WorldWideScience

Sample records for carlo particle transport

  1. Scalable Domain Decomposed Monte Carlo Particle Transport

    Energy Technology Data Exchange (ETDEWEB)

    O' Brien, Matthew Joseph [Univ. of California, Davis, CA (United States)

    2013-12-05

    In this dissertation, we present the parallel algorithms necessary to run domain decomposed Monte Carlo particle transport on large numbers of processors (millions of processors). Previous algorithms were not scalable, and the parallel overhead became more computationally costly than the numerical simulation.

  2. Particle-transport simulation with the Monte Carlo method

    International Nuclear Information System (INIS)

    Carter, L.L.; Cashwell, E.D.

    1975-01-01

    Attention is focused on the application of the Monte Carlo method to particle transport problems, with emphasis on neutron and photon transport. Topics covered include sampling methods, mathematical prescriptions for simulating particle transport, mechanics of simulating particle transport, neutron transport, and photon transport. A literature survey of 204 references is included. (GMT)

  3. Vectorization of Monte Carlo particle transport

    International Nuclear Information System (INIS)

    Burns, P.J.; Christon, M.; Schweitzer, R.; Lubeck, O.M.; Wasserman, H.J.; Simmons, M.L.; Pryor, D.V.

    1989-01-01

    This paper reports that fully vectorized versions of the Los Alamos National Laboratory benchmark code Gamteb, a Monte Carlo photon transport algorithm, were developed for the Cyber 205/ETA-10 and Cray X-MP/Y-MP architectures. Single-processor performance measurements of the vector and scalar implementations were modeled in a modified Amdahl's Law that accounts for additional data motion in the vector code. The performance and implementation strategy of the vector codes are related to architectural features of each machine. Speedups between fifteen and eighteen for Cyber 205/ETA-10 architectures, and about nine for CRAY X-MP/Y-MP architectures are observed. The best single processor execution time for the problem was 0.33 seconds on the ETA-10G, and 0.42 seconds on the CRAY Y-MP

  4. Weighted-delta-tracking for Monte Carlo particle transport

    International Nuclear Information System (INIS)

    Morgan, L.W.G.; Kotlyar, D.

    2015-01-01

    Highlights: • This paper presents an alteration to the Monte Carlo Woodcock tracking technique. • The alteration improves computational efficiency within regions of high absorbers. • The rejection technique is replaced by a statistical weighting mechanism. • The modified Woodcock method is shown to be faster than standard Woodcock tracking. • The modified Woodcock method achieves a lower variance, given a specified accuracy. - Abstract: Monte Carlo particle transport (MCPT) codes are incredibly powerful and versatile tools to simulate particle behavior in a multitude of scenarios, such as core/criticality studies, radiation protection, shielding, medicine and fusion research to name just a small subset applications. However, MCPT codes can be very computationally expensive to run when the model geometry contains large attenuation depths and/or contains many components. This paper proposes a simple modification to the Woodcock tracking method used by some Monte Carlo particle transport codes. The Woodcock method utilizes the rejection method for sampling virtual collisions as a method to remove collision distance sampling at material boundaries. However, it suffers from poor computational efficiency when the sample acceptance rate is low. The proposed method removes rejection sampling from the Woodcock method in favor of a statistical weighting scheme, which improves the computational efficiency of a Monte Carlo particle tracking code. It is shown that the modified Woodcock method is less computationally expensive than standard ray-tracing and rejection-based Woodcock tracking methods and achieves a lower variance, given a specified accuracy

  5. Adaptive multilevel splitting for Monte Carlo particle transport

    Directory of Open Access Journals (Sweden)

    Louvin Henri

    2017-01-01

    Full Text Available In the Monte Carlo simulation of particle transport, and especially for shielding applications, variance reduction techniques are widely used to help simulate realisations of rare events and reduce the relative errors on the estimated scores for a given computation time. Adaptive Multilevel Splitting (AMS is one of these variance reduction techniques that has recently appeared in the literature. In the present paper, we propose an alternative version of the AMS algorithm, adapted for the first time to the field of particle transport. Within this context, it can be used to build an unbiased estimator of any quantity associated with particle tracks, such as flux, reaction rates or even non-Boltzmann tallies like pulse-height tallies and other spectra. Furthermore, the efficiency of the AMS algorithm is shown not to be very sensitive to variations of its input parameters, which makes it capable of significant variance reduction without requiring extended user effort.

  6. Monte Carlo 2000 Conference : Advanced Monte Carlo for Radiation Physics, Particle Transport Simulation and Applications

    CERN Document Server

    Baräo, Fernando; Nakagawa, Masayuki; Távora, Luis; Vaz, Pedro

    2001-01-01

    This book focusses on the state of the art of Monte Carlo methods in radiation physics and particle transport simulation and applications, the latter involving in particular, the use and development of electron--gamma, neutron--gamma and hadronic codes. Besides the basic theory and the methods employed, special attention is paid to algorithm development for modeling, and the analysis of experiments and measurements in a variety of fields ranging from particle to medical physics.

  7. Modeling Dynamic Objects in Monte Carlo Particle Transport Calculations

    International Nuclear Information System (INIS)

    Yegin, G.

    2008-01-01

    In this study, the Multi-Geometry geometry modeling technique was improved in order to handle moving objects in a Monte Carlo particle transport calculation. In the Multi-Geometry technique, the geometry is a superposition of objects not surfaces. By using this feature, we developed a new algorithm which allows a user to make enable or disable geometry elements during particle transport. A disabled object can be ignored at a certain stage of a calculation and switching among identical copies of the same object located adjacent poins during a particle simulation corresponds to the movement of that object in space. We called this powerfull feature as Dynamic Multi-Geometry technique (DMG) which is used for the first time in Brachy Dose Monte Carlo code to simulate HDR brachytherapy treatment systems. Our results showed that having disabled objects in a geometry does not effect calculated dose values. This technique is also suitable to be used in other areas such as IMRT treatment planning systems

  8. 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.

  9. New features of the mercury Monte Carlo particle transport code

    International Nuclear Information System (INIS)

    Procassini, Richard; Brantley, Patrick; Dawson, Shawn

    2010-01-01

    Several new capabilities have been added to the Mercury Monte Carlo transport code over the past four years. The most important algorithmic enhancement is a general, extensible infrastructure to support source, tally and variance reduction actions. For each action, the user defines a phase space, as well as any number of responses that are applied to a specified event. Tallies are accumulated into a correlated, multi-dimensional. Cartesian-product result phase space. Our approach employs a common user interface to specify the data sets and distributions that define the phase, response and result for each action. Modifications to the particle trackers include the use of facet halos (instead of extrapolative fuzz) for robust tracking, and material interface reconstruction for use in shape overlaid meshes. Support for expected-value criticality eigenvalue calculations has also been implemented. Computer science enhancements include an in-line Python interface for user customization of problem setup and output. (author)

  10. The OpenMC Monte Carlo particle transport code

    International Nuclear Information System (INIS)

    Romano, Paul K.; Forget, Benoit

    2013-01-01

    Highlights: ► An open source Monte Carlo particle transport code, OpenMC, has been developed. ► Solid geometry and continuous-energy physics allow high-fidelity simulations. ► Development has focused on high performance and modern I/O techniques. ► OpenMC is capable of scaling up to hundreds of thousands of processors. ► Results on a variety of benchmark problems agree with MCNP5. -- Abstract: A new Monte Carlo code called OpenMC is currently under development at the Massachusetts Institute of Technology as a tool for simulation on high-performance computing platforms. Given that many legacy codes do not scale well on existing and future parallel computer architectures, OpenMC has been developed from scratch with a focus on high performance scalable algorithms as well as modern software design practices. The present work describes the methods used in the OpenMC code and demonstrates the performance and accuracy of the code on a variety of problems.

  11. Modified Monte Carlo procedure for particle transport problems

    International Nuclear Information System (INIS)

    Matthes, W.

    1978-01-01

    The simulation of photon transport in the atmosphere with the Monte Carlo method forms part of the EURASEP-programme. The specifications for the problems posed for a solution were such, that the direct application of the analogue Monte Carlo method was not feasible. For this reason the standard Monte Carlo procedure was modified in the sense that additional properly weighted branchings at each collision and transport process in a photon history were introduced. This modified Monte Carlo procedure leads to a clear and logical separation of the essential parts of a problem and offers a large flexibility for variance reducing techniques. More complex problems, as foreseen in the EURASEP-programme (e.g. clouds in the atmosphere, rough ocean-surface and chlorophyl-distribution in the ocean) can be handled by recoding some subroutines. This collision- and transport-splitting procedure can of course be performed differently in different space- and energy regions. It is applied here only for a homogeneous problem

  12. Antiproton annihilation physics annihilation physics in the Monte Carlo particle transport code particle transport code SHIELD-HIT12A

    DEFF Research Database (Denmark)

    Taasti, Vicki Trier; Knudsen, Helge; Holzscheiter, Michael

    2015-01-01

    The Monte Carlo particle transport code SHIELD-HIT12A is designed to simulate therapeutic beams for cancer radiotherapy with fast ions. SHIELD-HIT12A allows creation of antiproton beam kernels for the treatment planning system TRiP98, but first it must be benchmarked against experimental data. An...

  13. Particle Communication and Domain Neighbor Coupling: Scalable Domain Decomposed Algorithms for Monte Carlo Particle Transport

    Energy Technology Data Exchange (ETDEWEB)

    O' Brien, M. J.; Brantley, P. S.

    2015-01-20

    In order to run Monte Carlo particle transport calculations on new supercomputers with hundreds of thousands or millions of processors, care must be taken to implement scalable algorithms. This means that the algorithms must continue to perform well as the processor count increases. In this paper, we examine the scalability of:(1) globally resolving the particle locations on the correct processor, (2) deciding that particle streaming communication has finished, and (3) efficiently coupling neighbor domains together with different replication levels. We have run domain decomposed Monte Carlo particle transport on up to 221 = 2,097,152 MPI processes on the IBM BG/Q Sequoia supercomputer and observed scalable results that agree with our theoretical predictions. These calculations were carefully constructed to have the same amount of work on every processor, i.e. the calculation is already load balanced. We also examine load imbalanced calculations where each domain’s replication level is proportional to its particle workload. In this case we show how to efficiently couple together adjacent domains to maintain within workgroup load balance and minimize memory usage.

  14. Development of general-purpose particle and heavy ion transport monte carlo code

    International Nuclear Information System (INIS)

    Iwase, Hiroshi; Nakamura, Takashi; Niita, Koji

    2002-01-01

    The high-energy particle transport code NMTC/JAM, which has been developed at JAERI, was improved for the high-energy heavy ion transport calculation by incorporating the JQMD code, the SPAR code and the Shen formula. The new NMTC/JAM named PHITS (Particle and Heavy-Ion Transport code System) is the first general-purpose heavy ion transport Monte Carlo code over the incident energies from several MeV/nucleon to several GeV/nucleon. (author)

  15. PyMercury: Interactive Python for the Mercury Monte Carlo Particle Transport Code

    International Nuclear Information System (INIS)

    Iandola, F.N.; O'Brien, M.J.; Procassini, R.J.

    2010-01-01

    Monte Carlo particle transport applications are often written in low-level languages (C/C++) for optimal performance on clusters and supercomputers. However, this development approach often sacrifices straightforward usability and testing in the interest of fast application performance. To improve usability, some high-performance computing applications employ mixed-language programming with high-level and low-level languages. In this study, we consider the benefits of incorporating an interactive Python interface into a Monte Carlo application. With PyMercury, a new Python extension to the Mercury general-purpose Monte Carlo particle transport code, we improve application usability without diminishing performance. In two case studies, we illustrate how PyMercury improves usability and simplifies testing and validation in a Monte Carlo application. In short, PyMercury demonstrates the value of interactive Python for Monte Carlo particle transport applications. In the future, we expect interactive Python to play an increasingly significant role in Monte Carlo usage and testing.

  16. Variance analysis of the Monte-Carlo perturbation source method in inhomogeneous linear particle transport problems

    International Nuclear Information System (INIS)

    Noack, K.

    1982-01-01

    The perturbation source method may be a powerful Monte-Carlo means to calculate small effects in a particle field. In a preceding paper we have formulated this methos in inhomogeneous linear particle transport problems describing the particle fields by solutions of Fredholm integral equations and have derived formulae for the second moment of the difference event point estimator. In the present paper we analyse the general structure of its variance, point out the variance peculiarities, discuss the dependence on certain transport games and on generation procedures of the auxiliary particles and draw conclusions to improve this method

  17. Monte Carlo particle simulation and finite-element techniques for tandem mirror transport

    International Nuclear Information System (INIS)

    Rognlien, T.D.; Cohen, B.I.; Matsuda, Y.; Stewart, J.J. Jr.

    1987-01-01

    A description is given of numerical methods used in the study of axial transport in tandem mirrors owing to Coulomb collisions and rf diffusion. The methods are Monte Carlo particle simulations and direct solution to the Fokker-Planck equations by finite-element expansion. (author)

  18. Monte Carlo particle simulation and finite-element techniques for tandem mirror transport

    International Nuclear Information System (INIS)

    Rognlien, T.D.; Cohen, B.I.; Matsuda, Y.; Stewart, J.J. Jr.

    1985-12-01

    A description is given of numerical methods used in the study of axial transport in tandem mirrors owing to Coulomb collisions and rf diffusion. The methods are Monte Carlo particle simulations and direct solution to the Fokker-Planck equations by finite-element expansion. 11 refs

  19. Numerical computation of discrete differential scattering cross sections for Monte Carlo charged particle transport

    International Nuclear Information System (INIS)

    Walsh, Jonathan A.; Palmer, Todd S.; Urbatsch, Todd J.

    2015-01-01

    Highlights: • Generation of discrete differential scattering angle and energy loss cross sections. • Gauss–Radau quadrature utilizing numerically computed cross section moments. • Development of a charged particle transport capability in the Milagro IMC code. • Integration of cross section generation and charged particle transport capabilities. - Abstract: We investigate a method for numerically generating discrete scattering cross sections for use in charged particle transport simulations. We describe the cross section generation procedure and compare it to existing methods used to obtain discrete cross sections. The numerical approach presented here is generalized to allow greater flexibility in choosing a cross section model from which to derive discrete values. Cross section data computed with this method compare favorably with discrete data generated with an existing method. Additionally, a charged particle transport capability is demonstrated in the time-dependent Implicit Monte Carlo radiative transfer code, Milagro. We verify the implementation of charged particle transport in Milagro with analytic test problems and we compare calculated electron depth–dose profiles with another particle transport code that has a validated electron transport capability. Finally, we investigate the integration of the new discrete cross section generation method with the charged particle transport capability in Milagro.

  20. Chain segmentation for the Monte Carlo solution of particle transport problems

    International Nuclear Information System (INIS)

    Ragheb, M.M.H.

    1984-01-01

    A Monte Carlo approach is proposed where the random walk chains generated in particle transport simulations are segmented. Forward and adjoint-mode estimators are then used in conjunction with the firstevent source density on the segmented chains to obtain multiple estimates of the individual terms of the Neumann series solution at each collision point. The solution is then constructed by summation of the series. The approach is compared to the exact analytical and to the Monte Carlo nonabsorption weighting method results for two representative slowing down and deep penetration problems. Application of the proposed approach leads to unbiased estimates for limited numbers of particle simulations and is useful in suppressing an effective bias problem observed in some cases of deep penetration particle transport problems

  1. Data decomposition of Monte Carlo particle transport simulations via tally servers

    International Nuclear Information System (INIS)

    Romano, Paul K.; Siegel, Andrew R.; Forget, Benoit; Smith, Kord

    2013-01-01

    An algorithm for decomposing large tally data in Monte Carlo particle transport simulations is developed, analyzed, and implemented in a continuous-energy Monte Carlo code, OpenMC. The algorithm is based on a non-overlapping decomposition of compute nodes into tracking processors and tally servers. The former are used to simulate the movement of particles through the domain while the latter continuously receive and update tally data. A performance model for this approach is developed, suggesting that, for a range of parameters relevant to LWR analysis, the tally server algorithm should perform with minimal overhead on contemporary supercomputers. An implementation of the algorithm in OpenMC is then tested on the Intrepid and Titan supercomputers, supporting the key predictions of the model over a wide range of parameters. We thus conclude that the tally server algorithm is a successful approach to circumventing classical on-node memory constraints en route to unprecedentedly detailed Monte Carlo reactor simulations

  2. Design of sampling tools for Monte Carlo particle transport code JMCT

    International Nuclear Information System (INIS)

    Shangguan Danhua; Li Gang; Zhang Baoyin; Deng Li

    2012-01-01

    A class of sampling tools for general Monte Carlo particle transport code JMCT is designed. Two ways are provided to sample from distributions. One is the utilization of special sampling methods for special distribution; the other is the utilization of general sampling methods for arbitrary discrete distribution and one-dimensional continuous distribution on a finite interval. Some open source codes are included in the general sampling method for the maximum convenience of users. The sampling results show sampling correctly from distribution which are popular in particle transport can be achieved with these tools, and the user's convenience can be assured. (authors)

  3. Monte Carlo simulations of the particle transport in semiconductor detectors of fast neutrons

    International Nuclear Information System (INIS)

    Sedlačková, Katarína; Zaťko, Bohumír; Šagátová, Andrea; Nečas, Vladimír

    2013-01-01

    Several Monte Carlo all-particle transport codes are under active development around the world. In this paper we focused on the capabilities of the MCNPX code (Monte Carlo N-Particle eXtended) to follow the particle transport in semiconductor detector of fast neutrons. Semiconductor detector based on semi-insulating GaAs was the object of our investigation. As converter material capable to produce charged particles from the (n, p) interaction, a high-density polyethylene (HDPE) was employed. As the source of fast neutrons, the 239 Pu–Be neutron source was used in the model. The simulations were performed using the MCNPX code which makes possible to track not only neutrons but also recoiled protons at all interesting energies. Hence, the MCNPX code enables seamless particle transport and no other computer program is needed to process the particle transport. The determination of the optimal thickness of the conversion layer and the minimum thickness of the active region of semiconductor detector as well as the energy spectra simulation were the principal goals of the computer modeling. Theoretical detector responses showed that the best detection efficiency can be achieved for 500 μm thick HDPE converter layer. The minimum detector active region thickness has been estimated to be about 400 μm. -- Highlights: ► Application of the MCNPX code for fast neutron detector design is demonstrated. ► Simulations of the particle transport through conversion film of HDPE are presented. ► Simulations of the particle transport through detector active region are presented. ► The optimal thickness of the HDPE conversion film has been calculated. ► Detection efficiency of 0.135% was reached for 500 μm thick HDPE conversion film

  4. Optimal Spatial Subdivision method for improving geometry navigation performance in Monte Carlo particle transport simulation

    International Nuclear Information System (INIS)

    Chen, Zhenping; Song, Jing; Zheng, Huaqing; Wu, Bin; Hu, Liqin

    2015-01-01

    Highlights: • The subdivision combines both advantages of uniform and non-uniform schemes. • The grid models were proved to be more efficient than traditional CSG models. • Monte Carlo simulation performance was enhanced by Optimal Spatial Subdivision. • Efficiency gains were obtained for realistic whole reactor core models. - Abstract: Geometry navigation is one of the key aspects of dominating Monte Carlo particle transport simulation performance for large-scale whole reactor models. In such cases, spatial subdivision is an easily-established and high-potential method to improve the run-time performance. In this study, a dedicated method, named Optimal Spatial Subdivision, is proposed for generating numerically optimal spatial grid models, which are demonstrated to be more efficient for geometry navigation than traditional Constructive Solid Geometry (CSG) models. The method uses a recursive subdivision algorithm to subdivide a CSG model into non-overlapping grids, which are labeled as totally or partially occupied, or not occupied at all, by CSG objects. The most important point is that, at each stage of subdivision, a conception of quality factor based on a cost estimation function is derived to evaluate the qualities of the subdivision schemes. Only the scheme with optimal quality factor will be chosen as the final subdivision strategy for generating the grid model. Eventually, the model built with the optimal quality factor will be efficient for Monte Carlo particle transport simulation. The method has been implemented and integrated into the Super Monte Carlo program SuperMC developed by FDS Team. Testing cases were used to highlight the performance gains that could be achieved. Results showed that Monte Carlo simulation runtime could be reduced significantly when using the new method, even as cases reached whole reactor core model sizes

  5. Parallel processing of Monte Carlo code MCNP for particle transport problem

    Energy Technology Data Exchange (ETDEWEB)

    Higuchi, Kenji; Kawasaki, Takuji

    1996-06-01

    It is possible to vectorize or parallelize Monte Carlo codes (MC code) for photon and neutron transport problem, making use of independency of the calculation for each particle. Applicability of existing MC code to parallel processing is mentioned. As for parallel computer, we have used both vector-parallel processor and scalar-parallel processor in performance evaluation. We have made (i) vector-parallel processing of MCNP code on Monte Carlo machine Monte-4 with four vector processors, (ii) parallel processing on Paragon XP/S with 256 processors. In this report we describe the methodology and results for parallel processing on two types of parallel or distributed memory computers. In addition, we mention the evaluation of parallel programming environments for parallel computers used in the present work as a part of the work developing STA (Seamless Thinking Aid) Basic Software. (author)

  6. Solution of charged particle transport equation by Monte-Carlo method in the BRANDZ code system

    International Nuclear Information System (INIS)

    Artamonov, S.N.; Androsenko, P.A.; Androsenko, A.A.

    1992-01-01

    Consideration is given to the issues of Monte-Carlo employment for the solution of charged particle transport equation and its implementation in the BRANDZ code system under the conditions of real 3D geometry and all the data available on radiation-to-matter interaction in multicomponent and multilayer targets. For the solution of implantation problem the results of BRANDZ data comparison with the experiments and calculations by other codes in complexes systems are presented. The results of direct nuclear pumping process simulation for laser-active media by a proton beam are also included. 4 refs.; 7 figs

  7. Creating and using a type of free-form geometry in Monte Carlo particle transport

    International Nuclear Information System (INIS)

    Wessol, D.E.; Wheeler, F.J.

    1993-01-01

    While the reactor physicists were fine-tuning the Monte Carlo paradigm for particle transport in regular geometries, the computer scientists were developing rendering algorithms to display extremely realistic renditions of irregular objects ranging from the ubiquitous teakettle to dynamic Jell-O. Even though the modeling methods share a common basis, the initial strategies each discipline developed for variance reduction were remarkably different. Initially, the reactor physicist used Russian roulette, importance sampling, particle splitting, and rejection techniques. In the early stages of development, the computer scientist relied primarily on rejection techniques, including a very elegant hierarchical construction and sampling method. This sampling method allowed the computer scientist to viably track particles through irregular geometries in three-dimensional space, while the initial methods developed by the reactor physicists would only allow for efficient searches through analytical surfaces or objects. As time goes by, it appears there has been some merging of the variance reduction strategies between the two disciplines. This is an early (possibly first) incorporation of geometric hierarchical construction and sampling into the reactor physicists' Monte Carlo transport model that permits efficient tracking through nonuniform rational B-spline surfaces in three-dimensional space. After some discussion, the results from this model are compared with experiments and the model employing implicit (analytical) geometric representation

  8. Modelling of a general purpose irradiation chamber using a Monte Carlo particle transport code

    International Nuclear Information System (INIS)

    Dhiyauddin Ahmad Fauzi; Sheik, F.O.A.; Nurul Fadzlin Hasbullah

    2013-01-01

    Full-text: The aim of this research is to stimulate the effectiveness use of a general purpose irradiation chamber to contain pure neutron particles obtained from a research reactor. The secondary neutron and gamma particles dose discharge from the chamber layers will be used as a platform to estimate the safe dimension of the chamber. The chamber, made up of layers of lead (Pb), shielding, polyethylene (PE), moderator and commercial grade aluminium (Al) cladding is proposed for the use of interacting samples with pure neutron particles in a nuclear reactor environment. The estimation was accomplished through simulation based on general Monte Carlo N-Particle transport code using Los Alamos MCNPX software. Simulations were performed on the model of the chamber subjected to high neutron flux radiation and its gamma radiation product. The model of neutron particle used is based on the neutron source found in PUSPATI TRIGA MARK II research reactor which holds a maximum flux value of 1 x 10 12 neutron/ cm 2 s. The expected outcomes of this research are zero gamma dose in the core of the chamber and neutron dose rate of less than 10 μSv/ day discharge from the chamber system. (author)

  9. Load balancing in highly parallel processing of Monte Carlo code for particle transport

    International Nuclear Information System (INIS)

    Higuchi, Kenji; Takemiya, Hiroshi; Kawasaki, Takuji

    1998-01-01

    In parallel processing of Monte Carlo (MC) codes for neutron, photon and electron transport problems, particle histories are assigned to processors making use of independency of the calculation for each particle. Although we can easily parallelize main part of a MC code by this method, it is necessary and practically difficult to optimize the code concerning load balancing in order to attain high speedup ratio in highly parallel processing. In fact, the speedup ratio in the case of 128 processors remains in nearly one hundred times when using the test bed for the performance evaluation. Through the parallel processing of the MCNP code, which is widely used in the nuclear field, it is shown that it is difficult to attain high performance by static load balancing in especially neutron transport problems, and a load balancing method, which dynamically changes the number of assigned particles minimizing the sum of the computational and communication costs, overcomes the difficulty, resulting in nearly fifteen percentage of reduction for execution time. (author)

  10. Investigation of pattern recognition techniques for the indentification of splitting surfaces in Monte Carlo particle transport calculations

    International Nuclear Information System (INIS)

    Macdonald, J.L.

    1975-08-01

    Statistical and deterministic pattern recognition systems are designed to classify the state space of a Monte Carlo transport problem into importance regions. The surfaces separating the regions can be used for particle splitting and Russian roulette in state space in order to reduce the variance of the Monte Carlo tally. Computer experiments are performed to evaluate the performance of the technique using one and two dimensional Monte Carlo problems. Additional experiments are performed to determine the sensitivity of the technique to various pattern recognition and Monte Carlo problem dependent parameters. A system for applying the technique to a general purpose Monte Carlo code is described. An estimate of the computer time required by the technique is made in order to determine its effectiveness as a variance reduction device. It is recommended that the technique be further investigated in a general purpose Monte Carlo code. (auth)

  11. Evaluation and comparison of SN and Monte-Carlo charged particle transport calculations

    International Nuclear Information System (INIS)

    Hadad, K.

    2000-01-01

    A study was done to evaluate a 3-D S N charged particle transport code called SMARTEPANTS 1 and another 3-D Monte Carlo code called Integrated Tiger Series, ITS 2 . The evaluation study of SMARTEPANTS code was based on angular discretization and reflected boundary sensitivity whilst the evaluation of ITS was based on CPU time and variance reduction. The comparison of the two code was based on energy and charge deposition calculation in block of Gallium Arsenide with embedded gold cylinders. The result of evaluation tests shows that an S 8 calculation maintains both accuracy and speed and calculations with reflected boundaries geometry produces full symmetrical results. As expected for ITS evaluation, the CPU time and variance reduction are opposite to a point beyond which the history augmentation while increasing the CPU time do not result in variance reduction. The comparison test problem showed excellent agreement in total energy deposition calculations

  12. Transport methods: general. 2. Monte Carlo Particle Transport in Media with Exponentially Varying Time-Dependent Cross Sections

    International Nuclear Information System (INIS)

    Brown, Forrest B.; Martin, William R.

    2001-01-01

    We have investigated Monte Carlo schemes for analyzing particle transport through media with exponentially varying time-dependent cross sections. For such media, the cross sections are represented in the form Σ(t) = Σ 0 e -at (1) or equivalently as Σ(x) = Σ 0 e -bx (2) where b = av and v is the particle speed. For the following discussion, the parameters a and b may be either positive, for exponentially decreasing cross sections, or negative, for exponentially increasing cross sections. For most time-dependent Monte Carlo applications, the time and spatial variations of the cross-section data are handled by means of a stepwise procedure, holding the cross sections constant for each region over a small time interval Δt, performing the Monte Carlo random walk over the interval Δt, updating the cross sections, and then repeating for a series of time intervals. Continuously varying spatial- or time-dependent cross sections can be treated in a rigorous Monte Carlo fashion using delta-tracking, but inefficiencies may arise if the range of cross-section variation is large. In this paper, we present a new method for sampling collision distances directly for cross sections that vary exponentially in space or time. The method is exact and efficient and has direct application to Monte Carlo radiation transport methods. To verify that the probability density function (PDF) is correct and that the random-sampling procedure yields correct results, numerical experiments were performed using a one-dimensional Monte Carlo code. The physical problem consisted of a beam source impinging on a purely absorbing infinite slab, with a slab thickness of 1 cm and Σ 0 = 1 cm -1 . Monte Carlo calculations with 10 000 particles were run for a range of the exponential parameter b from -5 to +20 cm -1 . Two separate Monte Carlo calculations were run for each choice of b, a continuously varying case using the random-sampling procedures described earlier, and a 'conventional' case where the

  13. Advanced Monte Carlo for radiation physics, particle transport simulation and applications. Proceedings

    International Nuclear Information System (INIS)

    Kling, A.; Barao, F.J.C.; Nakagawa, M.; Tavora, L.

    2001-01-01

    The following topics were dealt with: Electron and photon interactions and transport mechanisms, random number generation, applications in medical physisc, microdosimetry, track structure, radiobiological modeling, Monte Carlo method in radiotherapy, dosimetry, and medical accelerator simulation, neutron transport, high-energy hadron transport. (HSI)

  14. Parallel Algorithms for Monte Carlo Particle Transport Simulation on Exascale Computing Architectures

    Science.gov (United States)

    Romano, Paul Kollath

    Monte Carlo particle transport methods are being considered as a viable option for high-fidelity simulation of nuclear reactors. While Monte Carlo methods offer several potential advantages over deterministic methods, there are a number of algorithmic shortcomings that would prevent their immediate adoption for full-core analyses. In this thesis, algorithms are proposed both to ameliorate the degradation in parallel efficiency typically observed for large numbers of processors and to offer a means of decomposing large tally data that will be needed for reactor analysis. A nearest-neighbor fission bank algorithm was proposed and subsequently implemented in the OpenMC Monte Carlo code. A theoretical analysis of the communication pattern shows that the expected cost is O( N ) whereas traditional fission bank algorithms are O(N) at best. The algorithm was tested on two supercomputers, the Intrepid Blue Gene/P and the Titan Cray XK7, and demonstrated nearly linear parallel scaling up to 163,840 processor cores on a full-core benchmark problem. An algorithm for reducing network communication arising from tally reduction was analyzed and implemented in OpenMC. The proposed algorithm groups only particle histories on a single processor into batches for tally purposes---in doing so it prevents all network communication for tallies until the very end of the simulation. The algorithm was tested, again on a full-core benchmark, and shown to reduce network communication substantially. A model was developed to predict the impact of load imbalances on the performance of domain decomposed simulations. The analysis demonstrated that load imbalances in domain decomposed simulations arise from two distinct phenomena: non-uniform particle densities and non-uniform spatial leakage. The dominant performance penalty for domain decomposition was shown to come from these physical effects rather than insufficient network bandwidth or high latency. The model predictions were verified with

  15. Development of CAD-Based Geometry Processing Module for a Monte Carlo Particle Transport Analysis Code

    International Nuclear Information System (INIS)

    Choi, Sung Hoon; Kwark, Min Su; Shim, Hyung Jin

    2012-01-01

    As The Monte Carlo (MC) particle transport analysis for a complex system such as research reactor, accelerator, and fusion facility may require accurate modeling of the complicated geometry. Its manual modeling by using the text interface of a MC code to define the geometrical objects is tedious, lengthy and error-prone. This problem can be overcome by taking advantage of modeling capability of the computer aided design (CAD) system. There have been two kinds of approaches to develop MC code systems utilizing the CAD data: the external format conversion and the CAD kernel imbedded MC simulation. The first approach includes several interfacing programs such as McCAD, MCAM, GEOMIT etc. which were developed to automatically convert the CAD data into the MCNP geometry input data. This approach makes the most of the existing MC codes without any modifications, but implies latent data inconsistency due to the difference of the geometry modeling system. In the second approach, a MC code utilizes the CAD data for the direct particle tracking or the conversion to an internal data structure of the constructive solid geometry (CSG) and/or boundary representation (B-rep) modeling with help of a CAD kernel. MCNP-BRL and OiNC have demonstrated their capabilities of the CAD-based MC simulations. Recently we have developed a CAD-based geometry processing module for the MC particle simulation by using the OpenCASCADE (OCC) library. In the developed module, CAD data can be used for the particle tracking through primitive CAD surfaces (hereafter the CAD-based tracking) or the internal conversion to the CSG data structure. In this paper, the performances of the text-based model, the CAD-based tracking, and the internal CSG conversion are compared by using an in-house MC code, McSIM, equipped with the developed CAD-based geometry processing module

  16. Design of tallying function for general purpose Monte Carlo particle transport code JMCT

    International Nuclear Information System (INIS)

    Shangguan Danhua; Li Gang; Deng Li; Zhang Baoyin

    2013-01-01

    A new postponed accumulation algorithm was proposed. Based on JCOGIN (J combinatorial geometry Monte Carlo transport infrastructure) framework and the postponed accumulation algorithm, the tallying function of the general purpose Monte Carlo neutron-photon transport code JMCT was improved markedly. JMCT gets a higher tallying efficiency than MCNP 4C by 28% for simple geometry model, and JMCT is faster than MCNP 4C by two orders of magnitude for complicated repeated structure model. The available ability of tallying function for JMCT makes firm foundation for reactor analysis and multi-step burnup calculation. (authors)

  17. Tripoli-3: monte Carlo transport code for neutral particles - version 3.5 - users manual

    International Nuclear Information System (INIS)

    Vergnaud, Th.; Nimal, J.C.; Chiron, M.

    2001-01-01

    The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)

  18. Randomly dispersed particle fuel model in the PSG Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Leppaenen, J.

    2007-01-01

    High-temperature gas-cooled reactor fuels are composed of thousands of microscopic fuel particles, randomly dispersed in a graphite matrix. The modelling of such geometry is complicated, especially using continuous-energy Monte Carlo codes, which are unable to apply any deterministic corrections in the calculation. This paper presents the geometry routine developed for modelling randomly dispersed particle fuels using the PSG Monte Carlo reactor physics code. The model is based on the delta-tracking method, and it takes into account the spatial self-shielding effects and the random dispersion of the fuel particles. The calculation routine is validated by comparing the results to reference MCNP4C calculations using uranium and plutonium based fuels. (authors)

  19. A probability-conserving cross-section biasing mechanism for variance reduction in Monte Carlo particle transport calculations

    OpenAIRE

    Mendenhall, Marcus H.; Weller, Robert A.

    2011-01-01

    In Monte Carlo particle transport codes, it is often important to adjust reaction cross sections to reduce the variance of calculations of relatively rare events, in a technique known as non-analogous Monte Carlo. We present the theory and sample code for a Geant4 process which allows the cross section of a G4VDiscreteProcess to be scaled, while adjusting track weights so as to mitigate the effects of altered primary beam depletion induced by the cross section change. This makes it possible t...

  20. Variance analysis of the Monte Carlo perturbation source method in inhomogeneous linear particle transport problems. Derivation of formulae

    International Nuclear Information System (INIS)

    Noack, K.

    1981-01-01

    The perturbation source method is used in the Monte Carlo method in calculating small effects in a particle field. It offers primising possibilities for introducing positive correlation between subtracting estimates even in the cases where other methods fail, in the case of geometrical variations of a given arrangement. The perturbation source method is formulated on the basis of integral equations for the particle fields. The formulae for the second moment of the difference of events are derived. Explicity a certain class of transport games and different procedures for generating the so-called perturbation particles are considered [ru

  1. A method for photon beam Monte Carlo multileaf collimator particle transport

    Science.gov (United States)

    Siebers, Jeffrey V.; Keall, Paul J.; Kim, Jong Oh; Mohan, Radhe

    2002-09-01

    Monte Carlo (MC) algorithms are recognized as the most accurate methodology for patient dose assessment. For intensity-modulated radiation therapy (IMRT) delivered with dynamic multileaf collimators (DMLCs), accurate dose calculation, even with MC, is challenging. Accurate IMRT MC dose calculations require inclusion of the moving MLC in the MC simulation. Due to its complex geometry, full transport through the MLC can be time consuming. The aim of this work was to develop an MLC model for photon beam MC IMRT dose computations. The basis of the MC MLC model is that the complex MLC geometry can be separated into simple geometric regions, each of which readily lends itself to simplified radiation transport. For photons, only attenuation and first Compton scatter interactions are considered. The amount of attenuation material an individual particle encounters while traversing the entire MLC is determined by adding the individual amounts from each of the simplified geometric regions. Compton scatter is sampled based upon the total thickness traversed. Pair production and electron interactions (scattering and bremsstrahlung) within the MLC are ignored. The MLC model was tested for 6 MV and 18 MV photon beams by comparing it with measurements and MC simulations that incorporate the full physics and geometry for fields blocked by the MLC and with measurements for fields with the maximum possible tongue-and-groove and tongue-or-groove effects, for static test cases and for sliding windows of various widths. The MLC model predicts the field size dependence of the MLC leakage radiation within 0.1% of the open-field dose. The entrance dose and beam hardening behind a closed MLC are predicted within +/-1% or 1 mm. Dose undulations due to differences in inter- and intra-leaf leakage are also correctly predicted. The MC MLC model predicts leaf-edge tongue-and-groove dose effect within +/-1% or 1 mm for 95% of the points compared at 6 MV and 88% of the points compared at 18 MV

  2. A method for photon beam Monte Carlo multileaf collimator particle transport

    Energy Technology Data Exchange (ETDEWEB)

    Siebers, Jeffrey V. [Department of Radiation Oncology, Medical College of Virginia Hospitals, Virginia Commonwealth University, Richmond, VA (United States)]. E-mail: jsiebers@vcu.edu; Keall, Paul J.; Kim, Jong Oh; Mohan, Radhe [Department of Radiation Oncology, Medical College of Virginia Hospitals, Virginia Commonwealth University, Richmond, VA (United States)

    2002-09-07

    Monte Carlo (MC) algorithms are recognized as the most accurate methodology for patient dose assessment. For intensity-modulated radiation therapy (IMRT) delivered with dynamic multileaf collimators (DMLCs), accurate dose calculation, even with MC, is challenging. Accurate IMRT MC dose calculations require inclusion of the moving MLC in the MC simulation. Due to its complex geometry, full transport through the MLC can be time consuming. The aim of this work was to develop an MLC model for photon beam MC IMRT dose computations. The basis of the MC MLC model is that the complex MLC geometry can be separated into simple geometric regions, each of which readily lends itself to simplified radiation transport. For photons, only attenuation and first Compton scatter interactions are considered. The amount of attenuation material an individual particle encounters while traversing the entire MLC is determined by adding the individual amounts from each of the simplified geometric regions. Compton scatter is sampled based upon the total thickness traversed. Pair production and electron interactions (scattering and bremsstrahlung) within the MLC are ignored. The MLC model was tested for 6 MV and 18 MV photon beams by comparing it with measurements and MC simulations that incorporate the full physics and geometry for fields blocked by the MLC and with measurements for fields with the maximum possible tongue-and-groove and tongue-or-groove effects, for static test cases and for sliding windows of various widths. The MLC model predicts the field size dependence of the MLC leakage radiation within 0.1% of the open-field dose. The entrance dose and beam hardening behind a closed MLC are predicted within {+-}1% or 1 mm. Dose undulations due to differences in inter- and intra-leaf leakage are also correctly predicted. The MC MLC model predicts leaf-edge tongue-and-groove dose effect within {+-}1% or 1 mm for 95% of the points compared at 6 MV and 88% of the points compared at 18 MV

  3. Neutron secondary-particle production cross sections and their incorporation into Monte-Carlo transport codes

    International Nuclear Information System (INIS)

    Brenner, D.J.; Prael, R.E.; Little, R.C.

    1987-01-01

    Realistic simulations of the passage of fast neutrons through tissue require a large quantity of cross-sectional data. What are needed are differential (in particle type, energy and angle) cross sections. A computer code is described which produces such spectra for neutrons above ∼14 MeV incident on light nuclei such as carbon and oxygen. Comparisons have been made with experimental measurements of double-differential secondary charged-particle production on carbon and oxygen at energies from 27 to 60 MeV; they indicate that the model is adequate in this energy range. In order to utilize fully the results of these calculations, they should be incorporated into a neutron transport code. This requires defining a generalized format for describing charged-particle production, putting the calculated results in this format, interfacing the neutron transport code with these data, and charged-particle transport. The design and development of such a program is described. 13 refs., 3 figs

  4. A probability-conserving cross-section biasing mechanism for variance reduction in Monte Carlo particle transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Mendenhall, Marcus H., E-mail: marcus.h.mendenhall@vanderbilt.edu [Vanderbilt University, Department of Electrical Engineering, P.O. Box 351824B, Nashville, TN 37235 (United States); Weller, Robert A., E-mail: robert.a.weller@vanderbilt.edu [Vanderbilt University, Department of Electrical Engineering, P.O. Box 351824B, Nashville, TN 37235 (United States)

    2012-03-01

    In Monte Carlo particle transport codes, it is often important to adjust reaction cross-sections to reduce the variance of calculations of relatively rare events, in a technique known as non-analog Monte Carlo. We present the theory and sample code for a Geant4 process which allows the cross-section of a G4VDiscreteProcess to be scaled, while adjusting track weights so as to mitigate the effects of altered primary beam depletion induced by the cross-section change. This makes it possible to increase the cross-section of nuclear reactions by factors exceeding 10{sup 4} (in appropriate cases), without distorting the results of energy deposition calculations or coincidence rates. The procedure is also valid for bias factors less than unity, which is useful in problems that involve the computation of particle penetration deep into a target (e.g. atmospheric showers or shielding studies).

  5. A probability-conserving cross-section biasing mechanism for variance reduction in Monte Carlo particle transport calculations

    International Nuclear Information System (INIS)

    Mendenhall, Marcus H.; Weller, Robert A.

    2012-01-01

    In Monte Carlo particle transport codes, it is often important to adjust reaction cross-sections to reduce the variance of calculations of relatively rare events, in a technique known as non-analog Monte Carlo. We present the theory and sample code for a Geant4 process which allows the cross-section of a G4VDiscreteProcess to be scaled, while adjusting track weights so as to mitigate the effects of altered primary beam depletion induced by the cross-section change. This makes it possible to increase the cross-section of nuclear reactions by factors exceeding 10 4 (in appropriate cases), without distorting the results of energy deposition calculations or coincidence rates. The procedure is also valid for bias factors less than unity, which is useful in problems that involve the computation of particle penetration deep into a target (e.g. atmospheric showers or shielding studies).

  6. Tripoli-4, a three-dimensional poly-kinetic particle transport Monte-Carlo code

    International Nuclear Information System (INIS)

    Both, J.P.; Lee, Y.K.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.; Soldevila, M.

    2003-01-01

    In this updated of the Monte-Carlo transport code Tripoli-4, we list and describe its current main features. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. Tripoli-4 enables the user to compute the following physical quantities: a flux, a multiplication factor, a current, a reaction rate, a dose equivalent rate as well as deposit of energy and recoil energies. For each interesting physical quantity, a Monte-Carlo simulation offers different types of estimators. Tripoli-4 has support for execution in parallel mode. Special features and applications are also presented

  7. Tripoli-4, a three-dimensional poly-kinetic particle transport Monte-Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Both, J P; Lee, Y K; Mazzolo, A; Peneliau, Y; Petit, O; Roesslinger, B; Soldevila, M [CEA Saclay, Dir. de l' Energie Nucleaire (DEN/DM2S/SERMA/LEPP), 91 - Gif sur Yvette (France)

    2003-07-01

    In this updated of the Monte-Carlo transport code Tripoli-4, we list and describe its current main features. The code computes coupled neutron-photon propagation as well as the electron-photon cascade shower. While providing the user with common biasing techniques, it also implements an automatic weighting scheme. Tripoli-4 enables the user to compute the following physical quantities: a flux, a multiplication factor, a current, a reaction rate, a dose equivalent rate as well as deposit of energy and recoil energies. For each interesting physical quantity, a Monte-Carlo simulation offers different types of estimators. Tripoli-4 has support for execution in parallel mode. Special features and applications are also presented.

  8. Unbiased estimators of coincidence and correlation in non-analogous Monte Carlo particle transport

    International Nuclear Information System (INIS)

    Szieberth, M.; Kloosterman, J.L.

    2014-01-01

    Highlights: • The history splitting method was developed for non-Boltzmann Monte Carlo estimators. • The method allows variance reduction for pulse-height and higher moment estimators. • It works in highly multiplicative problems but Russian roulette has to be replaced. • Estimation of higher moments allows the simulation of neutron noise measurements. • Biased sampling of fission helps the effective simulation of neutron noise methods. - Abstract: The conventional non-analogous Monte Carlo methods are optimized to preserve the mean value of the distributions. Therefore, they are not suited to non-Boltzmann problems such as the estimation of coincidences or correlations. This paper presents a general method called history splitting for the non-analogous estimation of such quantities. The basic principle of the method is that a non-analogous particle history can be interpreted as a collection of analogous histories with different weights according to the probability of their realization. Calculations with a simple Monte Carlo program for a pulse-height-type estimator prove that the method is feasible and provides unbiased estimation. Different variance reduction techniques have been tried with the method and Russian roulette turned out to be ineffective in high multiplicity systems. An alternative history control method is applied instead. Simulation results of an auto-correlation (Rossi-α) measurement show that even the reconstruction of the higher moments is possible with the history splitting method, which makes the simulation of neutron noise measurements feasible

  9. Enhancements to the Combinatorial Geometry Particle Tracker in the Mercury Monte Carlo Transport Code: Embedded Meshes and Domain Decomposition

    International Nuclear Information System (INIS)

    Greenman, G.M.; O'Brien, M.J.; Procassini, R.J.; Joy, K.I.

