Simulated spatial resolution was fitted to experimental data for 18 F point sources at different locations within the FOV with an analytical blurring function for electronic processing effects. Simulated and measured sensitivities differed by less than 3%, while scatter fractions agreed

  17. EMIC wave parameterization in the long-term VERB code simulation

    Science.gov (United States)

    Drozdov, A. Y.; Shprits, Y. Y.; Usanova, M. E.; Aseev, N. A.; Kellerman, A. C.; Zhu, H.

    2017-08-01

    Electromagnetic ion cyclotron (EMIC) waves play an important role in the dynamics of ultrarelativistic electron population in the radiation belts. However, as EMIC waves are very sporadic, developing a parameterization of such wave properties is a challenging task. Currently, there are no dynamic, activity-dependent models of EMIC waves that can be used in the long-term (several months) simulations, which makes the quantitative modeling of the radiation belt dynamics incomplete. In this study, we investigate Kp, Dst, and AE indices, solar wind speed, and dynamic pressure as possible parameters of EMIC wave presence. The EMIC waves are included in the long-term simulations (1 year, including different geomagnetic activity) performed with the Versatile Electron Radiation Belt code, and we compare results of the simulation with the Van Allen Probes observations. The comparison shows that modeling with EMIC waves, parameterized by solar wind dynamic pressure, provides a better agreement with the observations among considered parameterizations. The simulation with EMIC waves improves the dynamics of ultrarelativistic fluxes and reproduces the formation of the local minimum in the phase space density profiles.

  18. EMIC waves parameterization in the long-term VERB code simulation

    Science.gov (United States)

    Drozdov, Alexander; Shprits, Yuri; Usanova, Maria; Kellerman, Adam; Aseev, Nikita; Zhu, Hui

    2017-04-01

    EMIC waves play an important role in the dynamics of ultra-relativistic electron population in the radiation belts. Recently we showed that, a long-term 3-D simulation of electron distribution function based on the Fokker-Planck equation significantly overestimates electron fluxes without considering EMIC waves activity. As EMIC waves are very sporadic, developing a parametrization of such wave properties is a challenging task. Currently, there are no dynamic, activity-dependent models of EMIC waves that can be used in the long-term simulations, which makes the quantitative modeling of the radiation belts dynamics incomplete. In this study, we compared long-term simulations performed with the Versatile Electron Radiation Belt (VERB) code and the Van Allen Probes observations. The model includes radial, energy, pitch-angle and mixed diffusion, losses into the atmosphere, and magnetopause shadowing. We included scattering by hiss and chorus based on a recently developed statistical models of VLF/ELF waves. We considered the relativistic ( 0.5-1 MeV) and ultra-relativistic (>3 MeV) electrons. One year of relativistic electron measurements were well reproduced by the simulation during a period of the various geomagnetic activity. However, in order to accurately reproduce the dynamics of ultra-relativistic population, additional loss mechanisms are required. We investigated EMIC wave occurrence depending on solar wind parameters and geomagnetic indices. Using the obtained dependence we found that modeling with EMIC waves provided a better agreement with the observations.

  19. Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

    International Nuclear Information System (INIS)

    Lee, Y. G.; Kim, J. W.; Yoon, S. J.; Park, G. C.

    2010-10-01

    Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

  20. Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

    Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

  1. Assessment of the IVA3 code for multifield flow simulation. Formal report

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, H.B.

    1995-07-01

    This report presents an assessment of the IVA3 computer code for multifield flow simulation, as applied to the premixing phase of a hypothetical steam explosion in a water-cooled power reactor. The first section of this report reviews the derivation of the basic partial differential equations of multifield modeling, with reference to standard practices in the multiphase flow literature. Basic underlying assumptions and approximations are highlighted, and comparison is made between IVA3 and other codes in current use. Although Kolev`s derivation of these equations is outside the mainstream of the multiphase literature, the basic partial differential equations are in fact nearly equivalent to those in other codes. In the second section, the assumptions and approximations required to pass from generic differential equations to a specific working form are detailed. Some modest improvements to the IVA3 model are suggested. In Section 3, the finite difference approximations to the differential equations are described. The discretization strategy is discussed with reference to numerical stability, accuracy, and the role of various physical phenomena - material convection, sonic propagation, viscous stress, and interfacial exchanges - in the choice of discrete approximations. There is also cause for concern about the approximations of time evolution in some heat transfer terms, which might be adversely affecting numerical accuracy. The fourth section documents the numerical solution method used in IVA3. An explanation for erratic behavior sometimes observed in the first outer iteration is suggested, along with possible remedies. Finally, six recommendations for future assessment and improvement of the IVA3 model and code are made.

  2. Tornado missile simulation and design methodology. Volume 1: simulation methodology, design applications, and TORMIS computer code. Final report

    International Nuclear Information System (INIS)

    Twisdale, L.A.; Dunn, W.L.

    1981-08-01

    A probabilistic methodology has been developed to predict the probabilities of tornado-propelled missiles impacting and damaging nuclear power plant structures. Mathematical models of each event in the tornado missile hazard have been developed and sequenced to form an integrated, time-history simulation methodology. The models are data based where feasible. The data include documented records of tornado occurrence, field observations of missile transport, results of wind tunnel experiments, and missile impact tests. Probabilistic Monte Carlo techniques are used to estimate the risk probabilities. The methodology has been encoded in the TORMIS computer code to facilitate numerical analysis and plant-specific tornado missile probability assessments. Sensitivity analyses have been performed on both the individual models and the integrated methodology, and risk has been assessed for a hypothetical nuclear power plant design case study

  3. COOL: A code for Dynamic Monte Carlo Simulation of molecular dynamics

    Science.gov (United States)

    Barletta, Paolo

    2012-02-01

    Cool is a program to simulate evaporative and sympathetic cooling for a mixture of two gases co-trapped in an harmonic potential. The collisions involved are assumed to be exclusively elastic, and losses are due to evaporation from the trap. Each particle is followed individually in its trajectory, consequently properties such as spatial densities or energy distributions can be readily evaluated. The code can be used sequentially, by employing one output as input for another run. The code can be easily generalised to describe more complicated processes, such as the inclusion of inelastic collisions, or the possible presence of more than two species in the trap. New version program summaryProgram title: COOL Catalogue identifier: AEHJ_v2_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEHJ_v2_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 1 097 733 No. of bytes in distributed program, including test data, etc.: 18 425 722 Distribution format: tar.gz Programming language: C++ Computer: Desktop Operating system: Linux RAM: 500 Mbytes Classification: 16.7, 23 Catalogue identifier of previous version: AEHJ_v1_0 Journal reference of previous version: Comput. Phys. Comm. 182 (2011) 388 Does the new version supersede the previous version?: Yes Nature of problem: Simulation of the sympathetic process occurring for two molecular gases co-trapped in a deep optical trap. Solution method: The Direct Simulation Monte Carlo method exploits the decoupling, over a short time period, of the inter-particle interaction from the trapping potential. The particle dynamics is thus exclusively driven by the external optical field. The rare inter-particle collisions are considered with an acceptance/rejection mechanism, that is, by comparing a random number to the collisional probability

  4. The effect of boron dilution transient on the VVER-1000 reactor core using MCNP and COBRA-EN codes

    Energy Technology Data Exchange (ETDEWEB)

    Jafari, Naser; Talebi, Saeed [Amirkabir Univ. of Technology, Tehran Polytechnic (Iran, Islamic Republic of). Dept. of Energy Engineering and Physics

    2017-07-15

    In this paper, the effect of boron dilution transient, as a consequence of the malfunction of the boron control system, was investigated in a VVER-1000 reactor, and then an appropriate setpoint was determined for the actuation of the emergency protection system to the reactor shutdown. In order to simulate the boron dilution, first, the whole reactor core was simulated by MCNPX code to compute the radial and axial power distribution. Then, the COBRA-EN code was employed using calculated power distribution for analyzing the thermal-hydraulic of hot fuel assembly and for extracting the safety parameters. For the safe operation of the reactor, certain parameters must be in defined specified ranges. Comparison between our results and FSARs data shows that the present modeling provides a good prediction of boron dilution transient with the maximum relative difference about 4%.

  5. Study of a scintillation neutron detector of 1OB+ZnS(Ag) as alternative to the 3He detectors: model MCNPX and validation

    International Nuclear Information System (INIS)

    Guzman G, K. A.; Gallego D, E.; Lorente F, A.; Ibanez F, S.; Vega C, H. R.; Mendez V, R.; Gonzalez, J. A.

    2015-10-01

    Using Monte Carlo methods with the code MCNPX, was estimated the response of a scintillation neutron detector of Zn S(Ag) with a mixture of 10 B high enrichment. The detector consists of four plates of Poly (methyl methacrylate) (PMMA) and five layers of ∼0, 017 cm 10 B+ZnS(Ag) in contact with PMMA. The naked detector response was calculated and with different thicknesses of high density polyethylene moderator, for 29 monoenergetic sources and for sources of 241 AmBe and 252 Cf of neutrons. In these calculations the reactions 10 B(n,α) 7 Li and neutron fluence in the sensitive area of detector 10 B+ZnS(Ag) were estimated. Measurements were performed in the Laboratory of Neutron Measurement to quantify detections in counts per second to a neutron source of 252 Cf to 200 cm on the bench, modeling with MCNPX, these measures were compared to validate the model and the Zn S(Ag) efficiency of α detection was estimated. Calculations in the LPN-CIEMAT were realized. Starting from the validation new models were carried out with geometries that improve the detector response, trying reaching the detection of 2, 5 cps-ng of 252 Cf comparable requirement for responding to the installed equipment of 3 He in the radiation portal monitor. This type of detector can be considered an alternative to detectors of 3 He for detecting special nuclear material. (Author)

  6. Wavelet subband coding of computer simulation output using the A++ array class library

    Energy Technology Data Exchange (ETDEWEB)

    Bradley, J.N.; Brislawn, C.M.; Quinlan, D.J.; Zhang, H.D. [Los Alamos National Lab., NM (United States); Nuri, V. [Washington State Univ., Pullman, WA (United States). School of EECS

    1995-07-01

    The goal of the project is to produce utility software for off-line compression of existing data and library code that can be called from a simulation program for on-line compression of data dumps as the simulation proceeds. Naturally, we would like the amount of CPU time required by the compression algorithm to be small in comparison to the requirements of typical simulation codes. We also want the algorithm to accomodate a wide variety of smooth, multidimensional data types. For these reasons, the subband vector quantization (VQ) approach employed in has been replaced by a scalar quantization (SQ) strategy using a bank of almost-uniform scalar subband quantizers in a scheme similar to that used in the FBI fingerprint image compression standard. This eliminates the considerable computational burdens of training VQ codebooks for each new type of data and performing nearest-vector searches to encode the data. The comparison of subband VQ and SQ algorithms in indicated that, in practice, there is relatively little additional gain from using vector as opposed to scalar quantization on DWT subbands, even when the source imagery is from a very homogeneous population, and our subjective experience with synthetic computer-generated data supports this stance. It appears that a careful study is needed of the tradeoffs involved in selecting scalar vs. vector subband quantization, but such an analysis is beyond the scope of this paper. Our present work is focused on the problem of generating wavelet transform/scalar quantization (WSQ) implementations that can be ported easily between different hardware environments. This is an extremely important consideration given the great profusion of different high-performance computing architectures available, the high cost associated with learning how to map algorithms effectively onto a new architecture, and the rapid rate of evolution in the world of high-performance computing.

  7. Use of a general-purpose heat-transfer code for casting simulation

    International Nuclear Information System (INIS)

    Erickson, W.C.

    1975-07-01

    The practical use of numerical techniques in simulating casting solidification dictate that a general purpose heat transfer code be used and that results be obtained in an easy-to-analyze format. Color film plotting routines were developed for use with NASA's CINDA-3G heat transfer code; the combination of which meet the above criteria. The subroutine LQSLTR written for SINDA, the successor to CINDA-3G, was verified by comparing calculated results obtained using LQSLTR with those obtained using the specific heat method for handling the heat of fusion. Excellent agreement existed when similar data was used. When the more restrictive requirement of a 1 0 F melting range was used, comparable results were obtained. Uranium and lead rod castings were cast in instrumented graphite molds and the solidification sequence simulated using CINDA-3G. Discrepancies attributed to initial assumptions of instantaneous mold filling, uniform melt temperature, and intimate metal/mold contact were encountered. Further calculations using a model incorporating a gap between the mold and casting showed that the intimate contact assumption could not be used; a three-dimensional model also showed that the thermocouple assemblies used with the platinum--platinum-10 percent rhodium were a significant perturbation to the system. An L-shaped steel casting was simulated and the results compared to those reported in the literature. The experimental data for this casting were reproduced within the accuracy permitted by the thermal conductivity of the sand, thus demonstrating that agreement can be obtained when the mold material does not act as a chill. (U.S.)