    2009-01-01

    Two enhancements to the combinatorial geometry (CG) particle tracker in the Mercury Monte Carlo transport code are presented. The first enhancement is a hybrid particle tracker wherein a mesh region is embedded within a CG region. This method permits efficient calculations of problems with contain both large-scale heterogeneous and homogeneous regions. The second enhancement relates to the addition of parallelism within the CG tracker via spatial domain decomposition. This permits calculations of problems with a large degree of geometric complexity, which are not possible through particle parallelism alone. In this method, the cells are decomposed across processors and a particles is communicated to an adjacent processor when it tracks to an interprocessor boundary. Applications that demonstrate the efficacy of these new methods are presented

  10. Parallel Monte Carlo Particle Transport and the Quality of Random Number Generators: How Good is Good Enough?

    International Nuclear Information System (INIS)

    Procassini, R J; Beck, B R

    2004-01-01

    It might be assumed that use of a ''high-quality'' random number generator (RNG), producing a sequence of ''pseudo random'' numbers with a ''long'' repetition period, is crucial for producing unbiased results in Monte Carlo particle transport simulations. While several theoretical and empirical tests have been devised to check the quality (randomness and period) of an RNG, for many applications it is not clear what level of RNG quality is required to produce unbiased results. This paper explores the issue of RNG quality in the context of parallel, Monte Carlo transport simulations in order to determine how ''good'' is ''good enough''. This study employs the MERCURY Monte Carlo code, which incorporates the CNPRNG library for the generation of pseudo-random numbers via linear congruential generator (LCG) algorithms. The paper outlines the usage of random numbers during parallel MERCURY simulations, and then describes the source and criticality transport simulations which comprise the empirical basis of this study. A series of calculations for each test problem in which the quality of the RNG (period of the LCG) is varied provides the empirical basis for determining the minimum repetition period which may be employed without producing a bias in the mean integrated results

  11. Estimation of coincidence and correlation in non-analogous Monte Carlo particle transport - 159

    International Nuclear Information System (INIS)

    Szieberth, M.; Leen Kloosterman, J.

    2010-01-01

    The conventional non-analogous Monte Carlo methods are optimized to preserve the mean value of the distributions and therefore they are not suited for non-Boltzmann problems like the estimation of coincidences or correlations. This paper presents a general method called history splitting for the non-analogous estimation of such quantities. The basic principle of the method is that a non-analogous particle history can be interpreted as a collection of analogous histories with different weights according to the probability of their realization. Calculations with a simple Monte Carlo program for a pulse-height-type estimator prove that the method is feasible and provides unbiased estimation. Different variance reduction techniques have been tried with the method and Russian roulette turned out to be ineffective in high multiplicity systems. An alternative history control method is applied instead. Simulation results of a Feynman-α measurement shows that even the reconstruction of the higher moments is possible with the history splitting method, which makes the simulation of neutron noise measurements feasible. (authors)

  12. Simulation of neutron transport process, photons and charged particles within the Monte Carlo method

    International Nuclear Information System (INIS)

    Androsenko, A.A.; Androsenko, P.A.; Artamonov, S.N.; Bolonkina, G.V.; Lomtev, V.L.; Pupko, S.V.

    1991-01-01

    Description is given to the program system BRAND designed for the accurate solution of non-stationary transport equation of neutrons, photons and charged particles in the conditions of real three-dimensional geometry. An extensive set of local and non-local estimates provides an opportunity of calculating a great set of linear functionals normally being of interest in the calculation of reactors, radiation protection and experiment simulation. The process of particle interaction with substance is simulated on the basis of individual non-group data on each isotope of the composition. 24 refs

  13. Towards scalable parellelism in Monte Carlo particle transport codes using remote memory access

    Energy Technology Data Exchange (ETDEWEB)

    Romano, Paul K [Los Alamos National Laboratory; Brown, Forrest B [Los Alamos National Laboratory; Forget, Benoit [MIT

    2010-01-01

    One forthcoming challenge in the area of high-performance computing is having the ability to run large-scale problems while coping with less memory per compute node. In this work, they investigate a novel data decomposition method that would allow Monte Carlo transport calculations to be performed on systems with limited memory per compute node. In this method, each compute node remotely retrieves a small set of geometry and cross-section data as needed and remotely accumulates local tallies when crossing the boundary of the local spatial domain. initial results demonstrate that while the method does allow large problems to be run in a memory-limited environment, achieving scalability may be difficult due to inefficiencies in the current implementation of RMA operations.

  14. Towards scalable parallelism in Monte Carlo particle transport codes using remote memory access

    International Nuclear Information System (INIS)

    Romano, Paul K.; Forget, Benoit; Brown, Forrest

    2010-01-01

    One forthcoming challenge in the area of high-performance computing is having the ability to run large-scale problems while coping with less memory per compute node. In this work, we investigate a novel data decomposition method that would allow Monte Carlo transport calculations to be performed on systems with limited memory per compute node. In this method, each compute node remotely retrieves a small set of geometry and cross-section data as needed and remotely accumulates local tallies when crossing the boundary of the local spatial domain. Initial results demonstrate that while the method does allow large problems to be run in a memory-limited environment, achieving scalability may be difficult due to inefficiencies in the current implementation of RMA operations. (author)

  15. SHIELD-HIT12A - a Monte Carlo particle transport program for ion therapy research

    DEFF Research Database (Denmark)

    Bassler, Niels; Hansen, David Christoffer; Lühr, Armin

    2014-01-01

    . We experienced that new users quickly learn to use SHIELD-HIT12A and setup new geometries. Contrary to previous versions of SHIELD-HIT, the 12A distribution comes along with easy-to-use example files and an English manual. A new implementation of Vavilov straggling resulted in a massive reduction......Abstract. Purpose: The Monte Carlo (MC) code SHIELD-HIT simulates the transport of ions through matter. Since SHIELD-HIT08 we added numerous features that improves speed, usability and underlying physics and thereby the user experience. The “-A” fork of SHIELD-HIT also aims to attach SHIELD....... It supports native formats compatible with the heavy ion treatment planning system TRiP. Stopping power files follow ICRU standard and are generated using the libdEdx library, which allows the user to choose from a multitude of stopping power tables. Results: SHIELD-HIT12A runs on Linux and Windows platforms...

  16. Abstract ID: 240 A probabilistic-based nuclear reaction model for Monte Carlo ion transport in particle therapy.

    Science.gov (United States)

    Maria Jose, Gonzalez Torres; Jürgen, Henniger

    2018-01-01

    In order to expand the Monte Carlo transport program AMOS to particle therapy applications, the ion module is being developed in the radiation physics group (ASP) at the TU Dresden. This module simulates the three main interactions of ions in matter for the therapy energy range: elastic scattering, inelastic collisions and nuclear reactions. The simulation of the elastic scattering is based on the Binary Collision Approximation and the inelastic collisions on the Bethe-Bloch theory. The nuclear reactions, which are the focus of the module, are implemented according to a probabilistic-based model developed in the group. The developed model uses probability density functions to sample the occurrence of a nuclear reaction given the initial energy of the projectile particle as well as the energy at which this reaction will take place. The particle is transported until the reaction energy is reached and then the nuclear reaction is simulated. This approach allows a fast evaluation of the nuclear reactions. The theory and application of the proposed model will be addressed in this presentation. The results of the simulation of a proton beam colliding with tissue will also be presented. Copyright © 2017.

  17. Variational variance reduction for particle transport eigenvalue calculations using Monte Carlo adjoint simulation

    International Nuclear Information System (INIS)

    Densmore, Jeffery D.; Larsen, Edward W.

    2003-01-01

    The Variational Variance Reduction (VVR) method is an effective technique for increasing the efficiency of Monte Carlo simulations [Ann. Nucl. Energy 28 (2001) 457; Nucl. Sci. Eng., in press]. This method uses a variational functional, which employs first-order estimates of forward and adjoint fluxes, to yield a second-order estimate of a desired system characteristic - which, in this paper, is the criticality eigenvalue k. If Monte Carlo estimates of the forward and adjoint fluxes are used, each having global 'first-order' errors of O(1/√N), where N is the number of histories used in the Monte Carlo simulation, then the statistical error in the VVR estimation of k will in principle be O(1/N). In this paper, we develop this theoretical possibility and demonstrate with numerical examples that implementations of the VVR method for criticality problems can approximate O(1/N) convergence for significantly large values of N

  18. Multilevel parallel strategy on Monte Carlo particle transport for the large-scale full-core pin-by-pin simulations

    International Nuclear Information System (INIS)

    Zhang, B.; Li, G.; Wang, W.; Shangguan, D.; Deng, L.

    2015-01-01

    This paper introduces the Strategy of multilevel hybrid parallelism of JCOGIN Infrastructure on Monte Carlo Particle Transport for the large-scale full-core pin-by-pin simulations. The particle parallelism, domain decomposition parallelism and MPI/OpenMP parallelism are designed and implemented. By the testing, JMCT presents the parallel scalability of JCOGIN, which reaches the parallel efficiency 80% on 120,000 cores for the pin-by-pin computation of the BEAVRS benchmark. (author)

  19. Convergence acceleration in the Monte-Carlo particle transport code TRIPOLI-4 in criticality

    International Nuclear Information System (INIS)

    Dehaye, Benjamin

    2014-01-01

    Fields such as criticality studies need to compute some values of interest in neutron physics. Two kind of codes may be used: deterministic ones and stochastic ones. The stochastic codes do not require approximation and are thus more exact. However, they may require a lot of time to converge with a sufficient precision.The work carried out during this thesis aims to build an efficient acceleration strategy in the TRIPOLI-4. We wish to implement the zero variance game. To do so, the method requires to compute the adjoint flux. The originality of this work is to directly compute the adjoint flux directly from a Monte-Carlo simulation without using external codes thanks to the fission matrix method. This adjoint flux is then used as an importance map to bias the simulation. (author) [fr

  20. MCNP-DSP, Monte Carlo Neutron-Particle Transport Code with Digital Signal Processing

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description of program or function: MCNP-DSP is recommended only for experienced MCNP users working with subcritical measurements. It is a modification of the Los Alamos National Laboratory's Monte Carlo code MCNP4a that is used to simulate a variety of subcritical measurements. The DSP version was developed to simulate frequency analysis measurements, correlation (Rossi-) measurements, pulsed neutron measurements, Feynman variance measurements, and multiplicity measurements. CCC-700/MCNP4C is recommended for general purpose calculations. 2 - Methods:MCNP-DSP performs calculations very similarly to MCNP and uses the same generalized geometry capabilities of MCNP. MCNP-DSP can only be used with the continuous-energy cross-section data. A variety of source and detector options are available. However, unlike standard MCNP, the source and detector options are limited to those described in the manual because these options are specified in the MCNP-DSP extra data file. MCNP-DSP is used to obtain the time-dependent response of detectors that are modeled in the simulation geometry. The detectors represent actual detectors used in measurements. These time-dependent detector responses are used to compute a variety of quantities such as frequency analysis signatures, correlation signatures, multiplicity signatures, etc., between detectors or sources and detectors. Energy ranges are 0-60 MeV for neutrons (data generally only available up to 20 MeV) and 1 keV - 1 GeV for photons and electrons. 3 - Restrictions on the complexity of the problem: None noted

  1. A photon source model based on particle transport in a parameterized accelerator structure for Monte Carlo dose calculations.

    Science.gov (United States)

    Ishizawa, Yoshiki; Dobashi, Suguru; Kadoya, Noriyuki; Ito, Kengo; Chiba, Takahito; Takayama, Yoshiki; Sato, Kiyokazu; Takeda, Ken

    2018-05-17

    An accurate source model of a medical linear accelerator is essential for Monte Carlo (MC) dose calculations. This study aims to propose an analytical photon source model based on particle transport in parameterized accelerator structures, focusing on a more realistic determination of linac photon spectra compared to existing approaches. We designed the primary and secondary photon sources based on the photons attenuated and scattered by a parameterized flattening filter. The primary photons were derived by attenuating bremsstrahlung photons based on the path length in the filter. Conversely, the secondary photons were derived from the decrement of the primary photons in the attenuation process. This design facilitates these sources to share the free parameters of the filter shape and be related to each other through the photon interaction in the filter. We introduced two other parameters of the primary photon source to describe the particle fluence in penumbral regions. All the parameters are optimized based on calculated dose curves in water using the pencil-beam-based algorithm. To verify the modeling accuracy, we compared the proposed model with the phase space data (PSD) of the Varian TrueBeam 6 and 15 MV accelerators in terms of the beam characteristics and the dose distributions. The EGS5 Monte Carlo code was used to calculate the dose distributions associated with the optimized model and reference PSD in a homogeneous water phantom and a heterogeneous lung phantom. We calculated the percentage of points passing 1D and 2D gamma analysis with 1%/1 mm criteria for the dose curves and lateral dose distributions, respectively. The optimized model accurately reproduced the spectral curves of the reference PSD both on- and off-axis. The depth dose and lateral dose profiles of the optimized model also showed good agreement with those of the reference PSD. The passing rates of the 1D gamma analysis with 1%/1 mm criteria between the model and PSD were 100% for 4

  2. Calculation of absorbed fractions to human skeletal tissues due to alpha particles using the Monte Carlo and 3-d chord-based transport techniques

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, J.G. [Institute of Radiation Protection and Dosimetry, Av. Salvador Allende s/n, Recreio, Rio de Janeiro, CEP 22780-160 (Brazil); Watchman, C.J. [Department of Radiation Oncology, University of Arizona, Tucson, AZ, 85721 (United States); Bolch, W.E. [Department of Nuclear and Radiological Engineering, University of Florida, Gainesville, FL, 32611 (United States); Department of Biomedical Engineering, University of Florida, Gainesville, FL 32611 (United States)

    2007-07-01

    Absorbed fraction (AF) calculations to the human skeletal tissues due to alpha particles are of interest to the internal dosimetry of occupationally exposed workers and members of the public. The transport of alpha particles through the skeletal tissue is complicated by the detailed and complex microscopic histology of the skeleton. In this study, both Monte Carlo and chord-based techniques were applied to the transport of alpha particles through 3-D micro-CT images of the skeletal microstructure of trabecular spongiosa. The Monte Carlo program used was 'Visual Monte Carlo-VMC'. VMC simulates the emission of the alpha particles and their subsequent energy deposition track. The second method applied to alpha transport is the chord-based technique, which randomly generates chord lengths across bone trabeculae and the marrow cavities via alternate and uniform sampling of their cumulative density functions. This paper compares the AF of energy to two radiosensitive skeletal tissues, active marrow and shallow active marrow, obtained with these two techniques. (authors)

  3. Monte Carlo Simulation for Particle Detectors

    CERN Document Server

    Pia, Maria Grazia

    2012-01-01

    Monte Carlo simulation is an essential component of experimental particle physics in all the phases of its life-cycle: the investigation of the physics reach of detector concepts, the design of facilities and detectors, the development and optimization of data reconstruction software, the data analysis for the production of physics results. This note briefly outlines some research topics related to Monte Carlo simulation, that are relevant to future experimental perspectives in particle physics. The focus is on physics aspects: conceptual progress beyond current particle transport schemes, the incorporation of materials science knowledge relevant to novel detection technologies, functionality to model radiation damage, the capability for multi-scale simulation, quantitative validation and uncertainty quantification to determine the predictive power of simulation. The R&D on simulation for future detectors would profit from cooperation within various components of the particle physics community, and synerg...

  4. MC21 v.6.0 - A continuous-energy Monte Carlo particle transport code with integrated reactor feedback capabilities

    International Nuclear Information System (INIS)

    Grieshemer, D.P.; Gill, D.F.; Nease, B.R.; Carpenter, D.C.; Joo, H.; Millman, D.L.; Sutton, T.M.; Stedry, M.H.; Dobreff, P.S.; Trumbull, T.H.; Caro, E.

    2013-01-01

    MC21 is a continuous-energy Monte Carlo radiation transport code for the calculation of the steady-state spatial distributions of reaction rates in three-dimensional models. The code supports neutron and photon transport in fixed source problems, as well as iterated-fission-source (eigenvalue) neutron transport problems. MC21 has been designed and optimized to support large-scale problems in reactor physics, shielding, and criticality analysis applications. The code also supports many in-line reactor feedback effects, including depletion, thermal feedback, xenon feedback, eigenvalue search, and neutron and photon heating. MC21 uses continuous-energy neutron/nucleus interaction physics over the range from 10 -5 eV to 20 MeV. The code treats all common neutron scattering mechanisms, including fast-range elastic and non-elastic scattering, and thermal- and epithermal-range scattering from molecules and crystalline materials. For photon transport, MC21 uses continuous-energy interaction physics over the energy range from 1 keV to 100 GeV. The code treats all common photon interaction mechanisms, including Compton scattering, pair production, and photoelectric interactions. All of the nuclear data required by MC21 is provided by the NDEX system of codes, which extracts and processes data from EPDL-, ENDF-, and ACE-formatted source files. For geometry representation, MC21 employs a flexible constructive solid geometry system that allows users to create spatial cells from first- and second-order surfaces. The system also allows models to be built up as hierarchical collections of previously defined spatial cells, with interior detail provided by grids and template overlays. Results are collected by a generalized tally capability which allows users to edit integral flux and reaction rate information. Results can be collected over the entire problem or within specific regions of interest through the use of phase filters that control which particles are allowed to score each

  5. Angular Distribution of Particles Emerging from a Diffusive Region and its Implications for the Fleck-Canfield Random Walk Algorithm for Implicit Monte Carlo Radiation Transport

    CERN Document Server

    Cooper, M A

    2000-01-01

    We present various approximations for the angular distribution of particles emerging from an optically thick, purely isotropically scattering region into a vacuum. Our motivation is to use such a distribution for the Fleck-Canfield random walk method [1] for implicit Monte Carlo (IMC) [2] radiation transport problems. We demonstrate that the cosine distribution recommended in the original random walk paper [1] is a poor approximation to the angular distribution predicted by transport theory. Then we examine other approximations that more closely match the transport angular distribution.

  6. Weak second-order splitting schemes for Lagrangian Monte Carlo particle methods for the composition PDF/FDF transport equations

    International Nuclear Information System (INIS)

    Wang Haifeng; Popov, Pavel P.; Pope, Stephen B.

    2010-01-01

    We study a class of methods for the numerical solution of the system of stochastic differential equations (SDEs) that arises in the modeling of turbulent combustion, specifically in the Monte Carlo particle method for the solution of the model equations for the composition probability density function (PDF) and the filtered density function (FDF). This system consists of an SDE for particle position and a random differential equation for particle composition. The numerical methods considered advance the solution in time with (weak) second-order accuracy with respect to the time step size. The four primary contributions of the paper are: (i) establishing that the coefficients in the particle equations can be frozen at the mid-time (while preserving second-order accuracy), (ii) examining the performance of three existing schemes for integrating the SDEs, (iii) developing and evaluating different splitting schemes (which treat particle motion, reaction and mixing on different sub-steps), and (iv) developing the method of manufactured solutions (MMS) to assess the convergence of Monte Carlo particle methods. Tests using MMS confirm the second-order accuracy of the schemes. In general, the use of frozen coefficients reduces the numerical errors. Otherwise no significant differences are observed in the performance of the different SDE schemes and splitting schemes.

  7. Monte Carlo Transport for Electron Thermal Transport

    Science.gov (United States)

    Chenhall, Jeffrey; Cao, Duc; Moses, Gregory

    2015-11-01

    The iSNB (implicit Schurtz Nicolai Busquet multigroup electron thermal transport method of Cao et al. is adapted into a Monte Carlo transport method in order to better model the effects of non-local behavior. The end goal is a hybrid transport-diffusion method that combines Monte Carlo Transport with a discrete diffusion Monte Carlo (DDMC). The hybrid method will combine the efficiency of a diffusion method in short mean free path regions with the accuracy of a transport method in long mean free path regions. The Monte Carlo nature of the approach allows the algorithm to be massively parallelized. Work to date on the method will be presented. This work was supported by Sandia National Laboratory - Albuquerque and the University of Rochester Laboratory for Laser Energetics.

  8. Monte Carlo Particle Lists: MCPL

    DEFF Research Database (Denmark)

    Kittelmann, Thomas; Klinkby, Esben Bryndt; Bergbäck Knudsen, Erik

    2017-01-01

    A binary format with lists of particle state information, for interchanging particles between various Monte Carlo simulation applications, is presented. Portable C code for file manipulation is made available to the scientific community, along with converters and plugins for several popular...... simulation packages. Program summary: Program Title: MCPL. Program Files doi: http://dx.doi.org/10.17632/cby92vsv5g.1 Licensing provisions: CC0 for core MCPL, see LICENSE file for details. Programming language: C and C++ External routines/libraries: Geant4, MCNP, McStas, McXtrace Nature of problem: Saving...

  9. Improvement of neutron collimator design for thermal neutron radiography using Monte Carlo N-particle transport code version 5

    International Nuclear Information System (INIS)

    Thiagu Supramaniam

    2007-01-01

    The aim of this research was to propose a new neutron collimator design for thermal neutron radiography facility using tangential beam port of PUSPATI TRIGA Mark II reactor, Malaysia Institute of Nuclear Technology Research (MINT). Best geometry and materials for neutron collimator were chosen in order to obtain a uniform beam with maximum thermal neutron flux, high L/ D ratio, high neutron to gamma ratio and low beam divergence with high resolution. Monte Carlo N-particle Transport Code version 5 (MCNP 5) was used to optimize six neutron collimator components such as beam port medium, neutron scatterer, neutron moderator, gamma filter, aperture and collimator wall. The reactor and tangential beam port setup in MCNP5 was plotted according to its actual sizes. A homogeneous reactor core was assumed and population control method of variance reduction technique was applied by using cell importance. The comparison between experimental results and simulated results of the thermal neutron flux measurement of the bare tangential beam port, shows that both graph obtained had similar pattern. This directly suggests the reliability of MCNP5 in order to obtained optimal neutron collimator parameters. The simulated results of the optimal neutron medium, shows that vacuum was the best medium to transport neutrons followed by helium gas and air. The optimized aperture component was boral with 3 cm thickness. The optimal aperture center hole diameter was 2 cm which produces 88 L/ D ratio. Simulation also shows that graphite neutron scatterer improves thermal neutron flux while reducing fast neutron flux. Neutron moderator was used to moderate fast and epithermal neutrons in the beam port. Paraffin wax with 90 cm thick was bound to be the best neutron moderator material which produces the highest thermal neutron flux at the image plane. Cylindrical shape high density polyethylene neutron collimator produces the highest thermal neutron flux at the image plane rather than divergent

  10. Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

    International Nuclear Information System (INIS)

    White, Morgan C.

    2000-01-01

    The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V and V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to

  11. Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    White, Morgan C. [Univ. of Florida, Gainesville, FL (United States)

    2000-07-01

    The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second

  12. Monte Carlo electron/photon transport

    International Nuclear Information System (INIS)

    Mack, J.M.; Morel, J.E.; Hughes, H.G.

    1985-01-01

    A review of nonplasma coupled electron/photon transport using Monte Carlo method is presented. Remarks are mainly restricted to linerarized formalisms at electron energies from 1 keV to 1000 MeV. Applications involving pulse-height estimation, transport in external magnetic fields, and optical Cerenkov production are discussed to underscore the importance of this branch of computational physics. Advances in electron multigroup cross-section generation is reported, and its impact on future code development assessed. Progress toward the transformation of MCNP into a generalized neutral/charged-particle Monte Carlo code is described. 48 refs

  13. Simulation of transport equations with Monte Carlo

    International Nuclear Information System (INIS)

    Matthes, W.

    1975-09-01

    The main purpose of the report is to explain the relation between the transport equation and the Monte Carlo game used for its solution. The introduction of artificial particles carrying a weight provides one with high flexibility in constructing many different games for the solution of the same equation. This flexibility opens a way to construct a Monte Carlo game for the solution of the adjoint transport equation. Emphasis is laid mostly on giving a clear understanding of what to do and not on the details of how to do a specific game

  14. A kinetic theory for nonanalog Monte Carlo particle transport algorithms: Exponential transform with angular biasing in planar-geometry anisotropically scattering media

    International Nuclear Information System (INIS)

    Ueki, T.; Larsen, E.W.

    1998-01-01

    The authors show that Monte Carlo simulations of neutral particle transport in planargeometry anisotropically scattering media, using the exponential transform with angular biasing as a variance reduction device, are governed by a new Boltzman Monte Carlo (BMC) equation, which includes particle weight as an extra independent variable. The weight moments of the solution of the BMC equation determine the moments of the score and the mean number of collisions per history in the nonanalog Monte Carlo simulations. Therefore, the solution of the BMC equation predicts the variance of the score and the figure of merit in the simulation. Also, by (1) using an angular biasing function that is closely related to the ''asymptotic'' solution of the linear Boltzman equation and (2) requiring isotropic weight changes as collisions, they derive a new angular biasing scheme. Using the BMC equation, they propose a universal ''safe'' upper limit of the transform parameter, valid for any type of exponential transform. In numerical calculations, they demonstrate that the behavior of the Monte Carlo simulations and the performance predicted by deterministically solving the BMC equation agree well, and that the new angular biasing scheme is always advantageous

  15. Comparison of experimental and Monte-Carlo simulation of MeV particle transport through tapered/straight glass capillaries and circular collimators

    Energy Technology Data Exchange (ETDEWEB)

    Hespeels, F., E-mail: felicien.hespeels@unamur.be [University of Namur, PMR, 61 rue de Bruxelles, 5000 Namur (Belgium); Tonneau, R. [University of Namur, PMR, 61 rue de Bruxelles, 5000 Namur (Belgium); Ikeda, T. [RIKEN Nishina Center, 2-1 Hirosawa, Wako, Saitama 351-0198 (Japan); Lucas, S. [University of Namur, PMR, 61 rue de Bruxelles, 5000 Namur (Belgium)

    2015-11-01

    Highlights: • Monte-Carlo simulation for beam transportation through collimations devices. • We confirm the focusing effect of tapered glass capillary. • We confirm the feasibility of using passive collimation devices for ion beam analysis application. - Abstract: This study compares the capabilities of three different passive collimation devices to produce micrometer-sized beams for proton and alpha particle beams (1.7 MeV and 5.3 MeV respectively): classical platinum TEM-like collimators, straight glass capillaries and tapered glass capillaries. In addition, we developed a Monte-Carlo code, based on the Rutherford scattering theory, which simulates particle transportation through collimating devices. The simulation results match the experimental observations of beam transportation through collimators both in air and vacuum. This research shows the focusing effects of tapered capillaries which clearly enable higher transmission flux. Nevertheless, the capillaries alignment with an incident beam is a prerequisite but is tedious, which makes the TEM collimator the easiest way to produce a 50 μm microbeam.

  16. Tripoli-3: monte Carlo transport code for neutral particles - version 3.5 - users manual; Tripoli-3: code de transport des particules neutres par la methode de monte carlo - version 3.5 - manuel d'utilisation

    Energy Technology Data Exchange (ETDEWEB)

    Vergnaud, Th.; Nimal, J.C.; Chiron, M

    2001-07-01

    The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)

  17. Tripoli-3: monte Carlo transport code for neutral particles - version 3.5 - users manual; Tripoli-3: code de transport des particules neutres par la methode de monte carlo - version 3.5 - manuel d'utilisation

    Energy Technology Data Exchange (ETDEWEB)

    Vergnaud, Th; Nimal, J C; Chiron, M

    2001-07-01

    The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)

  18. User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code

    International Nuclear Information System (INIS)

    Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.

    2003-01-01

    This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k eff (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

  19. Monte Carlo method in radiation transport problems

    International Nuclear Information System (INIS)

    Dejonghe, G.; Nimal, J.C.; Vergnaud, T.

    1986-11-01

    In neutral radiation transport problems (neutrons, photons), two values are important: the flux in the phase space and the density of particles. To solve the problem with Monte Carlo method leads to, among other things, build a statistical process (called the play) and to provide a numerical value to a variable x (this attribution is called score). Sampling techniques are presented. Play biasing necessity is proved. A biased simulation is made. At last, the current developments (rewriting of programs for instance) are presented due to several reasons: two of them are the vectorial calculation apparition and the photon and neutron transport in vacancy media [fr

  20. Beta-particle dosimetry of the trabecular skeleton using Monte Carlo transport within 3D digital images

    International Nuclear Information System (INIS)

    Jokisch, D.W.; Bouchet, L.G.; Patton, P.W.; Rajon, D.A.; Bolch, W.E.

    2001-01-01

    Presently, skeletal dosimetry models utilized in clinical medicine simulate electron path lengths through skeletal regions based upon distributions of linear chords measured across bone trabeculae and marrow cavities. In this work, a human thoracic vertebra has been imaged via nuclear magnetic resonance (NMR) spectroscopy yielding a three-dimensional voxelized representation of this skeletal site. The image was then coupled to the radiation transport code EGS4 allowing for 3D tracing of electron paths within its true 3D structure. The macroscopic boundaries of the trabecular regions, as well as the cortex of cortical bone surrounding the bone site, were explicitly considered in the voxelized transport model. For the case of a thoracic vertebra, energy escape to the cortical bone became significant at source energies exceeding ∼2 MeV. Chord-length distributions were acquired from the same NMR image, and subsequently used as input for a chord-based dosimetry model. Differences were observed in the absorbed fractions given by the chord-based model and the voxel transport model, suggesting that some of the input chord distributions for the chord-based models may not be accurate. Finally, this work shows that skeletal mass estimates can be made from the same NMR image in which particle transport is performed. This feature allows one to determine a skeletal S-value using absorbed fraction and mass data taken from the same anatomical tissue sample. The techniques developed in this work may be applied to a variety of skeletal sites, thus allowing for the development of skeletal dosimetry models at all skeletal sites for both males and females and as a function of subject age

  1. Statistics of Monte Carlo methods used in radiation transport calculation

    International Nuclear Information System (INIS)

    Datta, D.

    2009-01-01

    Radiation transport calculation can be carried out by using either deterministic or statistical methods. Radiation transport calculation based on statistical methods is basic theme of the Monte Carlo methods. The aim of this lecture is to describe the fundamental statistics required to build the foundations of Monte Carlo technique for radiation transport calculation. Lecture note is organized in the following way. Section (1) will describe the introduction of Basic Monte Carlo and its classification towards the respective field. Section (2) will describe the random sampling methods, a key component of Monte Carlo radiation transport calculation, Section (3) will provide the statistical uncertainty of Monte Carlo estimates, Section (4) will describe in brief the importance of variance reduction techniques while sampling particles such as photon, or neutron in the process of radiation transport

  2. An integrated high-performance beam optics-nuclear processes framework with hybrid transfer map-Monte Carlo particle transport and optimization

    Energy Technology Data Exchange (ETDEWEB)

    Bandura, L., E-mail: bandura@msu.ed [Argonne National Laboratory, Argonne, IL 60439 (United States); Erdelyi, B. [Argonne National Laboratory, Argonne, IL 60439 (United States); Northern Illinois University, DeKalb, IL 60115 (United States); Nolen, J. [Argonne National Laboratory, Argonne, IL 60439 (United States)

    2010-12-01

    An integrated beam optics-nuclear processes framework is essential for accurate simulation of fragment separator beam dynamics. The code COSY INFINITY provides powerful differential algebraic methods for modeling and beam dynamics simulations in absence of beam-material interactions. However, these interactions are key for accurately simulating the dynamics of heavy ion fragmentation and fission. We have developed an extended version of the code that includes these interactions, and a set of new tools that allow efficient and accurate particle transport: by transfer map in vacuum and by Monte Carlo methods in materials. The new framework is presented, along with several examples from a preliminary layout of a fragment separator for a facility for rare isotope beams.

  3. An integrated high-performance beam optics-nuclear processes framework with hybrid transfer map-Monte Carlo particle transport and optimization

    International Nuclear Information System (INIS)

    Bandura, L.; Erdelyi, B.; Nolen, J.

    2010-01-01

    An integrated beam optics-nuclear processes framework is essential for accurate simulation of fragment separator beam dynamics. The code COSY INFINITY provides powerful differential algebraic methods for modeling and beam dynamics simulations in absence of beam-material interactions. However, these interactions are key for accurately simulating the dynamics of heavy ion fragmentation and fission. We have developed an extended version of the code that includes these interactions, and a set of new tools that allow efficient and accurate particle transport: by transfer map in vacuum and by Monte Carlo methods in materials. The new framework is presented, along with several examples from a preliminary layout of a fragment separator for a facility for rare isotope beams.

  4. Monte Carlo simulation for the transport beamline

    Energy Technology Data Exchange (ETDEWEB)

    Romano, F.; Cuttone, G.; Jia, S. B.; Varisano, A. [INFN, Laboratori Nazionali del Sud, Via Santa Sofia 62, Catania (Italy); Attili, A.; Marchetto, F.; Russo, G. [INFN, Sezione di Torino, Via P.Giuria, 1 10125 Torino (Italy); Cirrone, G. A. P.; Schillaci, F.; Scuderi, V. [INFN, Laboratori Nazionali del Sud, Via Santa Sofia 62, Catania, Italy and Institute of Physics Czech Academy of Science, ELI-Beamlines project, Na Slovance 2, Prague (Czech Republic); Carpinelli, M. [INFN Sezione di Cagliari, c/o Dipartimento di Fisica, Università di Cagliari, Cagliari (Italy); Tramontana, A. [INFN, Laboratori Nazionali del Sud, Via Santa Sofia 62, Catania, Italy and Università di Catania, Dipartimento di Fisica e Astronomia, Via S. Sofia 64, Catania (Italy)

    2013-07-26

    In the framework of the ELIMED project, Monte Carlo (MC) simulations are widely used to study the physical transport of charged particles generated by laser-target interactions and to preliminarily evaluate fluence and dose distributions. An energy selection system and the experimental setup for the TARANIS laser facility in Belfast (UK) have been already simulated with the GEANT4 (GEometry ANd Tracking) MC toolkit. Preliminary results are reported here. Future developments are planned to implement a MC based 3D treatment planning in order to optimize shots number and dose delivery.

  5. Monte Carlo simulation for the transport beamline

    International Nuclear Information System (INIS)

    Romano, F.; Cuttone, G.; Jia, S. B.; Varisano, A.; Attili, A.; Marchetto, F.; Russo, G.; Cirrone, G. A. P.; Schillaci, F.; Scuderi, V.; Carpinelli, M.; Tramontana, A.

    2013-01-01

    In the framework of the ELIMED project, Monte Carlo (MC) simulations are widely used to study the physical transport of charged particles generated by laser-target interactions and to preliminarily evaluate fluence and dose distributions. An energy selection system and the experimental setup for the TARANIS laser facility in Belfast (UK) have been already simulated with the GEANT4 (GEometry ANd Tracking) MC toolkit. Preliminary results are reported here. Future developments are planned to implement a MC based 3D treatment planning in order to optimize shots number and dose delivery

  6. Monte Carlo method for neutron transport problems

    International Nuclear Information System (INIS)

    Asaoka, Takumi

    1977-01-01

    Some methods for decreasing variances in Monte Carlo neutron transport calculations are presented together with the results of sample calculations. A general purpose neutron transport Monte Carlo code ''MORSE'' was used for the purpose. The first method discussed in this report is the method of statistical estimation. As an example of this method, the application of the coarse-mesh rebalance acceleration method to the criticality calculation of a cylindrical fast reactor is presented. Effective multiplication factor and its standard deviation are presented as a function of the number of histories and comparisons are made between the coarse-mesh rebalance method and the standard method. Five-group neutron fluxes at core center are also compared with the result of S4 calculation. The second method is the method of correlated sampling. This method was applied to the perturbation calculation of control rod worths in a fast critical assembly (FCA-V-3) Two methods of sampling (similar flight paths and identical flight paths) are tested and compared with experimental results. For every cases the experimental value lies within the standard deviation of the Monte Carlo calculations. The third method is the importance sampling. In this report a biased selection of particle flight directions discussed. This method was applied to the flux calculation in a spherical fast neutron system surrounded by a 10.16 cm iron reflector. Result-direction biasing, path-length stretching, and no biasing are compared with S8 calculation. (Aoki, K.)

  7. A punctual flux estimator and reactions rates optimization in neutral particles transport calculus by the Monte Carlo method

    International Nuclear Information System (INIS)

    Authier, N.

    1998-12-01

    One of the questions asked in radiation shielding problems is the estimation of the radiation level in particular to determine accessibility of working persons in controlled area (nuclear power plants, nuclear fuel reprocessing plants) or to study the dose gradients encountered in material (iron nuclear vessel, medical therapy, electronics in satellite). The flux and reaction rate estimators used in Monte Carlo codes give average values in volumes or on surfaces of the geometrical description of the system. But in certain configurations, the knowledge of punctual deposited energy and dose estimates are necessary. The Monte Carlo estimate of the flux at a point of interest is a calculus which presents an unbounded variance. The central limit theorem cannot be applied thus no easy confidence level may be calculated. The convergence rate is then very poor. We propose in this study a new solution for the photon flux at a point estimator. The method is based on the 'once more collided flux estimator' developed earlier for neutron calculations. It solves the problem of the unbounded variance and do not add any bias to the estimation. We show however that our new sampling schemes specially developed to treat the anisotropy of the photon coherent scattering is necessary for a good and regular behavior of the estimator. This developments integrated in the TRIPOLI-4 Monte Carlo code add the possibility of an unbiased punctual estimate on media interfaces. (author)

  8. 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

  9. Spot: a new Monte Carlo solver for fast alpha particles

    International Nuclear Information System (INIS)

    Schneider, M.; Eriksson, L.G.; Basiuk, V.; Imbeaux, F.

    2004-01-01

    The predictive transport code CRONOS has been augmented by an orbit following Monte Carlo code, SPOT (Simulation of Particle Orbits in a Tokamak). The SPOT code simulates the dynamics of nonthermal particles, and takes into account effects of finite orbit width and collisional transport of fast ions. Recent developments indicate that it might be difficult to avoid, at least transiently, current holes in a reactor. They occur already on existing tokamaks during advanced tokamak scenarios. The SPOT code has been used to study the alpha particle behaviour in the presence of current holes for both JET and ITER relevant parameters. (authors)

  10. Three-dimensional Monte Carlo simulations of W7-X plasma transport: density control and particle balance in steady-state operations

    International Nuclear Information System (INIS)

    Sharma, D.; Feng, Y.; Sardei, F.; Reiter, D.

    2005-01-01

    This paper presents self-consistent three-dimensional (3D) plasma transport simulations in the boundary of stellarator W7-X obtained with the Monte Carlo code EMC3-EIRENE for three typical island divertor configurations. The chosen 3D grid consists of relatively simple nested finite toroidal surfaces defined on a toroidal field period and covering the whole edge topology, which includes closed surfaces, islands and ergodic regions. Local grid refinements account for the required high resolution in the divertor region. The distribution of plasma density and temperature in the divertor region, as well as the power deposition profiles on the divertor plates, are shown to strongly depend on the island geometry, i.e. on the position and size of the dominant island chain. Configurations with strike-point positions closer to the gap of the divertor chamber generally favour the neutral compression in the divertor chamber and hence the pumping efficiency. The ratio of pumping to recycling fluxes is found to be roughly independent of the separatrix density and is thus a figure of merit for the quality of the configuration and of the divertor system in terms of density control. Lower limits for the achievable separatrix density, which determine the particle exhaust capabilities in stationary conditions, are compared for the three W7-X configurations

  11. Optimization of Monte Carlo particle transport parameters and validation of a novel high throughput experimental setup to measure the biological effects of particle beams.

    Science.gov (United States)

    Patel, Darshana; Bronk, Lawrence; Guan, Fada; Peeler, Christopher R; Brons, Stephan; Dokic, Ivana; Abdollahi, Amir; Rittmüller, Claudia; Jäkel, Oliver; Grosshans, David; Mohan, Radhe; Titt, Uwe

    2017-11-01

    Accurate modeling of the relative biological effectiveness (RBE) of particle beams requires increased systematic in vitro studies with human cell lines with care towards minimizing uncertainties in biologic assays as well as physical parameters. In this study, we describe a novel high-throughput experimental setup and an optimized parameterization of the Monte Carlo (MC) simulation technique that is universally applicable for accurate determination of RBE of clinical ion beams. Clonogenic cell-survival measurements on a human lung cancer cell line (H460) are presented using proton irradiation. Experiments were performed at the Heidelberg Ion Therapy Center (HIT) with support from the Deutsches Krebsforschungszentrum (DKFZ) in Heidelberg, Germany using a mono-energetic horizontal proton beam. A custom-made variable range selector was designed for the horizontal beam line using the Geant4 MC toolkit. This unique setup enabled a high-throughput clonogenic assay investigation of multiple, well defined dose and linear energy transfer (LETs) per irradiation for human lung cancer cells (H460) cultured in a 96-well plate. Sensitivity studies based on application of different physics lists in conjunction with different electromagnetic constructors and production threshold values to the MC simulations were undertaken for accurate assessment of the calculated dose and the dose-averaged LET (LET d ). These studies were extended to helium and carbon ion beams. Sensitivity analysis of the MC parameterization revealed substantial dependence of the dose and LET d values on both the choice of physics list and the production threshold values. While the dose and LET d calculations using FTFP_BERT_LIV, FTFP_BERT_EMZ, FTFP_BERT_PEN and QGSP_BIC_EMY physics lists agree well with each other for all three ions, they show large differences when compared to the FTFP_BERT physics list with the default electromagnetic constructor. For carbon ions, the dose corresponding to the largest LET d

  12. Vectorizing and macrotasking Monte Carlo neutral particle algorithms

    International Nuclear Information System (INIS)

    Heifetz, D.B.

    1987-04-01

    Monte Carlo algorithms for computing neutral particle transport in plasmas have been vectorized and macrotasked. The techniques used are directly applicable to Monte Carlo calculations of neutron and photon transport, and Monte Carlo integration schemes in general. A highly vectorized code was achieved by calculating test flight trajectories in loops over arrays of flight data, isolating the conditional branches to as few a number of loops as possible. A number of solutions are discussed to the problem of gaps appearing in the arrays due to completed flights, which impede vectorization. A simple and effective implementation of macrotasking is achieved by dividing the calculation of the test flight profile among several processors. A tree of random numbers is used to ensure reproducible results. The additional memory required for each task may preclude using a larger number of tasks. In future machines, the limit of macrotasking may be possible, with each test flight, and split test flight, being a separate task

  13. Monte Carlo simulation of particle-induced bit upsets

    Science.gov (United States)

    Wrobel, Frédéric; Touboul, Antoine; Vaillé, Jean-Roch; Boch, Jérôme; Saigné, Frédéric

    2017-09-01

    We investigate the issue of radiation-induced failures in electronic devices by developing a Monte Carlo tool called MC-Oracle. It is able to transport the particles in device, to calculate the energy deposited in the sensitive region of the device and to calculate the transient current induced by the primary particle and the secondary particles produced during nuclear reactions. We compare our simulation results with SRAM experiments irradiated with neutrons, protons and ions. The agreement is very good and shows that it is possible to predict the soft error rate (SER) for a given device in a given environment.

  14. Monte Carlo simulation of particle-induced bit upsets

    Directory of Open Access Journals (Sweden)

    Wrobel Frédéric

    2017-01-01

    Full Text Available We investigate the issue of radiation-induced failures in electronic devices by developing a Monte Carlo tool called MC-Oracle. It is able to transport the particles in device, to calculate the energy deposited in the sensitive region of the device and to calculate the transient current induced by the primary particle and the secondary particles produced during nuclear reactions. We compare our simulation results with SRAM experiments irradiated with neutrons, protons and ions. The agreement is very good and shows that it is possible to predict the soft error rate (SER for a given device in a given environment.

  15. The MC21 Monte Carlo Transport Code

    International Nuclear Information System (INIS)

    Sutton TM; Donovan TJ; Trumbull TH; Dobreff PS; Caro E; Griesheimer DP; Tyburski LJ; Carpenter DC; Joo H

    2007-01-01

    MC21 is a new Monte Carlo neutron and photon transport code currently under joint development at the Knolls Atomic Power Laboratory and the Bettis Atomic Power Laboratory. MC21 is the Monte Carlo transport kernel of the broader Common Monte Carlo Design Tool (CMCDT), which is also currently under development. The vision for CMCDT is to provide an automated, computer-aided modeling and post-processing environment integrated with a Monte Carlo solver that is optimized for reactor analysis. CMCDT represents a strategy to push the Monte Carlo method beyond its traditional role as a benchmarking tool or ''tool of last resort'' and into a dominant design role. This paper describes various aspects of the code, including the neutron physics and nuclear data treatments, the geometry representation, and the tally and depletion capabilities

  16. Monte Carlo simulations of the Galileo energetic particle detector

    International Nuclear Information System (INIS)

    Jun, I.; Ratliff, J.M.; Garrett, H.B.; McEntire, R.W.

    2002-01-01

    Monte Carlo radiation transport studies have been performed for the Galileo spacecraft energetic particle detector (EPD) in order to study its response to energetic electrons and protons. Three-dimensional Monte Carlo radiation transport codes, MCNP version 4B (for electrons) and MCNPX version 2.2.3 (for protons), were used throughout the study. The results are presented in the form of 'geometric factors' for the high-energy channels studied in this paper: B1, DC2, and DC3 for electrons and B0, DC0, and DC1 for protons. The geometric factor is the energy-dependent detector response function that relates the incident particle fluxes to instrument count rates. The trend of actual data measured by the EPD was successfully reproduced using the geometric factors obtained in this study

  17. Monte Carlo simulations of the Galileo energetic particle detector

    CERN Document Server

    Jun, I; Garrett, H B; McEntire, R W

    2002-01-01

    Monte Carlo radiation transport studies have been performed for the Galileo spacecraft energetic particle detector (EPD) in order to study its response to energetic electrons and protons. Three-dimensional Monte Carlo radiation transport codes, MCNP version 4B (for electrons) and MCNPX version 2.2.3 (for protons), were used throughout the study. The results are presented in the form of 'geometric factors' for the high-energy channels studied in this paper: B1, DC2, and DC3 for electrons and B0, DC0, and DC1 for protons. The geometric factor is the energy-dependent detector response function that relates the incident particle fluxes to instrument count rates. The trend of actual data measured by the EPD was successfully reproduced using the geometric factors obtained in this study.

  18. Parallel processing Monte Carlo radiation transport codes

    International Nuclear Information System (INIS)

    McKinney, G.W.

    1994-01-01

    Issues related to distributed-memory multiprocessing as applied to Monte Carlo radiation transport are discussed. Measurements of communication overhead are presented for the radiation transport code MCNP which employs the communication software package PVM, and average efficiency curves are provided for a homogeneous virtual machine

  19. Load Balancing of Parallel Monte Carlo Transport Calculations

    International Nuclear Information System (INIS)

    Procassini, R J; O'Brien, M J; Taylor, J M

    2005-01-01

    The performance of parallel Monte Carlo transport calculations which use both spatial and particle parallelism is increased by dynamically assigning processors to the most worked domains. Since he particle work load varies over the course of the simulation, this algorithm determines each cycle if dynamic load balancing would speed up the calculation. If load balancing is required, a small number of particle communications are initiated in order to achieve load balance. This method has decreased the parallel run time by more than a factor of three for certain criticality calculations

  20. Dynamic Load Balancing of Parallel Monte Carlo Transport Calculations

    International Nuclear Information System (INIS)

    O'Brien, M; Taylor, J; Procassini, R

    2004-01-01

    The performance of parallel Monte Carlo transport calculations which use both spatial and particle parallelism is increased by dynamically assigning processors to the most worked domains. Since the particle work load varies over the course of the simulation, this algorithm determines each cycle if dynamic load balancing would speed up the calculation. If load balancing is required, a small number of particle communications are initiated in order to achieve load balance. This method has decreased the parallel run time by more than a factor of three for certain criticality calculations

  1. Automatic modeling for the monte carlo transport TRIPOLI code

    International Nuclear Information System (INIS)

    Zhang Junjun; Zeng Qin; Wu Yican; Wang Guozhong; FDS Team

    2010-01-01

    TRIPOLI, developed by CEA, France, is Monte Carlo particle transport simulation code. It has been widely applied to nuclear physics, shielding design, evaluation of nuclear safety. However, it is time-consuming and error-prone to manually describe the TRIPOLI input file. This paper implemented bi-directional conversion between CAD model and TRIPOLI model. Its feasibility and efficiency have been demonstrated by several benchmarking examples. (authors)

  2. Parallel MCNP Monte Carlo transport calculations with MPI

    International Nuclear Information System (INIS)

    Wagner, J.C.; Haghighat, A.