  8. Implementation and evaluation of a simulation curriculum for paediatric residency programs including just-in-time in situ mock codes.

    Science.gov (United States)

    Sam, Jonathan; Pierse, Michael; Al-Qahtani, Abdullah; Cheng, Adam

    2012-02-01

    To develop, implement and evaluate a simulation-based acute care curriculum in a paediatric residency program using an integrated and longitudinal approach. Curriculum framework consisting of three modular, year-specific courses and longitudinal just-in-time, in situ mock codes. Paediatric residency program at BC Children's Hospital, Vancouver, British Columbia. The three year-specific courses focused on the critical first 5 min, complex medical management and crisis resource management, respectively. The just-in-time in situ mock codes simulated the acute deterioration of an existing ward patient, prepared the actual multidisciplinary code team, and primed the surrounding crisis support systems. Each curriculum component was evaluated with surveys using a five-point Likert scale. A total of 40 resident surveys were completed after each of the modular courses, and an additional 28 surveys were completed for the overall simulation curriculum. The highest Likert scores were for hands-on skill stations, immersive simulation environment and crisis resource management teaching. Survey results also suggested that just-in-time mock codes were realistic, reinforced learning, and prepared ward teams for patient deterioration. A simulation-based acute care curriculum was successfully integrated into a paediatric residency program. It provides a model for integrating simulation-based learning into other training programs, as well as a model for any hospital that wishes to improve paediatric resuscitation outcomes using just-in-time in situ mock codes.

  9. Simulation of the thermalhydraulic behavior of a molten core within a structure, with the three dimensions three components TOLBIAC code

    Energy Technology Data Exchange (ETDEWEB)

    Spindler, B.; Moreau, G.M.; Pigny S. [Centre d`Etudes Nucleaires de Grenoble (France)

    1995-09-01

    The TOLBIAC code is devoted to the simulation of the behavior of a molten core within a structure (pressure vessel of core catcher), taking into account the relative position of the core components, the wall ablation and the crust formation. The code is briefly described: 3D model, physical properties and constitutive laws. wall ablation and crust model. Two results are presented: the simulation of the COPO experiment (natural convection with water in a 1/2 scale elliptic pressure vessel), and the simulation of the behavior of a corium in a PWR pressure vessel, with ablation and crust formation.

  10. Long-term radiation belt simulation with the VERB 3-D code: Comparison with CRRES observations

    Science.gov (United States)

    Subbotin, D. A.; Shprits, Y. Y.; Ni, B.

    2011-12-01

    Highly energetic electrons in the Earth’s radiation belts are hazardous for satellite equipment. Fluxes of relativistic electrons can vary by orders of magnitude during geomagnetic storms. The evolution of relativistic electron fluxes in the radiation belts is described by the 3-D Fokker-Planck equation in terms of the radial distance, energy, and equatorial pitch angle. To better understand the mechanisms that control radiation belt acceleration and loss and particle flux dynamics, we present a long-term radiation belt simulation for 100 days from 29 July to 6 November 1990 with the 3-D Versatile Electron Radiation Belt (VERB) code and compare the results with the electron fluxes observed by the Combined Release and Radiation Effects Satellite (CRRES). We also perform a comparison of Phase Space Density with a multisatellite reanalysis obtained by using Kalman filtering of observations from CRRES, Geosynchronous (GEO), GPS, and Akebono satellites. VERB 3-D simulations include radial, energy, and pitch angle diffusion and mixed energy and pitch angle diffusion driven by electromagnetic waves inside the magnetosphere with losses to the atmosphere. Boundary conditions account for the convective source of electrons and loss to the magnetopause. The results of the simulation that include all of the above processes show a good agreement with the data. The agreement implies that these processes are important for the radiation belt electron dynamics and therefore should be accounted for in outer radiation belt simulations. We also show that the results are very sensitive to the assumed wave model. Our simulations are driven only by the variation of the Kp index and variations of the seed electron population around geosynchronous orbit, which allows the model to be used for forecasting and nowcasting.

  11. Modeling the reactor core of MNSR to simulate its dynamic behavior using the code PARET

    International Nuclear Information System (INIS)

    Hainoun, A.; Alhabet, F.

    2004-02-01

    Using the computer code PARET the core of the MNSR reactor was modelled and the neutronics and thermal hydraulic behaviour of the reactor core for the steady state and selected transients, that deal with step change of reactivity including control rod withdraw starting from steady state at various low power level, were simulated. For this purpose a PARET input model for the core of MNSR reactor has been developed enabling the simulation of neutron kinetic and thermal hydraulic of reactor core including reactivity feedback effects. The neutron kinetic model depends on the point kinetic with 15 groups delayed neutrons including photo neutrons of beryllium reflector. In this regard the effect of photo neutron on the dynamic behaviour has been analysed through two additional calculation. In the first the yield of photo neutrons was neglected completely and in the second its share was added to the sixth group of delayed neutrons. In the thermal hydraulic model the fuel elements with their cooling channels were distributed to 4 different groups with various radial power factors. The pressure lose factors for friction, flow direction change, expansion and contraction were estimated using suitable approaches. The post calculations of the relative neutron flux change and core average temperature were found to be consistent with the experimental measurements. Furthermore, the simulation has indicated the influence of photo neutrons of the Beryllium reflector on the neutron flux behaviour. For the reliability of the results sensitivity analysis was carried out to consider the uncertainty in some important parameters like temperature feedback coefficient and flow velocity. On the other hand the application of PARET in simulation of void formation in the subcooled boiling regime were tested. The calculation indicates the capability of PARET in modelling this phenomenon. However, big discrepancy between calculation results and measurement of axial void distribution were observed

  12. Simulation of KAEVER experiments on aerosol behavior in a nuclear power plant containment at accident conditions with the ASTEC code

    International Nuclear Information System (INIS)

    Kljenak, I.; Mavko, B.

    2006-01-01

    Experiments on aerosol behaviour in saturated and non-saturated atmosphere, which were performed in the KAEVER experimental facility, were simulated with the severe accident computer code ASTEC CPA V1.2. The specific purpose of the work was to assess the capability of the code to model aerosol condensation and deposition in the containment of a light-water-reactor nuclear power plant at severe accident conditions, if the atmosphere saturation conditions are simulated adequately. Five different tests were first simulated with boundary conditions, obtained from the experiments. In all five tests, a non-saturated atmosphere was simulated, although, in four tests, the atmosphere was allegedly saturated. The simulations were repeated with modified boundary conditions, to obtain a saturated atmosphere in all tests. Results of dry and wet aerosol concentrations in the test vessel atmosphere for both sets of simulations are compared to experimental results. (author)

  13. Parameters that affect parallel processing for computational electromagnetic simulation codes on high performance computing clusters

    Science.gov (United States)

    Moon, Hongsik

    What is the impact of multicore and associated advanced technologies on computational software for science? Most researchers and students have multicore laptops or desktops for their research and they need computing power to run computational software packages. Computing power was initially derived from Central Processing Unit (CPU) clock speed. That changed when increases in clock speed became constrained by power requirements. Chip manufacturers turned to multicore CPU architectures and associated technological advancements to create the CPUs for the future. Most software applications benefited by the increased computing power the same way that increases in clock speed helped applications run faster. However, for Computational ElectroMagnetics (CEM) software developers, this change was not an obvious benefit - it appeared to be a detriment. Developers were challenged to find a way to correctly utilize the advancements in hardware so that their codes could benefit. The solution was parallelization and this dissertation details the investigation to address these challenges. Prior to multicore CPUs, advanced computer technologies were compared with the performance using benchmark software and the metric was FLoting-point Operations Per Seconds (FLOPS) which indicates system performance for scientific applications that make heavy use of floating-point calculations. Is FLOPS an effective metric for parallelized CEM simulation tools on new multicore system? Parallel CEM software needs to be benchmarked not only by FLOPS but also by the performance of other parameters related to type and utilization of the hardware, such as CPU, Random Access Memory (RAM), hard disk, network, etc. The codes need to be optimized for more than just FLOPs and new parameters must be included in benchmarking. In this dissertation, the parallel CEM software named High Order Basis Based Integral Equation Solver (HOBBIES) is introduced. This code was developed to address the needs of the

  14. Comparison benchmark between tokamak simulation code and TokSys for Chinese Fusion Engineering Test Reactor vertical displacement control design

    International Nuclear Information System (INIS)

    Qiu Qing-Lai; Xiao Bing-Jia; Guo Yong; Liu Lei; Wang Yue-Hang

    2017-01-01

    Vertical displacement event (VDE) is a big challenge to the existing tokamak equipment and that being designed. As a Chinese next-step tokamak, the Chinese Fusion Engineering Test Reactor (CFETR) has to pay attention to the VDE study with full-fledged numerical codes during its conceptual design. The tokamak simulation code (TSC) is a free boundary time-dependent axisymmetric tokamak simulation code developed in PPPL, which advances the MHD equations describing the evolution of the plasma in a rectangular domain. The electromagnetic interactions between the surrounding conductor circuits and the plasma are solved self-consistently. The TokSys code is a generic modeling and simulation environment developed in GA. Its RZIP model treats the plasma as a fixed spatial distribution of currents which couple with the surrounding conductors through circuit equations. Both codes have been individually used for the VDE study on many tokamak devices, such as JT-60U, EAST, NSTX, DIII-D, and ITER. Considering the model differences, benchmark work is needed to answer whether they reproduce each other’s results correctly. In this paper, the TSC and TokSys codes are used for analyzing the CFETR vertical instability passive and active controls design simultaneously. It is shown that with the same inputs, the results from these two codes conform with each other. (paper)

  15. Simulation of two-phase flows in vertical tubes with the CTFD code FLUBOX

    International Nuclear Information System (INIS)

    Graf, Udo; Papadimitriou, Pavlos

    2007-01-01

    The computational two-fluid dynamics (CTFD) code FLUBOX is developed at GRS for the multidimensional simulation of two-phase flows. The single-pressure two-fluid model is used as basis of the simulation. A basic mathematical property of the two-fluid model of FLUBOX is the hyperbolic character of the advection. The numerical solution methods of FLUBOX make explicit use of the hyperbolic structure of the coefficient matrices. The simulation of two-phase flow phenomena needs, apart from the conservation equations for each phase, an additional transport equation for the interfacial area concentration. The concentration of the interfacial area is one of the key parameters for the modeling of interfacial friction forces and interfacial transfer terms. A new transport equation for the interfacial area concentration is in development. It describes the dynamic change of the interfacial area concentration due to mass exchange and a force balance at the phase boundary. Results from FLUBOX calculations for different experiments of two-phase flows in vertical tubes are presented as part of the validation

  16. Simulations of hydrogen distribution experiments using the PRESCON2 and GOTHIC codes

    International Nuclear Information System (INIS)

    Nguyen, T.H.; Collins, W.M.

    1994-01-01

    The main objective of this work is to develop modelling guidelines in the use of containment models to more accurately predict hydrogen distribution in the HDR facility and to assess the ability of both lumped and distributed parameter models in predicting natural convective flows within containment. Experiences learned from this exercise will be applied to present methodologies used in licensing analyses for CANDU containments. PRESCON2 simulations of hydrogen distribution experiments performed in the HDR facility show hydrogen and helium concentrations are under-predicted at high elevations and over predicted at low elevations. Acceptable predictions of the gas concentration are obtained in the vicinity of the release. Results obtained from GOTHIC simulations using lumped parameter models are very comparable to those predicted by PRESCON2. This indicates that lumped parameter codes tend to over-estimate the degree of mixing of fluids due to the inherent nodal atmospheric homogeneity assumption in their numerical formulation. Results obtained from the GOTHIC simulation using a simple distributed parameter model show little improvement compared to those predicted using the lumped parameter model. This indicates that a simple 3-D model will not be sufficient to make significant improvements in the results. More detailed modelling of the junction flows and finer grids should lead to more accurate results. More detailed investigations employing finer 3-D meshes is under investigation. (author)

  17. Simulations of a DIII-D plasma disruption with the NIMROD code

    International Nuclear Information System (INIS)

    Kruger, S.E.; Schnack, D.D.

    2005-01-01

    To investigate the dynamics of the disruption of DIII-D discharge number 87009, two types of initial-value simulations with the NIMROD code were performed to investigate different characteristics. In the first set of simulations, a conducting wall was placed on the last close flux surface, and an equilibrium that was close to the ideal-MHD marginal stability point was created. A heating source was added to the pressure equation to drive the plasma equilibrium through an ideal-MHD instability point. Excellent agreement with analytic theory was obtained. To investigate how the stored energy is deposited on the wall, free-boundary simulations were performed with an ideal-MHD unstable equilibria. The unstable modes grow until the magnetic islands overlap and the magnetic field is stochastic over a large part of the plasma domain. The rapid stochastization of the field allows the plasma to lose two thirds of its internal energy in approximately 200 microseconds in qualitative agreement with the experiment. The deposition of thermal energy on the wall is localized poloidally and toroidally on the wall due to helically-localized temperature gradients and the rapid parallel heat conduction which carries this heat flux to the wall. (author)

  18. Monte Carlo simulation using the PENELOPE code with an ant colony algorithm to study MOSFET detectors

    Energy Technology Data Exchange (ETDEWEB)

    Carvajal, M A; Palma, A J [Departamento de Electronica y Tecnologia de Computadores, Universidad de Granada, E-18071 Granada (Spain); Garcia-Pareja, S [Servicio de Radiofisica Hospitalaria, Hospital Regional Universitario ' Carlos Haya' , Avda Carlos Haya, s/n, E-29010 Malaga (Spain); Guirado, D [Servicio de RadiofIsica, Hospital Universitario ' San Cecilio' , Avda Dr Oloriz, 16, E-18012 Granada (Spain); Vilches, M [Servicio de Fisica y Proteccion Radiologica, Hospital Regional Universitario ' Virgen de las Nieves' , Avda Fuerzas Armadas, 2, E-18014 Granada (Spain); Anguiano, M; Lallena, A M [Departamento de Fisica Atomica, Molecular y Nuclear, Universidad de Granada, E-18071 Granada (Spain)], E-mail: carvajal@ugr.es, E-mail: garciapareja@gmail.com, E-mail: dguirado@ugr.es, E-mail: mvilches@ugr.es, E-mail: mangui@ugr.es, E-mail: ajpalma@ugr.es, E-mail: lallena@ugr.es

    2009-10-21

    In this work we have developed a simulation tool, based on the PENELOPE code, to study the response of MOSFET devices to irradiation with high-energy photons. The energy deposited in the extremely thin silicon dioxide layer has been calculated. To reduce the statistical uncertainties, an ant colony algorithm has been implemented to drive the application of splitting and Russian roulette as variance reduction techniques. In this way, the uncertainty has been reduced by a factor of {approx}5, while the efficiency is increased by a factor of above 20. As an application, we have studied the dependence of the response of the pMOS transistor 3N163, used as a dosimeter, with the incidence angle of the radiation for three common photons sources used in radiotherapy: a {sup 60}Co Theratron-780 and the 6 and 18 MV beams produced by a Mevatron KDS LINAC. Experimental and simulated results have been obtained for gantry angles of 0 deg., 15 deg., 30 deg., 45 deg., 60 deg. and 75 deg. The agreement obtained has permitted validation of the simulation tool. We have studied how to reduce the angular dependence of the MOSFET response by using an additional encapsulation made of brass in the case of the two LINAC qualities considered.

  19. Preliminary Results of 4-Dimensional Radiation Belt Simulation With the VERB Code

    Science.gov (United States)

    Subbotin, D.; Shprits, Y.

    2012-12-01

    Highly energetic electrons in the Earth's radiation belts are hazardous for satellite equipment. The dynamics of the radiation belt electrons can be described by the Fokker-Plank equation in terms of the radial distance, energy, and equatorial pitch-angle combined with magnetospheric convection model, which provides low energy electron seed population for Earth's radiation belts and is important for understanding of the radiation belts storm time and quite time variations. To better understand the global magnetospheric dynamics we present preliminary results of the combined convective-diffusive 4-dimentional simulations produced with 4-D Versatile Electron Radiation Belt (VERB) code. The simulations include radial diffusion due to ULF waves, energy, and pitch angle scattering due to day- and night-side chorus waves at high and low latitudes, plasmaspheric hiss waves, losses to magnetopause and atmosphere. Radial and azimuthal bounce-averaged electron drifts due to the influence of electric and magnetic fields are calculated to provide realistic low-energy electrons seed population during the simulation.

  20. Electromagnetic behaviour of a plasma in fluid and relativistic regimes: simulation code R H E A; Comportement electromagnetique d`un plasma en regimes hydrodynamique et relativiste: code de simulation R H E A

    Energy Technology Data Exchange (ETDEWEB)

    Bonnaud, G.; Dussy, S.; Lefebvre, E. [CEA Bruyeres-le-Chatel, 91 (France). Dept. de Physique Theorique et Appliquee; Bouchut, F. [Orleans Univ., 45 (France). Dept. de Mathematiques, UMR CNRS

    1998-12-31

    This report presents a numerical model to simulate the electromagnetic processes involved by electrically-charged relativistic fluids. The physical model is first given. Second, the numerical methods are explained with the various packages of the code RHEA, with indication methods are explained with the various packages of the code RHEA, with indication of its performances, within a 1.5.- dimensional framework. Results from test-simulations are shown to validate the use of the code, for both academic situations and realistic context of laser-plasma interaction, for which the code has been designed: the non-linear phenomena in the context of inertial confinement fusion and the ultra-intense laser pulses. (author) 25 refs.

  1. Verification of HYDRASTAR - A code for stochastic continuum simulation of groundwater flow

    International Nuclear Information System (INIS)

    Norman, S.

    1991-07-01

    HYDRASTAR is a code developed at Starprog AB for use in the SKB 91 performance assessment project with the following principal function: - Reads the actual conductivity measurements from a file created from the data base GEOTAB. - Regularizes the measurements to a user chosen calculation scale. - Generates three dimensional unconditional realizations of the conductivity field by using a supplied model of the conductivity field as a stochastic function. - Conditions the simulated conductivity field on the actual regularized measurements. - Reads the boundary conditions from a regional deterministic NAMMU computation. - Calculates the hydraulic head field, Darcy velocity field, stream lines and water travel times by solving the stationary hydrology equation and the streamline equation obtained with the velocities calculated from Darcy's law. - Generates visualizations of the realizations if desired. - Calculates statistics such as semivariograms and expectation values of the output fields by repeating the above procedure by iterations of the Monte Carlo type. When using computer codes for safety assessment purpose validation and verification of the codes are important. Thus this report describes a work performed with the goal of verifying parts of HYDRASTAR. The verification described in this report uses comparisons with two other solutions of related examples: A. Comparison with a so called perturbation solution of the stochastical stationary hydrology equation. This as an analytical approximation of the stochastical stationary hydrology equation valid in the case of small variability of the unconditional random conductivity field. B. Comparison with the (Hydrocoin, 1988), case 2. This is a classical example of a hydrology problem with a deterministic conductivity field. The principal feature of the problem is the presence of narrow fracture zones with high conductivity. the compared output are the hydraulic head field and a number of stream lines originating from a

  2. MCPT: A Monte Carlo code for simulation of photon transport in tomographic scanners

    International Nuclear Information System (INIS)

    Prettyman, T.H.; Gardner, R.P.; Verghese, K.

    1990-01-01

    MCPT is a special-purpose Monte Carlo code designed to simulate photon transport in tomographic scanners. Variance reduction schemes and sampling games present in MCPT were selected to characterize features common to most tomographic scanners. Combined splitting and biasing (CSB) games are used to systematically sample important detection pathways. An efficient splitting game is used to tally particle energy deposition in detection zones. The pulse height distribution of each detector can be found by convolving the calculated energy deposition distribution with the detector's resolution function. A general geometric modelling package, HERMETOR, is used to describe the geometry of the tomographic scanners and provide MCPT information needed for particle tracking. MCPT's modelling capabilites are described and preliminary experimental validation is presented. (orig.)

  3. Particle-in-cell/accelerator code for space-charge dominated beam simulation

    Energy Technology Data Exchange (ETDEWEB)

    2012-05-08

    Warp is a multidimensional discrete-particle beam simulation program designed to be applicable where the beam space-charge is non-negligible or dominant. It is being developed in a collaboration among LLNL, LBNL and the University of Maryland. It was originally designed and optimized for heave ion fusion accelerator physics studies, but has received use in a broader range of applications, including for example laser wakefield accelerators, e-cloud studies in high enery accelerators, particle traps and other areas. At present it incorporates 3-D, axisymmetric (r,z) planar (x-z) and transverse slice (x,y) descriptions, with both electrostatic and electro-magnetic fields, and a beam envelope model. The code is guilt atop the Python interpreter language.

  4. Algorithmes robustes en optimisation non convexe : codes et simulations numériques en grande dimension

    OpenAIRE

    Chine, Abderrazek

    1991-01-01

    Cette thèse est consacrée a l'étude des algorithmes en optimisation non convexe, a l'implémentation des codes a l'usage industriel et aux simulations numériques dans les problèmes de grande tailles. L'étude des problèmes quadratiques (convexes ou non convexes) sous contraintes linéaires et quadratiques ainsi que celle des méthodes de région de confiance pour minimisation d'une fonction de classe c#2, font l'objet de deux premiers chapitres. Les chapitres 3 et 4 sont réservés a l'optimisation ...

  5. A particle simulation code for analysis of nonlinear electron oscillations in a plasma waveguide

    International Nuclear Information System (INIS)

    Turikov, V.A.

    1978-08-01

    A description is given of a computer code for simulation of electron oscillations in a magnetized plasma in a cylindrical waveguide. The one-dimensional particle-in-cell method with the reverse linear interpolation of charge density is used. The program has options for treating nonlinear processes in a plasma with periodical and reflecting boundary conditions. For periodical conditions, Poisson's equation is solved by means of the Fourier method. For reflecting conditions, the double recursive procedure is used. The values of the potential derivatives at the space grid points are calculated by means of the parabolic interpolation. The main purpose of the program is to investigate nonlinear phenomena in a plasma column after applying a short localized impulse of an external electric field. (Auth.)

  6. Update on comparison of the particle production using Mars simulation code

    CERN Document Server

    Prior, G; Kirk, H G; Souchlas, N; Ding, X

    2011-01-01

    In the International Design Study for the Neutrino Factory (IDS-NF), a 5-15 GeV (kinetic energy) proton beam impinges a Hg jet target, in order to produce pions that will decay into muons. The muons are captured and transformed into a beam, then passed to the downstream acceleration system. The target sits in a solenoid eld tapering from 20 T down to below 2 T, over several meters, permitting an optimized capture of the pions that will produce useful muons for the machine. The target and pion capture systems have been simulated using MARS. This paper presents an updated comparison of the particles production using the MARS code versions m1507 and m1510 on different machines located at the European Organization for Nuclear Research (CERN) and Brookhaven National Laboratory (BNL).

  7. KOELSCH, a computer code simulating mechanical HCDA consequences in LMFBR primary containment

    International Nuclear Information System (INIS)

    Lange, L.; Gerling, W.; Schlein, I.

    1981-01-01

    The computer code KOELSCH (Kontinuumsmechanik in Euler-Lagrange Darstellung und Schalentheorie) calculates problems of continuum mechanics in Euler-Lagrangian representation and with the use of shell theory. It is being developed for the advanced simulation of the mechanical loading and the response of the primary containment of a LMFBR in the case of an HCDA. The essential features of the code and examples of calculations are presented. For the hydrodynamics the theoretical basis is the 3D Euler-Lagrangian formulation of the hydrodynamic equations in general curved coordinates. This starting point assure optimum adaptation of the physical problem to the geometry to be considered. The options available in KOELSCH are - H1: quasi- 1D Euler-Lagrangian - H2: 2D Euler - H3: 3D Euler (only rigid boundaries possible). The structural dynamics is treated by 2D Lagrangian formalism in cylindrical coordinates. Two options are foreseen: - thin shells: membrane or shell theory - massive structures: full tensor properties of materials are included. Flexibility with respect to complex geometries is gained by surface integration of the governing equations over finite elements with 3 - 8 corners. (orig./GL)

  8. Referential coding of steering-wheel button presses in a simulated driving cockpit.

    Science.gov (United States)

    Xiong, Aiping; Proctor, Robert W

    2015-12-01

    The present study investigated whether left and right pushbuttons on a steering wheel are coded relative to an "infotainment display" in a simulated driving cockpit. Participants performed a go/no-go Simon task in which they responded on trials for which a tone, presented from a left or right speaker, was 1 of 2 pitches (low or high) with a single button press (left in 1 trial block; right in another). Without the infotainment display in Experiment 1, both left and right responses showed Simon effects of similar size. In both Experiments 2 and 3, the infotainment display was located to the right or left, and the Simon effect was smaller for the response that was on the side of the infotainment display than for the response that was on the opposite side. The results indicate that in a driving cockpit environment, the pushbutton responses are coded as left and right with respect not only to the wheel-based frame but also to a salient object like the infotainment display. The general point for application is that the driver's spatial representation of responses, and consequently performance, can be influenced by multiple frames of reference. (c) 2015 APA, all rights reserved).