    1996-01-01

    The steady increase in computational performance has made Monte Carlo calculations for large/complex systems possible. However, in order to make these calculations practical, order of magnitude increases in performance are necessary. The Monte Carlo method is inherently parallel (particles are simulated independently) and thus has the potential for near-linear speedup with respect to the number of processors. Further, the ever-increasing accessibility of parallel computers, such as workstation clusters, facilitates the practical use of parallel Monte Carlo. Recognizing the nature of the Monte Carlo method and the trends in available computing, the code developers at Los Alamos National Laboratory implemented the message-passing general-purpose Monte Carlo radiation transport code MCNP (version 4A). The PVM package was chosen by the MCNP code developers because it supports a variety of communication networks, several UNIX platforms, and heterogeneous computer systems. This PVM version of MCNP has been shown to produce speedups that approach the number of processors and thus, is a very useful tool for transport analysis. Due to software incompatibilities on the local IBM SP2, PVM has not been available, and thus it is not possible to take advantage of this useful tool. Hence, it became necessary to implement an alternative message-passing library package into MCNP. Because the message-passing interface (MPI) is supported on the local system, takes advantage of the high-speed communication switches in the SP2, and is considered to be the emerging standard, it was selected

  3. Analysis of error in Monte Carlo transport calculations

    International Nuclear Information System (INIS)

    Booth, T.E.

    1979-01-01

    The Monte Carlo method for neutron transport calculations suffers, in part, because of the inherent statistical errors associated with the method. Without an estimate of these errors in advance of the calculation, it is difficult to decide what estimator and biasing scheme to use. Recently, integral equations have been derived that, when solved, predicted errors in Monte Carlo calculations in nonmultiplying media. The present work allows error prediction in nonanalog Monte Carlo calculations of multiplying systems, even when supercritical. Nonanalog techniques such as biased kernels, particle splitting, and Russian Roulette are incorporated. Equations derived here allow prediction of how much a specific variance reduction technique reduces the number of histories required, to be weighed against the change in time required for calculation of each history. 1 figure, 1 table

  4. Cost of splitting in Monte Carlo transport

    International Nuclear Information System (INIS)

    Everett, C.J.; Cashwell, E.D.

    1978-03-01

    In a simple transport problem designed to estimate transmission through a plane slab of x free paths by Monte Carlo methods, it is shown that m-splitting (m > or = 2) does not pay unless exp(x) > m(m + 3)/(m - 1). In such a case, the minimum total cost in terms of machine time is obtained as a function of m, and the optimal value of m is determined

  5. Memory bottlenecks and memory contention in multi-core Monte Carlo transport codes

    International Nuclear Information System (INIS)

    Tramm, J.R.; Siegel, A.R.

    2013-01-01

    The simulation of whole nuclear cores through the use of Monte Carlo codes requires an impracticably long time-to-solution. We have extracted a kernel that executes only the most computationally expensive steps of the Monte Carlo particle transport algorithm - the calculation of macroscopic cross sections - in an effort to expose bottlenecks within multi-core, shared memory architectures. (authors)

  6. Ripple enhanced transport of suprathermal alpha particles

    International Nuclear Information System (INIS)

    Tani, K.; Takizuka, T.; Azumi, M.

    1986-01-01

    The ripple enhanced transport of suprathermal alpha particles has been studied by the newly developed Monte-Carlo code in which the motion of banana orbit in a toroidal field ripple is described by a mapping method. The existence of ripple-resonance diffusion has been confirmed numerically. We have developed another new code in which the radial displacement of banana orbit is given by the diffusion coefficients from the mapping code or the orbit following Monte-Carlo code. The ripple loss of α particles during slowing down has been estimated by the mapping model code as well as the diffusion model code. From the comparison of the results with those from the orbit-following Monte-Carlo code, it has been found that all of them agree very well. (author)

  7. Monte Carlo methods in electron transport problems. Pt. 1

    International Nuclear Information System (INIS)

    Cleri, F.

    1989-01-01

    The condensed-history Monte Carlo method for charged particles transport is reviewed and discussed starting from a general form of the Boltzmann equation (Part I). The physics of the electronic interactions, together with some pedagogic example will be introduced in the part II. The lecture is directed to potential users of the method, for which it can be a useful introduction to the subject matter, and wants to establish the basis of the work on the computer code RECORD, which is at present in a developing stage

  8. Monte Carlo radiation transport: A revolution in science

    International Nuclear Information System (INIS)

    Hendricks, J.

    1993-01-01

    When Enrico Fermi, Stan Ulam, Nicholas Metropolis, John von Neuman, and Robert Richtmyer invented the Monte Carlo method fifty years ago, little could they imagine the far-flung consequences, the international applications, and the revolution in science epitomized by their abstract mathematical method. The Monte Carlo method is used in a wide variety of fields to solve exact computational models approximately by statistical sampling. It is an alternative to traditional physics modeling methods which solve approximate computational models exactly by deterministic methods. Modern computers and improved methods, such as variance reduction, have enhanced the method to the point of enabling a true predictive capability in areas such as radiation or particle transport. This predictive capability has contributed to a radical change in the way science is done: design and understanding come from computations built upon experiments rather than being limited to experiments, and the computer codes doing the computations have become the repository for physics knowledge. The MCNP Monte Carlo computer code effort at Los Alamos is an example of this revolution. Physicians unfamiliar with physics details can design cancer treatments using physics buried in the MCNP computer code. Hazardous environments and hypothetical accidents can be explored. Many other fields, from underground oil well exploration to aerospace, from physics research to energy production, from safety to bulk materials processing, benefit from MCNP, the Monte Carlo method, and the revolution in science

  9. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for PKA energy spectra and heating number under neutron irradiation

    International Nuclear Information System (INIS)

    Iwamoto, Y.; Ogawa, T.

    2016-01-01

    The modelling of the damage in materials irradiated by neutrons is needed for understanding the mechanism of radiation damage in fission and fusion reactor facilities. The molecular dynamics simulations of damage cascades with full atomic interactions require information about the energy distribution of the Primary Knock on Atoms (PKAs). The most common process to calculate PKA energy spectra under low-energy neutron irradiation is to use the nuclear data processing code NJOY2012. It calculates group-to-group recoil cross section matrices using nuclear data libraries in ENDF data format, which is energy and angular recoil distributions for many reactions. After the NJOY2012 process, SPKA6C is employed to produce PKA energy spectra combining recoil cross section matrices with an incident neutron energy spectrum. However, intercomparison with different processes and nuclear data libraries has not been studied yet. Especially, the higher energy (~5 MeV) of the incident neutrons, compared to fission, leads to many reaction channels, which produces a complex distribution of PKAs in energy and type. Recently, we have developed the event generator mode (EGM) in the Particle and Heavy Ion Transport code System PHITS for neutron incident reactions in the energy region below 20 MeV. The main feature of EGM is to produce PKA with keeping energy and momentum conservation in a reaction. It is used for event-by-event analysis in application fields such as soft error analysis in semiconductors, micro dosimetry in human body, and estimation of Displacement per Atoms (DPA) value in metals and so on. The purpose of this work is to specify differences of PKA spectra and heating number related with kerma between different calculation method using PHITS-EGM and NJOY2012+SPKA6C with different libraries TENDL-2015, ENDF/B-VII.1 and JENDL-4.0 for fusion relevant materials

  10. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for recoil cross section spectra under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Iwamoto, Yosuke, E-mail: iwamoto.yosuke@jaea.go.jp; Ogawa, Tatsuhiko

    2017-04-01

    Because primary knock-on atoms (PKAs) create point defects and clusters in materials that are irradiated with neutrons, it is important to validate the calculations of recoil cross section spectra that are used to estimate radiation damage in materials. Here, the recoil cross section spectra of fission- and fusion-relevant materials were calculated using the Event Generator Mode (EGM) of the Particle and Heavy Ion Transport code System (PHITS) and also using the data processing code NJOY2012 with the nuclear data libraries TENDL2015, ENDF/BVII.1, and JEFF3.2. The heating number, which is the integral of the recoil cross section spectra, was also calculated using PHITS-EGM and compared with data extracted from the ACE files of TENDL2015, ENDF/BVII.1, and JENDL4.0. In general, only a small difference was found between the PKA spectra of PHITS + TENDL2015 and NJOY + TENDL2015. From analyzing the recoil cross section spectra extracted from the nuclear data libraries using NJOY2012, we found that the recoil cross section spectra were incorrect for {sup 72}Ge, {sup 75}As, {sup 89}Y, and {sup 109}Ag in the ENDF/B-VII.1 library, and for {sup 90}Zr and {sup 55}Mn in the JEFF3.2 library. From analyzing the heating number, we found that the data extracted from the ACE file of TENDL2015 for all nuclides were problematic in the neutron capture region because of incorrect data regarding the emitted gamma energy. However, PHITS + TENDL2015 can calculate PKA spectra and heating numbers correctly.

  11. Exponential convergence on a continuous Monte Carlo transport problem

    International Nuclear Information System (INIS)

    Booth, T.E.

    1997-01-01

    For more than a decade, it has been known that exponential convergence on discrete transport problems was possible using adaptive Monte Carlo techniques. An adaptive Monte Carlo method that empirically produces exponential convergence on a simple continuous transport problem is described

  12. Heavy particle transport in sputtering systems

    Science.gov (United States)

    Trieschmann, Jan

    2015-09-01

    This contribution aims to discuss the theoretical background of heavy particle transport in plasma sputtering systems such as direct current magnetron sputtering (dcMS), high power impulse magnetron sputtering (HiPIMS), or multi frequency capacitively coupled plasmas (MFCCP). Due to inherently low process pressures below one Pa only kinetic simulation models are suitable. In this work a model appropriate for the description of the transport of film forming particles sputtered of a target material has been devised within the frame of the OpenFOAM software (specifically dsmcFoam). The three dimensional model comprises of ejection of sputtered particles into the reactor chamber, their collisional transport through the volume, as well as deposition of the latter onto the surrounding surfaces (i.e. substrates, walls). An angular dependent Thompson energy distribution fitted to results from Monte-Carlo simulations is assumed initially. Binary collisions are treated via the M1 collision model, a modified variable hard sphere (VHS) model. The dynamics of sputtered and background gas species can be resolved self-consistently following the direct simulation Monte-Carlo (DSMC) approach or, whenever possible, simplified based on the test particle method (TPM) with the assumption of a constant, non-stationary background at a given temperature. At the example of an MFCCP research reactor the transport of sputtered aluminum is specifically discussed. For the peculiar configuration and under typical process conditions with argon as process gas the transport of aluminum sputtered of a circular target is shown to be governed by a one dimensional interaction of the imposed and backscattered particle fluxes. The results are analyzed and discussed on the basis of the obtained velocity distribution functions (VDF). This work is supported by the German Research Foundation (DFG) in the frame of the Collaborative Research Centre TRR 87.

  13. Relativity primer for particle transport. A LASL monograph

    International Nuclear Information System (INIS)

    Everett, C.J.; Cashwell, E.D.

    1979-04-01

    The basic principles of special relativity involved in Monte Carlo transport problems are developed with emphasis on the possible transmutations of particles, and on computational methods. Charged particle ballistics and polarized scattering are included, as well as a discussion of colliding beams

  14. PEREGRINE: An all-particle Monte Carlo code for radiation therapy

    International Nuclear Information System (INIS)

    Hartmann Siantar, C.L.; Chandler, W.P.; Rathkopf, J.A.; Svatos, M.M.; White, R.M.

    1994-09-01

    The goal of radiation therapy is to deliver a lethal dose to the tumor while minimizing the dose to normal tissues. To carry out this task, it is critical to calculate correctly the distribution of dose delivered. Monte Carlo transport methods have the potential to provide more accurate prediction of dose distributions than currently-used methods. PEREGRINE is a new Monte Carlo transport code developed at Lawrence Livermore National Laboratory for the specific purpose of modeling the effects of radiation therapy. PEREGRINE transports neutrons, photons, electrons, positrons, and heavy charged-particles, including protons, deuterons, tritons, helium-3, and alpha particles. This paper describes the PEREGRINE transport code and some preliminary results for clinically relevant materials and radiation sources

  15. General particle transport equation. Final report

    International Nuclear Information System (INIS)

    Lafi, A.Y.; Reyes, J.N. Jr.

    1994-12-01

    The general objectives of this research are as follows: (1) To develop fundamental models for fluid particle coalescence and breakage rates for incorporation into statistically based (Population Balance Approach or Monte Carlo Approach) two-phase thermal hydraulics codes. (2) To develop fundamental models for flow structure transitions based on stability theory and fluid particle interaction rates. This report details the derivation of the mass, momentum and energy conservation equations for a distribution of spherical, chemically non-reacting fluid particles of variable size and velocity. To study the effects of fluid particle interactions on interfacial transfer and flow structure requires detailed particulate flow conservation equations. The equations are derived using a particle continuity equation analogous to Boltzmann's transport equation. When coupled with the appropriate closure equations, the conservation equations can be used to model nonequilibrium, two-phase, dispersed, fluid flow behavior. Unlike the Eulerian volume and time averaged conservation equations, the statistically averaged conservation equations contain additional terms that take into account the change due to fluid particle interfacial acceleration and fluid particle dynamics. Two types of particle dynamics are considered; coalescence and breakage. Therefore, the rate of change due to particle dynamics will consider the gain and loss involved in these processes and implement phenomenological models for fluid particle breakage and coalescence

  16. Monte Carlo charged-particle tracking and energy deposition on a Lagrangian mesh.

    Science.gov (United States)

    Yuan, J; Moses, G A; McKenty, P W

    2005-10-01

    A Monte Carlo algorithm for alpha particle tracking and energy deposition on a cylindrical computational mesh in a Lagrangian hydrodynamics code used for inertial confinement fusion (ICF) simulations is presented. The straight line approximation is used to follow propagation of "Monte Carlo particles" which represent collections of alpha particles generated from thermonuclear deuterium-tritium (DT) reactions. Energy deposition in the plasma is modeled by the continuous slowing down approximation. The scheme addresses various aspects arising in the coupling of Monte Carlo tracking with Lagrangian hydrodynamics; such as non-orthogonal severely distorted mesh cells, particle relocation on the moving mesh and particle relocation after rezoning. A comparison with the flux-limited multi-group diffusion transport method is presented for a polar direct drive target design for the National Ignition Facility. Simulations show the Monte Carlo transport method predicts about earlier ignition than predicted by the diffusion method, and generates higher hot spot temperature. Nearly linear speed-up is achieved for multi-processor parallel simulations.

  17. Monte Carlo simulation of the turbulent transport of airborne contaminants

    International Nuclear Information System (INIS)

    Watson, C.W.; Barr, S.

    1975-09-01

    A generalized, three-dimensional Monte Carlo model and computer code (SPOOR) are described for simulating atmospheric transport and dispersal of small pollutant clouds. A cloud is represented by a large number of particles that we track by statistically sampling simulated wind and turbulence fields. These fields are based on generalized wind data for large-scale flow and turbulent energy spectra for the micro- and mesoscales. The large-scale field can be input from a climatological data base, or by means of real-time analyses, or from a separate, subjectively defined data base. We introduce the micro- and mesoscale wind fluctuations through a power spectral density, to include effects from a broad spectrum of turbulent-energy scales. The role of turbulence is simulated in both meander and dispersal. Complex flow fields and time-dependent diffusion rates are accounted for naturally, and shear effects are simulated automatically in the ensemble of particle trajectories. An important adjunct has been the development of computer-graphics displays. These include two- and three-dimensional (perspective) snapshots and color motion pictures of particle ensembles, plus running displays of differential and integral cloud characteristics. The model's versatility makes it a valuable atmospheric research tool that we can adapt easily into broader, multicomponent systems-analysis codes. Removal, transformation, dry or wet deposition, and resuspension of contaminant particles can be readily included

  18. A 3D Monte Carlo code for plasma transport in island divertors

    International Nuclear Information System (INIS)

    Feng, Y.; Sardei, F.; Kisslinger, J.; Grigull, P.

    1997-01-01

    A fully 3D self-consistent Monte Carlo code EMC3 (edge Monte Carlo 3D) for modelling the plasma transport in island divertors has been developed. In a first step, the code solves a simplified version of the 3D time-independent plasma fluid equations. Coupled to the neutral transport code EIRENE, the EMC3 code has been used to study the particle, energy and neutral transport in W7-AS island divertor configurations. First results are compared with data from different diagnostics (Langmuir probes, H α cameras and thermography). (orig.)

  19. Advantages of Analytical Transformations in Monte Carlo Methods for Radiation Transport

    International Nuclear Information System (INIS)

    McKinley, M S; Brooks III, E D; Daffin, F

    2004-01-01

    Monte Carlo methods for radiation transport typically attempt to solve an integral by directly sampling analog or weighted particles, which are treated as physical entities. Improvements to the methods involve better sampling, probability games or physical intuition about the problem. We show that significant improvements can be achieved by recasting the equations with an analytical transform to solve for new, non-physical entities or fields. This paper looks at one such transform, the difference formulation for thermal photon transport, showing a significant advantage for Monte Carlo solution of the equations for time dependent transport. Other related areas are discussed that may also realize significant benefits from similar analytical transformations

  20. A Monte Carlo simulation code for calculating damage and particle transport in solids: The case for electron-bombarded solids for electron energies up to 900 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Yan, Qiang [College of Nuclear Science and Technology, Harbin Engineering University, Harbin 150001 (China); Shao, Lin, E-mail: lshao@tamu.edu [Department of Nuclear Engineering, Texas A& M University, College Station, TX 77843 (United States)

    2017-03-15

    Current popular Monte Carlo simulation codes for simulating electron bombardment in solids focus primarily on electron trajectories, instead of electron-induced displacements. Here we report a Monte Carol simulation code, DEEPER (damage creation and particle transport in matter), developed for calculating 3-D distributions of displacements produced by electrons of incident energies up to 900 MeV. Electron elastic scattering is calculated by using full-Mott cross sections for high accuracy, and primary-knock-on-atoms (PKAs)-induced damage cascades are modeled using ZBL potential. We compare and show large differences in 3-D distributions of displacements and electrons in electron-irradiated Fe. The distributions of total displacements are similar to that of PKAs at low electron energies. But they are substantially different for higher energy electrons due to the shifting of PKA energy spectra towards higher energies. The study is important to evaluate electron-induced radiation damage, for the applications using high flux electron beams to intentionally introduce defects and using an electron analysis beam for microstructural characterization of nuclear materials.

  1. A Fano cavity test for Monte Carlo proton transport algorithms

    International Nuclear Information System (INIS)

    Sterpin, Edmond; Sorriaux, Jefferson; Souris, Kevin; Vynckier, Stefaan; Bouchard, Hugo

    2014-01-01

    Purpose: In the scope of reference dosimetry of radiotherapy beams, Monte Carlo (MC) simulations are widely used to compute ionization chamber dose response accurately. Uncertainties related to the transport algorithm can be verified performing self-consistency tests, i.e., the so-called “Fano cavity test.” The Fano cavity test is based on the Fano theorem, which states that under charged particle equilibrium conditions, the charged particle fluence is independent of the mass density of the media as long as the cross-sections are uniform. Such tests have not been performed yet for MC codes simulating proton transport. The objectives of this study are to design a new Fano cavity test for proton MC and to implement the methodology in two MC codes: Geant4 and PENELOPE extended to protons (PENH). Methods: The new Fano test is designed to evaluate the accuracy of proton transport. Virtual particles with an energy ofE 0 and a mass macroscopic cross section of (Σ)/(ρ) are transported, having the ability to generate protons with kinetic energy E 0 and to be restored after each interaction, thus providing proton equilibrium. To perform the test, the authors use a simplified simulation model and rigorously demonstrate that the computed cavity dose per incident fluence must equal (ΣE 0 )/(ρ) , as expected in classic Fano tests. The implementation of the test is performed in Geant4 and PENH. The geometry used for testing is a 10 × 10 cm 2 parallel virtual field and a cavity (2 × 2 × 0.2 cm 3 size) in a water phantom with dimensions large enough to ensure proton equilibrium. Results: For conservative user-defined simulation parameters (leading to small step sizes), both Geant4 and PENH pass the Fano cavity test within 0.1%. However, differences of 0.6% and 0.7% were observed for PENH and Geant4, respectively, using larger step sizes. For PENH, the difference is attributed to the random-hinge method that introduces an artificial energy straggling if step size is not

  2. Interface methods for hybrid Monte Carlo-diffusion radiation-transport simulations

    International Nuclear Information System (INIS)

    Densmore, Jeffery D.

    2006-01-01

    Discrete diffusion Monte Carlo (DDMC) is a technique for increasing the efficiency of Monte Carlo simulations in diffusive media. An important aspect of DDMC is the treatment of interfaces between diffusive regions, where DDMC is used, and transport regions, where standard Monte Carlo is employed. Three previously developed methods exist for treating transport-diffusion interfaces: the Marshak interface method, based on the Marshak boundary condition, the asymptotic interface method, based on the asymptotic diffusion-limit boundary condition, and the Nth-collided source technique, a scheme that allows Monte Carlo particles to undergo several collisions in a diffusive region before DDMC is used. Numerical calculations have shown that each of these interface methods gives reasonable results as part of larger radiation-transport simulations. In this paper, we use both analytic and numerical examples to compare the ability of these three interface techniques to treat simpler, transport-diffusion interface problems outside of a more complex radiation-transport calculation. We find that the asymptotic interface method is accurate regardless of the angular distribution of Monte Carlo particles incident on the interface surface. In contrast, the Marshak boundary condition only produces correct solutions if the incident particles are isotropic. We also show that the Nth-collided source technique has the capacity to yield accurate results if spatial cells are optically small and Monte Carlo particles are allowed to undergo many collisions within a diffusive region before DDMC is employed. These requirements make the Nth-collided source technique impractical for realistic radiation-transport calculations

  3. Towards a Revised Monte Carlo Neutral Particle Surface Interaction Model

    International Nuclear Information System (INIS)

    Stotler, D.P.

    2005-01-01

    The components of the neutral- and plasma-surface interaction model used in the Monte Carlo neutral transport code DEGAS 2 are reviewed. The idealized surfaces and processes handled by that model are inadequate for accurately simulating neutral transport behavior in present day and future fusion devices. We identify some of the physical processes missing from the model, such as mixed materials and implanted hydrogen, and make some suggestions for improving the model

  4. Discrete Diffusion Monte Carlo for Electron Thermal Transport

    Science.gov (United States)

    Chenhall, Jeffrey; Cao, Duc; Wollaeger, Ryan; Moses, Gregory

    2014-10-01

    The iSNB (implicit Schurtz Nicolai Busquet electron thermal transport method of Cao et al. is adapted to a Discrete Diffusion Monte Carlo (DDMC) solution method for eventual inclusion in a hybrid IMC-DDMC (Implicit Monte Carlo) method. The hybrid method will combine the efficiency of a diffusion method in short mean free path regions with the accuracy of a transport method in long mean free path regions. The Monte Carlo nature of the approach allows the algorithm to be massively parallelized. Work to date on the iSNB-DDMC method will be presented. This work was supported by Sandia National Laboratory - Albuquerque.

  5. Particle-gamma and particle-particle correlations in nuclear reactions using Monte Carlo Hauser-Feshback model

    Energy Technology Data Exchange (ETDEWEB)

    Kawano, Toshihiko [Los Alamos National Laboratory; Talou, Patrick [Los Alamos National Laboratory; Watanabe, Takehito [Los Alamos National Laboratory; Chadwick, Mark [Los Alamos National Laboratory

    2010-01-01

    Monte Carlo simulations for particle and {gamma}-ray emissions from an excited nucleus based on the Hauser-Feshbach statistical theory are performed to obtain correlated information between emitted particles and {gamma}-rays. We calculate neutron induced reactions on {sup 51}V to demonstrate unique advantages of the Monte Carlo method. which are the correlated {gamma}-rays in the neutron radiative capture reaction, the neutron and {gamma}-ray correlation, and the particle-particle correlations at higher energies. It is shown that properties in nuclear reactions that are difficult to study with a deterministic method can be obtained with the Monte Carlo simulations.

  6. Monte Carlo methods for flux expansion solutions of transport problems

    International Nuclear Information System (INIS)

    Spanier, J.

    1999-01-01

    Adaptive Monte Carlo methods, based on the use of either correlated sampling or importance sampling, to obtain global solutions to certain transport problems have recently been described. The resulting learning algorithms are capable of achieving geometric convergence when applied to the estimation of a finite number of coefficients in a flux expansion representation of the global solution. However, because of the nonphysical nature of the random walk simulations needed to perform importance sampling, conventional transport estimators and source sampling techniques require modification to be used successfully in conjunction with such flux expansion methods. It is shown how these problems can be overcome. First, the traditional path length estimators in wide use in particle transport simulations are generalized to include rather general detector functions (which, in this application, are the individual basis functions chosen for the flus expansion). Second, it is shown how to sample from the signed probabilities that arise as source density functions in these applications, without destroying the zero variance property needed to ensure geometric convergence to zero error

  7. Development of Monte Carlo decay gamma-ray transport calculation system

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Kawasaki, Nobuo [Fujitsu Ltd., Tokyo (Japan); Kume, Etsuo [Japan Atomic Energy Research Inst., Center for Promotion of Computational Science and Engineering, Tokai, Ibaraki (Japan)

    2001-06-01

    In the DT fusion reactor, it is critical concern to evaluate the decay gamma-ray biological dose rates after the reactor shutdown exactly. In order to evaluate the decay gamma-ray biological dose rates exactly, three dimensional Monte Carlo decay gamma-ray transport calculation system have been developed by connecting the three dimensional Monte Carlo particle transport calculation code and the induced activity calculation code. The developed calculation system consists of the following four functions. (1) The operational neutron flux distribution is calculated by the three dimensional Monte Carlo particle transport calculation code. (2) The induced activities are calculated by the induced activity calculation code. (3) The decay gamma-ray source distribution is obtained from the induced activities. (4) The decay gamma-rays are generated by using the decay gamma-ray source distribution, and the decay gamma-ray transport calculation is conducted by the three dimensional Monte Carlo particle transport calculation code. In order to reduce the calculation time drastically, a biasing system for the decay gamma-ray source distribution has been developed, and the function is also included in the present system. In this paper, the outline and the detail of the system, and the execution example are reported. The evaluation for the effect of the biasing system is also reported. (author)

  8. Modelling of neutral particle transport in divertor plasma

    International Nuclear Information System (INIS)

    Kakizuka, Tomonori; Shimizu, Katsuhiro

    1995-01-01

    An outline of the modelling of neutral particle transport in the diverter plasma was described in the paper. The characteristic properties of divertor plasma were largely affected by interaction between neutral particles and divertor plasma. Accordingly, the behavior of neutral particle should be investigated quantitatively. Moreover, plasma and neutral gas should be traced consistently in the plasma simulation. There are Monte Carlo modelling and the neutral gas fluid modelling as the transport modelling. The former need long calculation time, but it is able to make the physical process modelling. A ultra-large parallel computer is good for the former. In spite of proposing some kinds of models, the latter has not been established. At the view point of reducing calculation time, a work station is good for the simulation of the latter, although some physical problems have not been solved. On the Monte Carlo method particle modelling, reducing the calculation time and introducing the interaction of particles are important subjects to develop 'the evolutional Monte Carlo Method'. To reduce the calculation time, two new methods: 'Implicit Monte Carlo method' and 'Free-and Diffusive-Motion Hybrid Monte-Carlo method' have been developing. (S.Y.)

  9. Monte Carlo transport of electrons and positrons through thin foils

    International Nuclear Information System (INIS)

    Legarda, F.; Idoeta, R.

    2000-01-01

    In the different measurements made with electrons traversing matter it becomes useful the knowledge of its transmission through that medium, their paths and their angular distribution through matter so as to process and get information about the traversed medium and to improve and innovate the techniques that employ electrons, as medical applications or materials irradiation. This work presents a simulation of the transport of beams of electrons and positrons through thin foils using an analog Monte Carlo code that simulates in a detailed way every electron movement or interaction in matter. As those particles penetrate thin absorbers it has been assumed that they interact with matter only through elastic scattering, with negligible energy loss. This type of interaction has been described quite precisely because its angular form influences very much the angular distribution of electrons and positrons in matter. With this code it has been calculated the number of particles, with energies between 100 and 3000 keV, that are transmitted through different media of various thicknesses as well as its angular distribution, showing a good agreement with experimental data. The discrepancies are less than 5% for thicknesses lower than about 30% of the corresponding range in the tested material. As elastic scattering is very anisotropic, angular distributions resemble a collimated incident beam for very thin foils becoming slowly more isotropic when absorber thickness is increased. (author)

  10. Monte Carlo simulations of interacting particle mixtures in ratchet potentials

    International Nuclear Information System (INIS)

    Fendrik, A J; Romanelli, L

    2012-01-01

    There are different models of devices for achieving a separation of mixtures of particles by using the ratchet effect. On the other hand, it has been proposed that one could also control the separation by means of appropriate interactions. Through Monte Carlo simulations, we show that inclusion of simple interactions leads to a decrease of the ratchet effect and therefore also a separation of the mixtures.

  11. Optix: A Monte Carlo scintillation light transport code

    Energy Technology Data Exchange (ETDEWEB)

    Safari, M.J., E-mail: mjsafari@aut.ac.ir [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Afarideh, H. [Department of Energy Engineering and Physics, Amir Kabir University of Technology, PO Box 15875-4413, Tehran (Iran, Islamic Republic of); Ghal-Eh, N. [School of Physics, Damghan University, PO Box 36716-41167, Damghan (Iran, Islamic Republic of); Davani, F. Abbasi [Nuclear Engineering Department, Shahid Beheshti University, PO Box 1983963113, Tehran (Iran, Islamic Republic of)

    2014-02-11

    The paper reports on the capabilities of Monte Carlo scintillation light transport code Optix, which is an extended version of previously introduced code Optics. Optix provides the user a variety of both numerical and graphical outputs with a very simple and user-friendly input structure. A benchmarking strategy has been adopted based on the comparison with experimental results, semi-analytical solutions, and other Monte Carlo simulation codes to verify various aspects of the developed code. Besides, some extensive comparisons have been made against the tracking abilities of general-purpose MCNPX and FLUKA codes. The presented benchmark results for the Optix code exhibit promising agreements. -- Highlights: • Monte Carlo simulation of scintillation light transport in 3D geometry. • Evaluation of angular distribution of detected photons. • Benchmark studies to check the accuracy of Monte Carlo simulations.

  12. A midway forward-adjoint coupling method for neutron and photon Monte Carlo transport

    International Nuclear Information System (INIS)

    Serov, I.V.; John, T.M.; Hoogenboom, J.E.

    1999-01-01

    The midway Monte Carlo method for calculating detector responses combines a forward and an adjoint Monte Carlo calculation. In both calculations, particle scores are registered at a surface to be chosen by the user somewhere between the source and detector domains. The theory of the midway response determination is developed within the framework of transport theory for external sources and for criticality theory. The theory is also developed for photons, which are generated at inelastic scattering or capture of neutrons. In either the forward or the adjoint calculation a so-called black absorber technique can be applied; i.e., particles need not be followed after passing the midway surface. The midway Monte Carlo method is implemented in the general-purpose MCNP Monte Carlo code. The midway Monte Carlo method is demonstrated to be very efficient in problems with deep penetration, small source and detector domains, and complicated streaming paths. All the problems considered pose difficult variance reduction challenges. Calculations were performed using existing variance reduction methods of normal MCNP runs and using the midway method. The performed comparative analyses show that the midway method appears to be much more efficient than the standard techniques in an overwhelming majority of cases and can be recommended for use in many difficult variance reduction problems of neutral particle transport

  13. A 3D particle Monte Carlo approach to studying nucleation

    DEFF Research Database (Denmark)

    Köhn, Christoph; Bødker Enghoff, Martin; Svensmark, Henrik

    2018-01-01

    The nucleation of sulphuric acid molecules plays a key role in the formation of aerosols. We here present a three dimensional particle Monte Carlo model to study the growth of sulphuric acid clusters as well as its dependence on the ambient temperature and the initial particle density. We initiate...... a swarm of sulphuric acid–water clusters with a size of 0.329 nm with densities between 107 and and 108 cm-3 at temperatures between 200 and 300 K and a relative humidity of 50%. After every time step, we update the position of particles as a function of size-dependent diffusion coefficients. If two...... particles encounter, we merge them and add their volumes and masses. Inversely, we check after every time step whether a polymer evaporates liberating a molecule. We present the spatial distribution as well as the size distribution calculated from individual clusters. We also calculate the nucleation rate...

  14. Confidence interval procedures for Monte Carlo transport simulations

    International Nuclear Information System (INIS)

    Pederson, S.P.

    1997-01-01

    The problem of obtaining valid confidence intervals based on estimates from sampled distributions using Monte Carlo particle transport simulation codes such as MCNP is examined. Such intervals can cover the true parameter of interest at a lower than nominal rate if the sampled distribution is extremely right-skewed by large tallies. Modifications to the standard theory of confidence intervals are discussed and compared with some existing heuristics, including batched means normality tests. Two new types of diagnostics are introduced to assess whether the conditions of central limit theorem-type results are satisfied: the relative variance of the variance determines whether the sample size is sufficiently large, and estimators of the slope of the right tail of the distribution are used to indicate the number of moments that exist. A simulation study is conducted to quantify the relationship between various diagnostics and coverage rates and to find sample-based quantities useful in indicating when intervals are expected to be valid. Simulated tally distributions are chosen to emulate behavior seen in difficult particle transport problems. Measures of variation in the sample variance s 2 are found to be much more effective than existing methods in predicting when coverage will be near nominal rates. Batched means tests are found to be overly conservative in this regard. A simple but pathological MCNP problem is presented as an example of false convergence using existing heuristics. The new methods readily detect the false convergence and show that the results of the problem, which are a factor of 4 too small, should not be used. Recommendations are made for applying these techniques in practice, using the statistical output currently produced by MCNP

  15. Modification to the Monte Carlo N-Particle (MCNP) Visual Editor (MCNPVised) to Read in Computer Aided Design (CAD) Files

    International Nuclear Information System (INIS)

    Randolph Schwarz; Leland L. Carter; Alysia Schwarz

    2005-01-01

    Monte Carlo N-Particle Transport Code (MCNP) is the code of choice for doing complex neutron/photon/electron transport calculations for the nuclear industry and research institutions. The Visual Editor for Monte Carlo N-Particle is internationally recognized as the best code for visually creating and graphically displaying input files for MCNP. The work performed in this grant was used to enhance the capabilities of the MCNP Visual Editor to allow it to read in both 2D and 3D Computer Aided Design (CAD) files, allowing the user to electronically generate a valid MCNP input geometry

  16. SPHERE: a spherical-geometry multimaterial electron/photon Monte Carlo transport code

    International Nuclear Information System (INIS)

    Halbleib, J.A. Sr.

    1977-06-01

    SPHERE provides experimenters and theorists with a method for the routine solution of coupled electron/photon transport through multimaterial configurations possessing spherical symmetry. Emphasis is placed upon operational simplicity without sacrificing the rigor of the model. SPHERE combines condensed-history electron Monte Carlo with conventional single-scattering photon Monte Carlo in order to describe the transport of all generations of particles from several MeV down to 1.0 and 10.0 keV for electrons and photons, respectively. The model is more accurate at the higher energies, with a less rigorous description of the particle cascade at energies where the shell structure of the transport media becomes important. Flexibility of construction permits the user to tailor the model to specific applications and to extend the capabilities of the model to more sophisticated applications through relatively simple update procedures. 8 figs., 3 tables

  17. A hybrid transport-diffusion Monte Carlo method for frequency-dependent radiative-transfer simulations

    International Nuclear Information System (INIS)

    Densmore, Jeffery D.; Thompson, Kelly G.; Urbatsch, Todd J.

    2012-01-01

    Discrete Diffusion Monte Carlo (DDMC) is a technique for increasing the efficiency of Implicit Monte Carlo radiative-transfer simulations in optically thick media. In DDMC, particles take discrete steps between spatial cells according to a discretized diffusion equation. Each discrete step replaces many smaller Monte Carlo steps, thus improving the efficiency of the simulation. In this paper, we present an extension of DDMC for frequency-dependent radiative transfer. We base our new DDMC method on a frequency-integrated diffusion equation for frequencies below a specified threshold, as optical thickness is typically a decreasing function of frequency. Above this threshold we employ standard Monte Carlo, which results in a hybrid transport-diffusion scheme. With a set of frequency-dependent test problems, we confirm the accuracy and increased efficiency of our new DDMC method.

  18. Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method

    CERN Document Server

    2002-01-01

    This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.

  19. Particle transport in urban dwellings

    International Nuclear Information System (INIS)

    Cannell, R.J.; Goddard, A.J.H.; ApSimon, H.M.

    1988-01-01

    A quantitative investigation of the potential for contamination of a dwelling by material carried in on the occupants' footwear has been completed. Data are now available on the transport capacity of different footwear for a small range of particle sizes and contamination source strengths. Additional information is also given on the rate of redistribution

  20. Methodology of Continuous-Energy Adjoint Monte Carlo for Neutron, Photon, and Coupled Neutron-Photon Transport

    International Nuclear Information System (INIS)

    Hoogenboom, J. Eduard

    2003-01-01

    Adjoint Monte Carlo may be a useful alternative to regular Monte Carlo calculations in cases where a small detector inhibits an efficient Monte Carlo calculation as only very few particle histories will cross the detector. However, in general purpose Monte Carlo codes, normally only the multigroup form of adjoint Monte Carlo is implemented. In this article the general methodology for continuous-energy adjoint Monte Carlo neutron transport is reviewed and extended for photon and coupled neutron-photon transport. In the latter cases the discrete photons generated by annihilation or by neutron capture or inelastic scattering prevent a direct application of the general methodology. Two successive reaction events must be combined in the selection process to accommodate the adjoint analog of a reaction resulting in a photon with a discrete energy. Numerical examples illustrate the application of the theory for some simplified problems

  1. PHITS-a particle and heavy ion transport code system

    International Nuclear Information System (INIS)

    Niita, Koji; Sato, Tatsuhiko; Iwase, Hiroshi; Nose, Hiroyuki; Nakashima, Hiroshi; Sihver, Lembit

    2006-01-01

    The paper presents a summary of the recent development of the multi-purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS. In particular, we discuss in detail the development of two new models, JAM and JQMD, for high energy particle interactions, incorporated in PHITS, and show comparisons between model calculations and experiments for the validations of these models. The paper presents three applications of the code including spallation neutron source, heavy ion therapy and space radiation. The results and examples shown indicate PHITS has great ability of carrying out the radiation transport analysis of almost all particles including heavy ions within a wide energy range

  2. A computer code package for electron transport Monte Carlo simulation

    International Nuclear Information System (INIS)

    Popescu, Lucretiu M.

    1999-01-01

    A computer code package was developed for solving various electron transport problems by Monte Carlo simulation. It is based on condensed history Monte Carlo algorithm. In order to get reliable results over wide ranges of electron energies and target atomic numbers, specific techniques of electron transport were implemented such as: Moliere multiscatter angular distributions, Blunck-Leisegang multiscatter energy distribution, sampling of electron-electron and Bremsstrahlung individual interactions. Path-length and lateral displacement corrections algorithms and the module for computing collision, radiative and total restricted stopping powers and ranges of electrons are also included. Comparisons of simulation results with experimental measurements are finally presented. (author)

  3. A hybrid transport-diffusion method for Monte Carlo radiative-transfer simulations

    International Nuclear Information System (INIS)

    Densmore, Jeffery D.; Urbatsch, Todd J.; Evans, Thomas M.; Buksas, Michael W.

    2007-01-01

    Discrete Diffusion Monte Carlo (DDMC) is a technique for increasing the efficiency of Monte Carlo particle-transport simulations in diffusive media. If standard Monte Carlo is used in such media, particle histories will consist of many small steps, resulting in a computationally expensive calculation. In DDMC, particles take discrete steps between spatial cells according to a discretized diffusion equation. Each discrete step replaces many small Monte Carlo steps, thus increasing the efficiency of the simulation. In addition, given that DDMC is based on a diffusion equation, it should produce accurate solutions if used judiciously. In practice, DDMC is combined with standard Monte Carlo to form a hybrid transport-diffusion method that can accurately simulate problems with both diffusive and non-diffusive regions. In this paper, we extend previously developed DDMC techniques in several ways that improve the accuracy and utility of DDMC for nonlinear, time-dependent, radiative-transfer calculations. The use of DDMC in these types of problems is advantageous since, due to the underlying linearizations, optically thick regions appear to be diffusive. First, we employ a diffusion equation that is discretized in space but is continuous in time. Not only is this methodology theoretically more accurate than temporally discretized DDMC techniques, but it also has the benefit that a particle's time is always known. Thus, there is no ambiguity regarding what time to assign a particle that leaves an optically thick region (where DDMC is used) and begins transporting by standard Monte Carlo in an optically thin region. Also, we treat the interface between optically thick and optically thin regions with an improved method, based on the asymptotic diffusion-limit boundary condition, that can produce accurate results regardless of the angular distribution of the incident Monte Carlo particles. Finally, we develop a technique for estimating radiation momentum deposition during the

  4. Monte Carlo simulation of neutron transport phenomena

    International Nuclear Information System (INIS)

    Srinivasan, P.

    2009-01-01

    Neutron transport is one of the central problems in nuclear reactor related studies and other applied sciences. Some of the major applications of neutron transport include nuclear reactor design and safety, criticality safety of fissile material handling, neutron detector design and development, nuclear medicine, assessment of radiation damage to materials, nuclear well logging, forensic analysis etc. Most reactor and dosimetry studies assume that neutrons diffuse from regions of high to low density just like gaseous molecules diffuse to regions of low concentration or heat flow from high to low temperature regions. However while treatment of gaseous or heat diffusion is quite accurately modeled, treatment of neutron transport as simple diffusion is quite limited. In simple diffusion, the neutron trajectories are irregular, random and zigzag - where as in neutron transport low reaction cross sections (1 barn- 10 -24 cm 2 ) lead to long mean free paths which again depend on the nature and irregularities of the medium. Hence a more accurate representation of the neutron transport evolved based on the Boltzmann equation of kinetic gas theory. In fact the neutron transport equation is a linearized version of the Boltzmann gas equation based on neutron conservation with seven independent variables. The transport equation is difficult to solve except in simple cases amenable to numerical methods. The diffusion (equation) approximation follows from removing the angular dependence of the neutron flux

  5. A 3D particle Monte Carlo approach to studying nucleation

    Science.gov (United States)

    Köhn, Christoph; Enghoff, Martin Bødker; Svensmark, Henrik

    2018-06-01

    The nucleation of sulphuric acid molecules plays a key role in the formation of aerosols. We here present a three dimensional particle Monte Carlo model to study the growth of sulphuric acid clusters as well as its dependence on the ambient temperature and the initial particle density. We initiate a swarm of sulphuric acid-water clusters with a size of 0.329 nm with densities between 107 and 108 cm-3 at temperatures between 200 and 300 K and a relative humidity of 50%. After every time step, we update the position of particles as a function of size-dependent diffusion coefficients. If two particles encounter, we merge them and add their volumes and masses. Inversely, we check after every time step whether a polymer evaporates liberating a molecule. We present the spatial distribution as well as the size distribution calculated from individual clusters. We also calculate the nucleation rate of clusters with a radius of 0.85 nm as a function of time, initial particle density and temperature. The nucleation rates obtained from the presented model agree well with experimentally obtained values and those of a numerical model which serves as a benchmark of our code. In contrast to previous nucleation models, we here present for the first time a code capable of tracing individual particles and thus of capturing the physics related to the discrete nature of particles.

  6. Review of Monte Carlo methods for particle multiplicity evaluation

    CERN Document Server

    Armesto-Pérez, Nestor

    2005-01-01

    I present a brief review of the existing models for particle multiplicity evaluation in heavy ion collisions which are at our disposal in the form of Monte Carlo simulators. Models are classified according to the physical mechanisms with which they try to describe the different stages of a high-energy collision between heavy nuclei. A comparison of predictions, as available at the beginning of year 2000, for multiplicities in central AuAu collisions at the BNL Relativistic Heavy Ion Collider (RHIC) and PbPb collisions at the CERN Large Hadron Collider (LHC) is provided.

  7. Review of Monte Carlo methods for particle multiplicity evaluation

    International Nuclear Information System (INIS)

    Armesto, Nestor

    2005-01-01

    I present a brief review of the existing models for particle multiplicity evaluation in heavy ion collisions which are at our disposal in the form of Monte Carlo simulators. Models are classified according to the physical mechanisms with which they try to describe the different stages of a high-energy collision between heavy nuclei. A comparison of predictions, as available at the beginning of year 2000, for multiplicities in central AuAu collisions at the BNL Relativistic Heavy Ion Collider (RHIC) and PbPb collisions at the CERN Large Hadron Collider (LHC) is provided

  8. TRIPOLI-4: Monte Carlo transport code functionalities and applications

    International Nuclear Information System (INIS)

    Both, J.P.; Lee, Y.K.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.

    2003-01-01

    Tripoli-4 is a three dimensional calculations code using the Monte Carlo method to simulate the transport of neutrons, photons, electrons and positrons. This code is used in four application fields: the protection studies, the criticality studies, the core studies and the instrumentation studies. Geometry, cross sections, description of sources, principle. (N.C.)

  9. MC++: A parallel, portable, Monte Carlo neutron transport code in C++

    International Nuclear Information System (INIS)

    Lee, S.R.; Cummings, J.C.; Nolen, S.D.

    1997-01-01

    MC++ is an implicit multi-group Monte Carlo neutron transport code written in C++ and based on the Parallel Object-Oriented Methods and Applications (POOMA) class library. MC++ runs in parallel on and is portable to a wide variety of platforms, including MPPs, SMPs, and clusters of UNIX workstations. MC++ is being developed to provide transport capabilities to the Accelerated Strategic Computing Initiative (ASCI). It is also intended to form the basis of the first transport physics framework (TPF), which is a C++ class library containing appropriate abstractions, objects, and methods for the particle transport problem. The transport problem is briefly described, as well as the current status and algorithms in MC++ for solving the transport equation. The alpha version of the POOMA class library is also discussed, along with the implementation of the transport solution algorithms using POOMA. Finally, a simple test problem is defined and performance and physics results from this problem are discussed on a variety of platforms

  10. Multilevel and quasi-Monte Carlo methods for uncertainty quantification in particle travel times through random heterogeneous porous media

    Science.gov (United States)

    Crevillén-García, D.; Power, H.