  9. Comparison of MHD simulation codes for understanding nonlinear ELMs dynamics in KSTAR H-mode plasma

    Science.gov (United States)

    Kim, M.; Lee, J.; Park, H. K.; Yun, G. S.; Xu, X.; Jardin, S. C.; Becoulet, M.

    2017-10-01

    KSTAR electron cyclotron emission imaging (ECEI) systems have contributed to understanding the fundamental physics of ELMs by high-quality 2D and quasi-3D images of ELMs. However, in the highly nonlinear phase of ELM dynamics, the interpretation of ECE signals becomes complicated intrinsically. Theoretical and numerical approaches are necessary to enhance the understanding of ELM physics. Well-established MHD codes (BOUT + + , JOREK, and M3D-C1) are introduced for comparative study with the observations. The nonlinear solutions are obtained using the same equilibrium of the KSTAR H-mode plasma. Each code shows the partial difference in mode evolution, probably, due to the difference in optimized operation window of initial conditions. The nonlinear simulation results show that low- n (n qualitatively matches with the recent ECEI observation just before ELM-crash, or excitation of non-modal solitary perturbation (typically, n = 1) which is highly localized in poloidal and toroidal. Regardless of differences in details, qualitative similarity can provide inspiration to understand the triggering of ELM-crash. This work is supported by NRF of Korea under Contract No. NRF-2014M1A7A1A03029865.

  10. Development of an object-oriented simulation code for repository performance assessment

    International Nuclear Information System (INIS)

    Tsujimoto, Keiichi; Ahn, J.

    1999-01-01

    As understanding for mechanisms of radioactivity confinement by a deep geologic repository improves at the individual process level, it has become imperative to evaluate consequences of individual processes to the performance of the whole repository system. For this goal, the authors have developed a model for radionuclide transport in, and release from, the repository region by incorporating multiple-member decay chains and multiple waste canisters. A computer code has been developed with C++, an object-oriented language. By utilizing the feature that a geologic repository consists of thousands of objects of the same kind, such as the waste canister, the repository region is divided into multiple compartments and objects for simulation of radionuclide transport. Massive computational tasks are distributed over, and executed by, multiple networked workstations, with the help of parallel virtual machine (PVM) technology. Temporal change of the mass distribution of 28 radionuclides in the repository region for the time period of 100 million yr has been successfully obtained by the code

  11. Comparison of a 3D multi‐group SN particle transport code with Monte Carlo for intercavitary brachytherapy of the cervix uteri

    Science.gov (United States)

    Wareing, Todd A.; Failla, Gregory; Horton, John L.; Eifel, Patricia J.; Mourtada, Firas

    2009-01-01

    A patient dose distribution was calculated by a 3D multi‐group SN particle transport code for intracavitary brachytherapy of the cervix uteri and compared to previously published Monte Carlo results. A Cs‐137 LDR intracavitary brachytherapy CT data set was chosen from our clinical database. MCNPX version 2.5.c, was used to calculate the dose distribution. A 3D multi‐group SN particle transport code, Attila version 6.1.1 was used to simulate the same patient. Each patient applicator was built in SolidWorks, a mechanical design package, and then assembled with a coordinate transformation and rotation for the patient. The SolidWorks exported applicator geometry was imported into Attila for calculation. Dose matrices were overlaid on the patient CT data set. Dose volume histograms and point doses were compared. The MCNPX calculation required 14.8 hours, whereas the Attila calculation required 22.2 minutes on a 1.8 GHz AMD Opteron CPU. Agreement between Attila and MCNPX dose calculations at the ICRU 38 points was within ±3%. Calculated doses to the 2 cc and 5 cc volumes of highest dose differed by not more than ±1.1% between the two codes. Dose and DVH overlays agreed well qualitatively. Attila can calculate dose accurately and efficiently for this Cs‐137 CT‐based patient geometry. Our data showed that a three‐group cross‐section set is adequate for Cs‐137 computations. Future work is aimed at implementing an optimized version of Attila for radiotherapy calculations. PACS number: 87.53.Jw

  12. Icarus: A 2-D Direct Simulation Monte Carlo (DSMC) Code for Multi-Processor Computers

    International Nuclear Information System (INIS)

    BARTEL, TIMOTHY J.; PLIMPTON, STEVEN J.; GALLIS, MICHAIL A.

    2001-01-01

    Icarus is a 2D Direct Simulation Monte Carlo (DSMC) code which has been optimized for the parallel computing environment. The code is based on the DSMC method of Bird[11.1] and models from free-molecular to continuum flowfields in either cartesian (x, y) or axisymmetric (z, r) coordinates. Computational particles, representing a given number of molecules or atoms, are tracked as they have collisions with other particles or surfaces. Multiple species, internal energy modes (rotation and vibration), chemistry, and ion transport are modeled. A new trace species methodology for collisions and chemistry is used to obtain statistics for small species concentrations. Gas phase chemistry is modeled using steric factors derived from Arrhenius reaction rates or in a manner similar to continuum modeling. Surface chemistry is modeled with surface reaction probabilities; an optional site density, energy dependent, coverage model is included. Electrons are modeled by either a local charge neutrality assumption or as discrete simulational particles. Ion chemistry is modeled with electron impact chemistry rates and charge exchange reactions. Coulomb collision cross-sections are used instead of Variable Hard Sphere values for ion-ion interactions. The electro-static fields can either be: externally input, a Langmuir-Tonks model or from a Green's Function (Boundary Element) based Poison Solver. Icarus has been used for subsonic to hypersonic, chemically reacting, and plasma flows. The Icarus software package includes the grid generation, parallel processor decomposition, post-processing, and restart software. The commercial graphics package, Tecplot, is used for graphics display. All of the software packages are written in standard Fortran

  13. Development of a severe accident training simulator using a MELCOR code

    International Nuclear Information System (INIS)

    Kim, Ko Ryu; Jeong, Kwang Sub; Ha, Jae Joo; Jung, Won Dae

    2002-03-01

    Nuclear power plants' severe accidents are, despite of their rareness, very important in safety aspects, because of their huge damages when occurred. For the appropriate execution of severe accident strategy, more information for decision-making are required because of the uncertainties included in severe accidents. Earlier NRC raised concerns over severe accident training in the report NUREC/CR-477, and accordingly, developing effective training tools for severe accident were emphasized. In fact the training tools were requested from industrial area, nevertheless, few training tools were developed due to the uncertainties in severe accidents, lacks of analysis computer codes and technical limitations. SATS, the severe accident training simulator, is developed as a multi-purpose tools for severe accident training. SATS uses the calculation results of MELCOR, an integral severe accident analysis code, and with the help of SL-GMS graphic tools, provides dynamic displays of severe accident phenomena on the terminal of IBM PC. It aimed to have two main features: one is to provide graphic displays to represent severe accident phenomena and the other is to process and simulate severe accident strategy given by plant operators and TSC staffs. Severe accident strategies are basically composed of series of operations of available pumps, valves and other equipments. Whenever executing strategies with SATS, the trainee should follow the HyperKAMG, the on line version of the recently developed severe accident guidance (KAMG). Severe accident strategies are closely related to accidents scenarios. TLOFW and LOCA , two representative severe accident scenarios of Uljin 3,4, are developed as a built-in scenarios of SATS. Although SATS has some minor problems at this time, we expect SATS will be a good severe accident training tool after the appropriate addition of accident scenarios. Moreover, we also expect SATS will be a good advisory tool for the severe accident research

  14. MCNPX calculations of dose rate distribution inside samples treated in the research gamma irradiating facility at CTEx

    Energy Technology Data Exchange (ETDEWEB)

    Rusin, Tiago; Rebello, Wilson F.; Vellozo, Sergio O.; Gomes, Renato G., E-mail: tiagorusin@ime.eb.b, E-mail: rebello@ime.eb.b, E-mail: vellozo@cbpf.b, E-mail: renatoguedes@ime.eb.b [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Dept. de Engenharia Nuclear; Vital, Helio C., E-mail: vital@ctex.eb.b [Centro Tecnologico do Exercito (CTEx), Rio de Janeiro, RJ (Brazil); Silva, Ademir X., E-mail: ademir@con.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear

    2011-07-01

    A cavity-type cesium-137 research irradiating facility at CTEx has been modeled by using the Monte Carlo code MCNPX. The irradiator has been daily used in experiments to optimize the use of ionizing radiation for conservation of many kinds of food and to improve materials properties. In order to correlate the effects of the treatment, average doses have been calculated for each irradiated sample, accounting for the measured dose rate distribution in the irradiating chambers. However that approach is only approximate, being subject to significant systematic errors due to the heterogeneous internal structure of most samples that can lead to large anisotropy in attenuation and Compton scattering properties across the media. Thus this work is aimed at further investigating such uncertainties by calculating the dose rate distribution inside the items treated such that a more accurate and representative estimate of the total absorbed dose can be determined for later use in the effects-versus-dose correlation curves. Samples of different simplified geometries and densities (spheres, cylinders, and parallelepipeds), have been modeled to evaluate internal dose rate distributions within the volume of the samples and the overall effect on the average dose. (author)

  15. MCNPX calculations of dose rate distribution inside samples treated in the research gamma irradiating facility at CTEx

    International Nuclear Information System (INIS)

    Rusin, Tiago; Rebello, Wilson F.; Vellozo, Sergio O.; Gomes, Renato G.; Silva, Ademir X.

    2011-01-01

    A cavity-type cesium-137 research irradiating facility at CTEx has been modeled by using the Monte Carlo code MCNPX. The irradiator has been daily used in experiments to optimize the use of ionizing radiation for conservation of many kinds of food and to improve materials properties. In order to correlate the effects of the treatment, average doses have been calculated for each irradiated sample, accounting for the measured dose rate distribution in the irradiating chambers. However that approach is only approximate, being subject to significant systematic errors due to the heterogeneous internal structure of most samples that can lead to large anisotropy in attenuation and Compton scattering properties across the media. Thus this work is aimed at further investigating such uncertainties by calculating the dose rate distribution inside the items treated such that a more accurate and representative estimate of the total absorbed dose can be determined for later use in the effects-versus-dose correlation curves. Samples of different simplified geometries and densities (spheres, cylinders, and parallelepipeds), have been modeled to evaluate internal dose rate distributions within the volume of the samples and the overall effect on the average dose. (author)

  16. Gas stratification break-up by a vertical jet: Simulations using the GOTHIC code

    International Nuclear Information System (INIS)

    Andreani, Michele; Kapulla, Ralf; Zboray, Robert

    2012-01-01

    Highlights: ► Simulations of experiments addressing helium stratification break-up with GOTHIC are presented. ► In the tests, the initial helium-rich layer in a large vessel is eroded by a vertical jet. ► A 3-D coarse mesh and various finer 2-D meshes have been used for the simulations. ► In general, the 3-D calculations predict too slow mixing in the vessel. ► A reasonable agreement between calculated and measured gas concentrations requires a fine mesh. - Abstract: The capability assessment of three-dimensional computational tools to predict the erosion and the break-up of stratified conditions that can build-up in a containment through the release of hydrogen during an early phase of a hypothetical severe accident is the focus of intense research worldwide. In conjunction with the OECD SETH-2 project, the GOTHIC code is assessed against experiments in which mass and/or heat sources or sinks cause mixing. This paper reports on simulation results of selected experiments where the initial helium stratification in a vessel is eroded by a vertical jet originating from an injection below the initial density interface. A 3-D coarse mesh, as well as various finer 2-D meshes, is used to simulate the evolution of the helium distribution generated by jets having different initial momentum. In general, the 3-D calculations predict too slow mixing in the vessel and a reasonable agreement between calculated and measured gas concentrations can only be achieved with a sufficiently fine mesh. These results can be explained by comparing the calculated velocity field with that measured using the PIV technique, which also provides valuable insight into the mechanisms of the interaction between the jet and the density interface.