    2017-08-01

    In this study, we apply four Monte Carlo simulation methods, namely, Monte Carlo, quasi-Monte Carlo, multilevel Monte Carlo and multilevel quasi-Monte Carlo to the problem of uncertainty quantification in the estimation of the average travel time during the transport of particles through random heterogeneous porous media. We apply the four methodologies to a model problem where the only input parameter, the hydraulic conductivity, is modelled as a log-Gaussian random field by using direct Karhunen-Loéve decompositions. The random terms in such expansions represent the coefficients in the equations. Numerical calculations demonstrating the effectiveness of each of the methods are presented. A comparison of the computational cost incurred by each of the methods for three different tolerances is provided. The accuracy of the approaches is quantified via the mean square error.

  11. Multilevel and quasi-Monte Carlo methods for uncertainty quantification in particle travel times through random heterogeneous porous media.

    Science.gov (United States)

    Crevillén-García, D; Power, H

    2017-08-01

    In this study, we apply four Monte Carlo simulation methods, namely, Monte Carlo, quasi-Monte Carlo, multilevel Monte Carlo and multilevel quasi-Monte Carlo to the problem of uncertainty quantification in the estimation of the average travel time during the transport of particles through random heterogeneous porous media. We apply the four methodologies to a model problem where the only input parameter, the hydraulic conductivity, is modelled as a log-Gaussian random field by using direct Karhunen-Loéve decompositions. The random terms in such expansions represent the coefficients in the equations. Numerical calculations demonstrating the effectiveness of each of the methods are presented. A comparison of the computational cost incurred by each of the methods for three different tolerances is provided. The accuracy of the approaches is quantified via the mean square error.

  12. TRIPOLI-4: Monte Carlo transport code functionalities and applications; TRIPOLI-4: code de transport Monte Carlo fonctionnalites et applications

    Energy Technology Data Exchange (ETDEWEB)

    Both, J P; Lee, Y K; Mazzolo, A; Peneliau, Y; Petit, O; Roesslinger, B [CEA Saclay, Dir. de l' Energie Nucleaire (DEN), Service d' Etudes de Reacteurs et de Modelisation Avancee, 91 - Gif sur Yvette (France)

    2003-07-01

    Tripoli-4 is a three dimensional calculations code using the Monte Carlo method to simulate the transport of neutrons, photons, electrons and positrons. This code is used in four application fields: the protection studies, the criticality studies, the core studies and the instrumentation studies. Geometry, cross sections, description of sources, principle. (N.C.)

  13. Transport calculation of medium-energy protons and neutrons by Monte Carlo method

    International Nuclear Information System (INIS)

    Ban, Syuuichi; Hirayama, Hideo; Katoh, Kazuaki.

    1978-09-01

    A Monte Carlo transport code, ARIES, has been developed for protons and neutrons at medium energy (25 -- 500 MeV). Nuclear data provided by R.G. Alsmiller, Jr. were used for the calculation. To simulate the cascade development in the medium, each generation was represented by a single weighted particle and an average number of emitted particles was used as the weight. Neutron fluxes were stored by the collisions density method. The cutoff energy was set to 25 MeV. Neutrons below the cutoff were stored to be used as the source for the low energy neutron transport calculation upon the discrete ordinates method. Then transport calculations were performed for both low energy neutrons (thermal -- 25 MeV) and secondary gamma-rays. Energy spectra of emitted neutrons were calculated and compared with those of published experimental and calculated results. The agreement was good for the incident particles of energy between 100 and 500 MeV. (author)

  14. MONTE CARLO SIMULATION MODEL OF ENERGETIC PROTON TRANSPORT THROUGH SELF-GENERATED ALFVEN WAVES

    Energy Technology Data Exchange (ETDEWEB)

    Afanasiev, A.; Vainio, R., E-mail: alexandr.afanasiev@helsinki.fi [Department of Physics, University of Helsinki (Finland)

    2013-08-15

    A new Monte Carlo simulation model for the transport of energetic protons through self-generated Alfven waves is presented. The key point of the model is that, unlike the previous ones, it employs the full form (i.e., includes the dependence on the pitch-angle cosine) of the resonance condition governing the scattering of particles off Alfven waves-the process that approximates the wave-particle interactions in the framework of quasilinear theory. This allows us to model the wave-particle interactions in weak turbulence more adequately, in particular, to implement anisotropic particle scattering instead of isotropic scattering, which the previous Monte Carlo models were based on. The developed model is applied to study the transport of flare-accelerated protons in an open magnetic flux tube. Simulation results for the transport of monoenergetic protons through the spectrum of Alfven waves reveal that the anisotropic scattering leads to spatially more distributed wave growth than isotropic scattering. This result can have important implications for diffusive shock acceleration, e.g., affect the scattering mean free path of the accelerated particles in and the size of the foreshock region.

  15. Charged-particle calculations using Boltzmann transport methods

    International Nuclear Information System (INIS)

    Hoffman, T.J.; Dodds, H.L. Jr.; Robinson, M.T.; Holmes, D.K.

    1981-01-01

    Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of charged particle range distributions, reflection coefficients, and sputtering yields. The Boltzmann transport approach can be implemented, with minor changes, in standard neutral particle computer codes. With the multigroup discrete ordinates code, ANISN, determination of ion and target atom distributions as functions of position, energy, and direction can be obtained without the stochastic error associated with atomistic computer codes such as MARLOWE and TRIM. With the multigroup Monte Carlo code, MORSE, charged particle effects can be obtained for problems associated with very complex geometries. Results are presented for several charged particle problems. Good agreement is obtained between quantities calculated with the multigroup approach and those obtained experimentally or by atomistic computer codes

  16. Importance estimation in Monte Carlo modelling of neutron and photon transport

    International Nuclear Information System (INIS)

    Mickael, M.W.

    1992-01-01

    The estimation of neutron and photon importance in a three-dimensional geometry is achieved using a coupled Monte Carlo and diffusion theory calculation. The parameters required for the solution of the multigroup adjoint diffusion equation are estimated from an analog Monte Carlo simulation of the system under investigation. The solution of the adjoint diffusion equation is then used as an estimate of the particle importance in the actual simulation. This approach provides an automated and efficient variance reduction method for Monte Carlo simulations. The technique has been successfully applied to Monte Carlo simulation of neutron and coupled neutron-photon transport in the nuclear well-logging field. The results show that the importance maps obtained in a few minutes of computer time using this technique are in good agreement with Monte Carlo generated importance maps that require prohibitive computing times. The application of this method to Monte Carlo modelling of the response of neutron porosity and pulsed neutron instruments has resulted in major reductions in computation time. (Author)

  17. User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code; Notice d'utilisation du code Tripoli-4, version 4.3: code de transport de particules par la methode de Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B

    2003-07-01

    This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k{sub eff} (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

  18. Transport methods: general. 1. The Analytical Monte Carlo Method for Radiation Transport Calculations

    International Nuclear Information System (INIS)

    Martin, William R.; Brown, Forrest B.

    2001-01-01

    We present an alternative Monte Carlo method for solving the coupled equations of radiation transport and material energy. This method is based on incorporating the analytical solution to the material energy equation directly into the Monte Carlo simulation for the radiation intensity. This method, which we call the Analytical Monte Carlo (AMC) method, differs from the well known Implicit Monte Carlo (IMC) method of Fleck and Cummings because there is no discretization of the material energy equation since it is solved as a by-product of the Monte Carlo simulation of the transport equation. Our method also differs from the method recently proposed by Ahrens and Larsen since they use Monte Carlo to solve both equations, while we are solving only the radiation transport equation with Monte Carlo, albeit with effective sources and cross sections to represent the emission sources. Our method bears some similarity to a method developed and implemented by Carter and Forest nearly three decades ago, but there are substantive differences. We have implemented our method in a simple zero-dimensional Monte Carlo code to test the feasibility of the method, and the preliminary results are very promising, justifying further extension to more realistic geometries. (authors)

  19. A New Monte Carlo Neutron Transport Code at UNIST

    International Nuclear Information System (INIS)

    Lee, Hyunsuk; Kong, Chidong; Lee, Deokjung

    2014-01-01

    Monte Carlo neutron transport code named MCS is under development at UNIST for the advanced reactor design and research purpose. This MC code can be used for fixed source calculation and criticality calculation. Continuous energy neutron cross section data and multi-group cross section data can be used for the MC calculation. This paper presents the overview of developed MC code and its calculation results. The real time fixed source calculation ability is also tested in this paper. The calculation results show good agreement with commercial code and experiment. A new Monte Carlo neutron transport code is being developed at UNIST. The MC codes are tested with several benchmark problems: ICSBEP, VENUS-2, and Hoogenboom-Martin benchmark. These benchmarks covers pin geometry to 3-dimensional whole core, and results shows good agreement with reference results

  20. Monte Carlo parametric importance sampling with particle tracks scaling

    International Nuclear Information System (INIS)

    Ragheb, M.M.H.

    1981-01-01

    A method for Monte Carlo importance sampling with parametric dependence is proposed. It depends upon obtaining over a single stage the overall functional dependence of the variance on the importance function parameter over a broad range of its values. Results corresponding to minimum variance are adopted and others rejected. The proposed method is applied to the finite slab penetration problem. When the exponential transformation is used, our method involves scaling of the generated particle tracks, and is a new application of Morton's method of similar trajectories. The method constitutes a generalization of Spanier's multistage importance sampling method, obtained by proper weighting over a single stage the curves he obtains over several stages, and preserves the statistical correlations between histories. It represents an extension of a theory by Frolov and Chentsov on Monte Carlo calculations of smooth curves to surfaces and to importance sampling calculations. By the proposed method, it seems possible to systematically arrive at minimum variance results and to avoid the infinite variances and effective biases sometimes observed in this type of calculation. (orig.) [de

  1. Automatic modeling for the Monte Carlo transport code Geant4

    International Nuclear Information System (INIS)

    Nie Fanzhi; Hu Liqin; Wang Guozhong; Wang Dianxi; Wu Yican; Wang Dong; Long Pengcheng; FDS Team

    2015-01-01

    Geant4 is a widely used Monte Carlo transport simulation package. Its geometry models could be described in Geometry Description Markup Language (GDML), but it is time-consuming and error-prone to describe the geometry models manually. This study implemented the conversion between computer-aided design (CAD) geometry models and GDML models. This method has been Studied based on Multi-Physics Coupling Analysis Modeling Program (MCAM). The tests, including FDS-Ⅱ model, demonstrated its accuracy and feasibility. (authors)

  2. Verification of Monte Carlo transport codes by activation experiments

    OpenAIRE

    Chetvertkova, Vera

    2013-01-01

    With the increasing energies and intensities of heavy-ion accelerator facilities, the problem of an excessive activation of the accelerator components caused by beam losses becomes more and more important. Numerical experiments using Monte Carlo transport codes are performed in order to assess the levels of activation. The heavy-ion versions of the codes were released approximately a decade ago, therefore the verification is needed to be sure that they give reasonable results. Present work is...

  3. Monte Carlo investigation of anomalous transport in presence of a discontinuity and of an advection field

    Science.gov (United States)

    Marseguerra, M.; Zoia, A.

    2007-04-01

    Anomalous diffusion has recently turned out to be almost ubiquitous in transport problems. When the physical properties of the medium where the transport process takes place are stationary and constant at each spatial location, anomalous transport has been successfully analysed within the Continuous Time Random Walk (CTRW) model. In this paper, within a Monte Carlo approach to CTRW, we focus on the particle transport through two regions characterized by different physical properties, in presence of an external driving action constituted by an additional advective field, modelled within both the Galilei invariant and Galilei variant schemes. Particular attention is paid to the interplay between the distributions of space and time across the discontinuity. The resident concentration and the flux of the particles are finally evaluated and it is shown that at the interface between the two regions the flux is continuous as required by mass conservation, while the concentration may reveal a neat discontinuity. This result could open the route to the Monte Carlo investigation of the effectiveness of a physical discontinuity acting as a filter on particle concentration.

  4. A numerical analysis of antithetic variates in Monte Carlo radiation transport with geometrical surface splitting

    International Nuclear Information System (INIS)

    Sarkar, P.K.; Prasad, M.A.

    1989-01-01

    A numerical study for effective implementation of the antithetic variates technique with geometric splitting/Russian roulette in Monte Carlo radiation transport calculations is presented. The study is based on the theory of Monte Carlo errors where a set of coupled integral equations are solved for the first and second moments of the score and for the expected number of flights per particle history. Numerical results are obtained for particle transmission through an infinite homogeneous slab shield composed of an isotropically scattering medium. Two types of antithetic transformations are considered. The results indicate that the antithetic transformations always lead to reduction in variance and increase in efficiency provided optimal antithetic parameters are chosen. A substantial gain in efficiency is obtained by incorporating antithetic transformations in rule of thumb splitting. The advantage gained for thick slabs (∼20 mfp) with low scattering probability (0.1-0.5) is attractively large . (author). 27 refs., 9 tabs

  5. The use of Monte Carlo radiation transport codes in radiation physics and dosimetry

    CERN Multimedia

    CERN. Geneva; Ferrari, Alfredo; Silari, Marco

    2006-01-01

    Transport and interaction of electromagnetic radiation Interaction models and simulation schemes implemented in modern Monte Carlo codes for the simulation of coupled electron-photon transport will be briefly reviewed. In these codes, photon transport is simulated by using the detailed scheme, i.e., interaction by interaction. Detailed simulation is easy to implement, and the reliability of the results is only limited by the accuracy of the adopted cross sections. Simulations of electron and positron transport are more difficult, because these particles undergo a large number of interactions in the course of their slowing down. Different schemes for simulating electron transport will be discussed. Condensed algorithms, which rely on multiple-scattering theories, are comparatively fast, but less accurate than mixed algorithms, in which hard interactions (with energy loss or angular deflection larger than certain cut-off values) are simulated individually. The reliability, and limitations, of electron-interacti...

  6. Verification and Validation of Monte Carlo n-Particle Code 6 (MCNP6) with Neutron Protection Factor Measurements of an Iron Box

    Science.gov (United States)

    2014-03-27

    Vehicle Code System (VCS), the Monte Carlo Adjoint SHielding (MASH), and the Monte Carlo n- Particle ( MCNP ) code. Of the three, the oldest and still most...widely utilized radiation transport code is MCNP . First created at Los Alamos National Laboratory (LANL) in 1957, the code simulated neutral...particle types, and previous versions of MCNP were repeatedly validated using both simple and complex 10 geometries [12, 13]. Much greater discussion and

  7. Monte Carlo perturbation theory in neutron transport calculations

    International Nuclear Information System (INIS)

    Hall, M.C.G.

    1980-01-01

    The need to obtain sensitivities in complicated geometrical configurations has resulted in the development of Monte Carlo sensitivity estimation. A new method has been developed to calculate energy-dependent sensitivities of any number of responses in a single Monte Carlo calculation with a very small time penalty. This estimation typically increases the tracking time per source particle by about 30%. The method of estimation is explained. Sensitivities obtained are compared with those calculated by discrete ordinates methods. Further theoretical developments, such as second-order perturbation theory and application to k/sub eff/ calculations, are discussed. The application of the method to uncertainty analysis and to the analysis of benchmark experiments is illustrated. 5 figures

  8. Efficiencies of dynamic Monte Carlo algorithms for off-lattice particle systems with a single impurity

    KAUST Repository

    Novotny, M.A.; Watanabe, Hiroshi; Ito, Nobuyasu

    2010-01-01

    The efficiency of dynamic Monte Carlo algorithms for off-lattice systems composed of particles is studied for the case of a single impurity particle. The theoretical efficiencies of the rejection-free method and of the Monte Carlo with Absorbing

  9. Efficiencies of dynamic Monte Carlo algorithms for off-lattice particle systems with a single impurity

    KAUST Repository

    Novotny, M.A.

    2010-02-01

    The efficiency of dynamic Monte Carlo algorithms for off-lattice systems composed of particles is studied for the case of a single impurity particle. The theoretical efficiencies of the rejection-free method and of the Monte Carlo with Absorbing Markov Chains method are given. Simulation results are presented to confirm the theoretical efficiencies. © 2010.

  10. grmonty: A MONTE CARLO CODE FOR RELATIVISTIC RADIATIVE TRANSPORT

    International Nuclear Information System (INIS)

    Dolence, Joshua C.; Gammie, Charles F.; Leung, Po Kin; Moscibrodzka, Monika

    2009-01-01

    We describe a Monte Carlo radiative transport code intended for calculating spectra of hot, optically thin plasmas in full general relativity. The version we describe here is designed to model hot accretion flows in the Kerr metric and therefore incorporates synchrotron emission and absorption, and Compton scattering. The code can be readily generalized, however, to account for other radiative processes and an arbitrary spacetime. We describe a suite of test problems, and demonstrate the expected N -1/2 convergence rate, where N is the number of Monte Carlo samples. Finally, we illustrate the capabilities of the code with a model calculation, a spectrum of the slowly accreting black hole Sgr A* based on data provided by a numerical general relativistic MHD model of the accreting plasma.

  11. High-energy particle Monte Carlo at Los Alamos

    International Nuclear Information System (INIS)

    Prael, R.E.

    1985-01-01

    A major computational effort at Los Alamos has been the development of a code system based on the HETC code for the transport of nucleons, pions, and muons. The Los Alamos National Laboratory version of HETC utilizes MCNP geometry and interfaces with MCNP for the transport of neutrons below 20 MeV and photons at any energy. A major recent effort has been the development of the PHT code for treating the gamma cascade in excited nuclei (the residual nuclei from an HETC calculation) by the Monte Carlo method to generate a photon source for MCNP. The HETC/MCNP code system has been extensively used for design studies of accelerator targets and shielding, including the design of LAMPF-II. It is extensively used for the design and analysis of accelerator experiments. Los Alamos National Laboratory has been an active member of the International Collaboration on Advanced Neutron Sources; as such we engage in shared code development and computational efforts. In the past few years, additional effort has been devoted to the development of a Chen-model intranuclear cascade code (INCA1) featuring a cluster model for the nucleus and deuteron pickup reactions. Concurrently, the INCA2 code for the breakup of light, excited nuclei using the Fermi breakup model has been developed. Together, they have been used for the calculation of neutron and proton cross sections in the energy ranges appropriate to medical accelerators, and for the computation of tissue kerma factors

  12. Condensed history Monte Carlo methods for photon transport problems

    International Nuclear Information System (INIS)

    Bhan, Katherine; Spanier, Jerome

    2007-01-01

    We study methods for accelerating Monte Carlo simulations that retain most of the accuracy of conventional Monte Carlo algorithms. These methods - called Condensed History (CH) methods - have been very successfully used to model the transport of ionizing radiation in turbid systems. Our primary objective is to determine whether or not such methods might apply equally well to the transport of photons in biological tissue. In an attempt to unify the derivations, we invoke results obtained first by Lewis, Goudsmit and Saunderson and later improved by Larsen and Tolar. We outline how two of the most promising of the CH models - one based on satisfying certain similarity relations and the second making use of a scattering phase function that permits only discrete directional changes - can be developed using these approaches. The main idea is to exploit the connection between the space-angle moments of the radiance and the angular moments of the scattering phase function. We compare the results obtained when the two CH models studied are used to simulate an idealized tissue transport problem. The numerical results support our findings based on the theoretical derivations and suggest that CH models should play a useful role in modeling light-tissue interactions

  13. A portable, parallel, object-oriented Monte Carlo neutron transport code in C++

    International Nuclear Information System (INIS)

    Lee, S.R.; Cummings, J.C.; Nolen, S.D.

    1997-01-01

    We have developed a multi-group Monte Carlo neutron transport code using C++ and the Parallel Object-Oriented Methods and Applications (POOMA) class library. This transport code, called MC++, currently computes k and α-eigenvalues and is portable to and runs parallel on a wide variety of platforms, including MPPs, clustered SMPs, and individual workstations. It contains appropriate classes and abstractions for particle transport and, through the use of POOMA, for portable parallelism. Current capabilities of MC++ are discussed, along with physics and performance results on a variety of hardware, including all Accelerated Strategic Computing Initiative (ASCI) hardware. Current parallel performance indicates the ability to compute α-eigenvalues in seconds to minutes rather than hours to days. Future plans and the implementation of a general transport physics framework are also discussed

  14. Adaptively Learning an Importance Function Using Transport Constrained Monte Carlo

    International Nuclear Information System (INIS)

    Booth, T.E.

    1998-01-01

    It is well known that a Monte Carlo estimate can be obtained with zero-variance if an exact importance function for the estimate is known. There are many ways that one might iteratively seek to obtain an ever more exact importance function. This paper describes a method that has obtained ever more exact importance functions that empirically produce an error that is dropping exponentially with computer time. The method described herein constrains the importance function to satisfy the (adjoint) Boltzmann transport equation. This constraint is provided by using the known form of the solution, usually referred to as the Case eigenfunction solution

  15. Reliability analysis of neutron transport simulation using Monte Carlo method

    International Nuclear Information System (INIS)

    Souza, Bismarck A. de; Borges, Jose C.

    1995-01-01

    This work presents a statistical and reliability analysis covering data obtained by computer simulation of neutron transport process, using the Monte Carlo method. A general description of the method and its applications is presented. Several simulations, corresponding to slowing down and shielding problems have been accomplished. The influence of the physical dimensions of the materials and of the sample size on the reliability level of results was investigated. The objective was to optimize the sample size, in order to obtain reliable results, optimizing computation time. (author). 5 refs, 8 figs

  16. A Study on Efficiency Improvement of the Hybrid Monte Carlo/Deterministic Method for Global Transport Problems

    International Nuclear Information System (INIS)

    Kim, Jong Woo; Woo, Myeong Hyeon; Kim, Jae Hyun; Kim, Do Hyun; Shin, Chang Ho; Kim, Jong Kyung

    2017-01-01

    In this study hybrid Monte Carlo/Deterministic method is explained for radiation transport analysis in global system. FW-CADIS methodology construct the weight window parameter and it useful at most global MC calculation. However, Due to the assumption that a particle is scored at a tally, less particles are transported to the periphery of mesh tallies. For compensation this space-dependency, we modified the module in the ADVANTG code to add the proposed method. We solved the simple test problem for comparing with result from FW-CADIS methodology, it was confirmed that a uniform statistical error was secured as intended. In the future, it will be added more practical problems. It might be useful to perform radiation transport analysis using the Hybrid Monte Carlo/Deterministic method in global transport problems.

  17. A computationally efficient moment-preserving Monte Carlo electron transport method with implementation in Geant4

    Energy Technology Data Exchange (ETDEWEB)

    Dixon, D.A., E-mail: ddixon@lanl.gov [Los Alamos National Laboratory, P.O. Box 1663, MS P365, Los Alamos, NM 87545 (United States); Prinja, A.K., E-mail: prinja@unm.edu [Department of Nuclear Engineering, MSC01 1120, 1 University of New Mexico, Albuquerque, NM 87131-0001 (United States); Franke, B.C., E-mail: bcfrank@sandia.gov [Sandia National Laboratories, Albuquerque, NM 87123 (United States)

    2015-09-15

    This paper presents the theoretical development and numerical demonstration of a moment-preserving Monte Carlo electron transport method. Foremost, a full implementation of the moment-preserving (MP) method within the Geant4 particle simulation toolkit is demonstrated. Beyond implementation details, it is shown that the MP method is a viable alternative to the condensed history (CH) method for inclusion in current and future generation transport codes through demonstration of the key features of the method including: systematically controllable accuracy, computational efficiency, mathematical robustness, and versatility. A wide variety of results common to electron transport are presented illustrating the key features of the MP method. In particular, it is possible to achieve accuracy that is statistically indistinguishable from analog Monte Carlo, while remaining up to three orders of magnitude more efficient than analog Monte Carlo simulations. Finally, it is shown that the MP method can be generalized to any applicable analog scattering DCS model by extending previous work on the MP method beyond analytical DCSs to the partial-wave (PW) elastic tabulated DCS data.

  18. Development of particle and heavy ion transport code system

    International Nuclear Information System (INIS)

    Niita, Koji

    2004-01-01

    Particle and heavy ion transport code system (PHITS) is 3 dimension general purpose Monte Carlo simulation codes for description of transport and reaction of particle and heavy ion in materials. It is developed on the basis of NMTC/JAM for design and safety of J-PARC. What is PHITS, it's physical process, physical models and development process of PHITC code are described. For examples of application, evaluation of neutron optics, cancer treatment by heavy particle ray and cosmic radiation are stated. JAM and JQMD model are used as the physical model. Neutron motion in six polar magnetic field and gravitational field, PHITC simulation of trace of C 12 beam and secondary neutron track of small model of cancer treatment device in HIMAC and neutron flux in Space Shuttle are explained. (S.Y.)

  19. Monte-Carlo simulation of complex vapor-transport systems for RIB applications

    International Nuclear Information System (INIS)

    Zhang, Y.; Alton, G.D.

    2005-01-01

    In order to minimize decay losses of short-lived radioactive species at ISOL based RIB facilities, effusive-flow particle transit times between target and ion source must be as short as practically achievable. A Monte-Carlo code has been developed for simulating the effusive-flow of neutral particles through vapor-transport systems independent of materials of construction. The code provides average distance traveled and time information associated with the transit of individual particles through a system. It offers a cost effective and accurate means for arriving at vapor-transport system designs. In this report, the code will be described and results obtained by its use in evaluating several prototype vapor-transport systems using specular reflection, cosine and isotropic particle re-emission about the normal to the surface models following adsorption. Simulation results obtained with an isotropic distribution are in close agreement with experimental measurements of the properties of prototype vapor-transport systems fabricated at the Holifield Radioactive Ion Beam Facility

  20. Monte Carlo modeling of transport in PbSe nanocrystal films

    Energy Technology Data Exchange (ETDEWEB)

    Carbone, I., E-mail: icarbone@ucsc.edu; Carter, S. A. [University of California, Santa Cruz, California 95060 (United States); Zimanyi, G. T. [University of California, Davis, California 95616 (United States)

    2013-11-21

    A Monte Carlo hopping model was developed to simulate electron and hole transport in nanocrystalline PbSe films. Transport is carried out as a series of thermally activated hopping events between neighboring sites on a cubic lattice. Each site, representing an individual nanocrystal, is assigned a size-dependent electronic structure, and the effects of particle size, charging, interparticle coupling, and energetic disorder on electron and hole mobilities were investigated. Results of simulated field-effect measurements confirm that electron mobilities and conductivities at constant carrier densities increase with particle diameter by an order of magnitude up to 5 nm and begin to decrease above 6 nm. We find that as particle size increases, fewer hops are required to traverse the same distance and that site energy disorder significantly inhibits transport in films composed of smaller nanoparticles. The dip in mobilities and conductivities at larger particle sizes can be explained by a decrease in tunneling amplitudes and by charging penalties that are incurred more frequently when carriers are confined to fewer, larger nanoparticles. Using a nearly identical set of parameter values as the electron simulations, hole mobility simulations confirm measurements that increase monotonically with particle size over two orders of magnitude.

  1. Verification of Monte Carlo transport codes by activation experiments

    Energy Technology Data Exchange (ETDEWEB)

    Chetvertkova, Vera

    2012-12-18

    With the increasing energies and intensities of heavy-ion accelerator facilities, the problem of an excessive activation of the accelerator components caused by beam losses becomes more and more important. Numerical experiments using Monte Carlo transport codes are performed in order to assess the levels of activation. The heavy-ion versions of the codes were released approximately a decade ago, therefore the verification is needed to be sure that they give reasonable results. Present work is focused on obtaining the experimental data on activation of the targets by heavy-ion beams. Several experiments were performed at GSI Helmholtzzentrum fuer Schwerionenforschung. The interaction of nitrogen, argon and uranium beams with aluminum targets, as well as interaction of nitrogen and argon beams with copper targets was studied. After the irradiation of the targets by different ion beams from the SIS18 synchrotron at GSI, the γ-spectroscopy analysis was done: the γ-spectra of the residual activity were measured, the radioactive nuclides were identified, their amount and depth distribution were detected. The obtained experimental results were compared with the results of the Monte Carlo simulations using FLUKA, MARS and SHIELD. The discrepancies and agreements between experiment and simulations are pointed out. The origin of discrepancies is discussed. Obtained results allow for a better verification of the Monte Carlo transport codes, and also provide information for their further development. The necessity of the activation studies for accelerator applications is discussed. The limits of applicability of the heavy-ion beam-loss criteria were studied using the FLUKA code. FLUKA-simulations were done to determine the most preferable from the radiation protection point of view materials for use in accelerator components.

  2. KAMCCO, a reactor physics Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Arnecke, G.; Borgwaldt, H.; Brandl, V.; Lalovic, M.

    1976-06-01

    KAMCCO is a 3-dimensional reactor Monte Carlo code for fast neutron physics problems. Two options are available for the solution of 1) the inhomogeneous time-dependent neutron transport equation (census time scheme), and 2) the homogeneous static neutron transport equation (generation cycle scheme). The user defines the desired output, e.g. estimates of reaction rates or neutron flux integrated over specified volumes in phase space and time intervals. Such primary quantities can be arbitrarily combined, also ratios of these quantities can be estimated with their errors. The Monte Carlo techniques are mostly analogue (exceptions: Importance sampling for collision processes, ELP/MELP, Russian roulette and splitting). Estimates are obtained from the collision and track length estimators. Elastic scattering takes into account first order anisotropy in the center of mass system. Inelastic scattering is processed via the evaporation model or via the excitation of discrete levels. For the calculation of cross sections, the energy is treated as a continuous variable. They are computed by a) linear interpolation, b) from optionally Doppler broadened single level Breit-Wigner resonances or c) from probability tables (in the region of statistically distributed resonances). (orig.) [de

  3. A punctual flux estimator and reactions rates optimization in neutral particles transport calculus by the Monte Carlo method; Mise au point d'un estimateur ponctuel du flux et des taux de reactions dans les calculs de transport de particules neutres par la methode de monte carlo

    Energy Technology Data Exchange (ETDEWEB)

    Authier, N

    1998-12-01

    One of the questions asked in radiation shielding problems is the estimation of the radiation level in particular to determine accessibility of working persons in controlled area (nuclear power plants, nuclear fuel reprocessing plants) or to study the dose gradients encountered in material (iron nuclear vessel, medical therapy, electronics in satellite). The flux and reaction rate estimators used in Monte Carlo codes give average values in volumes or on surfaces of the geometrical description of the system. But in certain configurations, the knowledge of punctual deposited energy and dose estimates are necessary. The Monte Carlo estimate of the flux at a point of interest is a calculus which presents an unbounded variance. The central limit theorem cannot be applied thus no easy confidencelevel may be calculated. The convergence rate is then very poor. We propose in this study a new solution for the photon flux at a point estimator. The method is based on the 'once more collided flux estimator' developed earlier for neutron calculations. It solves the problem of the unbounded variance and do not add any bias to the estimation. We show however that our new sampling schemes specially developed to treat the anisotropy of the photon coherent scattering is necessary for a good and regular behavior of the estimator. This developments integrated in the TRIPOLI-4 Monte Carlo code add the possibility of an unbiased punctual estimate on media interfaces. (author)

  4. A punctual flux estimator and reactions rates optimization in neutral particles transport calculus by the Monte Carlo method; Mise au point d'un estimateur ponctuel du flux et des taux de reactions dans les calculs de transport de particules neutres par la methode de monte carlo

    Energy Technology Data Exchange (ETDEWEB)

    Authier, N

    1998-12-01

    One of the questions asked in radiation shielding problems is the estimation of the radiation level in particular to determine accessibility of working persons in controlled area (nuclear power plants, nuclear fuel reprocessing plants) or to study the dose gradients encountered in material (iron nuclear vessel, medical therapy, electronics in satellite). The flux and reaction rate estimators used in Monte Carlo codes give average values in volumes or on surfaces of the geometrical description of the system. But in certain configurations, the knowledge of punctual deposited energy and dose estimates are necessary. The Monte Carlo estimate of the flux at a point of interest is a calculus which presents an unbounded variance. The central limit theorem cannot be applied thus no easy confidencelevel may be calculated. The convergence rate is then very poor. We propose in this study a new solution for the photon flux at a point estimator. The method is based on the 'once more collided flux estimator' developed earlier for neutron calculations. It solves the problem of the unbounded variance and do not add any bias to the estimation. We show however that our new sampling schemes specially developed to treat the anisotropy of the photon coherent scattering is necessary for a good and regular behavior of the estimator. This developments integrated in the TRIPOLI-4 Monte Carlo code add the possibility of an unbiased punctual estimate on media interfaces. (author)

  5. 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

  6. ITS - The integrated TIGER series of coupled electron/photon Monte Carlo transport codes

    International Nuclear Information System (INIS)

    Halbleib, J.A.; Mehlhorn, T.A.

    1985-01-01

    The TIGER series of time-independent coupled electron/photon Monte Carlo transport codes is a group of multimaterial, multidimensional codes designed to provide a state-of-the-art description of the production and transport of the electron/photon cascade. The codes follow both electrons and photons from 1.0 GeV down to 1.0 keV, and the user has the option of combining the collisional transport with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence. Source particles can be either electrons or photons. The most important output data are (a) charge and energy deposition profiles, (b) integral and differential escape coefficients for both electrons and photons, (c) differential electron and photon flux, and (d) pulse-height distributions for selected regions of the problem geometry. The base codes of the series differ from one another primarily in their dimensionality and geometric modeling. They include (a) a one-dimensional multilayer code, (b) a code that describes the transport in two-dimensional axisymmetric cylindrical material geometries with a fully three-dimensional description of particle trajectories, and (c) a general three-dimensional transport code which employs a combinatorial geometry scheme. These base codes were designed primarily for describing radiation transport for those situations in which the detailed atomic structure of the transport medium is not important. For some applications, it is desirable to have a more detailed model of the low energy transport. The system includes three additional codes that contain a more elaborate ionization/relaxation model than the base codes. Finally, the system includes two codes that combine the collisional transport of the multidimensional base codes with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence

  7. Particle transport in porous media

    Science.gov (United States)

    Corapcioglu, M. Yavuz; Hunt, James R.

    The migration and capture of particles (such as colloidal materials and microorganisms) through porous media occur in fields as diversified as water and wastewater treatment, well drilling, and various liquid-solid separation processes. In liquid waste disposal projects, suspended solids can cause the injection well to become clogged, and groundwater quality can be endangered by suspended clay and silt particles because of migration to the formation adjacent to the well bore. In addition to reducing the permeability of the soil, mobile particles can carry groundwater contaminants adsorbed onto their surfaces. Furthermore, as in the case of contamination from septic tanks, the particles themselves may be pathogens, i.e., bacteria and viruses.

  8. Advanced Monte Carlo methods for thermal radiation transport

    Science.gov (United States)

    Wollaber, Allan B.

    During the past 35 years, the Implicit Monte Carlo (IMC) method proposed by Fleck and Cummings has been the standard Monte Carlo approach to solving the thermal radiative transfer (TRT) equations. However, the IMC equations are known to have accuracy limitations that can produce unphysical solutions. In this thesis, we explicitly provide the IMC equations with a Monte Carlo interpretation by including particle weight as one of its arguments. We also develop and test a stability theory for the 1-D, gray IMC equations applied to a nonlinear problem. We demonstrate that the worst case occurs for 0-D problems, and we extend the results to a stability algorithm that may be used for general linearizations of the TRT equations. We derive gray, Quasidiffusion equations that may be deterministically solved in conjunction with IMC to obtain an inexpensive, accurate estimate of the temperature at the end of the time step. We then define an average temperature T* to evaluate the temperature-dependent problem data in IMC, and we demonstrate that using T* is more accurate than using the (traditional) beginning-of-time-step temperature. We also propose an accuracy enhancement to the IMC equations: the use of a time-dependent "Fleck factor". This Fleck factor can be considered an automatic tuning of the traditionally defined user parameter alpha, which generally provides more accurate solutions at an increased cost relative to traditional IMC. We also introduce a global weight window that is proportional to the forward scalar intensity calculated by the Quasidiffusion method. This weight window improves the efficiency of the IMC calculation while conserving energy. All of the proposed enhancements are tested in 1-D gray and frequency-dependent problems. These enhancements do not unconditionally eliminate the unphysical behavior that can be seen in the IMC calculations. However, for fixed spatial and temporal grids, they suppress them and clearly work to make the solution more

  9. Limits on the efficiency of event-based algorithms for Monte Carlo neutron transport

    Directory of Open Access Journals (Sweden)

    Paul K. Romano

    2017-09-01

    Full Text Available The traditional form of parallelism in Monte Carlo particle transport simulations, wherein each individual particle history is considered a unit of work, does not lend itself well to data-level parallelism. Event-based algorithms, which were originally used for simulations on vector processors, may offer a path toward better utilizing data-level parallelism in modern computer architectures. In this study, a simple model is developed for estimating the efficiency of the event-based particle transport algorithm under two sets of assumptions. Data collected from simulations of four reactor problems using OpenMC was then used in conjunction with the models to calculate the speedup due to vectorization as a function of the size of the particle bank and the vector width. When each event type is assumed to have constant execution time, the achievable speedup is directly related to the particle bank size. We observed that the bank size generally needs to be at least 20 times greater than vector size to achieve vector efficiency greater than 90%. When the execution times for events are allowed to vary, the vector speedup is also limited by differences in the execution time for events being carried out in a single event-iteration.

  10. Non deterministic methods for charged particle transport

    International Nuclear Information System (INIS)

    Besnard, D.C.; Buresi, E.; Hermeline, F.; Wagon, F.

    1985-04-01

    The coupling of Monte-Carlo methods for solving Fokker Planck equation with ICF inertial confinement fusion codes requires them to be economical and to preserve gross conservation properties. Besides, the presence in FPE Fokker-Planck equation of diffusion terms due to collisions between test particles and the background plasma challenges standard M.C. (Monte-Carlo) techniques if this phenomenon is dominant. We address these problems through the use of a fixed mesh in phase space which allows us to handle highly variable sources, avoiding any Russian Roulette for lowering the size of the sample. Also on this mesh are solved diffusion equations obtained from a splitting of FPE. Any non linear diffusion terms of FPE can be handled in this manner. Another method, also presented here is to use a direct particle method for solving the full FPE

  11. Acceleration of a Monte Carlo radiation transport code

    International Nuclear Information System (INIS)

    Hochstedler, R.D.; Smith, L.M.

    1996-01-01

    Execution time for the Integrated TIGER Series (ITS) Monte Carlo radiation transport code has been reduced by careful re-coding of computationally intensive subroutines. Three test cases for the TIGER (1-D slab geometry), CYLTRAN (2-D cylindrical geometry), and ACCEPT (3-D arbitrary geometry) codes were identified and used to benchmark and profile program execution. Based upon these results, sixteen top time-consuming subroutines were examined and nine of them modified to accelerate computations with equivalent numerical output to the original. The results obtained via this study indicate that speedup factors of 1.90 for the TIGER code, 1.67 for the CYLTRAN code, and 1.11 for the ACCEPT code are achievable. copyright 1996 American Institute of Physics

  12. bhlight: GENERAL RELATIVISTIC RADIATION MAGNETOHYDRODYNAMICS WITH MONTE CARLO TRANSPORT

    International Nuclear Information System (INIS)

    Ryan, B. R.; Gammie, C. F.; Dolence, J. C.

    2015-01-01

    We present bhlight, a numerical scheme for solving the equations of general relativistic radiation magnetohydrodynamics using a direct Monte Carlo solution of the frequency-dependent radiative transport equation. bhlight is designed to evolve black hole accretion flows at intermediate accretion rate, in the regime between the classical radiatively efficient disk and the radiatively inefficient accretion flow (RIAF), in which global radiative effects play a sub-dominant but non-negligible role in disk dynamics. We describe the governing equations, numerical method, idiosyncrasies of our implementation, and a suite of test and convergence results. We also describe example applications to radiative Bondi accretion and to a slowly accreting Kerr black hole in axisymmetry

  13. Error reduction techniques for Monte Carlo neutron transport calculations

    International Nuclear Information System (INIS)

    Ju, J.H.W.

    1981-01-01

    Monte Carlo methods have been widely applied to problems in nuclear physics, mathematical reliability, communication theory, and other areas. The work in this thesis is developed mainly with neutron transport applications in mind. For nuclear reactor and many other applications, random walk processes have been used to estimate multi-dimensional integrals and obtain information about the solution of integral equations. When the analysis is statistically based such calculations are often costly, and the development of efficient estimation techniques plays a critical role in these applications. All of the error reduction techniques developed in this work are applied to model problems. It is found that the nearly optimal parameters selected by the analytic method for use with GWAN estimator are nearly identical to parameters selected by the multistage method. Modified path length estimation (based on the path length importance measure) leads to excellent error reduction in all model problems examined. Finally, it should be pointed out that techniques used for neutron transport problems may be transferred easily to other application areas which are based on random walk processes. The transport problems studied in this dissertation provide exceptionally severe tests of the error reduction potential of any sampling procedure. It is therefore expected that the methods of this dissertation will prove useful in many other application areas

  14. OGRE, Monte-Carlo System for Gamma Transport Problems

    International Nuclear Information System (INIS)

    1984-01-01

    1 - Nature of physical problem solved: The OGRE programme system was designed to calculate, by Monte Carlo methods, any quantity related to gamma-ray transport. The system is represented by two examples - OGRE-P1 and OGRE-G. The OGRE-P1 programme is a simple prototype which calculates dose rate on one side of a slab due to a plane source on the other side. The OGRE-G programme, a prototype of a programme utilizing a general-geometry routine, calculates dose rate at arbitrary points. A very general source description in OGRE-G may be employed by reading a tape prepared by the user. 2 - Method of solution: Case histories of gamma rays in the prescribed geometry are generated and analyzed to produce averages of any desired quantity which, in the case of the prototypes, are gamma-ray dose rates. The system is designed to achieve generality by ease of modification. No importance sampling is built into the prototypes, a very general geometry subroutine permits the treatment of complicated geometries. This is essentially the same routine used in the O5R neutron transport system. Boundaries may be either planes or quadratic surfaces, arbitrarily oriented and intersecting in arbitrary fashion. Cross section data is prepared by the auxiliary master cross section programme XSECT which may be used to originate, update, or edit the master cross section tape. The master cross section tape is utilized in the OGRE programmes to produce detailed tables of macroscopic cross sections which are used during the Monte Carlo calculations. 3 - Restrictions on the complexity of the problem: Maximum cross-section array information may be estimated by a given formula for a specific problem. The number of regions must be less than or equal to 50

  15. Modelling of neutron and photon transport in iron and concrete radiation shieldings by the Monte Carlo method - Version 2

    CERN Document Server

    Žukauskaite, A; Plukiene, R; Plukis, A

    2007-01-01

    Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 – γ-ray beams (1-10 MeV), HIMAC and ISIS-800 – high energy neutrons (20-800 MeV) transport in iron and concrete. The results were then compared with experimental data.

  16. Optimal transport of particle beams

    International Nuclear Information System (INIS)

    Allen, C.K.; Reiser, M.

    1997-01-01

    The transport and matching problem for a low energy transport system is approached from a control theoretical viewpoint. We develop a model for a beam transport and matching section based on a multistage control network. To this model we apply the principles of optimal control to formulate techniques aiding in the design of the transport and matching section. Both nonlinear programming and dynamic programming techniques are used in the optimization. These techniques are implemented in a computer-aided design program called SPOT. Examples are presented to demonstrate the procedure and outline the results. (orig.)

  17. Particle and heat transport in Tokamaks

    International Nuclear Information System (INIS)

    Chatelier, M.

    1984-01-01

    A limitation to performances of tokamaks is heat transport through magnetic surfaces. Principles of ''classical'' or ''neoclassical'' transport -i.e. transport due to particle and heat fluxes due to Coulomb scattering of charged particle in a magnetic field- are exposed. It is shown that beside this classical effect, ''anomalous'' transport occurs; it is associated to the existence of fluctuating electric or magnetic fields which can appear in the plasma as a result of charge and current perturbations. Tearing modes and drift wave instabilities are taken as typical examples. Experimental features are presented which show that ions behave approximately in a classical way whereas electrons are strongly anomalous [fr

  18. 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.)

  19. A computer code package for Monte Carlo photon-electron transport simulation Comparisons with experimental benchmarks

    International Nuclear Information System (INIS)

    Popescu, Lucretiu M.

    2000-01-01

    A computer code package (PTSIM) for particle transport Monte Carlo simulation was developed using object oriented techniques of design and programming. A flexible system for simulation of coupled photon, electron transport, facilitating development of efficient simulation applications, was obtained. For photons: Compton and photo-electric effects, pair production and Rayleigh interactions are simulated, while for electrons, a class II condensed history scheme was considered, in which catastrophic interactions (Moeller electron-electron interaction, bremsstrahlung, etc.) are treated in detail and all other interactions with reduced individual effect on electron history are grouped together using continuous slowing down approximation and energy straggling theories. Electron angular straggling is simulated using Moliere theory or a mixed model in which scatters at large angles are treated as distinct events. Comparisons with experimentally benchmarks for electron transmission and bremsstrahlung emissions energy and angular spectra, and for dose calculations are presented

  20. Vectorization and parallelization of Monte-Carlo programs for calculation of radiation transport

    International Nuclear Information System (INIS)

    Seidel, R.

    1995-01-01

    The versatile MCNP-3B Monte-Carlo code written in FORTRAN77, for simulation of the radiation transport of neutral particles, has been subjected to vectorization and parallelization of essential parts, without touching its versatility. Vectorization is not dependent on a specific computer. Several sample tasks have been selected in order to test the vectorized MCNP-3B code in comparison to the scalar MNCP-3B code. The samples are a representative example of the 3-D calculations to be performed for simulation of radiation transport in neutron and reactor physics. (1) 4πneutron detector. (2) High-energy calorimeter. (3) PROTEUS benchmark (conversion rates and neutron multiplication factors for the HCLWR (High Conversion Light Water Reactor)). (orig./HP) [de

  1. New capabilities for Monte Carlo simulation of deuteron transport and secondary products generation

    International Nuclear Information System (INIS)

    Sauvan, P.; Sanz, J.; Ogando, F.

    2010-01-01

    Several important research programs are dedicated to the development of facilities based on deuteron accelerators. In designing these facilities, the definition of a validated computational approach able to simulate deuteron transport and evaluate deuteron interactions and production of secondary particles with acceptable precision is a very important issue. Current Monte Carlo codes, such as MCNPX or PHITS, when applied for deuteron transport calculations use built-in semi-analytical models to describe deuteron interactions. These models are found unreliable in predicting neutron and photon generated by low energy deuterons, typically present in those facilities. We present a new computational tool, resulting from an extension of the MCNPX code, which improve significantly the treatment of problems where any secondary product (neutrons, photons, tritons, etc.) generated by low energy deuterons reactions could play a major role. Firstly, it handles deuteron evaluated data libraries, which allow describing better low deuteron energy interactions. Secondly, it includes a reduction variance technique for production of secondary particles by charged particle-induced nuclear interactions, which allow reducing drastically the computing time needed in transport and nuclear response calculations. Verification of the computational tool is successfully achieved. This tool can be very helpful in addressing design issues such as selection of the dedicated neutron production target and accelerator radioprotection analysis. It can be also helpful to test the deuteron cross-sections under development in the frame of different international nuclear data programs.

  2. Stochastic transport of particles across single barriers

    International Nuclear Information System (INIS)

    Kreuter, Christian; Siems, Ullrich; Henseler, Peter; Nielaba, Peter; Leiderer, Paul; Erbe, Artur

    2012-01-01

    Transport phenomena of interacting particles are of high interest for many applications in biology and mesoscopic systems. Here we present measurements on colloidal particles, which are confined in narrow channels on a substrate and interact with a barrier, which impedes the motion along the channel. The substrate of the particle is tilted in order for the particles to be driven towards the barrier and, if the energy gained by the tilt is large enough, surpass the barrier by thermal activation. We therefore study the influence of this barrier as well as the influence of particle interaction on the particle transport through such systems. All experiments are supported with Brownian dynamics simulations in order to complement the experiments with tests of a large range of parameter space which cannot be accessed in experiments.

  3. Monte Carlo estimation of neoclassical transport for the TJ-II stellarator

    International Nuclear Information System (INIS)

    Tribaldos, V.

    2001-01-01

    The neoclassical transport properties of TJ-II stellarator [C. Alejaldre et al., Fusion Technol. 13, 521 (1988)] are studied with the monoenergetic Monte Carlo technique. A compromise between the number of modes and particles and the required computing time to obtain reliable estimates, from the computational point of view, of the monoenergetic diffusion coefficients is shown to be of one thousand particles and one hundred harmonics, because of the rich magnetic-field structure of TJ-II. Although, these requirements are probably too demanding in making the transport estimations. The data base containing the normalized monoenergetic diffusion coefficient for several radial positions, radial electric fields and collisionalities have been fitted using a neural network. This fit reduces the number of points necessary in the data base and allows a smooth interpolation and extrapolation to perform the convolutions of the monoenergetic coefficients with the Maxwellian. For two different typical TJ-II discharges the ambipolar radial electric field, and the neoclassical particle and heat fluxes are presented both showing rather large positive radial electric fields at the plasma core and small negative fields at the edge. The neoclassical particle and energy confinement time are in surprisingly good agreement with the experimental energy balance analysis and the international stellarator scaling. Although no satisfactory explanation is available yet the large neoclassical diffusion caused by the complex ripple structure of TJ-II magnetic field may be an important ingredient

  4. Distributed Monte Carlo Information Fusion and Distributed Particle Filtering

    Science.gov (United States)

    2014-08-24

    gossip procedure to arrive at a consensus about the likelihoods of particles. The selective gossip procedure shares particles based upon weights. This...focuses commu- nication on the particles that contain the most informa- tion. Oreshkin and Coates (2010) also use a gossip con- sensus approach to...Applied Probability. Chapman and Hall. Üstebay, D., Coates, M., and Rabbat, M. (2011). Dis- tributed auxiliary particle filters using selective gossip . In

  5. Development of a consistent Monte Carlo-deterministic transport methodology based on the method of characteristics and MCNP5

    International Nuclear Information System (INIS)

    Karriem, Z.; Ivanov, K.; Zamonsky, O.

    2011-01-01

    This paper presents work that has been performed to develop an integrated Monte Carlo- Deterministic transport methodology in which the two methods make use of exactly the same general geometry and multigroup nuclear data. The envisioned application of this methodology is in reactor lattice physics methods development and shielding calculations. The methodology will be based on the Method of Long Characteristics (MOC) and the Monte Carlo N-Particle Transport code MCNP5. Important initial developments pertaining to ray tracing and the development of an MOC flux solver for the proposed methodology are described. Results showing the viability of the methodology are presented for two 2-D general geometry transport problems. The essential developments presented is the use of MCNP as geometry construction and ray tracing tool for the MOC, verification of the ray tracing indexing scheme that was developed to represent the MCNP geometry in the MOC and the verification of the prototype 2-D MOC flux solver. (author)

  6. Temperature variance study in Monte-Carlo photon transport theory

    International Nuclear Information System (INIS)

    Giorla, J.

    1985-10-01

    We study different Monte-Carlo methods for solving radiative transfer problems, and particularly Fleck's Monte-Carlo method. We first give the different time-discretization schemes and the corresponding stability criteria. Then we write the temperature variance as a function of the variances of temperature and absorbed energy at the previous time step. Finally we obtain some stability criteria for the Monte-Carlo method in the stationary case [fr

  7. Hybrid transport and diffusion modeling using electron thermal transport Monte Carlo SNB in DRACO

    Science.gov (United States)

    Chenhall, Jeffrey; Moses, Gregory

    2017-10-01

    The iSNB (implicit Schurtz Nicolai Busquet) multigroup diffusion electron thermal transport method is adapted into an Electron Thermal Transport Monte Carlo (ETTMC) transport method to better model angular and long mean free path non-local effects. Previously, the ETTMC model had been implemented in the 2D DRACO multiphysics code and found to produce consistent results with the iSNB method. Current work is focused on a hybridization of the computationally slower but higher fidelity ETTMC transport method with the computationally faster iSNB diffusion method in order to maximize computational efficiency. Furthermore, effects on the energy distribution of the heat flux divergence are studied. Work to date on the hybrid method will be presented. This work was supported by Sandia National Laboratories and the Univ. of Rochester Laboratory for Laser Energetics.

  8. Monte Carlo impurity transport modeling in the DIII-D transport

    International Nuclear Information System (INIS)

    Evans, T.E.; Finkenthal, D.F.

    1998-04-01

    A description of the carbon transport and sputtering physics contained in the Monte Carlo Impurity (MCI) transport code is given. Examples of statistically significant carbon transport pathways are examined using MCI's unique tracking visualizer and a mechanism for enhanced carbon accumulation on the high field side of the divertor chamber is discussed. Comparisons between carbon emissions calculated with MCI and those measured in the DIII-D tokamak are described. Good qualitative agreement is found between 2D carbon emission patterns calculated with MCI and experimentally measured carbon patterns. While uncertainties in the sputtering physics, atomic data, and transport models have made quantitative comparisons with experiments more difficult, recent results using a physics based model for physical and chemical sputtering has yielded simulations with about 50% of the total carbon radiation measured in the divertor. These results and plans for future improvement in the physics models and atomic data are discussed

  9. Approximate models for neutral particle transport calculations in ducts

    International Nuclear Information System (INIS)

    Ono, Shizuca

    2000-01-01

    The problem of neutral particle transport in evacuated ducts of arbitrary, but axially uniform, cross-sectional geometry and isotropic reflection at the wall is studied. The model makes use of basis functions to represent the transverse and azimuthal dependences of the particle angular flux in the duct. For the approximation in terms of two basis functions, an improvement in the method is implemented by decomposing the problem into uncollided and collided components. A new quadrature set, more suitable to the problem, is developed and generated by one of the techniques of the constructive theory of orthogonal polynomials. The approximation in terms of three basis functions is developed and implemented to improve the precision of the results. For both models of two and three basis functions, the energy dependence of the problem is introduced through the multigroup formalism. The results of sample problems are compared to literature results and to results of the Monte Carlo code, MCNP. (author)

  10. An improved Monte Carlo (MC) dose simulation for charged particle cancer therapy

    Energy Technology Data Exchange (ETDEWEB)

    Ying, C. K. [Advanced Medical and Dental Institute, AMDI, Universiti Sains Malaysia, Penang, Malaysia and School of Medical Sciences, Universiti Sains Malaysia, Kota Bharu (Malaysia); Kamil, W. A. [Advanced Medical and Dental Institute, AMDI, Universiti Sains Malaysia, Penang, Malaysia and Radiology Department, Hospital USM, Kota Bharu (Malaysia); Shuaib, I. L. [Advanced Medical and Dental Institute, AMDI, Universiti Sains Malaysia, Penang (Malaysia); Matsufuji, Naruhiro [Research Centre of Charged Particle Therapy, National Institute of Radiological Sciences, NIRS, Chiba (Japan)

    2014-02-12

    Heavy-particle therapy such as carbon ion therapy are more popular nowadays because of the nature characteristics of charged particle and almost no side effect to patients. An effective treatment is achieved with high precision of dose calculation, in this research work, Geant4 based Monte Carlo simulation method has been used to calculate the radiation transport and dose distribution. The simulation have the same setting with the treatment room in Heavy Ion Medical Accelerator, HIMAC. The carbon ion beam at the isocentric gantry nozzle for the therapeutic energy of 290 MeV/u was simulated, experimental work was carried out in National Institute of Radiological Sciences, NIRS, Chiba, Japan by using the HIMAC to confirm the accuracy and qualities dose distribution by MC methods. The Geant4 based simulated dose distribution were verified with measurements for Bragg peak and spread out Bragg peak (SOBP) respectively. The verification of results shows that the Bragg peak depth-dose and SOBP distributions in simulation has good agreement with measurements. In overall, the study showed that Geant4 based can be fully applied in the heavy-ion therapy field for simulation, further works need to be carry on to refine and improve the Geant4 MC simulations.

  11. An improved Monte Carlo (MC) dose simulation for charged particle cancer therapy

    International Nuclear Information System (INIS)

    Ying, C. K.; Kamil, W. A.; Shuaib, I. L.; Matsufuji, Naruhiro

    2014-01-01

    Heavy-particle therapy such as carbon ion therapy are more popular nowadays because of the nature characteristics of charged particle and almost no side effect to patients. An effective treatment is achieved with high precision of dose calculation, in this research work, Geant4 based Monte Carlo simulation method has been used to calculate the radiation transport and dose distribution. The simulation have the same setting with the treatment room in Heavy Ion Medical Accelerator, HIMAC. The carbon ion beam at the isocentric gantry nozzle for the therapeutic energy of 290 MeV/u was simulated, experimental work was carried out in National Institute of Radiological Sciences, NIRS, Chiba, Japan by using the HIMAC to confirm the accuracy and qualities dose distribution by MC methods. The Geant4 based simulated dose distribution were verified with measurements for Bragg peak and spread out Bragg peak (SOBP) respectively. The verification of results shows that the Bragg peak depth-dose and SOBP distributions in simulation has good agreement with measurements. In overall, the study showed that Geant4 based can be fully applied in the heavy-ion therapy field for simulation, further works need to be carry on to refine and improve the Geant4 MC simulations

  12. An improved Monte Carlo (MC) dose simulation for charged particle cancer therapy

    International Nuclear Information System (INIS)

    Ying, C.K.; Kamil, W.A.; Shuaib, I.L.; Ying, C.K.; Kamil, W.A.

    2013-01-01

    Full-text: Heavy-particle therapy such as carbon ion therapy are more popular nowadays because of the nature characteristics of charged particle and almost no side effect to patients. An effective treatment is achieved with high precision of dose calculation, in this research work, Geant4 based Monte Carlo simulation method has been used to calculate the radiation transport and dose distribution. The simulation have the same setting with the treatment room in Heavy Ion Medical Accelerator, HIMAC. The carbon ion beam at the isocentric gantry nozzle for the therapeutic energy of 290 MeV/u was simulated, experimental work was carried out in National Institute of Radiological Sciences, NIRS, Chiba, Japan by using the HIMAC to confirm the accuracy and qualities dose distribution by MC methods. The Geant4 based simulated dose distribution were verified with measurements for Bragg peak and spread out Bragg peak (SOBP) respectively. The verification of results shows that the Bragg peak depth-dose and SOBP distributions in simulation has good agreement with measurements. In overall, the study showed that Geant4 based can be fully applied in the heavy ion therapy field for simulation, further works need to be carry on to refine and improve the Geant4 MC simulations. (author)

  13. Particle transport in inclined annuli

    Energy Technology Data Exchange (ETDEWEB)

    Kurtzhals, Erik

    1993-12-31

    A new model for the formation and behaviour of deposits in inclined wellbores is formulated. The annular space is divided into two layers, separated by a distinct plane boundary. While the lower layer is taken to consist of closely packed cuttings, the upper layer is presumed to behave as a pure fluid. A force balance for the lower layer decides whether it is stationary or slides in the upwards- or downwards direction. The position of the deposit surface is governed by the fluid shear stress at the deposit surface. The proposed model represents a major improvement compared to an earlier model. The predictions from the SCSB-model are in good qualitative agreement with experimental results obtained by the author, and results published by research groups in the U.S.A., United Kingdom and Germany. The quantitative agreement is variable, presumably because the SCSB-model is a somewhat simplified description of particle behaviour in inclined annuli. However, the model provides a clearer understanding of the physical background for previously published experimental results. In order to couple the theoretical work with experimental observations, an annular flow loop has been constructed. A characteristic feature in the flow loop design is the application of load cells, which permits determination of the annular particle content at steady state as well as under transient conditions. Due to delays in the constructional work, it has only been possible to perform a limited number of investigations in the loop. However, the results produced are in agreement with results published by other research groups. (au)

  14. Particle transport in inclined annuli

    Energy Technology Data Exchange (ETDEWEB)

    Kurtzhals, Erik

    1994-12-31

    A new model for the formation and behaviour of deposits in inclined wellbores is formulated. The annular space is divided into two layers, separated by a distinct plane boundary. While the lower layer is taken to consist of closely packed cuttings, the upper layer is presumed to behave as a pure fluid. A force balance for the lower layer decides whether it is stationary or slides in the upwards- or downwards direction. The position of the deposit surface is governed by the fluid shear stress at the deposit surface. The proposed model represents a major improvement compared to an earlier model. The predictions from the SCSB-model are in good qualitative agreement with experimental results obtained by the author, and results published by research groups in the U.S.A., United Kingdom and Germany. The quantitative agreement is variable, presumably because the SCSB-model is a somewhat simplified description of particle behaviour in inclined annuli. However, the model provides a clearer understanding of the physical background for previously published experimental results. In order to couple the theoretical work with experimental observations, an annular flow loop has been constructed. A characteristic feature in the flow loop design is the application of load cells, which permits determination of the annular particle content at steady state as well as under transient conditions. Due to delays in the constructional work, it has only been possible to perform a limited number of investigations in the loop. However, the results produced are in agreement with results published by other research groups. (au)

  15. SRNA-2K5, Proton Transport Using 3-D by Monte Carlo Techniques

    International Nuclear Information System (INIS)

    Ilic, Radovan D.

    2005-01-01

    1 - Description of program or function: SRNA-2K5 performs Monte Carlo transport simulation of proton in 3D source and 3D geometry of arbitrary materials. The proton transport based on condensed history model, and on model of compound nuclei decays that creates in nonelastic nuclear interaction by proton absorption. 2 - Methods: The SRNA-2K5 package is developed for time independent simulation of proton transport by Monte Carlo techniques for numerical experiments in complex geometry, using PENGEOM from PENELOPE with different material compositions, and arbitrary spectrum of proton generated from the 3D source. This package developed for 3D proton dose distribution in proton therapy and dosimetry, and it was based on the theory of multiple scattering. The compound nuclei decay was simulated by our and Russian MSDM models using ICRU 49 and ICRU 63 data. If protons trajectory is divided on great number of steps, protons passage can be simulated according to Berger's Condensed Random Walk model. Conditions of angular distribution and fluctuation of energy loss determinate step length. Physical picture of these processes is described by stopping power, Moliere's angular distribution, Vavilov's distribution with Sulek's correction per all electron orbits, and Chadwick's cross sections for nonelastic nuclear interactions, obtained by his GNASH code. According to physical picture of protons passage and with probabilities of protons transition from previous to next stage, which is prepared by SRNADAT program, simulation of protons transport in all SRNA codes runs according to usual Monte Carlo scheme: (i) proton from the spectrum prepared for random choice of energy, position and space angle is emitted from the source; (ii) proton is loosing average energy on the step; (iii) on that step, proton experience a great number of collisions, and it changes direction of movement randomly chosen from angular distribution; (iv) random fluctuation is added to average energy loss; (v

  16. Monte Carlo Studies of Electron Transport In Semiconductor Nanostructures

    Science.gov (United States)

    Tierney, Brian David

    An Ensemble Monte Carlo (EMC) computer code has been developed to simulate, semi-classically, spin-dependent electron transport in quasi two-dimensional (2D) III-V semiconductors. The code accounts for both three-dimensional (3D) and quasi-2D transport, utilizing either 3D or 2D scattering mechanisms, as appropriate. Phonon, alloy, interface roughness, and impurity scattering mechanisms are included, accounting for the Pauli Exclusion Principle via a rejection algorithm. The 2D carrier states are calculated via a self-consistent 1D Schrodinger-3D-Poisson solution in which the charge distribution of the 2D carriers in the quantization direction is taken as the spatial distribution of the squared envelope functions within the Hartree approximation. The wavefunctions, subband energies, and 2D scattering rates are updated periodically by solving a series of 1D Schrodinger wave equations (SWE) over the real-space domain of the device at fixed time intervals. The electrostatic potential is updated by periodically solving the 3D Poisson equation. Spin-polarized transport is modeled via a spin density-matrix formalism that accounts for D'yakanov-Perel (DP) scattering. Also, the code allows for the easy inclusion of additional scattering mechanisms and structural modifications to devices. As an application of the simulator, the current voltage characteristics of an InGaAs/InAlAs HEMT are simulated, corresponding to nanoscale III-V HEMTs currently being fabricated by Intel Corporation. The comparative effects of various scattering parameters, material properties and structural attributes are investigated and compared with experiments where reasonable agreement is obtained. The spatial evolution of spin-polarized carriers in prototypical Spin Field Effect Transistor (SpinFET) devices is then simulated. Studies of the spin coherence times in quasi-2D structures is first investigated and compared to experimental results. It is found that the simulated spin coherence times for

  17. The energetic alpha particle transport method EATM

    International Nuclear Information System (INIS)

    Kirkpatrick, R.C.

    1998-02-01

    The EATM method is an evolving attempt to find an efficient method of treating the transport of energetic charged particles in a dynamic magnetized (MHD) plasma for which the mean free path of the particles and the Larmor radius may be long compared to the gradient lengths in the plasma. The intent is to span the range of parameter space with the efficiency and accuracy thought necessary for experimental analysis and design of magnetized fusion targets

  18. Limits on the Efficiency of Event-Based Algorithms for Monte Carlo Neutron Transport

    Energy Technology Data Exchange (ETDEWEB)

    Romano, Paul K.; Siegel, Andrew R.

    2017-04-16

    The traditional form of parallelism in Monte Carlo particle transport simulations, wherein each individual particle history is considered a unit of work, does not lend itself well to data-level parallelism. Event-based algorithms, which were originally used for simulations on vector processors, may offer a path toward better utilizing data-level parallelism in modern computer architectures. In this study, a simple model is developed for estimating the efficiency of the event-based particle transport algorithm under two sets of assumptions. Data collected from simulations of four reactor problems using OpenMC was then used in conjunction with the models to calculate the speedup due to vectorization as a function of two parameters: the size of the particle bank and the vector width. When each event type is assumed to have constant execution time, the achievable speedup is directly related to the particle bank size. We observed that the bank size generally needs to be at least 20 times greater than vector size in order to achieve vector efficiency greater than 90%. When the execution times for events are allowed to vary, however, the vector speedup is also limited by differences in execution time for events being carried out in a single event-iteration. For some problems, this implies that vector effciencies over 50% may not be attainable. While there are many factors impacting performance of an event-based algorithm that are not captured by our model, it nevertheless provides insights into factors that may be limiting in a real implementation.

  19. A Monte Carlo transport code study of the space radiation environment using FLUKA and ROOT

    CERN Document Server

    Wilson, T; Carminati, F; Brun, R; Ferrari, A; Sala, P; Empl, A; MacGibbon, J

    2001-01-01

    We report on the progress of a current study aimed at developing a state-of-the-art Monte-Carlo computer simulation of the space radiation environment using advanced computer software techniques recently available at CERN, the European Laboratory for Particle Physics in Geneva, Switzerland. By taking the next-generation computer software appearing at CERN and adapting it to known problems in the implementation of space exploration strategies, this research is identifying changes necessary to bring these two advanced technologies together. The radiation transport tool being developed is tailored to the problem of taking measured space radiation fluxes impinging on the geometry of any particular spacecraft or planetary habitat and simulating the evolution of that flux through an accurate model of the spacecraft material. The simulation uses the latest known results in low-energy and high-energy physics. The output is a prediction of the detailed nature of the radiation environment experienced in space as well a...

  20. Improved cache performance in Monte Carlo transport calculations using energy banding

    Science.gov (United States)

    Siegel, A.; Smith, K.; Felker, K.; Romano, P.; Forget, B.; Beckman, P.

    2014-04-01

    We present an energy banding algorithm for Monte Carlo (MC) neutral particle transport simulations which depend on large cross section lookup tables. In MC codes, read-only cross section data tables are accessed frequently, exhibit poor locality, and are typically too much large to fit in fast memory. Thus, performance is often limited by long latencies to RAM, or by off-node communication latencies when the data footprint is very large and must be decomposed on a distributed memory machine. The proposed energy banding algorithm allows maximal temporal reuse of data in band sizes that can flexibly accommodate different architectural features. The energy banding algorithm is general and has a number of benefits compared to the traditional approach. In the present analysis we explore its potential to achieve improvements in time-to-solution on modern cache-based architectures.

  1. Comparative assessment of different approaches for the use of CAD geometry in Monte Carlo transport calculations

    International Nuclear Information System (INIS)

    Weinhorst, Bastian; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Wilson, Paul

    2015-01-01

    Highlights: • Comparison of different approaches for the use of CAD geometry for Monte Carlo transport calculations. • Comparison with regard to user-friendliness and computation performance. • Three approaches, namely conversion with McCad, unstructured mesh feature of MCN6 and DAGMC. • Installation most complex for DAGMC, model preparation worst for McCad, computation performance worst for MCNP6. • Installation easiest for McCad, model preparation best for MCNP6, computation speed fastest for McCad. - Abstract: Computer aided design (CAD) is an important industrial way to produce high quality designs. Therefore, CAD geometries are in general used for engineering and the design of complex facilities like the ITER tokamak. Although Monte Carlo codes like MCNP are well suited to handle the complex 3D geometry of ITER for transport calculations, they rely on their own geometry description and are in general not able to directly use the CAD geometry. In this paper, three different approaches for the use of CAD geometries with MCNP calculations are investigated and assessed with regard to calculation performance and user-friendliness. The first method is the conversion of the CAD geometry into MCNP geometry employing the conversion software McCad developed by KIT. The second approach utilizes the MCNP6 mesh geometry feature for the particle tracking and relies on the conversion of the CAD geometry into a mesh model. The third method employs DAGMC, developed by the University of Wisconsin-Madison, for the direct particle tracking on the CAD geometry using a patched version of MCNP. The obtained results show that each method has its advantages depending on the complexity and size of the model, the calculation problem considered, and the expertise of the user.

  2. Comparative assessment of different approaches for the use of CAD geometry in Monte Carlo transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Weinhorst, Bastian, E-mail: bastian.weinhorst@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology, Eggenstein-Leopoldshafen (Germany); Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology, Eggenstein-Leopoldshafen (Germany); Wilson, Paul [University of Wisconsin-Madison, Computational Nuclear Engineering Research Group, Madison, WI (United States)

    2015-10-15

    Highlights: • Comparison of different approaches for the use of CAD geometry for Monte Carlo transport calculations. • Comparison with regard to user-friendliness and computation performance. • Three approaches, namely conversion with McCad, unstructured mesh feature of MCN6 and DAGMC. • Installation most complex for DAGMC, model preparation worst for McCad, computation performance worst for MCNP6. • Installation easiest for McCad, model preparation best for MCNP6, computation speed fastest for McCad. - Abstract: Computer aided design (CAD) is an important industrial way to produce high quality designs. Therefore, CAD geometries are in general used for engineering and the design of complex facilities like the ITER tokamak. Although Monte Carlo codes like MCNP are well suited to handle the complex 3D geometry of ITER for transport calculations, they rely on their own geometry description and are in general not able to directly use the CAD geometry. In this paper, three different approaches for the use of CAD geometries with MCNP calculations are investigated and assessed with regard to calculation performance and user-friendliness. The first method is the conversion of the CAD geometry into MCNP geometry employing the conversion software McCad developed by KIT. The second approach utilizes the MCNP6 mesh geometry feature for the particle tracking and relies on the conversion of the CAD geometry into a mesh model. The third method employs DAGMC, developed by the University of Wisconsin-Madison, for the direct particle tracking on the CAD geometry using a patched version of MCNP. The obtained results show that each method has its advantages depending on the complexity and size of the model, the calculation problem considered, and the expertise of the user.

  3. A NEW MONTE CARLO METHOD FOR TIME-DEPENDENT NEUTRINO RADIATION TRANSPORT

    International Nuclear Information System (INIS)

    Abdikamalov, Ernazar; Ott, Christian D.; O'Connor, Evan; Burrows, Adam; Dolence, Joshua C.; Löffler, Frank; Schnetter, Erik

    2012-01-01

    Monte Carlo approaches to radiation transport have several attractive properties such as simplicity of implementation, high accuracy, and good parallel scaling. Moreover, Monte Carlo methods can handle complicated geometries and are relatively easy to extend to multiple spatial dimensions, which makes them potentially interesting in modeling complex multi-dimensional astrophysical phenomena such as core-collapse supernovae. The aim of this paper is to explore Monte Carlo methods for modeling neutrino transport in core-collapse supernovae. We generalize the Implicit Monte Carlo photon transport scheme of Fleck and Cummings and gray discrete-diffusion scheme of Densmore et al. to energy-, time-, and velocity-dependent neutrino transport. Using our 1D spherically-symmetric implementation, we show that, similar to the photon transport case, the implicit scheme enables significantly larger timesteps compared with explicit time discretization, without sacrificing accuracy, while the discrete-diffusion method leads to significant speed-ups at high optical depth. Our results suggest that a combination of spectral, velocity-dependent, Implicit Monte Carlo and discrete-diffusion Monte Carlo methods represents a robust approach for use in neutrino transport calculations in core-collapse supernovae. Our velocity-dependent scheme can easily be adapted to photon transport.

  4. A NEW MONTE CARLO METHOD FOR TIME-DEPENDENT NEUTRINO RADIATION TRANSPORT

    Energy Technology Data Exchange (ETDEWEB)

    Abdikamalov, Ernazar; Ott, Christian D.; O' Connor, Evan [TAPIR, California Institute of Technology, MC 350-17, 1200 E California Blvd., Pasadena, CA 91125 (United States); Burrows, Adam; Dolence, Joshua C. [Department of Astrophysical Sciences, Princeton University, Peyton Hall, Ivy Lane, Princeton, NJ 08544 (United States); Loeffler, Frank; Schnetter, Erik, E-mail: abdik@tapir.caltech.edu [Center for Computation and Technology, Louisiana State University, 216 Johnston Hall, Baton Rouge, LA 70803 (United States)

    2012-08-20

    Monte Carlo approaches to radiation transport have several attractive properties such as simplicity of implementation, high accuracy, and good parallel scaling. Moreover, Monte Carlo methods can handle complicated geometries and are relatively easy to extend to multiple spatial dimensions, which makes them potentially interesting in modeling complex multi-dimensional astrophysical phenomena such as core-collapse supernovae. The aim of this paper is to explore Monte Carlo methods for modeling neutrino transport in core-collapse supernovae. We generalize the Implicit Monte Carlo photon transport scheme of Fleck and Cummings and gray discrete-diffusion scheme of Densmore et al. to energy-, time-, and velocity-dependent neutrino transport. Using our 1D spherically-symmetric implementation, we show that, similar to the photon transport case, the implicit scheme enables significantly larger timesteps compared with explicit time discretization, without sacrificing accuracy, while the discrete-diffusion method leads to significant speed-ups at high optical depth. Our results suggest that a combination of spectral, velocity-dependent, Implicit Monte Carlo and discrete-diffusion Monte Carlo methods represents a robust approach for use in neutrino transport calculations in core-collapse supernovae. Our velocity-dependent scheme can easily be adapted to photon transport.

  5. Improvement of the symbolic Monte-Carlo method for the transport equation: P1 extension and coupling with diffusion

    International Nuclear Information System (INIS)

    Clouet, J.F.; Samba, G.

    2005-01-01

    We use asymptotic analysis to study the diffusion limit of the Symbolic Implicit Monte-Carlo (SIMC) method for the transport equation. For standard SIMC with piecewise constant basis functions, we demonstrate mathematically that the solution converges to the solution of a wrong diffusion equation. Nevertheless a simple extension to piecewise linear basis functions enables to obtain the correct solution. This improvement allows the calculation in opaque medium on a mesh resolving the diffusion scale much larger than the transport scale. Anyway, the huge number of particles which is necessary to get a correct answer makes this computation time consuming. Thus, we have derived from this asymptotic study an hybrid method coupling deterministic calculation in the opaque medium and Monte-Carlo calculation in the transparent medium. This method gives exactly the same results as the previous one but at a much lower price. We present numerical examples which illustrate the analysis. (authors)

  6. Particle and heavy ion transport code system; PHITS

    International Nuclear Information System (INIS)

    Niita, Koji

    2004-01-01

    Intermediate and high energy nuclear data are strongly required in design study of many facilities such as accelerator-driven systems, intense pulse spallation neutron sources, and also in medical and space technology. There is, however, few evaluated nuclear data of intermediate and high energy nuclear reactions. Therefore, we have to use some models or systematics for the cross sections, which are essential ingredients of high energy particle and heavy ion transport code to estimate neutron yield, heat deposition and many other quantities of the transport phenomena in materials. We have developed general purpose particle and heavy ion transport Monte Carlo code system, PHITS (Particle and Heavy Ion Transport code System), based on the NMTC/JAM code by the collaboration of Tohoku University, JAERI and RIST. The PHITS has three important ingredients which enable us to calculate (1) high energy nuclear reactions up to 200 GeV, (2) heavy ion collision and its transport in material, (3) low energy neutron transport based on the evaluated nuclear data. In the PHITS, the cross sections of high energy nuclear reactions are obtained by JAM model. JAM (Jet AA Microscopic Transport Model) is a hadronic cascade model, which explicitly treats all established hadronic states including resonances and all hadron-hadron cross sections parametrized based on the resonance model and string model by fitting the available experimental data. The PHITS can describe the transport of heavy ions and their collisions by making use of JQMD and SPAR code. The JQMD (JAERI Quantum Molecular Dynamics) is a simulation code for nucleus nucleus collisions based on the molecular dynamics. The SPAR code is widely used to calculate the stopping powers and ranges for charged particles and heavy ions. The PHITS has included some part of MCNP4C code, by which the transport of low energy neutron, photon and electron based on the evaluated nuclear data can be described. Furthermore, the high energy nuclear

  7. Monte Carlo analysis of radiative transport in oceanographic lidar measurements

    Energy Technology Data Exchange (ETDEWEB)

    Cupini, E.; Ferro, G. [ENEA, Divisione Fisica Applicata, Centro Ricerche Ezio Clementel, Bologna (Italy); Ferrari, N. [Bologna Univ., Bologna (Italy). Dipt. Ingegneria Energetica, Nucleare e del Controllo Ambientale

    2001-07-01

    The analysis of oceanographic lidar systems measurements is often carried out with semi-empirical methods, since there is only a rough understanding of the effects of many environmental variables. The development of techniques for interpreting the accuracy of lidar measurements is needed to evaluate the effects of various environmental situations, as well as of different experimental geometric configurations and boundary conditions. A Monte Carlo simulation model represents a tool that is particularly well suited for answering these important questions. The PREMAR-2F Monte Carlo code has been developed taking into account the main molecular and non-molecular components of the marine environment. The laser radiation interaction processes of diffusion, re-emission, refraction and absorption are treated. In particular are considered: the Rayleigh elastic scattering, produced by atoms and molecules with small dimensions with respect to the laser emission wavelength (i.e. water molecules), the Mie elastic scattering, arising from atoms or molecules with dimensions comparable to the laser wavelength (hydrosols), the Raman inelastic scattering, typical of water, the absorption of water, inorganic (sediments) and organic (phytoplankton and CDOM) hydrosols, the fluorescence re-emission of chlorophyll and yellow substances. PREMAR-2F is an extension of a code for the simulation of the radiative transport in atmospheric environments (PREMAR-2). The approach followed in PREMAR-2 was to combine conventional Monte Carlo techniques with analytical estimates of the probability of the receiver to have a contribution from photons coming back after an interaction in the field of view of the lidar fluorosensor collecting apparatus. This offers an effective mean for modelling a lidar system with realistic geometric constraints. The retrieved semianalytic Monte Carlo radiative transfer model has been developed in the frame of the Italian Research Program for Antarctica (PNRA) and it is

  8. FLUKA: A Multi-Particle Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Ferrari, A.; Sala, P.R.; /CERN /INFN, Milan; Fasso, A.; /SLAC; Ranft, J.; /Siegen U.

    2005-12-14

    This report describes the 2005 version of the Fluka particle transport code. The first part introduces the basic notions, describes the modular structure of the system, and contains an installation and beginner's guide. The second part complements this initial information with details about the various components of Fluka and how to use them. It concludes with a detailed history and bibliography.

  9. (U) Introduction to Monte Carlo Methods

    Energy Technology Data Exchange (ETDEWEB)

    Hungerford, Aimee L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-20

    Monte Carlo methods are very valuable for representing solutions to particle transport problems. Here we describe a “cook book” approach to handling the terms in a transport equation using Monte Carlo methods. Focus is on the mechanics of a numerical Monte Carlo code, rather than the mathematical foundations of the method.

  10. Study on MPI/OpenMP hybrid parallelism for Monte Carlo neutron transport code

    International Nuclear Information System (INIS)

    Liang Jingang; Xu Qi; Wang Kan; Liu Shiwen

    2013-01-01

    Parallel programming with mixed mode of messages-passing and shared-memory has several advantages when used in Monte Carlo neutron transport code, such as fitting hardware of distributed-shared clusters, economizing memory demand of Monte Carlo transport, improving parallel performance, and so on. MPI/OpenMP hybrid parallelism was implemented based on a one dimension Monte Carlo neutron transport code. Some critical factors affecting the parallel performance were analyzed and solutions were proposed for several problems such as contention access, lock contention and false sharing. After optimization the code was tested finally. It is shown that the hybrid parallel code can reach good performance just as pure MPI parallel program, while it saves a lot of memory usage at the same time. Therefore hybrid parallel is efficient for achieving large-scale parallel of Monte Carlo neutron transport. (authors)

  11. The electron transport problem sampling by Monte Carlo individual collision technique

    International Nuclear Information System (INIS)

    Androsenko, P.A.; Belousov, V.I.

    2005-01-01

    The problem of electron transport is of most interest in all fields of the modern science. To solve this problem the Monte Carlo sampling has to be used. The electron transport is characterized by a large number of individual interactions. To simulate electron transport the 'condensed history' technique may be used where a large number of collisions are grouped into a single step to be sampled randomly. Another kind of Monte Carlo sampling is the individual collision technique. In comparison with condensed history technique researcher has the incontestable advantages. For example one does not need to give parameters altered by condensed history technique like upper limit for electron energy, resolution, number of sub-steps etc. Also the condensed history technique may lose some very important tracks of electrons because of its limited nature by step parameters of particle movement and due to weakness of algorithms for example energy indexing algorithm. There are no these disadvantages in the individual collision technique. This report presents some sampling algorithms of new version BRAND code where above mentioned technique is used. All information on electrons was taken from Endf-6 files. They are the important part of BRAND. These files have not been processed but directly taken from electron information source. Four kinds of interaction like the elastic interaction, the Bremsstrahlung, the atomic excitation and the atomic electro-ionization were considered. In this report some results of sampling are presented after comparison with analogs. For example the endovascular radiotherapy problem (P2) of QUADOS2002 was presented in comparison with another techniques that are usually used. (authors)

  12. The electron transport problem sampling by Monte Carlo individual collision technique

    Energy Technology Data Exchange (ETDEWEB)

    Androsenko, P.A.; Belousov, V.I. [Obninsk State Technical Univ. of Nuclear Power Engineering, Kaluga region (Russian Federation)

    2005-07-01

    The problem of electron transport is of most interest in all fields of the modern science. To solve this problem the Monte Carlo sampling has to be used. The electron transport is characterized by a large number of individual interactions. To simulate electron transport the 'condensed history' technique may be used where a large number of collisions are grouped into a single step to be sampled randomly. Another kind of Monte Carlo sampling is the individual collision technique. In comparison with condensed history technique researcher has the incontestable advantages. For example one does not need to give parameters altered by condensed history technique like upper limit for electron energy, resolution, number of sub-steps etc. Also the condensed history technique may lose some very important tracks of electrons because of its limited nature by step parameters of particle movement and due to weakness of algorithms for example energy indexing algorithm. There are no these disadvantages in the individual collision technique. This report presents some sampling algorithms of new version BRAND code where above mentioned technique is used. All information on electrons was taken from Endf-6 files. They are the important part of BRAND. These files have not been processed but directly taken from electron information source. Four kinds of interaction like the elastic interaction, the Bremsstrahlung, the atomic excitation and the atomic electro-ionization were considered. In this report some results of sampling are presented after comparison with analogs. For example the endovascular radiotherapy problem (P2) of QUADOS2002 was presented in comparison with another techniques that are usually used. (authors)

  13. Simulation of neutron transport equation using parallel Monte Carlo for deep penetration problems

    International Nuclear Information System (INIS)

    Bekar, K. K.; Tombakoglu, M.; Soekmen, C. N.

    2001-01-01

    Neutron transport equation is simulated using parallel Monte Carlo method for deep penetration neutron transport problem. Monte Carlo simulation is parallelized by using three different techniques; direct parallelization, domain decomposition and domain decomposition with load balancing, which are used with PVM (Parallel Virtual Machine) software on LAN (Local Area Network). The results of parallel simulation are given for various model problems. The performances of the parallelization techniques are compared with each other. Moreover, the effects of variance reduction techniques on parallelization are discussed

  14. Particle transport in field-reversed configurations

    Energy Technology Data Exchange (ETDEWEB)

    Tuszewski, M.; Linford, R.K.

    1982-05-01

    Particle transport in field-reversed configurations is investigated using a one-dimensional, nondecaying, magnetic field structure. The radial profiles are constrained to satisfy an average ..beta.. condition from two-dimensional equilibrium and a boundary condition at the separatrix to model the balance between closed and open-field-line transport. When applied to the FRX-B experimental data and to the projected performance of the FRX-C device, this model suggests that the particle confinement times obtained with anomalous lower-hybrid-drift transport are in good agreement with the available numerical and experimental data. Larger values of confinement times can be achieved by increasing the ratio of the separatrix radius to the conducting wall radius. Even larger increases in lifetimes might be obtained by improving the open-field-line confinement.

  15. Particle transport in field-reversed configurations

    International Nuclear Information System (INIS)

    Tuszewski, M.; Linford, R.K.

    1982-01-01

    Particle transport in field-reversed configurations is investigated using a one-dimensional, nondecaying, magnetic field structure. The radial profiles are constrained to satisfy an average β condition from two-dimensional equilibrium and a boundary condition at the separatrix to model the balance between closed and open-field-line transport. When applied to the FRX-B experimental data and to the projected performance of the FRX-C device, this model suggests that the particle confinement times obtained with anomalous lower-hybrid-drift transport are in good agreement with the available numerical and experimental data. Larger values of confinement times can be achieved by increasing the ratio of the separatrix radius to the conducting wall radius. Even larger increases in lifetimes might be obtained by improving the open-field-line confinement

  16. Transport with three-particle interaction

    International Nuclear Information System (INIS)

    Morawetz, K.

    2000-01-01

    Starting from a point - like two - and three - particle interaction the kinetic equation is derived. While the drift term of the kinetic equation turns out to be determined by the known Skyrme mean field the collision integral appears in two - and three - particle parts. The cross section results from the same microscopic footing and is naturally density dependent due to the three - particle force. By this way no hybrid model for drift and cross section is needed for nuclear transport. The resulting equation of state has besides the mean field correlation energy also a two - and three - particle correlation energy which both are calculated analytically for the ground state. These energies contribute to the equation of state and lead to an occurrence of a maximum at 3 times nuclear density in the total energy. (author)

  17. SU-E-T-558: Monte Carlo Photon Transport Simulations On GPU with Quadric Geometry

    International Nuclear Information System (INIS)

    Chi, Y; Tian, Z; Jiang, S; Jia, X

    2015-01-01

    Purpose: Monte Carlo simulation on GPU has experienced rapid advancements over the past a few years and tremendous accelerations have been achieved. Yet existing packages were developed only in voxelized geometry. In some applications, e.g. radioactive seed modeling, simulations in more complicated geometry are needed. This abstract reports our initial efforts towards developing a quadric geometry module aiming at expanding the application scope of GPU-based MC simulations. Methods: We defined the simulation geometry consisting of a number of homogeneous bodies, each specified by its material composition and limiting surfaces characterized by quadric functions. A tree data structure was utilized to define geometric relationship between different bodies. We modified our GPU-based photon MC transport package to incorporate this geometry. Specifically, geometry parameters were loaded into GPU’s shared memory for fast access. Geometry functions were rewritten to enable the identification of the body that contains the current particle location via a fast searching algorithm based on the tree data structure. Results: We tested our package in an example problem of HDR-brachytherapy dose calculation for shielded cylinder. The dose under the quadric geometry and that under the voxelized geometry agreed in 94.2% of total voxels within 20% isodose line based on a statistical t-test (95% confidence level), where the reference dose was defined to be the one at 0.5cm away from the cylinder surface. It took 243sec to transport 100million source photons under this quadric geometry on an NVidia Titan GPU card. Compared with simulation time of 99.6sec in the voxelized geometry, including quadric geometry reduced efficiency due to the complicated geometry-related computations. Conclusion: Our GPU-based MC package has been extended to support photon transport simulation in quadric geometry. Satisfactory accuracy was observed with a reduced efficiency. Developments for charged

  18. SU-E-T-558: Monte Carlo Photon Transport Simulations On GPU with Quadric Geometry

    Energy Technology Data Exchange (ETDEWEB)

    Chi, Y; Tian, Z; Jiang, S; Jia, X [The University of Texas Southwestern Medical Ctr, Dallas, TX (United States)

    2015-06-15

    Purpose: Monte Carlo simulation on GPU has experienced rapid advancements over the past a few years and tremendous accelerations have been achieved. Yet existing packages were developed only in voxelized geometry. In some applications, e.g. radioactive seed modeling, simulations in more complicated geometry are needed. This abstract reports our initial efforts towards developing a quadric geometry module aiming at expanding the application scope of GPU-based MC simulations. Methods: We defined the simulation geometry consisting of a number of homogeneous bodies, each specified by its material composition and limiting surfaces characterized by quadric functions. A tree data structure was utilized to define geometric relationship between different bodies. We modified our GPU-based photon MC transport package to incorporate this geometry. Specifically, geometry parameters were loaded into GPU’s shared memory for fast access. Geometry functions were rewritten to enable the identification of the body that contains the current particle location via a fast searching algorithm based on the tree data structure. Results: We tested our package in an example problem of HDR-brachytherapy dose calculation for shielded cylinder. The dose under the quadric geometry and that under the voxelized geometry agreed in 94.2% of total voxels within 20% isodose line based on a statistical t-test (95% confidence level), where the reference dose was defined to be the one at 0.5cm away from the cylinder surface. It took 243sec to transport 100million source photons under this quadric geometry on an NVidia Titan GPU card. Compared with simulation time of 99.6sec in the voxelized geometry, including quadric geometry reduced efficiency due to the complicated geometry-related computations. Conclusion: Our GPU-based MC package has been extended to support photon transport simulation in quadric geometry. Satisfactory accuracy was observed with a reduced efficiency. Developments for charged

  19. Particle Transport in ECRH Plasmas of the TJ-II

    International Nuclear Information System (INIS)

    Vargas, V. I.; Lopez-Bruna, D.; Estrada, T.; Guasp, J.; Reynolds, J. M.; Velasco, J. L.; Herranz, J.

    2007-01-01

    We present a systematic study of particle transport in ECRH plasmas of TJ-II with different densities. The goal is to fi nd particle confinement time and electron diffusivity dependence with line-averaged density. The experimental information consists of electron temperature profiles, T e (Thomson Scattering TS) and electron density, n e , (TS and reflectometry) and measured puffing data in stationary discharges. The profile of the electron source, Se, was obtained by the 3D Monte-Carlo code EIRENE. The analysis of particle balance has been done by linking the results of the code EIRENE with the results of a model that reproduces ECRH plasmas in stationary conditions. In the range of densities studied (0.58 ≤n e > (10 1 9m - 3) ≤0.80) there are two regions of confinement separated by a threshold density, e > ∼0.65 10 1 9m - 3. Below this threshold density the particle confinement time is low, and vice versa. This is reflected in the effective diffusivity, D e , which in the range of validity of this study, 0.5 e are flat for ≥0,63(10 1 9m - 3). (Author) 35 refs

  20. On the Langevin approach to particle transport

    International Nuclear Information System (INIS)

    Bringuier, Eric

    2006-01-01

    In the Langevin description of Brownian motion, the action of the surrounding medium upon the Brownian particle is split up into a systematic friction force of Stokes type and a randomly fluctuating force, alternatively termed noise. That simple description accounts for several basic features of particle transport in a medium, making it attractive to teach at the undergraduate level, but its range of applicability is limited. The limitation is illustrated here by showing that the Langevin description fails to account realistically for the transport of a charged particle in a medium under crossed electric and magnetic fields and the ensuing Hall effect. That particular failure is rooted in the concept of the friction force rather than in the accompanying random force. It is then shown that the framework of kinetic theory offers a better account of the Hall effect. It is concluded that the Langevin description is nothing but an extension of Drude's transport model subsuming diffusion, and so it inherits basic limitations from that model. This paper thus describes the interrelationship of the Langevin approach, the Drude model and kinetic theory, in a specific transport problem of physical interest

  1. Empirical particle transport model for tokamaks

    International Nuclear Information System (INIS)

    Petravic, M.; Kuo-Petravic, G.