  17. Simulation of IST Turbomachinery Power-Neutral Tests with the ANL Plant Dynamics Code

    Energy Technology Data Exchange (ETDEWEB)

    Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-13

    The validation of the Plant Dynamics Code (PDC) developed at Argonne National Laboratory (ANL) for the steady-state and transient analysis of supercritical carbon dioxide (sCO2) systems has been continued with new test data from the Naval Nuclear Laboratory (operated by Bechtel Marine Propulsion Corporation) Integrated System Test (IST). Although data from three runs were provided to ANL, only two of the data sets were analyzed and described in this report. The common feature of these tests is the power-neutral operation of the turbine-compressor shaft, where no external power through the alternator was provided during the tests. Instead, the shaft speed was allowed to change dictated by the power balance between the turbine, the compressor, and the power losses in the shaft. The new test data turned out to be important for code validation for several reasons. First, the power-neutral operation of the shaft allows validation of the shaft dynamics equations in asynchronous mode, when the shaft is disconnected from the grid. Second, the shaft speed control with the compressor recirculation (CR) valve not only allows for testing the code control logic itself, but it also serves as a good test for validation of both the compressor surge control and the turbine bypass control actions, since the effect of the CR action on the loop conditions is similar for both of these controls. Third, the varying compressor-inlet temperature change test allows validation of the transient response of the precooler, a shell-and-tube heat exchanger. The first transient simulation of the compressor-inlet temperature variation Test 64661 showed a much slower calculated response of the precooler in the calculations than the test data. Further investigation revealed an error in calculating the heat exchanger tube mass for the PDC dynamic equations that resulted in a slower change in the tube wall temperature than measured. The transient calculations for both tests were done in two steps. The

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

    Science.gov (United States)

    2015-11-01

    data .  scaleFactor: Scale factor to convert SYSWELD units to units in Abaqus analysis for field of interest. A help description can be obtained by...Memorandum Simulation of Weld Mechanical Behavior to Include Welding-Induced Residual Stress and Distortion: Coupling of SYSWELD and Abaqus Codes...Weld Mechanical Behavior to Include Welding-Induced Residual Stress and Distortion: Coupling of SYSWELD and Abaqus Codes by Charles R. Fisher

  19. FUMEX III Project (Improvement of Computer Codes Used for Fuel Behaviour Simulation). ENEA Contribution

    International Nuclear Information System (INIS)

    Calabrese, R.

    2013-01-01

    In a liberalised electricity market the improvement of nuclear fuel performance in terms of burn-up, reliability and safety directly affects the economic performance a fundamental factor in the assessment of nuclear energy sustainability (IAEA 2011). In this perspective, it is therefore relevant to have available fuel performance codes deeply verified and validated to perform with due accuracy safety assessment, fuel rod design and studies on fuel rod behaviour under different operating conditions, both normal and accidental. Fuel performance codes can be classified into national, with origins and histories of independent development, and international, distributed to multiple partners and then specialized according to the needs of individual users. On one hand codes are therefore different for history and stage of development, on the other hand new experimental data provide to researchers further information useful to improve the knowledge of fuel behaviour in an extended burn-up domain. Furthermore improvements of materials properties in parallel with investigations on new materials for nuclear applications are steadily proceeding. In the area of fuel modelling the IAEA launched a series of coordinated research projects (CRPs) since 1981 with the first project called DCOM followed by FUMEX and FUMEX II projects held in 1993-1996 and 2002-2006 respectively (IAEA 1998; Killeen et al. 2007).In 2006 and in 2007 the Technical Working Group on Fuel Performance and Technology (TWGFPT) of the IAEA recognized the need for a FUMEX III with the aim of supporting Member States in refining their codes in particular in the high burn-up domain of LWR applications (UO2, MOX). The Improvement of Computer Codes Used for Fuel Behaviour Simulation project (FUMEX III), coordinated by the IAEA, was an important opportunity for Member States to develop and extend the validation of their fuel performance codes on the basis of experimental data made available by the OECD/NEA and the IAEA

  20. Methodology and Toolset for Model Verification, Hardware/Software co-simulation, Performance Optimisation and Customisable Source-code generation

    DEFF Research Database (Denmark)

    Berger, Michael Stübert; Soler, José; Yu, Hao

    2013-01-01

    The MODUS project aims to provide a pragmatic and viable solution that will allow SMEs to substantially improve their positioning in the embedded-systems development market. The MODUS tool will provide a model verification and Hardware/Software co-simulation tool (TRIAL) and a performance...... optimisation and customisable source-code generation tool (TUNE). The concept is depicted in automated modelling and optimisation of embedded-systems development. The tool will enable model verification by guiding the selection of existing open-source model verification engines, based on the automated analysis...... of system properties, and producing inputs to be fed into these engines, interfacing with standard (SystemC) simulation platforms for HW/SW co-simulation, customisable source-code generation towards respecting coding standards and conventions and software performance-tuning optimisation through automated...

  1. Simulated and measured neutron/gamma light output distribution for poly-energetic neutron/gamma sources

    Science.gov (United States)

    Hosseini, S. A.; Zangian, M.; Aghabozorgi, S.

    2018-03-01

    In the present paper, the light output distribution due to poly-energetic neutron/gamma (neutron or gamma) source was calculated using the developed MCNPX-ESUT-PE (MCNPX-Energy engineering of Sharif University of Technology-Poly Energetic version) computational code. The simulation of light output distribution includes the modeling of the particle transport, the calculation of scintillation photons induced by charged particles, simulation of the scintillation photon transport and considering the light resolution obtained from the experiment. The developed computational code is able to simulate the light output distribution due to any neutron/gamma source. In the experimental step of the present study, the neutron-gamma discrimination based on the light output distribution was performed using the zero crossing method. As a case study, 241Am-9Be source was considered and the simulated and measured neutron/gamma light output distributions were compared. There is an acceptable agreement between the discriminated neutron/gamma light output distributions obtained from the simulation and experiment.

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

    International Nuclear Information System (INIS)

    Demaziere, Christophe; Staalek, Mathias

    2004-12-01

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

  3. Environmental construction of nano-material design codes. The example of simulation codes used in the CMD workshop

    Energy Technology Data Exchange (ETDEWEB)

    Miyazaki, Mikiya [Japan Atomic Energy Research Inst., Center for Promotion of Computational Science and Engineering, Kizu, Kyoto (Japan)

    2003-05-01

    Generally it is well known that the R and D works on new materials or devices will play a central role on the evolution of future society. But, the old ways based on the empirical and experimental approach have already reached the limit, especially for dealing with a strange substance and material. The structure of a substance and material is needed to be dealt with in detail by quantum mechanics, because the limit on accuracy has come in sight in the calculation using a classical theory. The research on the latest electronic state calculation technique founded on quantum mechanics made a great advance as the technique of solving these problems as well as the technique of a computational materials design. It enables the prediction of material properties because it is based on First Principles. Therefore, in the future it is expected to have a very high possibility of becoming a breakthrough in such a situation. In this article, the example calculation results by PC cluster on the codes (MACHIKANEYAMA-2000, OSAKA-2000) used in the CMD (Computational Materials Design) workshop, held on Sep. 19-21, at ITBL-Building and International Institute for Advanced Studies under the auspices of the University of Osaka, are described. Furthermore, the graphical user interfaces on the codes are examined. (author)

  4. V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

    Energy Technology Data Exchange (ETDEWEB)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

    2010-07-15

    V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

  5. A simple method for simulation of coherent synchrotron radiation in a tracking code

    International Nuclear Information System (INIS)

    Borland, M.

    2000-01-01

    Coherent synchrotron radiation (CSR) is of great interest to those designing accelerators as drivers for free-electron lasers (FELs). Although experimental evidence is incomplete, CSR is predicted to have potentially severe effects on the emittance of high-brightness electron beams. The performance of an FEL depends critically on the emittance, current, and energy spread of the beam. Attempts to increase the current through magnetic bunch compression can lead to increased emittance and energy spread due to CSR in the dipoles of such a compressor. The code elegant was used for design and simulation of the bunch compressor for the Low-Energy Undulator Test Line (LEUTL) FEL at the Advanced Photon Source (APS). In order to facilitate this design, a fast algorithm was developed based on the 1-D formalism of Saldin and coworkers. In addition, a plausible method of including CSR effects in drift spaces following the chicane magnets was developed and implemented. The algorithm is fast enough to permit running hundreds of tolerance simulations including CSR for 50 thousand particles. This article describes the details of the implementation and shows results for the APS bunch compressor

  6. Multiphase three-dimensional direct numerical simulation of a rotating impeller with code Blue

    Science.gov (United States)

    Kahouadji, Lyes; Shin, Seungwon; Chergui, Jalel; Juric, Damir; Craster, Richard V.; Matar, Omar K.

    2017-11-01

    The flow driven by a rotating impeller inside an open fixed cylindrical cavity is simulated using code Blue, a solver for massively-parallel simulations of fully three-dimensional multiphase flows. The impeller is composed of four blades at a 45° inclination all attached to a central hub and tube stem. In Blue, solid forms are constructed through the definition of immersed objects via a distance function that accounts for the object's interaction with the flow for both single and two-phase flows. We use a moving frame technique for imposing translation and/or rotation. The variation of the Reynolds number, the clearance, and the tank aspect ratio are considered, and we highlight the importance of the confinement ratio (blade radius versus the tank radius) in the mixing process. Blue uses a domain decomposition strategy for parallelization with MPI. The fluid interface solver is based on a parallel implementation of a hybrid front-tracking/level-set method designed complex interfacial topological changes. Parallel GMRES and multigrid iterative solvers are applied to the linear systems arising from the implicit solution for the fluid velocities and pressure in the presence of strong density and viscosity discontinuities across fluid phases. EPSRC, UK, MEMPHIS program Grant (EP/K003976/1), RAEng Research Chair (OKM).

  7. A simulation of a pebble bed reactor core by the MCNP-4C computer code

    Directory of Open Access Journals (Sweden)

    Bakhshayesh Moshkbar Khalil

    2009-01-01

    Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

  8. The design and implementation of Los Alamos PLasma Simulation (LAPS) code

    Science.gov (United States)

    Corbetta, Alessandro; Missanelli, Maria; Pagliantini, Cecilia; Scarabosio, Laura; Delzanno, Gian Luca; Guo, Zehua; Srinivasan, Bhuvana; Tang, Xianzhu

    2012-03-01

    Los Alamos Plasma Simulation (LAPS) is an integrated modeling code based on a common-data framework for multiphysics simulation of both magnetic and inertial confinment fusion (ICF) plasmas. Its principal design goal is to provide a common data structure on computational grids and plasma states for different components of the multiphysics integration. LAPS provides an optimal mesh generation for one to three dimensional configuration space discretization and an adaptive mesh scheme that equi-distributes application-specified error. The plasma state is defined on this mesh. LAPS supports the solution of moment and kinetic equations using grids, particle-in-cell, Monte-Carlo, and molecular dynamics. The parallel data structure and (non)linear solvers for PDEs are based on PETSc, while the parallel data structure and communication for particle and Monte-Carlo method are native to LAPS. LAPS separates the numerical discretization from application PDEs. The initial focus is on spectral method, including the spectral element/volume and discontinuous Garlekin scheme for conservative PDEs. The initial set of applications for LAPS development include PIC modeling of plasma transport and rotation in field reversed configuration, fluid-moment and kinetic model of tokamak scrape-off layer.

  9. Simulation of dental intensifying screen for intraoral radiographic using MCNP5 code

    International Nuclear Information System (INIS)

    Ferreira, Vanessa M.; Oliveira, Renato C.M.; Barros, Graiciany P.; Oliveira, Arno H.; Veloso, M. Auxiliadora F.

    2011-01-01

    One of basic principles for radiological protection is the optimization of techniques for obtain radiographic images, in way that the dose in the patient is kept as low as reasonably achievable (ALARA). Intensifying screens are used in medical radiology, which reduce considerably the dose rates in the production of radiographic images, maintaining the quality of these, while in dental radiology, there is no a intensifying screen available for intraoral examinations. From this technological requirement, this paper evaluates a computational modeling of an intensifying screen for use in intraoral radiography. For this, it was used the Monte Carlo code MCNP5 that allows the radiography simulation through the transport of electrons and photons in the different materials present in this examination. The goal of an intensifying screen is the conversion of X-ray photons to photons in the visible spectrum, knowing that radiographic films are more sensitive to light photons than to X-ray photons. So the screen should be composed of an efficient material for converting x-rays photons in light photons, therefore was made simulations using different materials, thicknesses and positions possible for placing screen in radiographic film in order to find the way more technically feasible. (author)

  10. Simulation of the BNCT of Brain Tumors Using MCNP Code: Beam Designing and Dose Evaluation

    Directory of Open Access Journals (Sweden)

    Fatemeh Sadat Rasouli

    2012-09-01

    Full Text Available Introduction BNCT is an effective method to destroy brain tumoral cells while sparing the healthy tissues. The recommended flux for epithermal neutrons is 109 n/cm2s, which has the most effectiveness on deep-seated tumors. In this paper, it is indicated that using D-T neutron source and optimizing of Beam Shaping Assembly (BSA leads to treating brain tumors in a reasonable time where all IAEA recommended criteria are met. Materials and Methods The proposed BSA based on a D-T neutron generator consists of a neutron multiplier system, moderators, reflector, and collimator. The simulated Snyder head phantom is used to evaluate dose profiles in tissues due to the irradiation of designed beam. Monte Carlo Code, MCNP-4C, was used in order to perform these calculations.   Results The neutron beam associated with the designed and optimized BSA has an adequate epithermal flux at the beam port and neutron and gamma contaminations are removed as much as possible. Moreover, it was showed that increasing J/Φ, as a measure of beam directionality, leads to improvement of beam performance and survival of healthy tissues surrounding the tumor. Conclusion According to the simulation results, the proposed system based on D-T neutron source, which is suitable for in-hospital installations, satisfies all in-air parameters. Moreover, depth-dose curves investigate proper performance of designed beam in tissues. The results are comparable with the performances of other facilities.