    1986-08-01

    A simple empirical particle transport model has been constructed with the purpose of gaining insight into the L- to H-mode transition in tokamaks. The aim was to construct the simplest possible model which would reproduce the measured density profiles in the L-regime, and also produce a qualitatively correct transition to the H-regime without having to assume a completely different transport mode for the bulk of the plasma. Rather than using completely ad hoc constructions for the particle diffusion coefficient, we assume D = 1/5 chi/sub total/, where chi/sub total/ ≅ chi/sub e/ is the thermal diffusivity, and then use the κ/sub e/ = n/sub e/chi/sub e/ values derived from experiments. The observed temperature profiles are then automatically reproduced, but nontrivially, the correct density profiles are also obtained, for realistic fueling rates and profiles. Our conclusion is that it is sufficient to reduce the transport coefficients within a few centimeters of the surface to produce the H-mode behavior. An additional simple assumption, concerning the particle mean-free path, leads to a convective transport term which reverses sign a few centimeters inside the surface, as required by the H-mode density profiles

  2. ETRAN, Electron Transport and Gamma Transport with Secondary Radiation in Slab by Monte-Carlo

    International Nuclear Information System (INIS)

    1992-01-01

    A - Nature of physical problem solved: ETRAN computes the transport of electrons and photons through plane-parallel slab targets that have a finite thickness in one dimension and are unbound in the other two-dimensions. The incident radiation can consist of a beam of either electrons or photons with specified spectral and directional distribution. Options are available by which all orders of the electron-photon cascade can be included in the calculation. Thus electrons are allowed to give rise to secondary knock-on electrons, continuous Bremsstrahlung and characteristic x-rays; and photons are allowed to produce photo-electrons, Compton electrons, and electron- positron pairs. Annihilation quanta, fluorescence radiation, and Auger electrons are also taken into account. If desired, the Monte- Carlo histories of all generations of secondary radiations are followed. The information produced by ETRAN includes the following items: 1) reflection and transmission of electrons or photons, differential in energy and direction; 2) the production of continuous Bremsstrahlung and characteristic x-rays by electrons and the emergence of such radiations from the target (differential in photon energy and direction); 3) the spectrum of the amounts of energy left behind in a thick target by an incident electron beam; 4) the deposition of energy and charge by an electron beam as function of the depth in the target; 5) the flux of electrons, differential in energy, as function of the depth in the target. B - Method of solution: A programme called DATAPAC-4 takes data for a particular material from a library tape and further processes them. The function of DATAPAC-4 is to produce single-scattering and multiple-scattering data in the form of tabular arrays (again stored on magnetic tape) which facilitate the rapid sampling of electron and photon Monte Carlo histories in ETRAN. The photon component of the electron-photon cascade is calculated by conventional random sampling that imitates

  3. FOCUS, Neutron Transport System for Complex Geometry Reactor Core and Shielding Problems by Monte-Carlo

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.

    1980-01-01

    1 - Description of problem or function: FOCUS enables the calculation of any quantity related to neutron transport in reactor or shielding problems, but was especially designed to calculate differential quantities, such as point values at one or more of the space, energy, direction and time variables of quantities like neutron flux, detector response, reaction rate, etc. or averages of such quantities over a small volume of the phase space. Different types of problems can be treated: systems with a fixed neutron source which may be a mono-directional source located out- side the system, and Eigen function problems in which the neutron source distribution is given by the (unknown) fundamental mode Eigen function distribution. Using Monte Carlo methods complex 3- dimensional geometries and detailed cross section information can be treated. Cross section data are derived from ENDF/B, with anisotropic scattering and discrete or continuous inelastic scattering taken into account. Energy is treated as a continuous variable and time dependence may also be included. 2 - Method of solution: A transformed form of the adjoint Boltzmann equation in integral representation is solved for the space, energy, direction and time variables by Monte Carlo methods. Adjoint particles are defined with properties in some respects contrary to those of neutrons. Adjoint particle histories are constructed from which estimates are obtained of the desired quantity. Adjoint cross sections are defined with which the nuclide and reaction type are selected in a collision. The energy after a collision is selected from adjoint energy distributions calculated together with the adjoint cross sections in advance of the actual Monte Carlo calculation. For multiplying systems successive generations of adjoint particles are obtained which will die out for subcritical systems with a fixed neutron source and will be kept approximately stationary for Eigen function problems. Completely arbitrary problems can

  4. Particle transport due to magnetic fluctuations

    International Nuclear Information System (INIS)

    Stoneking, M.R.; Hokin, S.A.; Prager, S.C.; Fiksel, G.; Ji, H.; Den Hartog, D.J.

    1994-01-01

    Electron current fluctuations are measured with an electrostatic energy analyzer at the edge of the MST reversed-field pinch plasma. The radial flux of fast electrons (E>T e ) due to parallel streaming along a fluctuating magnetic field is determined locally by measuring the correlated product e B r >. Particle transport is small just inside the last closed flux surface (Γ e,mag e,total ), but can account for all observed particle losses inside r/a=0.8. Electron diffusion is found to increase with parallel velocity, as expected for diffusion in a region of field stochasticity

  5. The simulation status of particle transport system JPTS

    International Nuclear Information System (INIS)

    Deng, L.

    2015-01-01

    'Full text:' Particle transport system JPTS has been developed by IAPCM. It is based on the three support frustrations (JASMIN, JAUMIN and JCOGIN) and is used to simulate the reactor full core and radiation shielding problems. The system has been realized the high fidelity. In this presentation, analysis of the H-M, BEAVRS, VENUS-III and SG-III models are shown. Analyze HZP conditions of BEAVRS model with Monte Carlo code JMCT, MC21 and OpenMC to assess code accuracy against available data. Assess the feasibility of analysis of a PWR using JMCT. The large scale depletion solver is also shown. Assess the feasibility of analysis of radiation shielding using JSNT. JPTS has been proved with the capability of the full-core pin-by-pin and radiation shielding. (author)

  6. Discrete elements method of neutral particle transport

    International Nuclear Information System (INIS)

    Mathews, K.A.

    1983-01-01

    A new discrete elements (L/sub N/) transport method is derived and compared to the discrete ordinates S/sub N/ method, theoretically and by numerical experimentation. The discrete elements method is more accurate than discrete ordinates and strongly ameliorates ray effects for the practical problems studied. The discrete elements method is shown to be more cost effective, in terms of execution time with comparable storage to attain the same accuracy, for a one-dimensional test case using linear characteristic spatial quadrature. In a two-dimensional test case, a vacuum duct in a shield, L/sub N/ is more consistently convergent toward a Monte Carlo benchmark solution than S/sub N/, using step characteristic spatial quadrature. An analysis of the interaction of angular and spatial quadrature in xy-geometry indicates the desirability of using linear characteristic spatial quadrature with the L/sub N/ method

  7. Radiation protection studies for medical particle accelerators using FLUKA Monte Carlo code

    International Nuclear Information System (INIS)

    Infantino, Angelo; Mostacci, Domiziano; Cicoria, Gianfranco; Lucconi, Giulia; Pancaldi, Davide; Vichi, Sara; Zagni, Federico; Marengo, Mario

    2017-01-01

    Radiation protection (RP) in the use of medical cyclotrons involves many aspects both in the routine use and for the decommissioning of a site. Guidelines for site planning and installation, as well as for RP assessment, are given in international documents; however, the latter typically offer analytic methods of calculation of shielding and materials activation, in approximate or idealised geometry set-ups. The availability of Monte Carlo (MC) codes with accurate up-to-date libraries for transport and interaction of neutrons and charged particles at energies below 250 MeV, together with the continuously increasing power of modern computers, makes the systematic use of simulations with realistic geometries possible, yielding equipment and site-specific evaluation of the source terms, shielding requirements and all quantities relevant to RP at the same time. In this work, the well-known FLUKA MC code was used to simulate different aspects of RP in the use of biomedical accelerators, particularly for the production of medical radioisotopes. In the context of the Young Professionals Award, held at the IRPA 14 conference, only a part of the complete work is presented. In particular, the simulation of the GE PETtrace cyclotron (16.5 MeV) installed at S. Orsola-Malpighi University Hospital evaluated the effective dose distribution around the equipment; the effective number of neutrons produced per incident proton and their spectral distribution; the activation of the structure of the cyclotron and the vault walls; the activation of the ambient air, in particular the production of "4"1Ar. The simulations were validated, in terms of physical and transport parameters to be used at the energy range of interest, through an extensive measurement campaign of the neutron environmental dose equivalent using a rem-counter and TLD dosemeters. The validated model was then used in the design and the licensing request of a new Positron Emission Tomography facility. (authors)

  8. MCNP: a general Monte Carlo code for neutron and photon transport

    International Nuclear Information System (INIS)

    1979-11-01

    The general-purpose Monte Carlo code MCNP ca be used for neutron, photon, or coupled neutron-photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation are accounted for. Thermal neutrons are described by both the free-gas and S(α,β) models. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. MCNP includes an elaborate, interactive plotting capability that allows the user to view his input geometry to help check for setup errors. Standard features which are available to improve computational efficiency include geometry splitting and Russian roulette, weight cutoff with Russian roulette, correlated sampling, analog capture or capture by weight reduction, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point or ring detectors, deterministically transporting pseudo-particles to designated regions, track-length estimators, source biasing, and several parameter cutoffs. Extensive summary information is provided to help the user better understand the physics and Monte Carlo simulation of his problem. The standard, user-defined output of MCNP includes two-way current as a function of direction across any set of surfaces or surface segments in the problem. Flux across any set of surfaces or surface segments is available. 58 figures, 28 tables

  9. A unified Monte Carlo interpretation of particle simulations and applications to nonneutral plasmas

    International Nuclear Information System (INIS)

    Aydemir, A.Y.

    1993-09-01

    Using a ''Monte Carlo interpretation'' a particle simulations, a general description of low-noise techniques is developed in terms well-known Monte Carlo variance reduction methods. Some of these techniques then are applied to linear and nonlinear studies of pure electron plasmas in cylindrical geometry, with emphasis on the generation and nonlinear evolution of electron vortices. Long-lived l = 1 and l and l = 2 vortices, and others produced by unstable diocotron modes in hollow profiles, are studies. It is shown that low-noise techniques make it possible to follow the linear evolution and saturation of even the very weakly unstable resonant diocotron modes

  10. Solar energetic particle anisotropies and insights into particle transport

    Energy Technology Data Exchange (ETDEWEB)

    Leske, R. A., E-mail: ral@srl.caltech.edu; Cummings, A. C.; Cohen, C. M. S.; Mewaldt, R. A.; Labrador, A. W.; Stone, E. C. [California Institute of Technology, Pasadena, CA 91125 (United States); Wiedenbeck, M. E. [Jet Propulsion Laboratory, California Institute of Technology, Pasadena, CA 91109 (United States); Christian, E. R.; Rosenvinge, T. T. von [NASA/Goddard Space Flight Center, Greenbelt, MD 20771 (United States)

    2016-03-25

    As solar energetic particles (SEPs) travel through interplanetary space, their pitch-angle distributions are shaped by the competing effects of magnetic focusing and scattering. Measurements of SEP anisotropies can therefore reveal information about interplanetary conditions such as magnetic field strength, topology, and turbulence levels at remote locations from the observer. Onboard each of the two STEREO spacecraft, the Low Energy Telescope (LET) measures pitch-angle distributions for protons and heavier ions up to iron at energies of about 2-12 MeV/nucleon. Anisotropies observed using LET include bidirectional flows within interplanetary coronal mass ejections, sunward-flowing particles when STEREO was magnetically connected to the back side of a shock, and loss-cone distributions in which particles with large pitch angles underwent magnetic mirroring at an interplanetary field enhancement that was too weak to reflect particles with the smallest pitch angles. Unusual oscillations in the width of a beamed distribution at the onset of the 23 July 2012 SEP event were also observed and remain puzzling. We report LET anisotropy observations at both STEREO spacecraft and discuss their implications for SEP transport, focusing exclusively on the extreme event of 23 July 2012 in which a large variety of anisotropies were present at various times during the event.

  11. Solar energetic particle anisotropies and insights into particle transport

    Science.gov (United States)

    Leske, R. A.; Cummings, A. C.; Cohen, C. M. S.; Mewaldt, R. A.; Labrador, A. W.; Stone, E. C.; Wiedenbeck, M. E.; Christian, E. R.; Rosenvinge, T. T. von

    2016-03-01

    As solar energetic particles (SEPs) travel through interplanetary space, their pitch-angle distributions are shaped by the competing effects of magnetic focusing and scattering. Measurements of SEP anisotropies can therefore reveal information about interplanetary conditions such as magnetic field strength, topology, and turbulence levels at remote locations from the observer. Onboard each of the two STEREO spacecraft, the Low Energy Telescope (LET) measures pitch-angle distributions for protons and heavier ions up to iron at energies of about 2-12 MeV/nucleon. Anisotropies observed using LET include bidirectional flows within interplanetary coronal mass ejections, sunward-flowing particles when STEREO was magnetically connected to the back side of a shock, and loss-cone distributions in which particles with large pitch angles underwent magnetic mirroring at an interplanetary field enhancement that was too weak to reflect particles with the smallest pitch angles. Unusual oscillations in the width of a beamed distribution at the onset of the 23 July 2012 SEP event were also observed and remain puzzling. We report LET anisotropy observations at both STEREO spacecraft and discuss their implications for SEP transport, focusing exclusively on the extreme event of 23 July 2012 in which a large variety of anisotropies were present at various times during the event.

  12. Gyrokinetic particle simulation of neoclassical transport

    International Nuclear Information System (INIS)

    Lin, Z.; Tang, W.M.; Lee, W.W.

    1995-01-01

    A time varying weighting (δf ) scheme for gyrokinetic particle simulation is applied to a steady-state, multispecies simulation of neoclassical transport. Accurate collision operators conserving momentum and energy are developed and implemented. Simulation results using these operators are found to agree very well with neoclassical theory. For example, it is dynamically demonstrated that like-particle collisions produce no particle flux and that the neoclassical fluxes are ambipolar for an ion--electron plasma. An important physics feature of the present scheme is the introduction of toroidal flow to the simulations. Simulation results are in agreement with the existing analytical neoclassical theory. The poloidal electric field associated with toroidal mass flow is found to enhance density gradient-driven electron particle flux and the bootstrap current while reducing temperature gradient-driven flux and current. Finally, neoclassical theory in steep gradient profile relevant to the edge regime is examined by taking into account finite banana width effects. In general, in the present work a valuable new capability for studying important aspects of neoclassical transport inaccessible by conventional analytical calculation processes is demonstrated. copyright 1995 American Institute of Physics

  13. Modeling of neutron and photon transport in iron and concrete radiation shields by using Monte Carlo method

    CERN Document Server

    Žukauskaitėa, A; Plukienė, R; Ridikas, D

    2007-01-01

    Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 (AVF cyclotron of Research Center of Nuclear Physics, Osaka University, Japan) – γ-ray beams (1-10 MeV), HIMAC (heavy-ion synchrotron of the National Institute of Radiological Sciences in Chiba, Japan) and ISIS-800 (ISIS intensive spallation neutron source facility of the Rutherford Appleton laboratory, UK) – high energy neutron (20-800 MeV) transport in iron and concrete. The calculation results were then compared with experimental data.compared with experimental data.

  14. Meaningful timescales from Monte Carlo simulations of particle systems with hard-core interactions

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Liborio I., E-mail: liborio78@gmail.com

    2016-12-01

    A new Markov Chain Monte Carlo method for simulating the dynamics of particle systems characterized by hard-core interactions is introduced. In contrast to traditional Kinetic Monte Carlo approaches, where the state of the system is associated with minima in the energy landscape, in the proposed method, the state of the system is associated with the set of paths traveled by the atoms and the transition probabilities for an atom to be displaced are proportional to the corresponding velocities. In this way, the number of possible state-to-state transitions is reduced to a discrete set, and a direct link between the Monte Carlo time step and true physical time is naturally established. The resulting rejection-free algorithm is validated against event-driven molecular dynamics: the equilibrium and non-equilibrium dynamics of hard disks converge to the exact results with decreasing displacement size.

  15. A user's manual for the three-dimensional Monte Carlo transport code SPARTAN

    International Nuclear Information System (INIS)

    Bending, R.C.; Heffer, P.J.H.

    1975-09-01

    SPARTAN is a general-purpose Monte Carlo particle transport code intended for neutron or gamma transport problems in reactor physics, health physics, shielding, and safety studies. The code used a very general geometry system enabling a complex layout to be described and allows the user to obtain physics data from a number of different types of source library. Special tracking and scoring techniques are used to improve the quality of the results obtained. To enable users to run SPARTAN, brief descriptions of the facilities available in the code are given and full details of data input and job control language, as well as examples of complete calculations, are included. It is anticipated that changes may be made to SPARTAN from time to time, particularly in those parts of the code which deal with physics data processing. The load module is identified by a version number and implementation date, and updates of sections of this manual will be issued when significant changes are made to the code. (author)

  16. Adaptive sampling method in deep-penetration particle transport problem

    International Nuclear Information System (INIS)

    Wang Ruihong; Ji Zhicheng; Pei Lucheng

    2012-01-01

    Deep-penetration problem has been one of the difficult problems in shielding calculation with Monte Carlo method for several decades. In this paper, a kind of particle transport random walking system under the emission point as a sampling station is built. Then, an adaptive sampling scheme is derived for better solution with the achieved information. The main advantage of the adaptive scheme is to choose the most suitable sampling number from the emission point station to obtain the minimum value of the total cost in the process of the random walk. Further, the related importance sampling method is introduced. Its main principle is to define the importance function due to the particle state and to ensure the sampling number of the emission particle is proportional to the importance function. The numerical results show that the adaptive scheme under the emission point as a station could overcome the difficulty of underestimation of the result in some degree, and the adaptive importance sampling method gets satisfied results as well. (authors)

  17. MORSE-C, Neutron Transport, Gamma Transport for Criticality Calculation by Monte-Carlo Method

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description of program or function: MORSE-C is a Monte-Carlo code to solve the multiple energy group form of the Boltzmann transport equation in order to obtain the eigenvalue (multiplication) when fissionable materials are present. Cross sections for up to 100 energy groups may be employed. The angular scattering is treated by the usual Legendre expansion as used in the discrete ordinates codes. Up-scattering may be specified. The geometry is defined by relationships to general 1. or 2. degree surfaces. Array units may be specified. Output includes, besides the usual values of input quantities, plots of the geometry, calculated volumes and masses, and graphs of results to assist the user in determining the correctness of the problem's solution

  18. Monte Carlo calculations of electron transport on microcomputers

    International Nuclear Information System (INIS)

    Chung, Manho; Jester, W.A.; Levine, S.H.; Foderaro, A.H.

    1990-01-01

    In the work described in this paper, the Monte Carlo program ZEBRA, developed by Berber and Buxton, was converted to run on the Macintosh computer using Microsoft BASIC to reduce the cost of Monte Carlo calculations using microcomputers. Then the Eltran2 program was transferred to an IBM-compatible computer. Turbo BASIC and Microsoft Quick BASIC have been used on the IBM-compatible Tandy 4000SX computer. The paper shows the running speed of the Monte Carlo programs on the different computers, normalized to one for Eltran2 on the Macintosh-SE or Macintosh-Plus computer. Higher values refer to faster running times proportionally. Since Eltran2 is a one-dimensional program, it calculates energy deposited in a semi-infinite multilayer slab. Eltran2 has been modified to a two-dimensional program called Eltran3 to computer more accurately the case with a point source, a small detector, and a short source-to-detector distance. The running time of Eltran3 is about twice as long as that of Eltran2 for a similar case

  19. Particle modeling of transport of α-ray generated ion clusters in air

    International Nuclear Information System (INIS)

    Tong, Lizhu; Nanbu, Kenichi; Hirata, Yosuke; Izumi, Mikio; Miyamoto, Yasuaki; Yamaguchi, Hiromi

    2006-01-01

    A particle model is developed using the test-particle Monte Carlo method to study the transport properties of α-ray generated ion clusters in a flow of air. An efficient ion-molecule collision model is proposed to simulate the collisions between ion and air molecule. The simulations are performed for a steady state of ion transport in a circular pipe. In the steady state, generation of ions is balanced with such losses of ions as absorption of the measuring sensor or pipe wall and disappearance by positive-negative ion recombination. The calculated ion current to the measuring sensor agrees well with the previous measured data. (author)

  20. Microwave transport in EBT distribution manifolds using Monte Carlo ray-tracing techniques

    International Nuclear Information System (INIS)

    Lillie, R.A.; White, T.L.; Gabriel, T.A.; Alsmiller, R.G. Jr.

    1983-01-01

    Ray tracing Monte Carlo calculations have been carried out using an existing Monte Carlo radiation transport code to obtain estimates of the microsave power exiting the torus coupling links in EPT microwave manifolds. The microwave power loss and polarization at surface reflections were accounted for by treating the microwaves as plane waves reflecting off plane surfaces. Agreement on the order of 10% was obtained between the measured and calculated output power distribution for an existing EBT-S toroidal manifold. A cost effective iterative procedure utilizing the Monte Carlo history data was implemented to predict design changes which could produce increased manifold efficiency and improved output power uniformity

  1. Monte Carlo simulation of fast electrons and heavy particles in the CDS of nitrogen dc glow discharge

    International Nuclear Information System (INIS)

    Yu, W.; Zhang, L.Z.; Wang, J.L.; Han, L.; Fu, G.S.

    2001-01-01

    The characteristics of fast electrons (e - ) and heavy particles (N 2 + , N + , N 2f , N f ) in the cathode dark space (CDS) of nitrogen dc glow discharge are simultaneously studied by Monte Carlo simulation. The calculated energy and angular distributions of these particles at different positions from the cathode provide a clear picture of their transport behaviours within the CDS. The density and mean energy of these particles indicate that the electrons and the atomic ions (N + ) are the main high-energy species and the molecular ions (N 2 + ) are the major ions in the CDS. It can be seen from the energy distributions of the bombarding particles at the cathode surface that the molecular ions and the fast atoms (N f ) are the main active species participating in the cathode nitride material synthesis process. The influence of the backscattering of the electrons from the negative glow to the CDS is also investigated. All the calculated results provide good information on the spatial characteristics of the particles considered in this paper and also their internal connections in the CDS of nitrogen dc glow discharge. (author)

  2. Transport of Particle Swarms Through Fractures

    Science.gov (United States)

    Boomsma, E.; Pyrak-Nolte, L. J.

    2011-12-01

    The transport of engineered micro- and nano-scale particles through fractured rock is often assumed to occur as dispersions or emulsions. Another potential transport mechanism is the release of particle swarms from natural or industrial processes where small liquid drops, containing thousands to millions of colloidal-size particles, are released over time from seepage or leaks. Swarms have higher velocities than any individual colloid because the interactions among the particles maintain the cohesiveness of the swarm as it falls under gravity. Thus particle swarms give rise to the possibility that engineered particles may be transported farther and faster in fractures than predicted by traditional dispersion models. In this study, the effect of fractures on colloidal swarm cohesiveness and evolution was studied as a swarm falls under gravity and interacts with fracture walls. Transparent acrylic was used to fabricate synthetic fracture samples with either (1) a uniform aperture or (2) a converging aperture followed by a uniform aperture (funnel-shaped). The samples consisted of two blocks that measured 100 x 100 x 50 mm. The separation between these blocks determined the aperture (0.5 mm to 50 mm). During experiments, a fracture was fully submerged in water and swarms were released into it. The swarms consisted of dilute suspensions of either 25 micron soda-lime glass beads (2% by mass) or 3 micron polystyrene fluorescent beads (1% by mass) with an initial volume of 5μL. The swarms were illuminated with a green (525 nm) LED array and imaged optically with a CCD camera. In the uniform aperture fracture, the speed of the swarm prior to bifurcation increased with aperture up to a maximum at a fracture width of approximately 10 mm. For apertures greater than ~15 mm, the velocity was essentially constant with fracture width (but less than at 10 mm). This peak suggests that two competing mechanisms affect swarm velocity in fractures. The wall provides both drag, which

  3. Sawtooth driven particle transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Nicolas, T.

    2013-01-01

    The radial transport of particles in tokamaks is one of the most stringent issues faced by the magnetic confinement fusion community, because the fusion power is proportional to the square of the pressure, and also because accumulation of heavy impurities in the core leads to important power losses which can lead to a 'radiative collapse'. Sawteeth and the associated periodic redistribution of the core quantities can significantly impact the radial transport of electrons and impurities. In this thesis, we perform numerical simulations of sawteeth using a nonlinear tridimensional magnetohydrodynamic code called XTOR-2F to study the particle transport induced by sawtooth crashes. We show that the code recovers, after the crash, the fine structures of electron density that are observed with fast-sweeping reflectometry on the JET and TS tokamaks. The presence of these structure may indicate a low efficiency of the sawtooth in expelling the impurities from the core. However, applying the same code to impurity profiles, we show that the redistribution is quantitatively similar to that predicted by Kadomtsev's model, which could not be predicted a priori. Hence finally the sawtooth flushing is efficient in expelling impurities from the core. (author) [fr

  4. Particle-In-Cell/Monte Carlo Simulation of Ion Back Bombardment in Photoinjectors

    International Nuclear Information System (INIS)

    Qiang, Ji; Corlett, John; Staples, John

    2009-01-01

    In this paper, we report on studies of ion back bombardment in high average current dc and rf photoinjectors using a particle-in-cell/Monte Carlo method. Using H 2 ion as an example, we observed that the ion density and energy deposition on the photocathode in rf guns are order of magnitude lower than that in a dc gun. A higher rf frequency helps mitigate the ion back bombardment of the cathode in rf guns

  5. Correlation of electron transport and photocatalysis of nanocrystalline clusters studied by Monte-Carlo continuity random walking.

    Science.gov (United States)

    Liu, Baoshun; Li, Ziqiang; Zhao, Xiujian

    2015-02-21

    In this research, Monte-Carlo Continuity Random Walking (MC-RW) model was used to study the relation between electron transport and photocatalysis of nano-crystalline (nc) clusters. The effects of defect energy disorder, spatial disorder of material structure, electron density, and interfacial transfer/recombination on the electron transport and the photocatalysis were studied. Photocatalytic activity is defined as 1/τ from a statistical viewpoint with τ being the electron average lifetime. Based on the MC-RW simulation, a clear physical and chemical "picture" was given for the photocatalytic kinetic analysis of nc-clusters. It is shown that the increase of defect energy disorder and material spatial structural disorder, such as the decrease of defect trap number, the increase of crystallinity, the increase of particle size, and the increase of inter-particle connection, can enhance photocatalytic activity through increasing electron transport ability. The increase of electron density increases the electron Fermi level, which decreases the activation energy for electron de-trapping from traps to extending states, and correspondingly increases electron transport ability and photocatalytic activity. Reducing recombination of electrons and holes can increase electron transport through the increase of electron density and then increases the photocatalytic activity. In addition to the electron transport, the increase of probability for electrons to undergo photocatalysis can increase photocatalytic activity through the increase of the electron interfacial transfer speed.

  6. A fully coupled Monte Carlo/discrete ordinates solution to the neutron transport equation. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Baker, Randal Scott [Univ. of Arizona, Tucson, AZ (United States)

    1990-01-01

    The neutron transport equation is solved by a hybrid method that iteratively couples regions where deterministic (SN) and stochastic (Monte Carlo) methods are applied. Unlike previous hybrid methods, the Monte Carlo and SN regions are fully coupled in the sense that no assumption is made about geometrical separation or decoupling. The hybrid method provides a new means of solving problems involving both optically thick and optically thin regions that neither Monte Carlo nor SN is well suited for by themselves. The fully coupled Monte Carlo/SN technique consists of defining spatial and/or energy regions of a problem in which either a Monte Carlo calculation or an SN calculation is to be performed. The Monte Carlo region may comprise the entire spatial region for selected energy groups, or may consist of a rectangular area that is either completely or partially embedded in an arbitrary SN region. The Monte Carlo and SN regions are then connected through the common angular boundary fluxes, which are determined iteratively using the response matrix technique, and volumetric sources. The hybrid method has been implemented in the SN code TWODANT by adding special-purpose Monte Carlo subroutines to calculate the response matrices and volumetric sources, and linkage subrountines to carry out the interface flux iterations. The common angular boundary fluxes are included in the SN code as interior boundary sources, leaving the logic for the solution of the transport flux unchanged, while, with minor modifications, the diffusion synthetic accelerator remains effective in accelerating SN calculations. The special-purpose Monte Carlo routines used are essentially analog, with few variance reduction techniques employed. However, the routines have been successfully vectorized, with approximately a factor of five increase in speed over the non-vectorized version.

  7. MCNP: a general Monte Carlo code for neutron and photon transport

    International Nuclear Information System (INIS)

    1978-07-01

    The general-purpose Monte Carlo code MCNP can be used for neutron, photon, or coupled neutron--photon transport. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces (elliptical tori). Pointwise cross-section data are used. For neutrons, all reactions given in a particular cross-section evaluation (such as ENDF/B-IV) are accounted for. For photons, the code takes account of incoherent and coherent scattering, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. Standard optional variance reduction schemes include geometry splitting and Russian roulette, the exponential transformation, energy splitting, forced collisions in designated cells, flux estimates at point detectors, track-length estimators, and source biasing. The standard output of MCNP includes two-way current as a function of energy, time, and angle with the normal, across any subset of bounding surfaces in the problem. Fluxes across any set of bounding surfaces are available as a function of time and energy. Similarly, the flux at designated points and the average flux in a cell (track length per unit volume) are standard tallies. Reactions such as fissions or absorptions may be obtained in a subset of geometric cells. The heating tallies give the energy deposition per starting particle. In addition, particles may be flagged when they cross specified surfaces or enter designated cells, and the contributions of these flagged particles to certain of the tallies are listed separately. All quantities printed out have their relative errors listed also. 11 figures, 27 tables

  8. Particle filters, a quasi-Monte-Carlo-solution for segmentation of coronaries.

    Science.gov (United States)

    Florin, Charles; Paragios, Nikos; Williams, Jim

    2005-01-01

    In this paper we propose a Particle Filter-based approach for the segmentation of coronary arteries. To this end, successive planes of the vessel are modeled as unknown states of a sequential process. Such states consist of the orientation, position, shape model and appearance (in statistical terms) of the vessel that are recovered in an incremental fashion, using a sequential Bayesian filter (Particle Filter). In order to account for bifurcations and branchings, we consider a Monte Carlo sampling rule that propagates in parallel multiple hypotheses. Promising results on the segmentation of coronary arteries demonstrate the potential of the proposed approach.

  9. Particle Markov Chain Monte Carlo Techniques of Unobserved Component Time Series Models Using Ox

    DEFF Research Database (Denmark)

    Nonejad, Nima

    This paper details Particle Markov chain Monte Carlo techniques for analysis of unobserved component time series models using several economic data sets. PMCMC combines the particle filter with the Metropolis-Hastings algorithm. Overall PMCMC provides a very compelling, computationally fast...... and efficient framework for estimation. These advantages are used to for instance estimate stochastic volatility models with leverage effect or with Student-t distributed errors. We also model changing time series characteristics of the US inflation rate by considering a heteroskedastic ARFIMA model where...

  10. Introduction to the Latest Version of the Test-Particle Monte Carlo Code Molflow+

    CERN Document Server

    Ady, M

    2014-01-01

    The Test-Particle Monte Carlo code Molflow+ is getting more and more attention from the scientific community needing detailed 3D calculations of vacuum in the molecular flow regime mainly, but not limited to, the particle accelerator field. Substantial changes, bug fixes, geometry-editing and modelling features, and computational speed improvements have been made to the code in the last couple of years. This paper will outline some of these new features, and show examples of applications to the design and analysis of vacuum systems at CERN and elsewhere.

  11. Application of a Java-based, univel geometry, neutral particle Monte Carlo code to the searchlight problem

    International Nuclear Information System (INIS)

    Charles A. Wemple; Joshua J. Cogliati

    2005-01-01

    A univel geometry, neutral particle Monte Carlo transport code, written entirely in the Java programming language, is under development for medical radiotherapy applications. The code uses ENDF-VI based continuous energy cross section data in a flexible XML format. Full neutron-photon coupling, including detailed photon production and photonuclear reactions, is included. Charged particle equilibrium is assumed within the patient model so that detailed transport of electrons produced by photon interactions may be neglected. External beam and internal distributed source descriptions for mixed neutron-photon sources are allowed. Flux and dose tallies are performed on a univel basis. A four-tap, shift-register-sequence random number generator is used. Initial verification and validation testing of the basic neutron transport routines is underway. The searchlight problem was chosen as a suitable first application because of the simplicity of the physical model. Results show excellent agreement with analytic solutions. Computation times for similar numbers of histories are comparable to other neutron MC codes written in C and FORTRAN

  12. Classical trajectory Monte Carlo simulations of particle confinement using dual levitated coils

    Directory of Open Access Journals (Sweden)

    R. A. Lane

    2014-07-01

    Full Text Available The particle confinement properties of plasma confinement systems that employ dual levitated magnetic coils are investigated using classical trajectory Monte Carlo simulations. Two model systems are examined. In one, two identical current-carrying loops are coaxial and separated axially. In the second, two concentric and coplanar loops have different radii and carry equal currents. In both systems, a magnetic null circle is present between the current loops. Simulations are carried out for seven current loop separations for each system and at numerous values of magnetic field strength. Particle confinement is investigated at three locations between the loops at different distances from the magnetic null circle. Each simulated particle that did not escape the system exhibited one of four modes of confinement. Reduced results are given for both systems as the lowest magnetic field strength that exhibits complete confinement of all simulated particles for a particular loop separation.

  13. Monte Carlo calculation of secondary electron emission from carbon-surface by obliquely incident particles

    International Nuclear Information System (INIS)

    Ohya, Kaoru; Kawata, Jun; Mori, Ichiro

    1990-01-01

    Incidence angle dependences of secondary electron emission from a carbon surface by low energy electron and hydrogen atom are calculated using Monte Carlo simulations on the kinetic emission model. The calculation shows very small increase or rather decrease of the secondary electron yield with oblique incidence. It is explained in terms of not only multiple elastic collisions of incident particles with the carbon atoms but also small penetration depth of the particles comparable with the escape depth of secondary electrons. In addition, the two types of secondary electron emission are distinguished by using the secondary electron yield statistics; one is the emission due to trapped particles in the carbon, and the other is that due to backscattered particles. The high-yield component of the statistics on oblique incidence is more suppressed than those on normal incidence. (author)

  14. Particle transport in breathing quantum graph

    International Nuclear Information System (INIS)

    Matrasulov, D.U.; Yusupov, J.R.; Sabirov, K.K.; Sobirov, Z.A.

    2012-01-01

    Full text: Particle transport in nanoscale networks and discrete structures is of fundamental and practical importance. Usually such systems are modeled by so-called quantum graphs, the systems attracting much attention in physics and mathematics during past two decades [1-5]. During last two decades quantum graphs found numerous applications in modeling different discrete structures and networks in nanoscale and mesoscopic physics (e.g., see reviews [1-3]). Despite considerable progress made in the study of particle dynamics most of the problems deal with unperturbed case and the case of time-dependent perturbation has not yet be explored. In this work we treat particle dynamics for quantum star graph with time-dependent bonds. In particular, we consider harmonically breathing quantum star graphs, the cases of monotonically contracting and expanding graphs. The latter can be solved exactly analytically. Edge boundaries are considered to be time-dependent, while branching point is assumed to be fixed. Quantum dynamics of a particle in such graphs is studied by solving Schrodinger equation with time-dependent boundary conditions given on a star graph. Time-dependence of the average kinetic energy is analyzed. Space-time evolution of the Gaussian wave packet is treated for harmonically breathing star graph. It is found that for certain frequencies energy is a periodic function of time, while for others it can be non-monotonically growing function of time. Such a feature can be caused by possible synchronization of the particles motion and the motions of the moving edges of graph bonds. (authors) References: [1] Tsampikos Kottos and Uzy Smilansky, Ann. Phys., 76, 274 (1999). [2] Sven Gnutzmann and Uzy Smilansky, Adv. Phys. 55, 527 (2006). [3] S. GnutzmannJ.P. Keating, F. Piotet, Ann. Phys., 325, 2595 (2010). [4] P.Exner, P.Seba, P.Stovicek, J. Phys. A: Math. Gen. 21, 4009 (1988). [5] J. Boman, P. Kurasov, Adv. Appl. Math., 35, 58 (2005)

  15. Material motion corrections for implicit Monte Carlo radiation transport

    International Nuclear Information System (INIS)

    Gentile, N.A.; Morel, Jim E.

    2011-01-01

    We describe changes to the Implicit Monte Carlo (IMC) algorithm to include the effects of material motion. These changes assume that the problem can be embedded in a global Lorentz frame. We also assume that the material in each zone can be characterized by a single velocity. With this approximation, we show how to make IMC Lorentz invariant, so that the material motion corrections are correct to all orders of v/c. We develop thermal emission and face sources in moving material and discuss the coupling of IMC to the non- relativistic hydrodynamics equations via operator splitting. We discuss the effect of this coupling on the value of the 'Fleck factor' in IMC. (author)

  16. Spokes and charged particle transport in HiPIMS magnetrons

    International Nuclear Information System (INIS)

    Brenning, N; Lundin, D; Minea, T; Vitelaru, C; Costin, C

    2013-01-01

    Two separate scientific communities are shown to have studied one common phenomenon, azimuthally rotating dense plasma structures, also called spokes, in pulsed-power E × B discharges, starting from quite different approaches. The first body of work is motivated by fundamental plasma science and concerns a phenomenon called the critical ionization velocity, CIV, while the other body of work is motivated by the applied plasma science of high power impulse magnetron sputtering (HiPIMS). Here we make use of this situation by applying experimental observations, and theoretical analysis, from the CIV literature to HiPIMS discharges. For a practical example, we take data from observed spokes in HiPIMS discharges and focus on their role in charged particle transport, and in electron energization. We also touch upon the closely related questions of how they channel the cross-B discharge current, how they maintain their internal potential structure and how they influence the energy spectrum of the ions? New particle-in-cell Monte Carlo collisional simulations that shed light on the azimuthal drift and expansion of the spokes are also presented. (paper)

  17. A Monte Carlo Green's function method for three-dimensional neutron transport

    International Nuclear Information System (INIS)

    Gamino, R.G.; Brown, F.B.; Mendelson, M.R.

    1992-01-01

    This paper describes a Monte Carlo transport kernel capability, which has recently been incorporated into the RACER continuous-energy Monte Carlo code. The kernels represent a Green's function method for neutron transport from a fixed-source volume out to a particular volume of interest. This method is very powerful transport technique. Also, since kernels are evaluated numerically by Monte Carlo, the problem geometry can be arbitrarily complex, yet exact. This method is intended for problems where an ex-core neutron response must be determined for a variety of reactor conditions. Two examples are ex-core neutron detector response and vessel critical weld fast flux. The response is expressed in terms of neutron transport kernels weighted by a core fission source distribution. In these types of calculations, the response must be computed for hundreds of source distributions, but the kernels only need to be calculated once. The advance described in this paper is that the kernels are generated with a highly accurate three-dimensional Monte Carlo transport calculation instead of an approximate method such as line-of-sight attenuation theory or a synthesized three-dimensional discrete ordinates solution

  18. Monte Carlo simulation of VHTR particle fuel with chord length sampling

    International Nuclear Information System (INIS)

    Ji, W.; Martin, W. R.

    2007-01-01

    The Very High Temperature Gas-Cooled Reactor (VHTR) poses a problem for neutronic analysis due to the double heterogeneity posed by the particle fuel and either the fuel compacts in the case of the prismatic block reactor or the fuel pebbles in the case of the pebble bed reactor. Direct Monte Carlo simulation has been used in recent years to analyze these VHTR configurations but is computationally challenged when space dependent phenomena are considered such as depletion or temperature feedback. As an alternative approach, we have considered chord length sampling to reduce the computational burden of the Monte Carlo simulation. We have improved on an existing method called 'limited chord length sampling' and have used it to analyze stochastic media representative of either pebble bed or prismatic VHTR fuel geometries. Based on the assumption that the PDF had an exponential form, a theoretical chord length distribution is derived and shown to be an excellent model for a wide range of packing fractions. This chord length PDF was then used to analyze a stochastic medium that was constructed using the RSA (Random Sequential Addition) algorithm and the results were compared to a benchmark Monte Carlo simulation of the actual stochastic geometry. The results are promising and suggest that the theoretical chord length PDF can be used instead of a full Monte Carlo random walk simulation in the stochastic medium, saving orders of magnitude in computational time (and memory demand) to perform the simulation. (authors)

  19. The effects of particle recycling on the divertor plasma: A particle-in-cell with Monte Carlo collision simulation

    Science.gov (United States)

    Chang, Mingyu; Sang, Chaofeng; Sun, Zhenyue; Hu, Wanpeng; Wang, Dezhen

    2018-05-01

    A Particle-In-Cell (PIC) with Monte Carlo Collision (MCC) model is applied to study the effects of particle recycling on divertor plasma in the present work. The simulation domain is the scrape-off layer of the tokamak in one-dimension along the magnetic field line. At the divertor plate, the reflected deuterium atoms (D) and thermally released deuterium molecules (D2) are considered. The collisions between the plasma particles (e and D+) and recycled neutral particles (D and D2) are described by the MCC method. It is found that the recycled neutral particles have a great impact on divertor plasma. The effects of different collisions on the plasma are simulated and discussed. Moreover, the impacts of target materials on the plasma are simulated by comparing the divertor with Carbon (C) and Tungsten (W) targets. The simulation results show that the energy and momentum losses of the C target are larger than those of the W target in the divertor region even without considering the impurity particles, whereas the W target has a more remarkable influence on the core plasma.

  20. A study of Monte Carlo methods for weak approximations of stochastic particle systems in the mean-field?

    KAUST Repository

    Haji Ali, Abdul Lateef

    2016-01-01

    I discuss using single level and multilevel Monte Carlo methods to compute quantities of interests of a stochastic particle system in the mean-field. In this context, the stochastic particles follow a coupled system of Ito stochastic differential equations (SDEs). Moreover, this stochastic particle system converges to a stochastic mean-field limit as the number of particles tends to infinity. I start by recalling the results of applying different versions of Multilevel Monte Carlo (MLMC) for particle systems, both with respect to time steps and the number of particles and using a partitioning estimator. Next, I expand on these results by proposing the use of our recent Multi-index Monte Carlo method to obtain improved convergence rates.

  1. A study of Monte Carlo methods for weak approximations of stochastic particle systems in the mean-field?

    KAUST Repository

    Haji Ali, Abdul Lateef

    2016-01-08

    I discuss using single level and multilevel Monte Carlo methods to compute quantities of interests of a stochastic particle system in the mean-field. In this context, the stochastic particles follow a coupled system of Ito stochastic differential equations (SDEs). Moreover, this stochastic particle system converges to a stochastic mean-field limit as the number of particles tends to infinity. I start by recalling the results of applying different versions of Multilevel Monte Carlo (MLMC) for particle systems, both with respect to time steps and the number of particles and using a partitioning estimator. Next, I expand on these results by proposing the use of our recent Multi-index Monte Carlo method to obtain improved convergence rates.

  2. Robust volume calculations for Constructive Solid Geometry (CSG) components in Monte Carlo transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Millman, D. L. [Dept. of Computer Science, Univ. of North Carolina at Chapel Hill (United States); Griesheimer, D. P.; Nease, B. R. [Bechtel Marine Propulsion Corporation, Bertis Atomic Power Laboratory (United States); Snoeyink, J. [Dept. of Computer Science, Univ. of North Carolina at Chapel Hill (United States)

    2012-07-01

    In this paper we consider a new generalized algorithm for the efficient calculation of component object volumes given their equivalent constructive solid geometry (CSG) definition. The new method relies on domain decomposition to recursively subdivide the original component into smaller pieces with volumes that can be computed analytically or stochastically, if needed. Unlike simpler brute-force approaches, the proposed decomposition scheme is guaranteed to be robust and accurate to within a user-defined tolerance. The new algorithm is also fully general and can handle any valid CSG component definition, without the need for additional input from the user. The new technique has been specifically optimized to calculate volumes of component definitions commonly found in models used for Monte Carlo particle transport simulations for criticality safety and reactor analysis applications. However, the algorithm can be easily extended to any application which uses CSG representations for component objects. The paper provides a complete description of the novel volume calculation algorithm, along with a discussion of the conjectured error bounds on volumes calculated within the method. In addition, numerical results comparing the new algorithm with a standard stochastic volume calculation algorithm are presented for a series of problems spanning a range of representative component sizes and complexities. (authors)

  3. Robust volume calculations for Constructive Solid Geometry (CSG) components in Monte Carlo transport calculations

    International Nuclear Information System (INIS)

    Millman, D. L.; Griesheimer, D. P.; Nease, B. R.; Snoeyink, J.

    2012-01-01

    In this paper we consider a new generalized algorithm for the efficient calculation of component object volumes given their equivalent constructive solid geometry (CSG) definition. The new method relies on domain decomposition to recursively subdivide the original component into smaller pieces with volumes that can be computed analytically or stochastically, if needed. Unlike simpler brute-force approaches, the proposed decomposition scheme is guaranteed to be robust and accurate to within a user-defined tolerance. The new algorithm is also fully general and can handle any valid CSG component definition, without the need for additional input from the user. The new technique has been specifically optimized to calculate volumes of component definitions commonly found in models used for Monte Carlo particle transport simulations for criticality safety and reactor analysis applications. However, the algorithm can be easily extended to any application which uses CSG representations for component objects. The paper provides a complete description of the novel volume calculation algorithm, along with a discussion of the conjectured error bounds on volumes calculated within the method. In addition, numerical results comparing the new algorithm with a standard stochastic volume calculation algorithm are presented for a series of problems spanning a range of representative component sizes and complexities. (authors)

  4. Particle rejuvenation of Rao-Blackwellized sequential Monte Carlo smoothers for conditionally linear and Gaussian models

    Science.gov (United States)

    Nguyen, Ngoc Minh; Corff, Sylvain Le; Moulines, Éric

    2017-12-01

    This paper focuses on sequential Monte Carlo approximations of smoothing distributions in conditionally linear and Gaussian state spaces. To reduce Monte Carlo variance of smoothers, it is typical in these models to use Rao-Blackwellization: particle approximation is used to sample sequences of hidden regimes while the Gaussian states are explicitly integrated conditional on the sequence of regimes and observations, using variants of the Kalman filter/smoother. The first successful attempt to use Rao-Blackwellization for smoothing extends the Bryson-Frazier smoother for Gaussian linear state space models using the generalized two-filter formula together with Kalman filters/smoothers. More recently, a forward-backward decomposition of smoothing distributions mimicking the Rauch-Tung-Striebel smoother for the regimes combined with backward Kalman updates has been introduced. This paper investigates the benefit of introducing additional rejuvenation steps in all these algorithms to sample at each time instant new regimes conditional on the forward and backward particles. This defines particle-based approximations of the smoothing distributions whose support is not restricted to the set of particles sampled in the forward or backward filter. These procedures are applied to commodity markets which are described using a two-factor model based on the spot price and a convenience yield for crude oil data.