  11. Simulation of dental intensifying screen for intraoral radiographic using MCNP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Vanessa M.; Oliveira, Renato C.M., E-mail: vanessamachado@ufmg.br [Curso Superior de Tecnologia em Radiologia. Faculdade de Medicina da Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil); Barros, Graiciany P.; Oliveira, Arno H.; Veloso, M. Auxiliadora F. [Departamento de Engenharia Nuclear. Escola de Engenharia. Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil)

    2011-07-01

    One of basic principles for radiological protection is the optimization of techniques for obtain radiographic images, in way that the dose in the patient is kept as low as reasonably achievable (ALARA). Intensifying screens are used in medical radiology, which reduce considerably the dose rates in the production of radiographic images, maintaining the quality of these, while in dental radiology, there is no a intensifying screen available for intraoral examinations. From this technological requirement, this paper evaluates a computational modeling of an intensifying screen for use in intraoral radiography. For this, it was used the Monte Carlo code MCNP5 that allows the radiography simulation through the transport of electrons and photons in the different materials present in this examination. The goal of an intensifying screen is the conversion of X-ray photons to photons in the visible spectrum, knowing that radiographic films are more sensitive to light photons than to X-ray photons. So the screen should be composed of an efficient material for converting x-rays photons in light photons, therefore was made simulations using different materials, thicknesses and positions possible for placing screen in radiographic film in order to find the way more technically feasible. (author)

  12. Laser-Plasma Modeling Using PERSEUS Extended-MHD Simulation Code for HED Plasmas

    Science.gov (United States)

    Hamlin, Nathaniel; Seyler, Charles

    2017-10-01

    We discuss the use of the PERSEUS extended-MHD simulation code for high-energy-density (HED) plasmas in modeling the influence of Hall and electron inertial physics on laser-plasma interactions. By formulating the extended-MHD equations as a relaxation system in which the current is semi-implicitly time-advanced using the Generalized Ohm's Law, PERSEUS enables modeling of extended-MHD phenomena (Hall and electron inertial physics) without the need to resolve the smallest electron time scales, which would otherwise be computationally prohibitive in HED plasma simulations. We first consider a laser-produced plasma plume pinched by an applied magnetic field parallel to the laser axis in axisymmetric cylindrical geometry, forming a conical shock structure and a jet above the flow convergence. The Hall term produces low-density outer plasma, a helical field structure, flow rotation, and field-aligned current, rendering the shock structure dispersive. We then model a laser-foil interaction by explicitly driving the oscillating laser fields, and examine the essential physics governing the interaction. This work is supported by the National Nuclear Security Administration stewardship sciences academic program under Department of Energy cooperative agreements DE-FOA-0001153 and DE-NA0001836.

  13. Parallel deposition, sorting, and reordering methods in the Hybrid Ordered Plasma Simulation (HOPS) code

    International Nuclear Information System (INIS)

    Anderson, D.V.; Shumaker, D.E.

    1993-01-01

    From a computational standpoint, particle simulation calculations for plasmas have not adapted well to the transitions from scalar to vector processing nor from serial to parallel environments. They have suffered from inordinate and excessive accessing of computer memory and have been hobbled by relatively inefficient gather-scatter constructs resulting from the use of indirect indexing. Lastly, the many-to-one mapping characteristic of the deposition phase has made it difficult to perform this in parallel. The authors' code sorts and reorders the particles in a spatial order. This allows them to greatly reduce the memory references, to run in directly indexed vector mode, and to employ domain decomposition to achieve parallelization. In this hybrid simulation the electrons are modeled as a fluid and the field equations solved are obtained from the electron momentum equation together with the pre-Maxwell equations (displacement current neglected). Either zero or finite electron mass can be used in the electron model. The resulting field equations are solved with an iteratively explicit procedure which is thus trivial to parallelize. Likewise, the field interpolations and the particle pushing is simple to parallelize. The deposition, sorting, and reordering phases are less simple and it is for these that the authors present detailed algorithms. They have now successfully tested the parallel version of HOPS in serial mode and it is now being readied for parallel execution on the Cray C-90. They will then port HOPS to a massively parallel computer, in the next year

  14. Rn3D: A finite element code for simulating gas flow and radon transport in variably saturated, nonisothermal porous media

    International Nuclear Information System (INIS)

    Holford, D.J.

    1994-01-01

    This document is a user's manual for the Rn3D finite element code. Rn3D was developed to simulate gas flow and radon transport in variably saturated, nonisothermal porous media. The Rn3D model is applicable to a wide range of problems involving radon transport in soil because it can simulate either steady-state or transient flow and transport in one-, two- or three-dimensions (including radially symmetric two-dimensional problems). The porous materials may be heterogeneous and anisotropic. This manual describes all pertinent mathematics related to the governing, boundary, and constitutive equations of the model, as well as the development of the finite element equations used in the code. Instructions are given for constructing Rn3D input files and executing the code, as well as a description of all output files generated by the code. Five verification problems are given that test various aspects of code operation, complete with example input files, FORTRAN programs for the respective analytical solutions, and plots of model results. An example simulation is presented to illustrate the type of problem Rn3D is designed to solve. Finally, instructions are given on how to convert Rn3D to simulate systems other than radon, air, and water

  15. Simulation and study on the γ response spectrum of BGO detector by the application of monte carlo code MOCA

    International Nuclear Information System (INIS)

    Jia Wenbao; Chen Xiaowen; Xu Aiguo; Li Anmin

    2010-01-01

    Application of Monte Carlo method to build spectra library is useful to reduce experiment workload in Prompt Gamma Neutron Activation Analysis (PGNAA). The new Monte Carlo Code MOCA was used to simulate the response spectra of BGO detector for gamma rays from 137 Cs, 60 Co and neutron induced gamma rays from S and Ti. The results were compared with general code MCNP, show that the agreement of MOCA between simulation and experiment is better than MCNP. This research indicates that building spectra library by Monte Carlo method is feasible. (authors)

  16. Simulation of the turbine discharge transient with the code Trace; Simulacion del transitorio disparo de turbina con el codigo TRACE

    Energy Technology Data Exchange (ETDEWEB)

    Mejia S, D. M.; Filio L, C., E-mail: dulcemaria.mejia@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2014-10-15

    In this paper the results of the simulation of the turbine discharge transient are shown, occurred in Unit 1 of nuclear power plant of Laguna Verde (NPP-L V), carried out with the model of this unit for the best estimate code Trace. The results obtained by the code Trace are compared with those obtained from the Process Information Integral System (PIIS) of the NPP-L V. The reactor pressure, level behavior in the down-comer, steam flow and flow rate through the recirculation circuits are compared. The results of the simulation for the operation power of 2027 MWt, show concordance with the system PIIS. (Author)

  17. Accurate simulation of ionisation chamber response with the Monte Carlo code PENELOPE

    International Nuclear Information System (INIS)

    Sempau, Josep; Andreo, Pedro

    2011-01-01

    Ionisation chambers (IC) are routinely used in hospitals for the dosimetry of the photon and electron beams used for radiotherapy treatments. The determination of absorbed dose to water from the absorbed dose to the air filling the cavity requires the introduction of stopping power ratios and perturbation factors, which account for the disturbance caused by the presence of the chamber. Although this may seem a problem readily amenable to Monte Carlo simulation, the fact is that the accurate determination of IC response has been, for various decades, one of the most important challenges of the simulation of electromagnetic showers. The main difficulty stems from the use of condensed history techniques for electron and positron transport. This approach, which involves grouping a large number of interactions into a single artificial event, is known to produce the so-called interface effects when particles travel across surfaces separating different media. These effects can be sizeable when the electron step length is not negligible compared to the size of the region being crossed, as it is the case with the cavity of an IC. The artefact, which becomes apparent when the chamber response shows a marked dependence on the adopted step size, can be palliated with the use of sophisticated electron transport algorithms. These topics are discussed in the context of the transport model implemented in the PENELOPE code. The degree of violation of the Fano theorem for a simple, planar geometry, is used as a measure of the stability of the algorithm with respect to variations of the electron step length, thus assessing the 'quality' of its condensed history scheme. It is shown that, with a suitable choice of transport parameters, PENELOPE simulates IC response with an accuracy of the order of 0.1%.

  18. SU-E-T-254: Optimization of GATE and PHITS Monte Carlo Code Parameters for Uniform Scanning Proton Beam Based On Simulation with FLUKA General-Purpose Code

    International Nuclear Information System (INIS)

    Kurosu, K; Takashina, M; Koizumi, M; Das, I; Moskvin, V

    2014-01-01

    Purpose: Monte Carlo codes are becoming important tools for proton beam dosimetry. However, the relationships between the customizing parameters and percentage depth dose (PDD) of GATE and PHITS codes have not been reported which are studied for PDD and proton range compared to the FLUKA code and the experimental data. Methods: The beam delivery system of the Indiana University Health Proton Therapy Center was modeled for the uniform scanning beam in FLUKA and transferred identically into GATE and PHITS. This computational model was built from the blue print and validated with the commissioning data. Three parameters evaluated are the maximum step size, cut off energy and physical and transport model. The dependence of the PDDs on the customizing parameters was compared with the published results of previous studies. Results: The optimal parameters for the simulation of the whole beam delivery system were defined by referring to the calculation results obtained with each parameter. Although the PDDs from FLUKA and the experimental data show a good agreement, those of GATE and PHITS obtained with our optimal parameters show a minor discrepancy. The measured proton range R90 was 269.37 mm, compared to the calculated range of 269.63 mm, 268.96 mm, and 270.85 mm with FLUKA, GATE and PHITS, respectively. Conclusion: We evaluated the dependence of the results for PDDs obtained with GATE and PHITS Monte Carlo generalpurpose codes on the customizing parameters by using the whole computational model of the treatment nozzle. The optimal parameters for the simulation were then defined by referring to the calculation results. The physical model, particle transport mechanics and the different geometrybased descriptions need accurate customization in three simulation codes to agree with experimental data for artifact-free Monte Carlo simulation. This study was supported by Grants-in Aid for Cancer Research (H22-3rd Term Cancer Control-General-043) from the Ministry of Health

  19. GETRAN: A generic, modularly structured computer code for simulation of dynamic behavior of aero- and power generation gas turbine engines

    Science.gov (United States)

    Schobeiri, M. T.; Attia, M.; Lippke, C.

    1994-07-01

    The design concept, the theoretical background essential for the development of the modularly structured simulation code GETRAN, and several critical simulation cases are presented in this paper. The code being developed under contract with NASA Lewis Research Center is capable of simulating the nonlinear dynamic behavior of single- and multispool core engines, turbofan engines, and power generation gas turbine engines under adverse dynamic operating conditions. The modules implemented into GETRAN correspond to components of existing and new-generation aero- and stationary gas turbine engines with arbitrary configuration and arrangement. For precise simulation of turbine and compressor components, row-by-row diabatic and adiabatic calculation procedures are implemented that account for the specific turbine and compressor cascade, blade geometry, and characteristics. The nonlinear, dynamic behavior of the subject engine is calculated solving a number of systems of partial differential equations, which describe the unsteady behavior of each component individually. To identify each differential equation system unambiguously, special attention is paid to the addressing of each component. The code is capable of executing the simulation procedure at four levels, which increase with the degree of complexity of the system and dynamic event. As representative simulations, four different transient cases with single- and multispool thrust and power generation engines were simulated. These transient cases vary from throttling the exit nozzle area, operation with fuel schedule, rotor speed control, to rotating stall and surge.

  20. Comparison and validation of the results of the AZNHEX v.1.0 code with the MCNP code simulating the core of a fast reactor cooled with sodium

    International Nuclear Information System (INIS)

    Galicia A, J.; Francois L, J. L.; Bastida O, G. E.; Esquivel E, J.

    2016-09-01

    The development of the AZTLAN platform for the analysis and design of nuclear reactors is led by Instituto Nacional de Investigaciones Nucleares (ININ) and divided into four working groups, which have well-defined activities to achieve significant progress in this project individually and jointly. Within these working groups is the users group, whose main task is to use the codes that make up the AZTLAN platform to provide feedback to the developers, and in this way to make the final versions of the codes are efficient and at the same time reliable and easy to understand. In this paper we present the results provided by the AZNHEX v.1.0 code when simulating the core of a fast reactor cooled with sodium at steady state. The validation of these results is a fundamental part of the platform development and responsibility of the users group, so in this research the results obtained with AZNHEX are compared and analyzed with those provided by the Monte Carlo code MCNP-5, software worldwide used and recognized. A description of the methodology used with MCNP-5 is also presented for the calculation of the interest variables and the difference that is obtained with respect to the calculated with AZNHEX. (Author)

  1. Numerical simulation of self-priming phenomena in venturi scrubber by two-phase flow simulation code TPFIT

    International Nuclear Information System (INIS)

    Horiguchi, Naoki; Kanagawa, Tetsuya; Kaneko, Akiko; Abe, Yutaka; Yoshida, Hiroyuki

    2015-01-01

    In the wake of Fukushima Daiichi nuclear disaster, reviews of the safety of nuclear facilities have been conducted in the world beginning with Japan. Countermeasures against severe accidents in nuclear power plants are an urgent need. In particular, from the viewpoint of protecting containment and suppressing diffusion of the radioactive materials, it is important to install filtered venting devices to release high pressure pollutant gas to the atmosphere with elimination radioactive materials in the gas. One of the devices for the filtered venting is a Multi venturi scrubber system (MVSS), which is used to realize filtered venting without any power supply in European reactors. The MVSS is composed of a “venturi Scrubbers” part, in which there are hundreds of the venturi scrubbers, and a “bubble column” part. In the MVSS, all of the venturi scrubbers is branched off from a vent line which connect between the containment and the MVSS. In an operation mode of the MVSS, the radioactive materials are eliminated through the gas-liquid interface from the pollutant gas to the liquid phase of a dispersed flow in the venturi scrubber and a bubbly flow in the bubble column part. The dispersed flow is formed from the liquid, which is suctioned from around the venturi scrubber through the hole for suction (called self-priming). In previous studies, an evaluation method for the scrubbing performance of the venturi scrubber was developed. However, actual hydraulic behavior in it is too complicated, the previous evaluation was not validated the hydraulic behavior and studied the effect of differences between the simulated hydraulic behavior and an actual one on the performance of the venturi scrubber. To develop a validated evaluation method for the scrubbing performance, it is important to develop detailed evaluation method for the hydraulic behavior in the venturi scrubber. To simulate the complicated hydraulic behavior, we consider to use analysis code TPFIT. Then, the

  2. Neutron H*(10) inside a proton therapy facility: comparison between Monte Carlo simulations and WENDI-2 measurements

    International Nuclear Information System (INIS)

    De Smet, V.; Stichelbaut, F.; Mathot, G.; Vanaudenhove, T.; De Lentdecker, G.; Dubus, A.; Pauly, N.; Gerardy, I.

    2014-01-01

    Inside an IBA proton therapy centre, secondary neutrons are produced due to nuclear interactions of the proton beam with matter mainly inside the cyclotron, the beam line, the treatment nozzle and the patient. Accurate measurements of the neutron ambient dose equivalent H*(10) in such a facility require the use of a detector that has a good sensitivity for neutrons ranging from thermal energies up to 230 MeV, such as for instance the WENDI-2 detector. WENDI-2 measurements have been performed at the Westdeutsches Protonentherapiezentrum Essen, at several positions around the cyclotron room and around a gantry treatment room operated in two different beam delivery modes: Pencil Beam Scanning and Double Scattering. These measurements are compared with Monte Carlo simulation results for the neutron H*(10) obtained with MCNPX 2.5.0 and GEANT4 9.6. In proton therapy, proton beams with energies up to typically 230 MeV are used to treat cancerous tumours very efficiently while sparing surrounding healthy tissues as much as possible. Due to nuclear interactions of the proton beams with matter, mainly inside the cyclotron, the beam line, the treatment nozzle and the patient, secondary neutrons with energies up to 230 MeV are unfortunately produced, as well as photons up to ∼10 MeV. Behind the thick concrete shielding walls which are necessary to attenuate the stray radiation fields, the total ambient dose equivalent H*(10) is very large due to the neutron component. In shielding studies for proton therapy facilities, the neutron H*(10) component is often evaluated using the Monte Carlo codes MCNPX(5), FLUKA(6) or PHITS(7). Recent benchmark simulations performed with GEANT4 have shown that this code would also be a suitable tool for the shielding studies of proton therapy centres. The experimental validation of such shielding studies requires the use of a detector with a good sensitivity for neutrons ranging from thermal energies up to 230 MeV, such as for example the

  3. Simulation of thermal fluid dynamics in parabolic trough receiver tubes with direct steam generation using the computer code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, Alexander; Merk, Bruno [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Hirsch, Tobias; Pitz-Paal, Robert [DLR Deutsches Zentrum fuer Luft- und Raumfahrt e.V., Stuttgart (Germany). Inst. fuer Solarforschung

    2014-06-15

    In the present feasibility study the system code ATHLET, which originates from nuclear engineering, is applied to a parabolic trough test facility. A model of the DISS (DIrect Solar Steam) test facility at Plataforma Solar de Almeria in Spain is assembled and the results of the simulations are compared to measured data and the simulation results of the Modelica library 'DissDyn'. A profound comparison between ATHLET Mod 3.0 Cycle A and the 'DissDyn' library reveals the capabilities of these codes. The calculated mass and energy balance in the ATHLET simulations are in good agreement with the results of the measurements and confirm the applicability for thermodynamic simulations of DSG processes in principle. Supplementary, the capabilities of the 6-equation model with transient momentum balances in ATHLET are used to study the slip between liquid and gas phases and to investigate pressure wave oscillations after a sudden valve closure. (orig.)

  4. Development of three-dimensional neoclassical transport simulation code with high performance Fortran on a vector-parallel computer

    International Nuclear Information System (INIS)

    Satake, Shinsuke; Okamoto, Masao; Nakajima, Noriyoshi; Takamaru, Hisanori

    2005-11-01

    A neoclassical transport simulation code (FORTEC-3D) applicable to three-dimensional configurations has been developed using High Performance Fortran (HPF). Adoption of computing techniques for parallelization and a hybrid simulation model to the δf Monte-Carlo method transport simulation, including non-local transport effects in three-dimensional configurations, makes it possible to simulate the dynamism of global, non-local transport phenomena with a self-consistent radial electric field within a reasonable computation time. In this paper, development of the transport code using HPF is reported. Optimization techniques in order to achieve both high vectorization and parallelization efficiency, adoption of a parallel random number generator, and also benchmark results, are shown. (author)

  5. Development and application of Siton, a new fuel cycle simulation code

    International Nuclear Information System (INIS)

    Brolly, Aron; Szieberth, Mate; Halasz, Mate; Nagy, Lajos; Feher, Sandor

    2015-01-01

    As the result of the co-operation between the Centre for Energy Research (EK) and the Institute of Nuclear Techniques (NTI) a new fuel cycle simulation code called SITON was developed. Physical model of the code takes into account six facilities of the nuclear fuel cycle namely material stocks, spent fuel interim storages, plants for uranium enrichment, fuel fabrication, spent fuel reprocessing and reactors. Facilities can be linked in a flexible manner and their number is not limited. Lag time of the facilities and cooling time of the spent fuel, which are the two main parameters to introduce lag time into the fuel cycle, are taken into account. Material transfer between the facilities is modelled in a discrete manner tracking 52 nuclides and their short-lived decay daughters. Composition of the discharged fuel is determined by means of burn-up tables except for the 2400 MWth design of gas cooled fast reactor (GFR2400) which has a separate burn-up module developed at the NTI. To demonstrate the capabilities of SITON introduction of a GFR2400 into the Hungarian reactor park using the legacy spent fuel of the four presently operating VVER-440 units was simulated. 2040 was assumed as the commissioning date of the GFR2400 and recycling of its fuel was started as soon as possible. It was found that the plutonium content of the legacy spent fuel is sufficient to the start-up of only one GFR2400. There is an intermediate period between the commissioning of the reactor and the recycling of its first discharged fuel. Plutonium need of this period can be covered by the legacy spent fuel if the cooling time of the spent GFR2400 fuel is 2 years. If the cooling time is 5 years there will be a lack of plutonium in this period. To counterbalance this lack an EPR was started before the GFR2400 and its spent fuel was accumulated and reprocessed. Cooling time of the spent EPR fuel was also varied. Finally, an EPR only scenario is presented using two EPRs as a reference case

  6. Proton absorbed dose distribution in human eye simulated by SRNA-2KG code

    International Nuclear Information System (INIS)

    Ilic, R. D.; Pavlovic, R.

    2004-01-01

    The model of Monte Carlo SRNA code is described together with some numerical experiments to show feasibility of this code to be used in proton therapy, especially for tree dimensional proton absorption dose calculation in human eye. (author) [sr

  7. Study of a scintillation neutron detector of {sup 1O}B+ZnS(Ag) as alternative to the {sup 3}He detectors: model MCNPX and validation; Estudio de un detector de neutrones de centelleo de {sup 10}B+ZnS(Ag) como alternativa a los detectores de {sup 3}He: modelo MCNPX y validacion

    Energy Technology Data Exchange (ETDEWEB)

    Guzman G, K. A.; Gallego D, E.; Lorente F, A.; Ibanez F, S. [Universidad Politecnica de Madrid, Departamento de Ingenieria Energetica, E.T.S. Ing. Industriales, Jose Gutierrez Abascal 2, 28006 Madrid (Spain); Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Mendez V, R. [CIEMAT, Av. Complutense 40, 28040 Madrid (Spain); Gonzalez, J. A., E-mail: karen.guzman.garcia@alumnos.upm.es [Universidad Politecnica de Madrid, Laboratorio de Ingenieria Nuclear, ETSI Caminos, Canales y Puertos, Ciudad Universitaria, C. Profesor Aranguren 3, 28040 Madrid (Spain)

    2015-10-15

    Using Monte Carlo methods with the code MCNPX, was estimated the response of a scintillation neutron detector of Zn S(Ag) with a mixture of {sup 10}B high enrichment. The detector consists of four plates of Poly (methyl methacrylate) (PMMA) and five layers of ∼0, 017 cm {sup 10}B+ZnS(Ag) in contact with PMMA. The naked detector response was calculated and with different thicknesses of high density polyethylene moderator, for 29 monoenergetic sources and for sources of {sup 241}AmBe and {sup 252}Cf of neutrons. In these calculations the reactions {sup 10}B(n,α){sup 7}Li and neutron fluence in the sensitive area of detector {sup 10}B+ZnS(Ag) were estimated. Measurements were performed in the Laboratory of Neutron Measurement to quantify detections in counts per second to a neutron source of {sup 252}Cf to 200 cm on the bench, modeling with MCNPX, these measures were compared to validate the model and the Zn S(Ag) efficiency of α detection was estimated. Calculations in the LPN-CIEMAT were realized. Starting from the validation new models were carried out with geometries that improve the detector response, trying reaching the detection of 2, 5 cps-ng of {sup 252}Cf comparable requirement for responding to the installed equipment of {sup 3}He in the radiation portal monitor. This type of detector can be considered an alternative to detectors of {sup 3}He for detecting special nuclear material. (Author)

  8. Generation of initial geometries for the simulation of the physical system in the DualPHYsics code

    International Nuclear Information System (INIS)

    Segura Q, E.

    2013-01-01

    In the diverse research areas of the Instituto Nacional de Investigaciones Nucleares (ININ) are different activities related to science and technology, one of great interest is the study and treatment of the collection and storage of radioactive waste. Therefore at ININ the draft on the simulation of the pollutants diffusion in the soil through a porous medium (third stage) has this problem inherent aspects, hence a need for such a situation is to generate the initial geometry of the physical system For the realization of the simulation method is implemented smoothed particle hydrodynamics (SPH). This method runs in DualSPHysics code, which has great versatility and ability to simulate phenomena of any physical system where hydrodynamic aspects combine. In order to simulate a physical system DualSPHysics code, you need to preset the initial geometry of the system of interest, then this is included in the input file of the code. The simulation sets the initial geometry through regular geometric bodies positioned at different points in space. This was done through a programming language (Fortran, C + +, Java, etc..). This methodology will provide the basis to simulate more complex geometries future positions and form. (Author)

  9. Development of integrated SOL/Divertor code and simulation study of the JT-60U/JT-60SA tokamaks

    International Nuclear Information System (INIS)

    Kawashima, H.; Shimizu, K.; Takizuka, T.