  5. Monte Carlo method for neutron transport calculations in graphics processing units (GPUs)

    International Nuclear Information System (INIS)

    Pellegrino, Esteban

    2011-01-01

    Monte Carlo simulation is well suited for solving the Boltzmann neutron transport equation in an inhomogeneous media for complicated geometries. However, routine applications require the computation time to be reduced to hours and even minutes in a desktop PC. The interest in adopting Graphics Processing Units (GPUs) for Monte Carlo acceleration is rapidly growing. This is due to the massive parallelism provided by the latest GPU technologies which is the most promising solution to the challenge of performing full-size reactor core analysis on a routine basis. In this study, Monte Carlo codes for a fixed-source neutron transport problem were developed for GPU environments in order to evaluate issues associated with computational speedup using GPUs. Results obtained in this work suggest that a speedup of several orders of magnitude is possible using the state-of-the-art GPU technologies. (author) [es

  6. Academic Training - The use of Monte Carlo radiation transport codes in radiation physics and dosimetry

    CERN Multimedia

    Françoise Benz

    2006-01-01

    2005-2006 ACADEMIC TRAINING PROGRAMME LECTURE SERIES 27, 28, 29 June 11:00-12:00 - TH Conference Room, bldg. 4 The use of Monte Carlo radiation transport codes in radiation physics and dosimetry F. Salvat Gavalda,Univ. de Barcelona, A. FERRARI, CERN-AB, M. SILARI, CERN-SC Lecture 1. Transport and interaction of electromagnetic radiation F. Salvat Gavalda,Univ. de Barcelona Interaction models and simulation schemes implemented in modern Monte Carlo codes for the simulation of coupled electron-photon transport will be briefly reviewed. Different schemes for simulating electron transport will be discussed. Condensed algorithms, which rely on multiple-scattering theories, are comparatively fast, but less accurate than mixed algorithms, in which hard interactions (with energy loss or angular deflection larger than certain cut-off values) are simulated individually. The reliability, and limitations, of electron-interaction models and multiple-scattering theories will be analyzed. Benchmark comparisons of simu...

  7. Modeling pollutant transport using a meshless-lagrangian particle model

    International Nuclear Information System (INIS)

    Carrington, D.B.; Pepper, D.W.

    2002-01-01

    A combined meshless-Lagrangian particle transport model is used to predict pollutant transport over irregular terrain. The numerical model for initializing the velocity field is based on a meshless approach utilizing multiquadrics established by Kansa. The Lagrangian particle transport technique uses a random walk procedure to depict the advection and dispersion of pollutants over any type of surface, including street and city canyons

  8. BALTORO a general purpose code for coupling discrete ordinates and Monte-Carlo radiation transport calculations

    International Nuclear Information System (INIS)

    Zazula, J.M.

    1983-01-01

    The general purpose code BALTORO was written for coupling the three-dimensional Monte-Carlo /MC/ with the one-dimensional Discrete Ordinates /DO/ radiation transport calculations. The quantity of a radiation-induced /neutrons or gamma-rays/ nuclear effect or the score from a radiation-yielding nuclear effect can be analysed in this way. (author)

  9. General-purpose Monte Carlo codes for neutron and photon transport calculations. MVP version 3

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu

    2017-01-01

    JAEA has developed a general-purpose neutron/photon transport Monte Carlo code MVP. This paper describes the recent development of the MVP code and reviews the basic features and capabilities. In addition, capabilities implemented in Version 3 are also described. (author)

  10. FMCEIR: a Monte Carlo program for solving the stationary neutron and gamma transport equation

    International Nuclear Information System (INIS)

    Taormina, A.

    1978-05-01

    FMCEIR is a three-dimensional Monte Carlo program for solving the stationary neutron and gamma transport equation. It is used to study the problem of neutron and gamma streaming in the GCFR and HHT reactor channels. (G.T.H.)

  11. Application of artificial intelligence techniques to the acceleration of Monte Carlo transport calculations

    International Nuclear Information System (INIS)

    Maconald, J.L.; Cashwell, E.D.

    1978-09-01

    The techniques of learning theory and pattern recognition are used to learn splitting surface locations for the Monte Carlo neutron transport code MCN. A study is performed to determine default values for several pattern recognition and learning parameters. The modified MCN code is used to reduce computer cost for several nontrivial example problems

  12. Analysis of Monte Carlo methods for the simulation of photon transport

    International Nuclear Information System (INIS)

    Carlsson, G.A.; Kusoffsky, L.

    1975-01-01

    In connection with the transport of low-energy photons (30 - 140 keV) through layers of water of different thicknesses, various aspects of Monte Carlo methods are examined in order to improve their effectivity (to produce statistically more reliable results with shorter computer times) and to bridge the gap between more physical methods and more mathematical ones. The calculations are compared with results of experiments involving the simulation of photon transport, using direct methods and collision density ones (J.S.)

  13. Application of an efficient materials perturbation technique to Monte Carlo photon transport calculations in borehole logging

    International Nuclear Information System (INIS)

    Picton, D.J.; Harris, R.G.; Randle, K.; Weaver, D.R.

    1995-01-01

    This paper describes a simple, accurate and efficient technique for the calculation of materials perturbation effects in Monte Carlo photon transport calculations. It is particularly suited to the application for which it was developed, namely the modelling of a dual detector density tool as used in borehole logging. However, the method would be appropriate to any photon transport calculation in the energy range 0.1 to 2 MeV, in which the predominant processes are Compton scattering and photoelectric absorption. The method enables a single set of particle histories to provide results for an array of configurations in which material densities or compositions vary. It can calculate the effects of small perturbations very accurately, but is by no means restricted to such cases. For the borehole logging application described here the method has been found to be efficient for a moderate range of variation in the bulk density (of the order of ±30% from a reference value) or even larger changes to a limited portion of the system (e.g. a low density mudcake of the order of a few tens of mm in thickness). The effective speed enhancement over an equivalent set of individual calculations is in the region of an order of magnitude or more. Examples of calculations on a dual detector density tool are given. It is demonstrated that the method predicts, to a high degree of accuracy, the variation of detector count rates with formation density, and that good results are also obtained for the effects of mudcake layers. An interesting feature of the results is that relative count rates (the ratios of count rates obtained with different configurations) can usually be determined more accurately than the absolute values of the count rates. (orig.)

  14. MCSLTT, Monte Carlo Simulation of Light Transport in Tissue

    International Nuclear Information System (INIS)

    2008-01-01

    Description of program or function: Understanding light-tissue interaction is fundamental in the field of Biomedical Optics. It has important implications for both therapeutic and diagnostic technologies. In this program, light transport in scattering tissue is modeled by absorption and scattering events as each photon travels through the tissue. The path of each photon is determined statistically by calculating probabilities of scattering and absorption. Other measured quantities are total reflected light, total transmitted light, and total heat absorbed

  15. Response matrix Monte Carlo based on a general geometry local calculation for electron transport

    International Nuclear Information System (INIS)

    Ballinger, C.T.; Rathkopf, J.A.; Martin, W.R.

    1991-01-01

    A Response Matrix Monte Carlo (RMMC) method has been developed for solving electron transport problems. This method was born of the need to have a reliable, computationally efficient transport method for low energy electrons (below a few hundred keV) in all materials. Today, condensed history methods are used which reduce the computation time by modeling the combined effect of many collisions but fail at low energy because of the assumptions required to characterize the electron scattering. Analog Monte Carlo simulations are prohibitively expensive since electrons undergo coulombic scattering with little state change after a collision. The RMMC method attempts to combine the accuracy of an analog Monte Carlo simulation with the speed of the condensed history methods. Like condensed history, the RMMC method uses probability distributions functions (PDFs) to describe the energy and direction of the electron after several collisions. However, unlike the condensed history method the PDFs are based on an analog Monte Carlo simulation over a small region. Condensed history theories require assumptions about the electron scattering to derive the PDFs for direction and energy. Thus the RMMC method samples from PDFs which more accurately represent the electron random walk. Results show good agreement between the RMMC method and analog Monte Carlo. 13 refs., 8 figs

  16. A Monte Carlo method using octree structure in photon and electron transport

    International Nuclear Information System (INIS)

    Ogawa, K.; Maeda, S.

    1995-01-01

    Most of the early Monte Carlo calculations in medical physics were used to calculate absorbed dose distributions, and detector responses and efficiencies. Recently, data acquisition in Single Photon Emission CT (SPECT) has been simulated by a Monte Carlo method to evaluate scatter photons generated in a human body and a collimator. Monte Carlo simulations in SPECT data acquisition are generally based on the transport of photons only because the photons being simulated are low energy, and therefore the bremsstrahlung productions by the electrons generated are negligible. Since the transport calculation of photons without electrons is much simpler than that with electrons, it is possible to accomplish the high-speed simulation in a simple object with one medium. Here, object description is important in performing the photon and/or electron transport using a Monte Carlo method efficiently. The authors propose a new description method using an octree representation of an object. Thus even if the boundaries of each medium are represented accurately, high-speed calculation of photon transport can be accomplished because the number of voxels is much fewer than that of the voxel-based approach which represents an object by a union of the voxels of the same size. This Monte Carlo code using the octree representation of an object first establishes the simulation geometry by reading octree string, which is produced by forming an octree structure from a set of serial sections for the object before the simulation; then it transports photons in the geometry. Using the code, if the user just prepares a set of serial sections for the object in which he or she wants to simulate photon trajectories, he or she can perform the simulation automatically using the suboptimal geometry simplified by the octree representation without forming the optimal geometry by handwriting

  17. A benchmark study of the Signed-particle Monte Carlo algorithm for the Wigner equation

    Directory of Open Access Journals (Sweden)

    Muscato Orazio

    2017-12-01

    Full Text Available The Wigner equation represents a promising model for the simulation of electronic nanodevices, which allows the comprehension and prediction of quantum mechanical phenomena in terms of quasi-distribution functions. During these years, a Monte Carlo technique for the solution of this kinetic equation has been developed, based on the generation and annihilation of signed particles. This technique can be deeply understood in terms of the theory of pure jump processes with a general state space, producing a class of stochastic algorithms. One of these algorithms has been validated successfully by numerical experiments on a benchmark test case.

  18. Continuous energy adjoint Monte Carlo for coupled neutron-photon transport

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J.E. [Delft Univ. of Technology (Netherlands). Interfaculty Reactor Inst.

    2001-07-01

    Although the theory for adjoint Monte Carlo calculations with continuous energy treatment for neutrons as well as for photons is known, coupled neutron-photon transport problems present fundamental difficulties because of the discrete energies of the photons produced by neutron reactions. This problem was solved by forcing the energy of the adjoint photon to the required discrete value by an adjoint Compton scattering reaction or an adjoint pair production reaction. A mathematical derivation shows the exact procedures to follow for the generation of an adjoint neutron and its statistical weight. A numerical example demonstrates that correct detector responses are obtained compared to a standard forward Monte Carlo calculation. (orig.)

  19. Particle transport simulation for spaceborne, NaI gamma-ray spectrometers

    International Nuclear Information System (INIS)

    Dyer, C.S.; Truscott, P.R.; Sims, A.J.; Comber, C.; Hammond, N.D.A.

    1988-11-01

    Radioactivity induced in detectors by protons and secondary neutrons limits the sensitivity of spaceborne gamma-ray spectrometers. Three dimensional Monte Carlo transport codes have been employed to simulate particle transport of cosmic rays and inner-belt protons in various representations of the Gamma Ray Observatory Spacecraft and the Oriented Scintillation Spectrometer Experiment. Results are used to accurately quantify the contributions to the radioactive background, assess shielding options and examine the effect of detector and space-craft orientation in anisotropic trapped proton fluxes. (author)

  20. Proposal of flexible atomic and molecular process management for Monte Carlo impurity transport code based on object oriented method

    International Nuclear Information System (INIS)

    Asano, K.; Ohno, N.; Takamura, S.

    2001-01-01

    Monte Carlo simulation code on impurity transport has been developed by several groups to be utilized mainly for fusion related edge plasmas. State of impurity particle is determined by atomic and molecular processes such as ionization, charge exchange in plasma. A lot of atomic and molecular processes have been considered because the edge plasma has not only impurity atoms, but also impurity molecules mainly related to chemical erosion of carbon materials, and their cross sections have been given experimentally and theoretically. We need to reveal which process is essential in a given edge plasma condition. Monte Carlo simulation code, which takes such various atomic and molecular processes into account, is necessary to investigate the behavior of impurity particle in plasmas. Usually, the impurity transport simulation code has been intended for some specific atomic and molecular processes so that the introduction of a new process forces complicated programming work. In order to evaluate various proposed atomic and molecular processes, a flexible management of atomic and molecular reaction should be established. We have developed the impurity transport simulation code based on object-oriented method. By employing object-oriented programming, we can handle each particle as 'object', which enfolds data and procedure function itself. A user (notice, not programmer) can define property of each particle species and the related atomic and molecular processes and then each 'object' is defined by analyzing this information. According to the relation among plasma particle species, objects are connected with each other and change their state by themselves. Dynamic allocation of these objects to program memory is employed to adapt for arbitrary number of species and atomic/molecular reactions. Thus we can treat arbitrary species and process starting from, for instance, methane and acetylene. Such a software procedure would be useful also for industrial application plasmas

  1. Systematic vacuum study of the ITER model cryopump by test particle Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Luo, Xueli; Haas, Horst; Day, Christian [Institute for Technical Physics, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2011-07-01

    The primary pumping systems on the ITER torus are based on eight tailor-made cryogenic pumps because not any standard commercial vacuum pump can meet the ITER working criteria. This kind of cryopump can provide high pumping speed, especially for light gases, by the cryosorption on activated charcoal at 4.5 K. In this paper we will present the systematic Monte Carlo simulation results of the model pump in a reduced scale by ProVac3D, a new Test Particle Monte Carlo simulation program developed by KIT. The simulation model has included the most important mechanical structures such as sixteen cryogenic panels working at 4.5 K, the 80 K radiation shield envelope with baffles, the pump housing, inlet valve and the TIMO (Test facility for the ITER Model Pump) test facility. Three typical gas species, i.e., deuterium, protium and helium are simulated. The pumping characteristics have been obtained. The result is in good agreement with the experiment data up to the gas throughput of 1000 sccm, which marks the limit for free molecular flow. This means that ProVac3D is a useful tool in the design of the prototype cryopump of ITER. Meanwhile, the capture factors at different critical positions are calculated. They can be used as the important input parameters for a follow-up Direct Simulation Monte Carlo (DSMC) simulation for higher gas throughput.

  2. An Analysis of Spherical Particles Distribution Randomly Packed in a Medium for the Monte Carlo Implicit Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Yong; Kim, Song Hyun; Shin, Chang Ho; Kim, Jong Kyung [Hanyang Univ., Seoul (Korea, Republic of)

    2014-05-15

    In this study, as a preliminary study to develop an implicit method having high accuracy, the distribution characteristics of spherical particles were evaluated by using explicit modeling techniques in various volume packing fractions. This study was performed to evaluate implicitly simulated distribution of randomly packed spheres in a medium. At first, an explicit modeling method to simulate random packed spheres in a hexahedron medium was proposed. The distributed characteristics of l{sub p} and r{sub p}, which are used in the particle position sampling, was estimated. It is analyzed that the use of the direct exponential distribution, which is generally used in the implicit modeling, can cause the distribution bias of the spheres. It is expected that the findings in this study can be utilized for improving the accuracy in using the implicit method. Spherical particles, which are randomly distributed in medium, are utilized for the radiation shields, fusion reactor blanket, fuels of VHTR reactors. Due to the difficulty on the simulation of the stochastic distribution, Monte Carlo (MC) method has been mainly considered as the tool for the analysis of the particle transport. For the MC modeling of the spherical particles, three methods are known; repeated structure, explicit modeling, and implicit modeling. Implicit method (called as the track length sampling method) is a modeling method that is the sampling based modeling technique of each spherical geometry (or track length of the sphere) during the MC simulation. Implicit modeling method has advantages in high computational efficiency and user convenience. However, it is noted that the implicit method has lower modeling accuracy in various finite mediums.

  3. High performance stream computing for particle beam transport simulations

    International Nuclear Information System (INIS)

    Appleby, R; Bailey, D; Higham, J; Salt, M

    2008-01-01

    Understanding modern particle accelerators requires simulating charged particle transport through the machine elements. These simulations can be very time consuming due to the large number of particles and the need to consider many turns of a circular machine. Stream computing offers an attractive way to dramatically improve the performance of such simulations by calculating the simultaneous transport of many particles using dedicated hardware. Modern Graphics Processing Units (GPUs) are powerful and affordable stream computing devices. The results of simulations of particle transport through the booster-to-storage-ring transfer line of the DIAMOND synchrotron light source using an NVidia GeForce 7900 GPU are compared to the standard transport code MAD. It is found that particle transport calculations are suitable for stream processing and large performance increases are possible. The accuracy and potential speed gains are compared and the prospects for future work in the area are discussed

  4. Mechanism of travelling-wave transport of particles

    International Nuclear Information System (INIS)

    Kawamoto, Hiroyuki; Seki, Kyogo; Kuromiya, Naoyuki

    2006-01-01

    Numerical and experimental investigations have been carried out on transport of particles in an electrostatic travelling field. A three-dimensional hard-sphere model of the distinct element method was developed to simulate the dynamics of particles. Forces applied to particles in the model were the Coulomb force, the dielectrophoresis force on polarized dipole particles in a non-uniform field, the image force, gravity and the air drag. Friction and repulsion between particle-particle and particle-conveyer were included in the model to replace initial conditions after mechanical contacts. Two kinds of experiments were performed to confirm the model. One was the measurement of charge of particles that is indispensable to determine the Coulomb force. Charge distribution was measured from the locus of free-fallen particles in a parallel electrostatic field. The averaged charge of the bulk particle was confirmed by measurement with a Faraday cage. The other experiment was measurements of the differential dynamics of particles on a conveyer consisting of parallel electrodes to which a four-phase travelling electrostatic wave was applied. Calculated results agreed with measurements, and the following characteristics were clarified. (1) The Coulomb force is the predominant force to drive particles compared with the other kinds of forces, (2) the direction of particle transport did not always coincide with that of the travelling wave but changed partially. It depended on the frequency of the travelling wave, the particle diameter and the electric field, (3) although some particles overtook the travelling wave at a very low frequency, the motion of particles was almost synchronized with the wave at the low frequency and (4) the transport of some particles was delayed to the wave at medium frequency; the majority of particles were transported backwards at high frequency and particles were not transported but only vibrated at very high frequency

  5. Electron transport in radiotherapy using local-to-global Monte Carlo

    International Nuclear Information System (INIS)

    Svatos, M.M.; Chandler, W.P.; Siantar, C.L.H.; Rathkopf, J.A.; Ballinger, C.T.

    1994-09-01

    Local-to-Global (L-G) Monte Carlo methods are a way to make three-dimensional electron transport both fast and accurate relative to other Monte Carlo methods. This is achieved by breaking the simulation into two stages: a local calculation done over small geometries having the size and shape of the ''steps'' to be taken through the mesh; and a global calculation which relies on a stepping code that samples the stored results of the local calculation. The increase in speed results from taking fewer steps in the global calculation than required by ordinary Monte Carlo codes and by speeding up the calculation per step. The potential for accuracy comes from the ability to use long runs of detailed codes to compile probability distribution functions (PDFs) in the local calculation. Specific examples of successful Local-to-Global algorithms are given

  6. Enhancing hydrologic data assimilation by evolutionary Particle Filter and Markov Chain Monte Carlo

    Science.gov (United States)

    Abbaszadeh, Peyman; Moradkhani, Hamid; Yan, Hongxiang

    2018-01-01

    Particle Filters (PFs) have received increasing attention by researchers from different disciplines including the hydro-geosciences, as an effective tool to improve model predictions in nonlinear and non-Gaussian dynamical systems. The implication of dual state and parameter estimation using the PFs in hydrology has evolved since 2005 from the PF-SIR (sampling importance resampling) to PF-MCMC (Markov Chain Monte Carlo), and now to the most effective and robust framework through evolutionary PF approach based on Genetic Algorithm (GA) and MCMC, the so-called EPFM. In this framework, the prior distribution undergoes an evolutionary process based on the designed mutation and crossover operators of GA. The merit of this approach is that the particles move to an appropriate position by using the GA optimization and then the number of effective particles is increased by means of MCMC, whereby the particle degeneracy is avoided and the particle diversity is improved. In this study, the usefulness and effectiveness of the proposed EPFM is investigated by applying the technique on a conceptual and highly nonlinear hydrologic model over four river basins located in different climate and geographical regions of the United States. Both synthetic and real case studies demonstrate that the EPFM improves both the state and parameter estimation more effectively and reliably as compared with the PF-MCMC.

  7. Monte Carlo simulation of the spectral response of beta-particle emitters in LSC systems

    International Nuclear Information System (INIS)

    Ortiz, F.; Los Arcos, J.M.; Grau, A.; Rodriguez, L.

    1992-01-01

    This paper presents a new method to evaluate the counting efficiency and the effective spectra at the output of any dynodic stage, for any pure beta-particle emitter, measured in a liquid scintillation counting system with two photomultipliers working in sum-coincidence mode. The process is carried out by a Monte Carlo simulation procedure that gives the electron distribution, and consequently the counting efficiency, at any dynode, in response to the beta particles emitted, as a function of the figure of merit of the system and the dynodic gains. The spectral outputs for 3 H and 14 C have been computed and compared with experimental data obtained with two sets of quenched radioactive standards of these nuclides. (orig.)

  8. A Generalized Boltzmann Fokker-Planck Method for Coupled Charged Particle Transport

    Energy Technology Data Exchange (ETDEWEB)

    Prinja, Anil K

    2012-01-09

    The goal of this project was to develop and investigate the performance of reduced-physics formulations of high energy charged particle (electrons, protons and heavier ions) transport that are computationally more efficient than not only analog Monte Carlo methods but also the established condensed history Monte Carlo technique. Charged particles interact with matter by Coulomb collisions with target nuclei and electrons, by bremsstrahlung radiation loss and by nuclear reactions such as spallation and fission. Of these, inelastic electronic collisions and elastic nuclear collisions are the dominant cause of energy-loss straggling and angular deflection or range straggling of a primary particle. These collisions are characterized by extremely short mean free paths (sub-microns) and highly peaked, near-singular differential cross sections about forward directions and zero energy loss, with the situation for protons and heavier ions more extreme than for electrons. For this reason, analog or truephysics single-event Monte Carlo simulation, while possible in principle, is computationally prohibitive for routine calculation of charged particle interaction phenomena.

  9. Ant colony algorithm implementation in electron and photon Monte Carlo transport: Application to the commissioning of radiosurgery photon beams

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Pareja, S.; Galan, P.; Manzano, F.; Brualla, L.; Lallena, A. M. [Servicio de Radiofisica Hospitalaria, Hospital Regional Universitario ' ' Carlos Haya' ' , Avda. Carlos Haya s/n, E-29010 Malaga (Spain); Unidad de Radiofisica Hospitalaria, Hospital Xanit Internacional, Avda. de los Argonautas s/n, E-29630 Benalmadena (Malaga) (Spain); NCTeam, Strahlenklinik, Universitaetsklinikum Essen, Hufelandstr. 55, D-45122 Essen (Germany); Departamento de Fisica Atomica, Molecular y Nuclear, Universidad de Granada, E-18071 Granada (Spain)

    2010-07-15

    Purpose: In this work, the authors describe an approach which has been developed to drive the application of different variance-reduction techniques to the Monte Carlo simulation of photon and electron transport in clinical accelerators. Methods: The new approach considers the following techniques: Russian roulette, splitting, a modified version of the directional bremsstrahlung splitting, and the azimuthal particle redistribution. Their application is controlled by an ant colony algorithm based on an importance map. Results: The procedure has been applied to radiosurgery beams. Specifically, the authors have calculated depth-dose profiles, off-axis ratios, and output factors, quantities usually considered in the commissioning of these beams. The agreement between Monte Carlo results and the corresponding measurements is within {approx}3%/0.3 mm for the central axis percentage depth dose and the dose profiles. The importance map generated in the calculation can be used to discuss simulation details in the different parts of the geometry in a simple way. The simulation CPU times are comparable to those needed within other approaches common in this field. Conclusions: The new approach is competitive with those previously used in this kind of problems (PSF generation or source models) and has some practical advantages that make it to be a good tool to simulate the radiation transport in problems where the quantities of interest are difficult to obtain because of low statistics.

  10. Ant colony algorithm implementation in electron and photon Monte Carlo transport: Application to the commissioning of radiosurgery photon beams

    International Nuclear Information System (INIS)

    Garcia-Pareja, S.; Galan, P.; Manzano, F.; Brualla, L.; Lallena, A. M.

    2010-01-01

    Purpose: In this work, the authors describe an approach which has been developed to drive the application of different variance-reduction techniques to the Monte Carlo simulation of photon and electron transport in clinical accelerators. Methods: The new approach considers the following techniques: Russian roulette, splitting, a modified version of the directional bremsstrahlung splitting, and the azimuthal particle redistribution. Their application is controlled by an ant colony algorithm based on an importance map. Results: The procedure has been applied to radiosurgery beams. Specifically, the authors have calculated depth-dose profiles, off-axis ratios, and output factors, quantities usually considered in the commissioning of these beams. The agreement between Monte Carlo results and the corresponding measurements is within ∼3%/0.3 mm for the central axis percentage depth dose and the dose profiles. The importance map generated in the calculation can be used to discuss simulation details in the different parts of the geometry in a simple way. The simulation CPU times are comparable to those needed within other approaches common in this field. Conclusions: The new approach is competitive with those previously used in this kind of problems (PSF generation or source models) and has some practical advantages that make it to be a good tool to simulate the radiation transport in problems where the quantities of interest are difficult to obtain because of low statistics.

  11. Ant colony algorithm implementation in electron and photon Monte Carlo transport: application to the commissioning of radiosurgery photon beams.

    Science.gov (United States)

    García-Pareja, S; Galán, P; Manzano, F; Brualla, L; Lallena, A M

    2010-07-01

    In this work, the authors describe an approach which has been developed to drive the application of different variance-reduction techniques to the Monte Carlo simulation of photon and electron transport in clinical accelerators. The new approach considers the following techniques: Russian roulette, splitting, a modified version of the directional bremsstrahlung splitting, and the azimuthal particle redistribution. Their application is controlled by an ant colony algorithm based on an importance map. The procedure has been applied to radiosurgery beams. Specifically, the authors have calculated depth-dose profiles, off-axis ratios, and output factors, quantities usually considered in the commissioning of these beams. The agreement between Monte Carlo results and the corresponding measurements is within approximately 3%/0.3 mm for the central axis percentage depth dose and the dose profiles. The importance map generated in the calculation can be used to discuss simulation details in the different parts of the geometry in a simple way. The simulation CPU times are comparable to those needed within other approaches common in this field. The new approach is competitive with those previously used in this kind of problems (PSF generation or source models) and has some practical advantages that make it to be a good tool to simulate the radiation transport in problems where the quantities of interest are difficult to obtain because of low statistics.

  12. Overview and applications of the Monte Carlo radiation transport kit at LLNL

    International Nuclear Information System (INIS)

    Sale, K. E.

    1999-01-01

    Modern Monte Carlo radiation transport codes can be applied to model most applications of radiation, from optical to TeV photons, from thermal neutrons to heavy ions. Simulations can include any desired level of detail in three-dimensional geometries using the right level of detail in the reaction physics. The technology areas to which we have applied these codes include medical applications, defense, safety and security programs, nuclear safeguards and industrial and research system design and control. The main reason such applications are interesting is that by using these tools substantial savings of time and effort (i.e. money) can be realized. In addition it is possible to separate out and investigate computationally effects which can not be isolated and studied in experiments. In model calculations, just as in real life, one must take care in order to get the correct answer to the right question. Advancing computing technology allows extensions of Monte Carlo applications in two directions. First, as computers become more powerful more problems can be accurately modeled. Second, as computing power becomes cheaper Monte Carlo methods become accessible more widely. An overview of the set of Monte Carlo radiation transport tools in use a LLNL will be presented along with a few examples of applications and future directions

  13. Entropic Ratchet transport of interacting active Brownian particles

    International Nuclear Information System (INIS)

    Ai, Bao-Quan; He, Ya-Feng; Zhong, Wei-Rong

    2014-01-01

    Directed transport of interacting active (self-propelled) Brownian particles is numerically investigated in confined geometries (entropic barriers). The self-propelled velocity can break thermodynamical equilibrium and induce the directed transport. It is found that the interaction between active particles can greatly affect the ratchet transport. For attractive particles, on increasing the interaction strength, the average velocity first decreases to its minima, then increases, and finally decreases to zero. For repulsive particles, when the interaction is very weak, there exists a critical interaction at which the average velocity is minimal, nearly tends to zero, however, for the strong interaction, the average velocity is independent of the interaction

  14. Entropic Ratchet transport of interacting active Brownian particles

    Energy Technology Data Exchange (ETDEWEB)

    Ai, Bao-Quan, E-mail: aibq@hotmail.com [Laboratory of Quantum Engineering and Quantum Materials, School of Physics and Telecommunication Engineering, South China Normal University, 510006 Guangzhou (China); He, Ya-Feng [College of Physics Science and Technology, Hebei University, 071002 Baoding (China); Zhong, Wei-Rong, E-mail: wrzhong@jnu.edu.cn [Department of Physics and Siyuan Laboratory, College of Science and Engineering, Jinan University, 510632 Guangzhou (China)

    2014-11-21

    Directed transport of interacting active (self-propelled) Brownian particles is numerically investigated in confined geometries (entropic barriers). The self-propelled velocity can break thermodynamical equilibrium and induce the directed transport. It is found that the interaction between active particles can greatly affect the ratchet transport. For attractive particles, on increasing the interaction strength, the average velocity first decreases to its minima, then increases, and finally decreases to zero. For repulsive particles, when the interaction is very weak, there exists a critical interaction at which the average velocity is minimal, nearly tends to zero, however, for the strong interaction, the average velocity is independent of the interaction.

  15. Implementation, capabilities, and benchmarking of Shift, a massively parallel Monte Carlo radiation transport code

    International Nuclear Information System (INIS)

    Pandya, Tara M.; Johnson, Seth R.; Evans, Thomas M.; Davidson, Gregory G.; Hamilton, Steven P.; Godfrey, Andrew T.

    2015-01-01

    This paper discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package developed and maintained at Oak Ridge National Laboratory. It has been developed to scale well from laptop to small computing clusters to advanced supercomputers. Special features of Shift include hybrid capabilities for variance reduction such as CADIS and FW-CADIS, and advanced parallel decomposition and tally methods optimized for scalability on supercomputing architectures. Shift has been validated and verified against various reactor physics benchmarks and compares well to other state-of-the-art Monte Carlo radiation transport codes such as MCNP5, CE KENO-VI, and OpenMC. Some specific benchmarks used for verification and validation include the CASL VERA criticality test suite and several Westinghouse AP1000 ® problems. These benchmark and scaling studies show promising results

  16. Implicit Monte Carlo methods and non-equilibrium Marshak wave radiative transport

    International Nuclear Information System (INIS)

    Lynch, J.E.

    1985-01-01

    Two enhancements to the Fleck implicit Monte Carlo method for radiative transport are described, for use in transparent and opaque media respectively. The first introduces a spectral mean cross section, which applies to pseudoscattering in transparent regions with a high frequency incident spectrum. The second provides a simple Monte Carlo random walk method for opaque regions, without the need for a supplementary diffusion equation formulation. A time-dependent transport Marshak wave problem of radiative transfer, in which a non-equilibrium condition exists between the radiation and material energy fields, is then solved. These results are compared to published benchmark solutions and to new discrete ordinate S-N results, for both spatially integrated radiation-material energies versus time and to new spatially dependent temperature profiles. Multigroup opacities, which are independent of both temperature and frequency, are used in addition to a material specific heat which is proportional to the cube of the temperature. 7 refs., 4 figs

  17. Transport of the moving barrier driven by chiral active particles

    Science.gov (United States)

    Liao, Jing-jing; Huang, Xiao-qun; Ai, Bao-quan

    2018-03-01

    Transport of a moving V-shaped barrier exposed to a bath of chiral active particles is investigated in a two-dimensional channel. Due to the chirality of active particles and the transversal asymmetry of the barrier position, active particles can power and steer the directed transport of the barrier in the longitudinal direction. The transport of the barrier is determined by the chirality of active particles. The moving barrier and active particles move in the opposite directions. The average velocity of the barrier is much larger than that of active particles. There exist optimal parameters (the chirality, the self-propulsion speed, the packing fraction, and the channel width) at which the average velocity of the barrier takes its maximal value. In particular, tailoring the geometry of the barrier and the active concentration provides novel strategies to control the transport properties of micro-objects or cargoes in an active medium.

  18. Monte Carlo simulation of nonlinear reactive contaminant transport in unsaturated porous media

    International Nuclear Information System (INIS)

    Giacobbo, F.; Patelli, E.

    2007-01-01

    In the current proposed solutions of radioactive waste repositories, the protective function against the radionuclide water-driven transport back to the biosphere is to be provided by an integrated system of engineered and natural geologic barriers. The occurrence of several nonlinear interactions during the radionuclide migration process may render burdensome the classical analytical-numerical approaches. Moreover, the heterogeneity of the barriers' media forces approximations to the classical analytical-numerical models, thus reducing their fidelity to reality. In an attempt to overcome these difficulties, in the present paper we adopt a Monte Carlo simulation approach, previously developed on the basis of the Kolmogorov-Dmitriev theory of branching stochastic processes. The approach is here extended for describing transport through unsaturated porous media under transient flow conditions and in presence of nonlinear interchange phenomena between the liquid and solid phases. This generalization entails the determination of the functional dependence of the parameters of the proposed transport model from the water content and from the contaminant concentration, which change in space and time during the water infiltration process. The corresponding Monte Carlo simulation approach is verified with respect to a case of nonreactive transport under transient unsaturated flow and to a case of nonlinear reactive transport under stationary saturated flow. Numerical applications regarding linear and nonlinear reactive transport under transient unsaturated flow are reported

  19. Monte Carlo study of the mechanisms of transport of fast neutrons in various media

    International Nuclear Information System (INIS)

    Ku, L.

    1976-01-01

    The technique of analyzing Monte Carlo histories was used to study the details of neutron transport and slowing down mechanisms. The statistical properties of life histories of ''exceptional'' neutrons, i.e., those staying closer to the source or penetrating to larger distances from the source, were compared to those of the general population. The macroscopic behavior of ''exceptional'' neutrons was also related to the interaction mechanics and to the microscopic properties of the medium

  20. Automatic modeling for the Monte Carlo transport code Geant4 in MCAM

    International Nuclear Information System (INIS)

    Nie Fanzhi; Hu Liqin; Wang Guozhong; Wang Dianxi; Wu Yican; Wang Dong; Long Pengcheng; FDS Team

    2014-01-01

    Geant4 is a widely used Monte Carlo transport simulation package. Its geometry models could be described in geometry description markup language (GDML), but it is time-consuming and error-prone to describe the geometry models manually. This study implemented the conversion between computer-aided design (CAD) geometry models and GDML models. The conversion program was integrated into Multi-Physics Coupling Analysis Modeling Program (MCAM). The tests, including FDS-Ⅱ model, demonstrated its accuracy and feasibility. (authors)

  1. Systems guide to MCNP (Monte Carlo Neutron and Photon Transport Code)

    International Nuclear Information System (INIS)

    Kirk, B.L.; West, J.T.

    1984-06-01

    The subject of this report is the implementation of the Los Alamos National Laboratory Monte Carlo Neutron and Photon Transport Code - Version 3 (MCNP) on the different types of computer systems, especially the IBM MVS system. The report supplements the documentation of the RSIC computer code package CCC-200/MCNP. Details of the procedure to follow in executing MCNP on the IBM computers, either in batch mode or interactive mode, are provided

  2. Modular, object-oriented redesign of a large-scale Monte Carlo neutron transport program

    International Nuclear Information System (INIS)

    Moskowitz, B.S.

    2000-01-01

    This paper describes the modular, object-oriented redesign of a large-scale Monte Carlo neutron transport program. This effort represents a complete 'white sheet of paper' rewrite of the code. In this paper, the motivation driving this project, the design objectives for the new version of the program, and the design choices and their consequences will be discussed. The design itself will also be described, including the important subsystems as well as the key classes within those subsystems

  3. Modeling of Particle Transport on Channels and Gaps Exposed to Plasma Fluxes

    International Nuclear Information System (INIS)

    Nieto-Perez, Martin

    2008-01-01

    Many problems in particle transport in fusion devices involve the transport of plasma or eroded particles through channels or gaps, such as in the case of trying to assess damage to delicate optical diagnostics collecting light through a slit or determining the deposition and codeposition on the gaps between tiles of plasma-facing components. A dynamic-composition Monte Carlo code in the spirit of TRIDYN, previously developed to study composition changes on optical mirrors subject to ion bombardment, has been upgraded to include motion of particles through a volume defined by sets of plane surfaces. Particles sputtered or reflected from the walls of the channel/gap can be tracked as well, allowing the calculation of wall impurity transport, either back to the plasma (for the case of a gap) or to components separated from the plasma by a channel/slit (for the case of optical diagnostics). Two examples of the code application to particle transport in fusion devices will be presented in this work: one will evaluate the erosion/impurity deposition rate on a mirror separated from a plasma source by a slit; the other case will look at the enhanced emission of tile material in the region of the gap between two tiles

  4. A vectorized Monte Carlo code for modeling photon transport in SPECT

    International Nuclear Information System (INIS)

    Smith, M.F.; Floyd, C.E. Jr.; Jaszczak, R.J.

    1993-01-01

    A vectorized Monte Carlo computer code has been developed for modeling photon transport in single photon emission computed tomography (SPECT). The code models photon transport in a uniform attenuating region and photon detection by a gamma camera. It is adapted from a history-based Monte Carlo code in which photon history data are stored in scalar variables and photon histories are computed sequentially. The vectorized code is written in FORTRAN77 and uses an event-based algorithm in which photon history data are stored in arrays and photon history computations are performed within DO loops. The indices of the DO loops range over the number of photon histories, and these loops may take advantage of the vector processing unit of our Stellar GS1000 computer for pipelined computations. Without the use of the vector processor the event-based code is faster than the history-based code because of numerical optimization performed during conversion to the event-based algorithm. When only the detection of unscattered photons is modeled, the event-based code executes 5.1 times faster with the use of the vector processor than without; when the detection of scattered and unscattered photons is modeled the speed increase is a factor of 2.9. Vectorization is a valuable way to increase the performance of Monte Carlo code for modeling photon transport in SPECT

  5. Exponentially-convergent Monte Carlo for the 1-D transport equation

    International Nuclear Information System (INIS)

    Peterson, J. R.; Morel, J. E.; Ragusa, J. C.

    2013-01-01

    We define a new exponentially-convergent Monte Carlo method for solving the one-speed 1-D slab-geometry transport equation. This method is based upon the use of a linear discontinuous finite-element trial space in space and direction to represent the transport solution. A space-direction h-adaptive algorithm is employed to restore exponential convergence after stagnation occurs due to inadequate trial-space resolution. This methods uses jumps in the solution at cell interfaces as an error indicator. Computational results are presented demonstrating the efficacy of the new approach. (authors)

  6. Efficiency of rejection-free methods for dynamic Monte Carlo studies of off-lattice interacting particles

    KAUST Repository

    Guerra, Marta L.; Novotny, M. A.; Watanabe, Hiroshi; Ito, Nobuyasu

    2009-01-01

    We calculate the efficiency of a rejection-free dynamic Monte Carlo method for d -dimensional off-lattice homogeneous particles interacting through a repulsive power-law potential r-p. Theoretically we find the algorithmic efficiency in the limit of low temperatures and/or high densities is asymptotically proportional to ρ (p+2) /2 T-d/2 with the particle density ρ and the temperature T. Dynamic Monte Carlo simulations are performed in one-, two-, and three-dimensional systems with different powers p, and the results agree with the theoretical predictions. © 2009 The American Physical Society.

  7. Efficiency of rejection-free methods for dynamic Monte Carlo studies of off-lattice interacting particles

    KAUST Repository

    Guerra, Marta L.

    2009-02-23

    We calculate the efficiency of a rejection-free dynamic Monte Carlo method for d -dimensional off-lattice homogeneous particles interacting through a repulsive power-law potential r-p. Theoretically we find the algorithmic efficiency in the limit of low temperatures and/or high densities is asymptotically proportional to ρ (p+2) /2 T-d/2 with the particle density ρ and the temperature T. Dynamic Monte Carlo simulations are performed in one-, two-, and three-dimensional systems with different powers p, and the results agree with the theoretical predictions. © 2009 The American Physical Society.

  8. Massively parallel Monte Carlo for many-particle simulations on GPUs

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Joshua A.; Jankowski, Eric [Department of Chemical Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Grubb, Thomas L. [Department of Materials Science and Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Engel, Michael [Department of Chemical Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Glotzer, Sharon C., E-mail: sglotzer@umich.edu [Department of Chemical Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Department of Materials Science and Engineering, University of Michigan, Ann Arbor, MI 48109 (United States)

    2013-12-01

    Current trends in parallel processors call for the design of efficient massively parallel algorithms for scientific computing. Parallel algorithms for Monte Carlo simulations of thermodynamic ensembles of particles have received little attention because of the inherent serial nature of the statistical sampling. In this paper, we present a massively parallel method that obeys detailed balance and implement it for a system of hard disks on the GPU. We reproduce results of serial high-precision Monte Carlo runs to verify the method. This is a good test case because the hard disk equation of state over the range where the liquid transforms into the solid is particularly sensitive to small deviations away from the balance conditions. On a Tesla K20, our GPU implementation executes over one billion trial moves per second, which is 148 times faster than on a single Intel Xeon E5540 CPU core, enables 27 times better performance per dollar, and cuts energy usage by a factor of 13. With this improved performance we are able to calculate the equation of state for systems of up to one million hard disks. These large system sizes are required in order to probe the nature of the melting transition, which has been debated for the last forty years. In this paper we present the details of our computational method, and discuss the thermodynamics of hard disks separately in a companion paper.

  9. Alpha particle density and energy distributions in tandem mirrors using Monte-Carlo techniques

    International Nuclear Information System (INIS)

    Kerns, J.A.

    1986-05-01

    We have simulated the alpha thermalization process using a Monte-Carlo technique, in which the alpha guiding center is followed between simulated collisions and Spitzer's collision model is used for the alpha-plasma interaction. Monte-Carlo techniques are used to determine the alpha radial birth position, the alpha particle position at a collision, and the angle scatter and dispersion at a collision. The plasma is modeled as a hot reacting core, surrounded by a cold halo plasma (T approx.50 eV). Alpha orbits that intersect the halo lose 90% of their energy to the halo electrons because of the halo drag, which is ten times greater than the drag in the core. The uneven drag across the alpha orbit also produces an outward, radial, guiding center drift. This drag drift is dependent on the plasma density and temperature radial profiles. We have modeled these profiles and have specifically studied a single-scale-length model, in which the density scale length (r/sub pD/) equals the temperature scale length (r/sub pT/), and a two-scale-length model, in which r/sub pD//r/sub pT/ = 1.1

  10. Multilevel Monte Carlo methods using ensemble level mixed MsFEM for two-phase flow and transport simulations

    KAUST Repository

    Efendiev, Yalchin R.; Iliev, Oleg; Kronsbein, C.

    2013-01-01

    In this paper, we propose multilevel Monte Carlo (MLMC) methods that use ensemble level mixed multiscale methods in the simulations of multiphase flow and transport. The contribution of this paper is twofold: (1) a design of ensemble level mixed

  11. Energy and particle core transport in tokamaks and stellarators compared

    Energy Technology Data Exchange (ETDEWEB)

    Beurskens, Marc; Angioni, Clemente; Beidler, Craig; Dinklage, Andreas; Fuchert, Golo; Hirsch, Matthias; Puetterich, Thomas; Wolf, Robert [Max-Planck-Institut fuer Plasmaphysik, Greifswald/Garching (Germany)

    2016-07-01

    The paper discusses expectations for core transport in the Wendelstein 7-X stellarator (W7-X) and presents a comparison to tokamaks. In tokamaks, the neoclassical trapped-particle-driven losses are small and turbulence dominates the energy and particle transport. At reactor relevant low collisionality, the heat transport is limited by ion temperature gradient limited turbulence, clamping the temperature gradient. The particle transport is set by an anomalous inward pinch, yielding peaked profiles. A strong edge pedestal adds to the good confinement properties. In traditional stellarators the 3D geometry cause increased trapped orbit losses. At reactor relevant low collisionality and high temperatures, these neoclassical losses would be well above the turbulent transport losses. The W7-X design minimizes neoclassical losses and turbulent transport can become dominant. Moreover, the separation of regions of bad curvature and that of trapped particle orbits in W7-X may have favourable implications on the turbulent electron heat transport. The neoclassical particle thermodiffusion is outward. Without core particle sources the density profile is flat or even hollow. The presence of a turbulence driven inward anomalous particle pinch in W7-X (like in tokamaks) is an open topic of research.

  12. Monte Carlo simulations of ultra high vacuum and synchrotron radiation for particle accelerators

    CERN Document Server

    AUTHOR|(CDS)2082330; Leonid, Rivkin

    With preparation of Hi-Lumi LHC fully underway, and the FCC machines under study, accelerators will reach unprecedented energies and along with it very large amount of synchrotron radiation (SR). This will desorb photoelectrons and molecules from accelerator walls, which contribute to electron cloud buildup and increase the residual pressure - both effects reducing the beam lifetime. In current accelerators these two effects are among the principal limiting factors, therefore precise calculation of synchrotron radiation and pressure properties are very important, desirably in the early design phase. This PhD project shows the modernization and a major upgrade of two codes, Molflow and Synrad, originally written by R. Kersevan in the 1990s, which are based on the test-particle Monte Carlo method and allow ultra-high vacuum and synchrotron radiation calculations. The new versions contain new physics, and are built as an all-in-one package - available to the public. Existing vacuum calculation methods are overvi...

  13. The motion of discs and spherical fuel particles in combustion burners based on Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Granada, E.; Patino, D.; Porteiro, J.; Collazo, J.; Miguez, J.L.; Moran, J. [University of Vigo, E.T.S. Ingenieros Industriales, Lagoas-Marcosende s/n, 36200-Vigo (Spain)

    2010-04-15

    The position of pellet fuel particles in a burner largely determines their combustion behaviour. This paper addresses the simulated motion of circles and spheres, equivalent to pellet, and their final position in a packed bed subject to a gravitational field confined inside rigid cylindrical walls. A simplified Monte Carlo statistical technique has been described and applied with the standard Metropolis method for the simulation of movement. This simplification provides an easier understanding of the method when applied to solid fuels in granular form, provided that they are only under gravitational forces. Not only have we contrasted one parameter, as other authors, but three, which are radial, bulk and local porosities, via Voronoi tessellation. Our simulations reveal a structural order near the walls, which declines towards the centre of the container, and no pattern was found in local porosity via Voronoi. Results with this simplified method are in agreement with more complex previously published studies. (author)

  14. The motion of discs and spherical fuel particles in combustion burners based on Monte Carlo simulation

    International Nuclear Information System (INIS)

    Granada, E.; Patino, D.; Porteiro, J.; Collazo, J.; Miguez, J.L.; Moran, J.