    2007-01-01

    To predict the particle and heat controllability in the divertor of tokamak reactors such as ITER and to optimize the divertor design, comprehensive simulations by integrated modelling with taking in various physical processes are indispensable. For the design study of ITER divertor, the modelling codes such as B2, UEDGE and EDGE2D have been developed, and their results have contributed to the evolution of the divertor concept. In Japan Atomic Energy Agency (JAEA), SOL/divertor codes have also been developed for the interpretation and the prediction on behaviours of plasmas, neutrals and impurities in the SOL/divertor regions. The code development is originally carried out since physics models can be verified quickly and flexibly under the circumstance of close collaboration with JT-60 team. Figure 1 shows our code system, which consists of the 2 dimensional fluid code SOLDOR, the neutral Monte Carlo (MC) code NEUT2D, and the impurity MC code IMPMC. The particle simulation code PARASOL has also been developed in order to establish the physics modelling used in fluid simulations. Integration of SOLDOR, NEUT2D and IMPMC, called the '' SONIC '' code, is being carried out to simulate self-consistently the SOL/divertor plasmas in present tokamaks and in future devices. Combination of the SOLDOR and NEUT2D was completed, which has the features such as 1) high-resolution oscillation-free scheme in solving fluid equations, 2) neutral transport calculation under the fine meshes, 3) success in reduction of MC noise, 4) optimization on the massive parallel computer, etc. The simulation reproduces the X-point MARFE in the JT-60U experiment. It is found that the chemically sputtered carbon at the dome causes the radiation peaking near the X-point. The performance of divertor pumping in JT-60U is evaluated from the particle balances. We also present the divertor designing of JT-60SA, which is the modification program of JT-60U to establish high beta steady-state operation. To

  10. Improvement of implicit finite element code performance in deep drawing simulations by dynamics contributions

    NARCIS (Netherlands)

    Meinders, Vincent T.; van den Boogaard, Antonius H.; Huetink, Han

    2003-01-01

    To intensify the use of implicit finite element codes for solving large scale problems, the computation time of these codes has to be decreased drastically. A method is developed which decreases the computational time of implicit codes by factors. The method is based on introducing inertia effects

  11. Testing and Modeling of a 3-MW Wind Turbine Using Fully Coupled Simulation Codes (Poster)

    Energy Technology Data Exchange (ETDEWEB)

    LaCava, W.; Guo, Y.; Van Dam, J.; Bergua, R.; Casanovas, C.; Cugat, C.

    2012-06-01

    This poster describes the NREL/Alstom Wind testing and model verification of the Alstom 3-MW wind turbine located at NREL's National Wind Technology Center. NREL,in collaboration with ALSTOM Wind, is studying a 3-MW wind turbine installed at the National Wind Technology Center(NWTC). The project analyzes the turbine design using a state-of-the-art simulation code validated with detailed test data. This poster describes the testing and the model validation effort, and provides conclusions about the performance of the unique drive train configuration used in this wind turbine. The 3-MW machine has been operating at the NWTC since March 2011, and drive train measurements will be collected through the spring of 2012. The NWTC testing site has particularly turbulent wind patterns that allow for the measurement of large transient loads and the resulting turbine response. This poster describes the 3-MW turbine test project, the instrumentation installed, and the load cases captured. The design of a reliable wind turbine drive train increasingly relies on the use of advanced simulation to predict structural responses in a varying wind field. This poster presents a fully coupled, aero-elastic and dynamic model of the wind turbine. It also shows the methodology used to validate the model, including the use of measured tower modes, model-to-model comparisons of the power curve, and mainshaft bending predictions for various load cases. The drivetrain is designed to only transmit torque to the gearbox, eliminating non-torque moments that are known to cause gear misalignment. Preliminary results show that the drivetrain is able to divert bending loads in extreme loading cases, and that a significantly smaller bending moment is induced on the mainshaft compared to a three-point mounting design.

  12. Implementation of the Actuator Cylinder Flow Model in the HAWC2 code for Aeroelastic Simulations on Vertical Axis Wind Turbines

    DEFF Research Database (Denmark)

    Aagaard Madsen, Helge; Larsen, Torben J.; Schmidt Paulsen, Uwe

    2013-01-01

    The paper presents the implementation of the Actuator Cylinder (AC) flow model in the HAWC2 aeroelastic code originally developed for simulation of Horizontal Axis Wind Turbine (HAWT) aeroelasticity. This is done within the DeepWind project where the main objective is to explore the competitivene...

  13. Improvement of spallation reaction simulation codes NMTC/JAERI and NUCLEUS

    International Nuclear Information System (INIS)

    Nishida, T.; Takada, H.; Kanno, I.; Nakahara, Y.

    1990-01-01

    To make evaluations of theoretical models for nuclear spallation reaction, simulation codes are modified and a new mass formula is used to improve the accuracy of Monte Carlo calculations. The following conclusions are made from analyses of calculated distributions of nuclear spallation products. A difference is found between the Cameron's old and the Uno and Yamada's new mass formula, which is due to the difference in the method used to fit their shell energy terms to measured data for selected nuclei and in data themselves. For nuclides with an atomic number larger than 70, mass excesses calculated by the Camerons's mass formula are greater than those by the Uno and Yamada's one, whereas the reverse tendency is seen for ones with atomic numbers smaller than 70. Analysis shows that the distributions of produced nuclei have patterns that appear natural from a physical point of view when artificial restrictions are removed in counting the nuclide production events. The new mass formula can reproduce fairly well the experimental product yield distributions, especially in the neutron excess side. It is also found that the old mass formula gives lower estimations for the number of produced nuclei than the new one, especially in the nuclide region far from the beta stable line. (N.K.)

  14. Development and application of a multi-fluid simulation code for modeling interpenetrating plasmas

    Science.gov (United States)

    Khodak, M.; Berger, R. L.; Chapman, T.; Hittinger, J. A. F.

    2015-11-01

    A multi-fluid model, with independent velocities for all species, is developed and implemented for the numerical simulation of the interpenetration of colliding plasmas. The Euler equations for fluid flow, coupled through electron-ion and ion-ion collisional drag terms, thermal equilibration terms, and the electric field, are solved for each ion species with the electrons treated under a quasineutrality assumption. Fourth-order spatial convergence in smooth regions is achieved using flux-conservative iterative time integration and a Weighted Essentially Non-Oscillatory (WENO) finite volume scheme employing an approximate Riemann solver. Analytic solutions of well-known shock tube tests and spectral solutions of the linearized coupled system are used to test the implementation, and the model is further numerically compared to interpenetration experiments such as those of J.S. Ross et al. [Phys. Rev. Lett. 110 145005 (2013)]. This work has applications to laser-plasma interactions, specifically to hohlraum physics, as well as to modeling laboratory experiments of collisionless shocks important in astrophysical plasmas. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract No. DE-AC52-07NA27344 and funded by the Laboratory Research and Development Program at LLNL under project code 15-ERD-038.

  15. PERIGEE computer codes for reactor simulation in 3 dimensions, using 1 or 2 neutron velocity groups

    International Nuclear Information System (INIS)

    Olson, A.P.

    1964-02-01

    PERIGEE is a code written in SNAP for the G-20 computer. It solves the one- or two-group neutron diffusion equations by finite-difference methods on a three-dimensional, uniform mesh having a common spacing in the two directions normal to the fuel channels. The positions of mesh points along a fuel channel, relative to points in adjacent channels, may correspond to either NPD or CANDU fuel bundle positions. The extrapolated flux boundary may be specified in sufficient detail to represent a tapered or stepped circumferential reflector, a variable axial length and, for a reactor with axis horizontal, a variable moderator level and a variable plane bottom surface equivalent to the CANDU dump structure. The neutron flux may be normalized to give a specified power output from the hottest fuel bundle or hottest channel, or to give a total thermal power limited by the turbine and generator. Reactor operation may be simulated in finite time steps, taking into account any fuel shifts, any changes in moderator level and the change in nuclear properties of the fuel with increasing irradiation. The appropriate properties are obtained by interpolation from tables supplied for as many as 8 types of fuel bundle. The mean fuel exit burnup can be calculated at equilibrium for a reactor in which the exit burnups for two zones may be adjusted to give radial power flattening and the fuelling schedules may be designed to give axial power flattening in one or both zones. (author)

  16. Penelope - a code system for Monte Carlo simulation of electron and photon transport

    International Nuclear Information System (INIS)

    2003-01-01

    Radiation is used in many applications of modern technology. Its proper handling requires competent knowledge of the basic physical laws governing its interaction with matter. To ensure its safe use, appropriate tools for predicting radiation fields and doses, as well as pertinent regulations, are required. One area of radiation physics that has received much attention concerns electron-photon transport in matter. PENELOPE is a modern, general-purpose Monte Carlo tool for simulating the transport of electrons and photons, which is applicable for arbitrary materials and in a wide energy range. PENELOPE provides quantitative guidance for many practical situations and techniques, including electron and X-ray spectroscopies, electron microscopy and microanalysis, biophysics, dosimetry, medical diagnostics and radiotherapy, as well as radiation damage and shielding. These proceedings contain the extensively revised teaching notes of the second workshop/training course on PENELOPE held in 2003, along with a detailed description of the improved physic models, numerical algorithms and structure of the code system. (author)

  17. Recent Developments in the VISRAD 3-D Target Design and Radiation Simulation Code

    Science.gov (United States)

    Macfarlane, Joseph; Golovkin, Igor; Sebald, James

    2017-10-01

    The 3-D view factor code VISRAD is widely used in designing HEDP experiments at major laser and pulsed-power facilities, including NIF, OMEGA, OMEGA-EP, ORION, Z, and LMJ. It simulates target designs by generating a 3-D grid of surface elements, utilizing a variety of 3-D primitives and surface removal algorithms, and can be used to compute the radiation flux throughout the surface element grid by computing element-to-element view factors and solving power balance equations. Target set-up and beam pointing are facilitated by allowing users to specify positions and angular orientations using a variety of coordinates systems (e.g., that of any laser beam, target component, or diagnostic port). Analytic modeling for laser beam spatial profiles for OMEGA DPPs and NIF CPPs is used to compute laser intensity profiles throughout the grid of surface elements. VISRAD includes a variety of user-friendly graphics for setting up targets and displaying results, can readily display views from any point in space, and can be used to generate image sequences for animations. We will discuss recent improvements to conveniently assess beam capture on target and beam clearance of diagnostic components, as well as plans for future developments.

  18. The GENGA code: gravitational encounters in N-body simulations with GPU acceleration

    International Nuclear Information System (INIS)

    Grimm, Simon L.; Stadel, Joachim G.

    2014-01-01

    We describe an open source GPU implementation of a hybrid symplectic N-body integrator, GENGA (Gravitational ENcounters with Gpu Acceleration), designed to integrate planet and planetesimal dynamics in the late stage of planet formation and stability analyses of planetary systems. GENGA uses a hybrid symplectic integrator to handle close encounters with very good energy conservation, which is essential in long-term planetary system integration. We extended the second-order hybrid integration scheme to higher orders. The GENGA code supports three simulation modes: integration of up to 2048 massive bodies, integration with up to a million test particles, or parallel integration of a large number of individual planetary systems. We compare the results of GENGA to Mercury and pkdgrav2 in terms of energy conservation and performance and find that the energy conservation of GENGA is comparable to Mercury and around two orders of magnitude better than pkdgrav2. GENGA runs up to 30 times faster than Mercury and up to 8 times faster than pkdgrav2. GENGA is written in CUDA C and runs on all NVIDIA GPUs with a computing capability of at least 2.0.

  19. The GENGA code: gravitational encounters in N-body simulations with GPU acceleration

    Energy Technology Data Exchange (ETDEWEB)

    Grimm, Simon L.; Stadel, Joachim G., E-mail: sigrimm@physik.uzh.ch [Institute for Computational Science, University of Zürich, Winterthurerstrasse 190, CH-8057 Zürich (Switzerland)

    2014-11-20

    We describe an open source GPU implementation of a hybrid symplectic N-body integrator, GENGA (Gravitational ENcounters with Gpu Acceleration), designed to integrate planet and planetesimal dynamics in the late stage of planet formation and stability analyses of planetary systems. GENGA uses a hybrid symplectic integrator to handle close encounters with very good energy conservation, which is essential in long-term planetary system integration. We extended the second-order hybrid integration scheme to higher orders. The GENGA code supports three simulation modes: integration of up to 2048 massive bodies, integration with up to a million test particles, or parallel integration of a large number of individual planetary systems. We compare the results of GENGA to Mercury and pkdgrav2 in terms of energy conservation and performance and find that the energy conservation of GENGA is comparable to Mercury and around two orders of magnitude better than pkdgrav2. GENGA runs up to 30 times faster than Mercury and up to 8 times faster than pkdgrav2. GENGA is written in CUDA C and runs on all NVIDIA GPUs with a computing capability of at least 2.0.

  20. Further development of the fast beam dynamics simulation tool V-code

    Energy Technology Data Exchange (ETDEWEB)

    Franke, Sylvain; Ackermann, Wolfgang; Weiland, Thomas [Institut fuer Theorie Elektromagnetischer Felder, TU Darmstadt (Germany)

    2010-07-01

    The Vlasov equation describes the evolution of a particle density under the effects of electromagnetic fields. It is derived from the fact that the volume occupied by a given number of particles in the six-dimensional phase space remains constant when only long-range interaction as for example Coulomb forces are relevant and ot