    2010-01-01

    The position of pellet fuel particles in a burner largely determines their combustion behaviour. This paper addresses the simulated motion of circles and spheres, equivalent to pellet, and their final position in a packed bed subject to a gravitational field confined inside rigid cylindrical walls. A simplified Monte Carlo statistical technique has been described and applied with the standard Metropolis method for the simulation of movement. This simplification provides an easier understanding of the method when applied to solid fuels in granular form, provided that they are only under gravitational forces. Not only have we contrasted one parameter, as other authors, but three, which are radial, bulk and local porosities, via Voronoi tessellation. Our simulations reveal a structural order near the walls, which declines towards the centre of the container, and no pattern was found in local porosity via Voronoi. Results with this simplified method are in agreement with more complex previously published studies.

  15. Subgroup approximation in Monte Carlo neutron transport calculation in resonance region; Primena podgrupe aproksimacije u Monte Carlo proracunima transporta neutrona u rezonantnoj oblasti energije

    Energy Technology Data Exchange (ETDEWEB)

    Belicev, P [Vojnotehnicki Inst., Belgrade (Yugoslavia)

    1988-07-01

    An outline of the problems encountered in the multigroup calculations of the neutron transport in the resonance region is given. The difference between subgroup and multigroup approximation is described briefly. The features of the Monte Carlo code SUBGR are presented. The results of the calculations of the neutron transmission and albedo for infinite iron slabs are given. (author)

  16. The energy band memory server algorithm for parallel Monte Carlo transport calculations

    International Nuclear Information System (INIS)

    Felker, K.G.; Siegel, A.R.; Smith, K.S.; Romano, P.K.; Forget, B.

    2013-01-01

    An algorithm is developed to significantly reduce the on-node footprint of cross section memory in Monte Carlo particle tracking algorithms. The classic method of per-node replication of cross section data is replaced by a memory server model, in which the read-only lookup tables reside on a remote set of disjoint processors. The main particle tracking algorithm is then modified in such a way as to enable efficient use of the remotely stored data in the particle tracking algorithm. Results of a prototype code on a Blue Gene/Q installation reveal that the penalty for remote storage is reasonable in the context of time scales for real-world applications, thus yielding a path forward for a broad range of applications that are memory bound using current techniques. (authors)

  17. AREVA Developments for an Efficient and Reliable use of Monte Carlo codes for Radiation Transport Applications

    Science.gov (United States)

    Chapoutier, Nicolas; Mollier, François; Nolin, Guillaume; Culioli, Matthieu; Mace, Jean-Reynald

    2017-09-01

    In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics). Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition) has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.

  18. AREVA Developments for an Efficient and Reliable use of Monte Carlo codes for Radiation Transport Applications

    Directory of Open Access Journals (Sweden)

    Chapoutier Nicolas

    2017-01-01

    Full Text Available In the context of the rising of Monte Carlo transport calculations for any kind of application, AREVA recently improved its suite of engineering tools in order to produce efficient Monte Carlo workflow. Monte Carlo codes, such as MCNP or TRIPOLI, are recognized as reference codes to deal with a large range of radiation transport problems. However the inherent drawbacks of theses codes - laboring input file creation and long computation time - contrast with the maturity of the treatment of the physical phenomena. The goals of the recent AREVA developments were to reach similar efficiency as other mature engineering sciences such as finite elements analyses (e.g. structural or fluid dynamics. Among the main objectives, the creation of a graphical user interface offering CAD tools for geometry creation and other graphical features dedicated to the radiation field (source definition, tally definition has been reached. The computations times are drastically reduced compared to few years ago thanks to the use of massive parallel runs, and above all, the implementation of hybrid variance reduction technics. From now engineering teams are capable to deliver much more prompt support to any nuclear projects dealing with reactors or fuel cycle facilities from conceptual phase to decommissioning.

  19. Progress on RMC: a Monte Carlo neutron transport code for reactor analysis

    International Nuclear Information System (INIS)

    Wang, Kan; Li, Zeguang; She, Ding; Liu, Yuxuan; Xu, Qi; Shen, Huayun; Yu, Ganglin

    2011-01-01

    This paper presents a new 3-D Monte Carlo neutron transport code named RMC (Reactor Monte Carlo code), specifically intended for reactor physics analysis. This code is being developed by Department of Engineering Physics in Tsinghua University and written in C++ and Fortran 90 language with the latest version of RMC 2.5.0. The RMC code uses the method known as the delta-tracking method to simulate neutron transport, the advantages of which include fast simulation in complex geometries and relatively simple handling of complicated geometrical objects. Some other techniques such as computational-expense oriented method and hash-table method have been developed and implemented in RMC to speedup the calculation. To meet the requirements of reactor analysis, the RMC code has the calculational functions including criticality calculation, burnup calculation and also kinetics simulation. In this paper, comparison calculations of criticality problems, burnup problems and transient problems are carried out using RMC code and other Monte Carlo codes, and the results show that RMC performs quite well in these kinds of problems. Based on MPI, RMC succeeds in parallel computation and represents a high speed-up. This code is still under intensive development and the further work directions are mentioned at the end of this paper. (author)

  20. Ratchet Transport of Chiral Particles Caused by the Transversal Asymmetry: Current Reversals and Particle Separation

    Science.gov (United States)

    Liu, Jian-li; Lu, Shi-cai; Ai, Bao-quan

    2018-06-01

    Due to the chirality of active particles, the transversal asymmetry can induce the the longitudinal directed transport. The transport of chiral active particles in a periodic channel is investigated in the presence of two types of the transversal asymmetry, the transverse force and the transverse rigid half-circle obstacles. For all cases, the counterclockwise and clockwise particles move to the opposite directions. For the case of the only transverse force, the chiral active particles can reverse their directions when increasing the transverse force. When the transverse rigid half-circle obstacles are introduced, the transport behavior of particles becomes more complex and multiple current reversals occur. The direction of the transport is determined by the competition between two types of the transversal asymmetry. For a given chirality, by suitably tailoring parameters, particles with different self-propulsion speed can move in different directions and can be separated.

  1. Microstripes for transport and separation of magnetic particles

    DEFF Research Database (Denmark)

    Donolato, Marco; Dalslet, Bjarke Thomas; Hansen, Mikkel Fougt

    2012-01-01

    We present a simple technique for creating an on-chip magnetic particle conveyor based on exchange-biased permalloy microstripes. The particle transportation relies on an array of stripes with a spacing smaller than their width in conjunction with a periodic sequence of four different externally...... applied magnetic fields. We demonstrate the controlled transportation of a large population of particles over several millimeters of distance as well as the spatial separation of two populations of magnetic particles with different magnetophoretic mobilities. The technique can be used for the controlled...... selective manipulation and separation of magnetically labelled species. (C) 2012 American Institute of Physics....

  2. Modeling and analysis of surface roughness effects on sputtering, reflection, and sputtered particle transport

    International Nuclear Information System (INIS)

    Brooks, J.N.; Ruzic, D.N.

    1990-01-01

    The microstructure of the redeposited surface in tokamaks may affect sputtering and reflection properties and subsequent particle transport. This subject has been studied numerically using coupled models/codes for near-surface plasma particle kinetic transport (WBC code) and rough surface sputtering (fractal-TRIM). The coupled codes provide an overall Monte Carlo calculation of the sputtering cascade resulting from an initial flux of hydrogen ions. Beryllium, carbon, and tungsten surfaces are analyzed for typical high recycling, oblique magnetic field, divertor conditions. Significant variations in computed sputtering rates are found with surface roughness. Beryllium exhibits high D-T and self-sputtering coefficients for the plasma regime studied (T e = 30-75 eV). Carbon and tungsten sputtering is significantly lower. 9 refs., 6 figs., 1 tab

  3. Thermal transport in nanocrystalline Si and SiGe by ab initio based Monte Carlo simulation.

    Science.gov (United States)

    Yang, Lina; Minnich, Austin J

    2017-03-14

    Nanocrystalline thermoelectric materials based on Si have long been of interest because Si is earth-abundant, inexpensive, and non-toxic. However, a poor understanding of phonon grain boundary scattering and its effect on thermal conductivity has impeded efforts to improve the thermoelectric figure of merit. Here, we report an ab-initio based computational study of thermal transport in nanocrystalline Si-based materials using a variance-reduced Monte Carlo method with the full phonon dispersion and intrinsic lifetimes from first-principles as input. By fitting the transmission profile of grain boundaries, we obtain excellent agreement with experimental thermal conductivity of nanocrystalline Si [Wang et al. Nano Letters 11, 2206 (2011)]. Based on these calculations, we examine phonon transport in nanocrystalline SiGe alloys with ab-initio electron-phonon scattering rates. Our calculations show that low energy phonons still transport substantial amounts of heat in these materials, despite scattering by electron-phonon interactions, due to the high transmission of phonons at grain boundaries, and thus improvements in ZT are still possible by disrupting these modes. This work demonstrates the important insights into phonon transport that can be obtained using ab-initio based Monte Carlo simulations in complex nanostructured materials.

  4. Dynamical theory of anomalous particle transport

    International Nuclear Information System (INIS)

    Meiss, J.D.; Cary, J.R.; Escande, D.F.; MacKay, R.S.; Percival, I.C.; Tennyson, J.L.

    1985-01-01

    The quasi-linear theory of transport applies only in a restricted parameter range, which does not necessarily correspond to experimental conditions. Theories are developed which extend transport calculations to the regimes of marginal stochasticity and strong turbulence. Near the stochastic threshold the description of transport involves the leakage through destroyed invariant surfaces, and the dynamical scaling theory is used to obtain a universal form for transport coefficients. In the strong-turbulence regime, there is an adiabatic invariant which is preserved except near separatrices. Breakdown of this invariant leads to a new form for the diffusion coefficient. (author)

  5. New model for mines and transportation tunnels external dose calculation using Monte Carlo simulation

    International Nuclear Information System (INIS)

    Allam, Kh. A.

    2017-01-01

    In this work, a new methodology is developed based on Monte Carlo simulation for tunnels and mines external dose calculation. Tunnels external dose evaluation model of a cylindrical shape of finite thickness with an entrance and with or without exit. A photon transportation model was applied for exposure dose calculations. A new software based on Monte Carlo solution was designed and programmed using Delphi programming language. The variation of external dose due to radioactive nuclei in a mine tunnel and the corresponding experimental data lies in the range 7.3 19.9%. The variation of specific external dose rate with position in, tunnel building material density and composition were studied. The given new model has more flexible for real external dose in any cylindrical tunnel structure calculations. (authors)

  6. Turbulent transport of large particles in the atmospheric boundary layer

    Science.gov (United States)

    Richter, D. H.; Chamecki, M.

    2017-12-01

    To describe the transport of heavy dust particles in the atmosphere, assumptions must typically be made in order to connect the micro-scale emission processes with the larger-scale atmospheric motions. In the context of numerical models, this can be thought of as the transport process which occurs between the domain bottom and the first vertical grid point. For example, in the limit of small particles (both low inertia and low settling velocity), theory built upon Monin-Obukhov similarity has proven effective in relating mean dust concentration profiles to surface emission fluxes. For increasing particle mass, however, it becomes more difficult to represent dust transport as a simple extension of the transport of a passive scalar due to issues such as the crossing trajectories effect. This study focuses specifically on the problem of large particle transport and dispersion in the turbulent boundary layer by utilizing direct numerical simulations with Lagrangian point-particle tracking to determine under what, if any, conditions the large dust particles (larger than 10 micron in diameter) can be accurately described in a simplified Eulerian framework. In particular, results will be presented detailing the independent contributions of both particle inertia and particle settling velocity relative to the strength of the surrounding turbulent flow, and consequences of overestimating surface fluxes via traditional parameterizations will be demonstrated.

  7. ASYMPTOTICS OF a PARTICLES TRANSPORT PROBLEM

    Directory of Open Access Journals (Sweden)

    Kuzmina Ludmila Ivanovna

    2017-11-01

    Full Text Available Subject: a groundwater filtration affects the strength and stability of underground and hydro-technical constructions. Research objectives: the study of one-dimensional problem of displacement of suspension by the flow of pure water in a porous medium. Materials and methods: when filtering a suspension some particles pass through the porous medium, and some of them are stuck in the pores. It is assumed that size distributions of the solid particles and the pores overlap. In this case, the main mechanism of particle retention is a size-exclusion: the particles pass freely through the large pores and get stuck at the inlet of the tiny pores that are smaller than the particle diameter. The concentrations of suspended and retained particles satisfy two quasi-linear differential equations of the first order. To solve the filtration problem, methods of nonlinear asymptotic analysis are used. Results: in a mathematical model of filtration of suspensions, which takes into account the dependence of the porosity and permeability of the porous medium on concentration of retained particles, the boundary between two phases is moving with variable velocity. The asymptotic solution to the problem is constructed for a small filtration coefficient. The theorem of existence of the asymptotics is proved. Analytical expressions for the principal asymptotic terms are presented for the case of linear coefficients and initial conditions. The asymptotics of the boundary of two phases is given in explicit form. Conclusions: the filtration problem under study can be solved analytically.

  8. The application of Monte Carlo method to electron and photon beams transport; Zastosowanie metody Monte Carlo do analizy transportu elektronow i fotonow

    Energy Technology Data Exchange (ETDEWEB)

    Zychor, I. [Soltan Inst. for Nuclear Studies, Otwock-Swierk (Poland)

    1994-12-31

    The application of a Monte Carlo method to study a transport in matter of electron and photon beams is presented, especially for electrons with energies up to 18 MeV. The SHOWME Monte Carlo code, a modified version of GEANT3 code, was used on the CONVEX C3210 computer at Swierk. It was assumed that an electron beam is mono directional and monoenergetic. Arbitrary user-defined, complex geometries made of any element or material can be used in calculation. All principal phenomena occurring when electron beam penetrates the matter are taken into account. The use of calculation for a therapeutic electron beam collimation is presented. (author). 20 refs, 29 figs.

  9. McSnow: A Monte-Carlo Particle Model for Riming and Aggregation of Ice Particles in a Multidimensional Microphysical Phase Space

    Science.gov (United States)

    Brdar, S.; Seifert, A.

    2018-01-01

    We present a novel Monte-Carlo ice microphysics model, McSnow, to simulate the evolution of ice particles due to deposition, aggregation, riming, and sedimentation. The model is an application and extension of the super-droplet method of Shima et al. (2009) to the more complex problem of rimed ice particles and aggregates. For each individual super-particle, the ice mass, rime mass, rime volume, and the number of monomers are predicted establishing a four-dimensional particle-size distribution. The sensitivity of the model to various assumptions is discussed based on box model and one-dimensional simulations. We show that the Monte-Carlo method provides a feasible approach to tackle this high-dimensional problem. The largest uncertainty seems to be related to the treatment of the riming processes. This calls for additional field and laboratory measurements of partially rimed snowflakes.

  10. WE-H-BRA-08: A Monte Carlo Cell Nucleus Model for Assessing Cell Survival Probability Based On Particle Track Structure Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B [Northwestern Memorial Hospital, Chicago, IL (United States); Georgia Institute of Technology, Atlanta, GA (Georgia); Wang, C [Georgia Institute of Technology, Atlanta, GA (Georgia)

    2016-06-15

    Purpose: To correlate the damage produced by particles of different types and qualities to cell survival on the basis of nanodosimetric analysis and advanced DNA structures in the cell nucleus. Methods: A Monte Carlo code was developed to simulate subnuclear DNA chromatin fibers (CFs) of 30nm utilizing a mean-free-path approach common to radiation transport. The cell nucleus was modeled as a spherical region containing 6000 chromatin-dense domains (CDs) of 400nm diameter, with additional CFs modeled in a sparser interchromatin region. The Geant4-DNA code was utilized to produce a particle track database representing various particles at different energies and dose quantities. These tracks were used to stochastically position the DNA structures based on their mean free path to interaction with CFs. Excitation and ionization events intersecting CFs were analyzed using the DBSCAN clustering algorithm for assessment of the likelihood of producing DSBs. Simulated DSBs were then assessed based on their proximity to one another for a probability of inducing cell death. Results: Variations in energy deposition to chromatin fibers match expectations based on differences in particle track structure. The quality of damage to CFs based on different particle types indicate more severe damage by high-LET radiation than low-LET radiation of identical particles. In addition, the model indicates more severe damage by protons than of alpha particles of same LET, which is consistent with differences in their track structure. Cell survival curves have been produced showing the L-Q behavior of sparsely ionizing radiation. Conclusion: Initial results indicate the feasibility of producing cell survival curves based on the Monte Carlo cell nucleus method. Accurate correlation between simulated DNA damage to cell survival on the basis of nanodosimetric analysis can provide insight into the biological responses to various radiation types. Current efforts are directed at producing cell

  11. METHES: A Monte Carlo collision code for the simulation of electron transport in low temperature plasmas

    Science.gov (United States)

    Rabie, M.; Franck, C. M.

    2016-06-01

    We present a freely available MATLAB code for the simulation of electron transport in arbitrary gas mixtures in the presence of uniform electric fields. For steady-state electron transport, the program provides the transport coefficients, reaction rates and the electron energy distribution function. The program uses established Monte Carlo techniques and is compatible with the electron scattering cross section files from the open-access Plasma Data Exchange Project LXCat. The code is written in object-oriented design, allowing the tracing and visualization of the spatiotemporal evolution of electron swarms and the temporal development of the mean energy and the electron number due to attachment and/or ionization processes. We benchmark our code with well-known model gases as well as the real gases argon, N2, O2, CF4, SF6 and mixtures of N2 and O2.

  12. Directed Transport of Brownian Particles in a Periodic Channel

    International Nuclear Information System (INIS)

    Jiang Jie; Ai Bao-Quan; Wu Jian-Chun

    2015-01-01

    The transport of Brownian particles in the infinite channel within an external force along the axis of the channel has been studied. In this paper, we study the transport of Brownian particle in the infinite channel within an external force along the axis of the channel and an external force in the transversal direction. In this more sophisticated situation, some property is similar to the simple situation, but some interesting property also appears. (paper)

  13. Numerical study of the particle transport in fast neutron detectors with conversion layer

    International Nuclear Information System (INIS)

    Sedlackova, K.; Zatko, B.; Necas, V.

    2012-01-01

    This paper deals with fast neutron and recoil proton transport simulation using statistical analysis of Monte Carlo radiation transport code (MCNPX). Its possibilities in the detector design and optimization are presented. MCNPX proved as a very advantageous self-contained simulation program for fast neutron and secondary proton tracking. Simulations of respective particle transport through conversion layer of HDPE and further in the active volume of detector let us to follow important characteristics as neutron/proton flux density, reaction rate of elastic scattering on hydrogen nuclei and deposited energy as well as their dependencies on incident neutron energy and conversion layer/active region thickness. The efficiency of neutrons to protons conversion has been calculated and its maximum was reached for 500 μm thick conversion layer. The minimum active region thickness has been estimated to be about 300 μm.(authors)

  14. Monte Carlo closure for moment-based transport schemes in general relativistic radiation hydrodynamic simulations

    Science.gov (United States)

    Foucart, Francois

    2018-04-01

    General relativistic radiation hydrodynamic simulations are necessary to accurately model a number of astrophysical systems involving black holes and neutron stars. Photon transport plays a crucial role in radiatively dominated accretion discs, while neutrino transport is critical to core-collapse supernovae and to the modelling of electromagnetic transients and nucleosynthesis in neutron star mergers. However, evolving the full Boltzmann equations of radiative transport is extremely expensive. Here, we describe the implementation in the general relativistic SPEC code of a cheaper radiation hydrodynamic method that theoretically converges to a solution of Boltzmann's equation in the limit of infinite numerical resources. The algorithm is based on a grey two-moment scheme, in which we evolve the energy density and momentum density of the radiation. Two-moment schemes require a closure that fills in missing information about the energy spectrum and higher order moments of the radiation. Instead of the approximate analytical closure currently used in core-collapse and merger simulations, we complement the two-moment scheme with a low-accuracy Monte Carlo evolution. The Monte Carlo results can provide any or all of the missing information in the evolution of the moments, as desired by the user. As a first test of our methods, we study a set of idealized problems demonstrating that our algorithm performs significantly better than existing analytical closures. We also discuss the current limitations of our method, in particular open questions regarding the stability of the fully coupled scheme.

  15. Development of a Monte Carlo software to photon transportation in voxel structures using graphic processing units

    International Nuclear Information System (INIS)

    Bellezzo, Murillo

    2014-01-01

    As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo Method (MCM) has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this thesis, the CUBMC code is presented, a GPU-based MC photon transport algorithm for dose calculation under the Compute Unified Device Architecture (CUDA) platform. The simulation of physical events is based on the algorithm used in PENELOPE, and the cross section table used is the one generated by the MATERIAL routine, also present in PENELOPE code. Photons are transported in voxel-based geometries with different compositions. There are two distinct approaches used for transport simulation. The rst of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon ignores the existence of borders and travels in homogeneous fictitious media. The CUBMC code aims to be an alternative of Monte Carlo simulator code that, by using the capability of parallel processing of graphics processing units (GPU), provide high performance simulations in low cost compact machines, and thus can be applied in clinical cases and incorporated in treatment planning systems for radiotherapy. (author)

  16. 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.

  17. Time-dependent 2-stream particle transport

    International Nuclear Information System (INIS)

    Corngold, Noel

    2015-01-01

    Highlights: • We consider time-dependent transport in the 2-stream or “rod” model via an attractive matrix formalism. • After reviewing some classical problems in homogeneous media we discuss transport in materials with whose density may vary. • There we achieve a significant contraction of the underlying Telegrapher’s equation. • We conclude with a discussion of stochastics, treated by the “first-order smoothing approximation.” - Abstract: We consider time-dependent transport in the 2-stream or “rod” model via an attractive matrix formalism. After reviewing some classical problems in homogeneous media we discuss transport in materials whose density may vary. There we achieve a significant contraction of the underlying Telegrapher’s equation. We conclude with a discussion of stochastics, treated by the “first-order smoothing approximation.”

  18. Spatiotemporal Structure of Aeolian Particle Transport on Flat Surface

    Science.gov (United States)

    Niiya, Hirofumi; Nishimura, Kouichi

    2017-05-01

    We conduct numerical simulations based on a model of blowing snow to reveal the long-term properties and equilibrium state of aeolian particle transport from 10-5 to 10 m above the flat surface. The numerical results are as follows. (i) Time-series data of particle transport are divided into development, relaxation, and equilibrium phases, which are formed by rapid wind response below 10 cm and gradual wind response above 10 cm. (ii) The particle transport rate at equilibrium is expressed as a power function of friction velocity, and the index of 2.35 implies that most particles are transported by saltation. (iii) The friction velocity below 100 µm remains roughly constant and lower than the fluid threshold at equilibrium. (iv) The mean particle speed above 300 µm is less than the wind speed, whereas that below 300 µm exceeds the wind speed because of descending particles. (v) The particle diameter increases with height in the saltation layer, and the relationship is expressed as a power function. Through comparisons with the previously reported random-flight model, we find a crucial problem that empirical splash functions cannot reproduce particle dynamics at a relatively high wind speed.

  19. Experimental characterization of solid particle transport by slug flow using Particle Image Velocimetry

    International Nuclear Information System (INIS)

    Goharzadeh, A; Rodgers, P

    2009-01-01

    This paper presents an experimental study of gas-liquid slug flow on solid particle transport inside a horizontal pipe with two types of experiments conducted. The influence of slug length on solid particle transportation is characterized using high speed photography. Using combined Particle Image Velocimetry (PIV) with Refractive Index Matching (RIM) and fluorescent tracers (two-phase oil-air loop) the velocity distribution inside the slug body is measured. Combining these experimental analyses, an insight is provided into the physical mechanism of solid particle transportation due to slug flow. It was observed that the slug body significantly influences solid particle mobility. The physical mechanism of solid particle transportation was found to be discontinuous. The inactive region (in terms of solid particle transport) upstream of the slug nose was quantified as a function of gas-liquid composition and solid particle size. Measured velocity distributions showed a significant drop in velocity magnitude immediately upstream of the slug nose and therefore the critical velocity for solid particle lifting is reached further upstream.

  20. Implementation of Japanese male and female tomographic phantoms to multi-particle Monte Carlo code for ionizing radiation dosimetry

    International Nuclear Information System (INIS)

    Lee, Choonsik; Nagaoka, Tomoaki; Lee, Jai-Ki

    2006-01-01

    Japanese male and female tomographic phantoms, which have been developed for radio-frequency electromagnetic-field dosimetry, were implemented into multi-particle Monte Carlo transport code to evaluate realistic dose distribution in human body exposed to radiation field. Japanese tomographic phantoms, which were developed from the whole body magnetic resonance images of Japanese average adult male and female, were processed as follows to be implemented into general purpose multi-particle Monte Carlo code, MCNPX2.5. Original array size of Japanese male and female phantoms, 320 x 160 x 866 voxels and 320 x 160 x 804 voxels, respectively, were reduced into 320 x 160 x 433 voxels and 320 x 160 x 402 voxels due to the limitation of memory use in MCNPX2.5. The 3D voxel array of the phantoms were processed by using the built-in repeated structure algorithm, where the human anatomy was described by the repeated lattice of tiny cube containing the information of material composition and organ index number. Original phantom data were converted into ASCII file, which can be directly ported into the lattice card of MCNPX2.5 input deck by using in-house code. A total of 30 material compositions obtained from International Commission on Radiation Units and Measurement (ICRU) report 46 were assigned to 54 and 55 organs and tissues in the male and female phantoms, respectively, and imported into the material card of MCNPX2.5 along with the corresponding cross section data. Illustrative calculation of absorbed doses for 26 internal organs and effective dose were performed for idealized broad parallel photon and neutron beams in anterior-posterior irradiation geometry, which is typical for workers at nuclear power plant. The results were compared with the data from other Japanese and Caucasian tomographic phantom, and International Commission on Radiological Protection (ICRP) report 74. The further investigation of the difference in organ dose and effective dose among tomographic

  1. Practical adjoint Monte Carlo technique for fixed-source and eigenfunction neutron transport problems

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.

    1981-01-01

    An adjoint Monte Carlo technique is described for the solution of neutron transport problems. The optimum biasing function for a zero-variance collision estimator is derived. The optimum treatment of an analog of a non-velocity thermal group has also been derived. The method is extended to multiplying systems, especially for eigenfunction problems to enable the estimate of averages over the unknown fundamental neutron flux distribution. A versatile computer code, FOCUS, has been written, based on the described theory. Numerical examples are given for a shielding problem and a critical assembly, illustrating the performance of the FOCUS code. 19 refs

  2. PBMC: Pre-conditioned Backward Monte Carlo code for radiative transport in planetary atmospheres

    Science.gov (United States)

    García Muñoz, A.; Mills, F. P.

    2017-08-01

    PBMC (Pre-Conditioned Backward Monte Carlo) solves the vector Radiative Transport Equation (vRTE) and can be applied to planetary atmospheres irradiated from above. The code builds the solution by simulating the photon trajectories from the detector towards the radiation source, i.e. in the reverse order of the actual photon displacements. In accounting for the polarization in the sampling of photon propagation directions and pre-conditioning the scattering matrix with information from the scattering matrices of prior (in the BMC integration order) photon collisions, PBMC avoids the unstable and biased solutions of classical BMC algorithms for conservative, optically-thick, strongly-polarizing media such as Rayleigh atmospheres.

  3. Accuracy estimation for intermediate and low energy neutron transport calculation with Monte Carlo code MCNP

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi

    1987-02-01

    Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)

  4. Monte Carlo investigation of minority electron transport in InP

    International Nuclear Information System (INIS)

    Osman, M.A.; Grubin, H.L.

    1989-01-01

    This paper discusses the investigation of the transport of minority electrons in p-type InP for acceptor doping level of 10 18 cm 3 using Monte Carlo procedures. It is found that the velocity of minority electrons are significantly lower than that of majority electrons for fields below 15 kV/cm and slightly higher at higher fields. The study shows that the interaction between the electrons and majority holes leads to reducing the mobility of electrons from 2000 cm 2 /Vs to 1500 cm 2 /Vs

  5. Monte Carlo simulation of radioactive contaminant transport in unsaturated porous media

    International Nuclear Information System (INIS)

    Giacobbo, F.; Patelli, E.; Zio, E.

    2005-01-01

    In the current proposed solutions of radioactive waste repositories, the protective function against the radionuclide water-driven transport back to the biosphere is to be provided by an integrated system of artificial and natural geologic barriers. The complexity of the transport process in the barriers' heterogeneous media forces approximations to the classical analytical-numerical models, thus reducing their adherence to reality. In an attempt to overcome these difficulties, in the present paper we adopt a Monte Carlo simulation approach, previously developed on the basis of the Kolmogorov and Dmitriev theory of branching stochastic processes. The approach is here extended for describing transport through unsaturated porous media under unsteady flow conditions. This generalization entails the determination of the functional dependence of the parameters of the proposed transport model from the water content, which changes in space and time during the water infiltration process. The approach is verified with respect to a case of non-reactive transport under transient unsaturated field conditions by a comparison with a standard code based on the classical advection-dispersion equations. An application regarding linear reactive transport is then presented. (authors)

  6. Advective isotope transport by mixing cell and particle tracking algorithms

    International Nuclear Information System (INIS)

    Tezcan, L.; Meric, T.

    1999-01-01

    The 'mixing cell' algorithm of the environmental isotope data evaluation is integrated with the three dimensional finite difference ground water flow model (MODFLOW) to simulate the advective isotope transport and the approach is compared with the 'particle tracking' algorithm of the MOC3D, that simulates three-dimensional solute transport with the method of characteristics technique

  7. Aerosol and particle transport in biomass furnaces

    NARCIS (Netherlands)

    Kemenade, van H.P.; Obernberger, G.

    2005-01-01

    The particulate emissions of solid fuel fired furnaces typically exhibit a bimodal distribution: a small peak in the range of 0.1 mm and a larger one above 10 mm. The particles with sizes above 10 mm are formed by a mechanical process like disintegration of the fuel after combustion, or erosion,

  8. Fueling profile sensitivities of trapped particle mode transport to TNS

    International Nuclear Information System (INIS)

    Mense, A.T.; Attenberger, S.E.; Houlberg, W.A.

    1977-01-01

    A key factor in the plasma thermal behavior is the anticipated existence of dissipative trapped particle modes. A possible scheme for controlling the strength of these modes was found. The scheme involves varying the cold fueling profile. A one dimensional multifluid transport code was used to simulate plasma behavior. A multiregime model for particle and energy transport was incorporated based on pseudoclassical, trapped electron, and trapped ion regimes used elsewhere in simulation of large tokamaks. Fueling profiles peaked toward the plasma edge may provide a means for reducing density-gradient-driven trapped particle modes, thus reducing diffusion and conduction losses

  9. Evaluation of the Neutron Detector Response for Cosmic Ray Energy Spectrum by Monte Carlo Transport Simulation

    International Nuclear Information System (INIS)

    Pazianotto, Mauricio T.; Carlson, Brett V.; Federico, Claudio A.; Gonzalez, Odair L.

    2011-01-01

    Neutrons generated by the interaction of cosmic rays with the atmosphere make an important contribution to the dose accumulated in electronic circuits and aircraft crew members at flight altitude. High-energy neutrons are produced in spallation reactions and intranuclear cascade processes by primary cosmic-ray particle interactions with atoms in the atmosphere. These neutrons can produce secondary neutrons and also undergo a moderation process due to atmosphere interactions, resulting in a wider energy spectrum, ranging from thermal energies (0.025 eV) to energies of several hundreds of MeV. The Long-Counter (LC) detector is a widely used neutron detector designed to measure the directional flux of neutrons with about constant response over a wide energy range (thermal to 20 MeV). ). Its calibration process and the determination of its energy response for the wide-energy of cosmic ray induced neutron spectrum is a very difficult process due to the lack of installations with these capabilities. The goal of this study is to assess the behavior of the response of a Long Counter using the Monte Carlo (MC) computational code MCNPX (Monte Carlo N-Particle eXtended). The dependence of the Long Counter response on the angle of incidence, as well as on the neutron energy, will be carefully investigated, compared with the experimental data previously obtained with 241 Am-Be and 252 Cf neutron sources and extended to the neutron spectrum produced by cosmic rays. (Author)

  10. Modelling of an industrial environment, part 1.: Monte Carlo simulations of photon transport

    International Nuclear Information System (INIS)

    Kis, Z.; Eged, K.; Meckbach, R.; Voigt, G.

    2002-01-01

    After a nuclear accident releasing radioactive material into the environment the external exposures may contribute significantly to the radiation exposure of the population (UNSCEAR 1988, 2000). For urban populations the external gamma exposure from radionuclides deposited on the surfaces of the urban-industrial environments yields the dominant contributions to the total dose to the public (Kelly 1987; Jacob and Meckbach 1990). The radiation field is naturally influenced by the environment around the sources. For calculations of the shielding effect of the structures in complex and realistic urban environments Monte Carlo methods turned out to be useful tools (Jacob and Meckbach 1987; Meckbach et al. 1988). Using these methods a complex environment can be set up in which the photon transport can be solved on a reliable way. The accuracy of the methods is in principle limited only by the knowledge of the atomic cross sections and the computational time. Several papers using Monte Carlo results for calculating doses from the external gamma exposures were published (Jacob and Meckbach 1987, 1990; Meckbach et al. 1988; Rochedo et al. 1996). In these papers the Monte Carlo simulations were run in urban environments and for different photon energies. The industrial environment can be defined as such an area where productive and/or commercial activity is carried out. A good example can be a factory or a supermarket. An industrial environment can rather be different from the urban ones as for the types and structures of the buildings and their dimensions. These variations will affect the radiation field of this environment. Hence there is a need to run new Monte Carlo simulations designed specially for the industrial environments

  11. Description of a neutron field perturbed by a probe using coupled Monte Carlo and discrete ordinates radiation transport calculations

    International Nuclear Information System (INIS)

    Zazula, J.M.

    1984-01-01

    This work concerns calculation of a neutron response, caused by a neutron field perturbed by materials surrounding the source or the detector. Solution of a problem is obtained using coupling of the Monte Carlo radiation transport computation for the perturbed region and the discrete ordinates transport computation for the unperturbed system. (author). 62 refs

  12. Monte Carlo techniques in radiation therapy

    CERN Document Server

    Verhaegen, Frank

    2013-01-01

    Modern cancer treatment relies on Monte Carlo simulations to help radiotherapists and clinical physicists better understand and compute radiation dose from imaging devices as well as exploit four-dimensional imaging data. With Monte Carlo-based treatment planning tools now available from commercial vendors, a complete transition to Monte Carlo-based dose calculation methods in radiotherapy could likely take place in the next decade. Monte Carlo Techniques in Radiation Therapy explores the use of Monte Carlo methods for modeling various features of internal and external radiation sources, including light ion beams. The book-the first of its kind-addresses applications of the Monte Carlo particle transport simulation technique in radiation therapy, mainly focusing on external beam radiotherapy and brachytherapy. It presents the mathematical and technical aspects of the methods in particle transport simulations. The book also discusses the modeling of medical linacs and other irradiation devices; issues specific...

  13. Comparison of stochastic models in Monte Carlo simulation of coated particle fuels

    International Nuclear Information System (INIS)

    Yu Hui; Nam Zin Cho

    2013-01-01

    There is growing interest worldwide in very high temperature gas cooled reactors as candidates for next generation reactor systems. For design and analysis of such reactors with double heterogeneity introduced by the coated particle fuels that are randomly distributed in graphite pebbles, stochastic transport models are becoming essential. Several models were reported in the literature, such as coarse lattice models, fine lattice stochastic (FLS) models, random sequential addition (RSA) models, metropolis models. The principles and performance of these stochastic models are described and compared in this paper. Compared with the usual fixed lattice methods, sub-FLS modeling allows more realistic stochastic distribution of fuel particles and thus results in more accurate criticality calculation. Compared with the basic RSA method, sub-FLS modeling requires simpler and more efficient overlapping checking procedure. (authors)

  14. Ensemble Monte Carlo particle investigation of hot electron induced source-drain burnout characteristics of GaAs field-effect transistors

    Science.gov (United States)

    Moglestue, C.; Buot, F. A.; Anderson, W. T.

    1995-08-01

    The lattice heating rate has been calculated for GaAs field-effect transistors of different source-drain channel design by means of the ensemble Monte Carlo particle model. Transport of carriers in the substrate and the presence of free surface charges are also included in our simulation. The actual heat generation was obtained by accounting for the energy exchanged with the lattice of the semiconductor during phonon scattering. It was found that the maximum heating rate takes place below the surface near the drain end of the gate. The results correlate well with a previous hydrodynamic energy transport estimate of the electronic energy density, but shifted slightly more towards the drain. These results further emphasize the adverse effects of hot electrons on the Ohmic contacts.

  15. Transport appraisal and Monte Carlo simulation by use of the CBA-DK model

    DEFF Research Database (Denmark)

    Salling, Kim Bang; Leleur, Steen

    2011-01-01

    calculation, where risk analysis is carried out using Monte Carlo simulation. Special emphasis has been placed on the separation between inherent randomness in the modeling system and lack of knowledge. These two concepts have been defined in terms of variability (ontological uncertainty) and uncertainty......This paper presents the Danish CBA-DK software model for assessment of transport infrastructure projects. The assessment model is based on both a deterministic calculation following the cost-benefit analysis (CBA) methodology in a Danish manual from the Ministry of Transport and on a stochastic...... (epistemic uncertainty). After a short introduction to deterministic calculation resulting in some evaluation criteria a more comprehensive evaluation of the stochastic calculation is made. Especially, the risk analysis part of CBA-DK, with considerations about which probability distributions should be used...

  16. Lateral electron transport in monolayers of short chains at interfaces: A Monte Carlo study

    International Nuclear Information System (INIS)

    George, Christopher B.; Szleifer, Igal; Ratner, Mark A.

    2010-01-01

    Graphical abstract: Electron hopping between electroactive sites in a monolayer composed of redox-active and redox-passive molecules. - Abstract: Using Monte Carlo simulations, we study lateral electronic diffusion in dense monolayers composed of a mixture of redox-active and redox-passive chains tethered to a surface. Two charge transport mechanisms are considered: the physical diffusion of electroactive chains and electron hopping between redox-active sites. Results indicate that by varying the monolayer density, the mole fraction of electroactive chains, and the electron hopping range, the dominant charge transport mechanism can be changed. For high density monolayers in a semi-crystalline phase, electron diffusion proceeds via electron hopping almost exclusively, leading to static percolation behavior. In fluid monolayers, the diffusion of chains may contribute more to the overall electronic diffusion, reducing the observed static percolation effects.

  17. Monte Carlo Simulation of Electron Transport in 4H- and 6H-SiC

    International Nuclear Information System (INIS)

    Sun, C. C.; You, A. H.; Wong, E. K.

    2010-01-01

    The Monte Carlo (MC) simulation of electron transport properties at high electric field region in 4H- and 6H-SiC are presented. This MC model includes two non-parabolic conduction bands. Based on the material parameters, the electron scattering rates included polar optical phonon scattering, optical phonon scattering and acoustic phonon scattering are evaluated. The electron drift velocity, energy and free flight time are simulated as a function of applied electric field at an impurity concentration of 1x10 18 cm 3 in room temperature. The simulated drift velocity with electric field dependencies is in a good agreement with experimental results found in literature. The saturation velocities for both polytypes are close, but the scattering rates are much more pronounced for 6H-SiC. Our simulation model clearly shows complete electron transport properties in 4H- and 6H-SiC.

  18. Transport of large particles released in a nuclear accident

    International Nuclear Information System (INIS)

    Poellaenen, R.; Toivonen, H.; Lahtinen, J.; Ilander, T.

    1995-10-01

    Highly radioactive particulate material may be released in a nuclear accident or sometimes during normal operation of a nuclear power plant. However, consequence analyses related to radioactive releases are often performed neglecting the particle nature of the release. The properties of the particles have an important role in the radiological hazard. A particle deposited on the skin may cause a large and highly non-uniform skin beta dose. Skin dose limits may be exceeded although the overall activity concentration in air is below the level of countermeasures. For sheltering purposes it is crucial to find out the transport range, i.e. the travel distance of the particles. A method for estimating the transport range of large particles (aerodynamic diameter d a > 20 μm) in simplified meteorological conditions is presented. A user-friendly computer code, known as TROP, is developed for fast range calculations in a nuclear emergency. (orig.) (23 refs., 13 figs.)

  19. Transport of large particles released in a nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Poellaenen, R; Toivonen, H; Lahtinen, J; Ilander, T

    1995-10-01

    Highly radioactive particulate material may be released in a nuclear accident or sometimes during normal operation of a nuclear power plant. However, consequence analyses related to radioactive releases are often performed neglecting the particle nature of the release. The properties of the particles have an important role in the radiological hazard. A particle deposited on the skin may cause a large and highly non-uniform skin beta dose. Skin dose limits may be exceeded although the overall activity concentration in air is below the level of countermeasures. For sheltering purposes it is crucial to find out the transport range, i.e. the travel distance of the particles. A method for estimating the transport range of large particles (aerodynamic diameter d{sub a} > 20 {mu}m) in simplified meteorological conditions is presented. A user-friendly computer code, known as TROP, is developed for fast range calculations in a nuclear emergency. (orig.) (23 refs., 13 figs.).

  20. 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

  1. 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)

  2. SANDYL, 3-D Time-Dependent and Space-Dependent Gamma Electron Cascade Transport by Monte-Carlo

    International Nuclear Information System (INIS)

    Haggmark, L.G.

    1980-01-01

    1 - Description of problem or function: SANDYL performs three- dimensional, time and space dependent Monte Carlo transport calculations for photon-electron cascades in complex systems. 2 - Method of solution: The problem geometry is divided into zones of homogeneous atomic composition bounded by sections of planes and quadrics. The material of each zone is a specified element or combination of elements. For a photon history, the trajectory is generated by following the photon from scattering to scattering using the various probability distributions to find distances between collisions, types of collisions, types of secondaries, and their energies and scattering angles. The photon interactions are photoelectric absorption (atomic ionization), coherent scattering, incoherent scattering, and pair production. The secondary photons which are followed include Bremsstrahlung, fluorescence photons, and positron-electron annihilation radiation. The condensed-history Monte Carlo method is used for the electron transport. In a history, the spatial steps taken by an electron are pre-computed and may include the effects of a number of collisions. The corresponding scattering angle and energy loss in the step are found from the multiple scattering distributions of these quantities. Atomic ionization and secondary particles are generated with the step according to the probabilities for their occurrence. Electron energy loss is through inelastic electron-electron collisions, Bremsstrahlung generation, and polarization of the medium (density effect). Included in the loss is the fluctuation due to the variation in the number of energy-loss collisions in a given Monte Carlo step (straggling). Scattering angular distributions are determined from elastic nuclear-collision cross sections corrected for electron-electron interactions. The secondary electrons which are followed included knock-on, pair, Auger (through atomic ionizations), Compton, and photoelectric electrons. 3

  3. Semi-analytic modeling of tokamak particle transport

    International Nuclear Information System (INIS)

    Shi Bingren; Long Yongxing; Li Jiquan

    2000-01-01

    The linear particle transport equation of tokamak plasma is analyzed. Particle flow consists of an outward diffusion and an inward convection. General solution is expressed in terms of a Green function constituted by eigen-functions of corresponding Sturm-Liouville problem. For a particle source near the plasma edge (shadow fueling), a well-behaved solution in terms of Fourier series can be constituted by using the complementarity relation. It can be seen from the lowest eigen-function that the particle density becomes peaked when the wall recycling reduced. For a transient point source in the inner region, a well-behaved solution can be obtained by the complementarity as well

  4. Monte Carlo N-particle simulation of neutron-based sterilisation of anthrax contamination.

    Science.gov (United States)

    Liu, B; Xu, J; Liu, T; Ouyang, X

    2012-10-01

    To simulate the neutron-based sterilisation of anthrax contamination by Monte Carlo N-particle (MCNP) 4C code. Neutrons are elementary particles that have no charge. They are 20 times more effective than electrons or γ-rays in killing anthrax spores on surfaces and inside closed containers. Neutrons emitted from a (252)Cf neutron source are in the 100 keV to 2 MeV energy range. A 2.5 MeV D-D neutron generator can create neutrons at up to 10(13) n s(-1) with current technology. All these enable an effective and low-cost method of killing anthrax spores. There is no effect on neutron energy deposition on the anthrax sample when using a reflector that is thicker than its saturation thickness. Among all three reflecting materials tested in the MCNP simulation, paraffin is the best because it has the thinnest saturation thickness and is easy to machine. The MCNP radiation dose and fluence simulation calculation also showed that the MCNP-simulated neutron fluence that is needed to kill the anthrax spores agrees with previous analytical estimations very well. The MCNP simulation indicates that a 10 min neutron irradiation from a 0.5 g (252)Cf neutron source or a 1 min neutron irradiation from a 2.5 MeV D-D neutron generator may kill all anthrax spores in a sample. This is a promising result because a 2.5 MeV D-D neutron generator output >10(13) n s(-1) should be attainable in the near future. This indicates that we could use a D-D neutron generator to sterilise anthrax contamination within several seconds.

  5. Numerical heating in Particle-In-Cell simulations with Monte Carlo binary collisions

    Science.gov (United States)

    Alves, E. Paulo; Mori, Warren; Fiuza, Frederico

    2017-10-01

    The binary Monte Carlo collision (BMCC) algorithm is a robust and popular method to include Coulomb collision effects in Particle-in-Cell (PIC) simulations of plasmas. While a number of works have focused on extending the validity of the model to different physical regimes of temperature and density, little attention has been given to the fundamental coupling between PIC and BMCC algorithms. Here, we show that the coupling between PIC and BMCC algorithms can give rise to (nonphysical) numerical heating of the system, that can be far greater than that observed when these algorithms operate independently. This deleterious numerical heating effect can significantly impact the evolution of the simulated system particularly for long simulation times. In this work, we describe the source of this numerical heating, and derive scaling laws for the numerical heating rates based on the numerical parameters of PIC-BMCC simulations. We compare our theoretical scalings with PIC-BMCC numerical experiments, and discuss strategies to minimize this parasitic effect. This work is supported by DOE FES under FWP 100237 and 100182.

  6. Monte Carlo Simulations of New 2D Ripple Filters for Particle Therapy Facilities

    DEFF Research Database (Denmark)

    Ringbæk, Toke Printz; Weber, Uli; Petersen, Jørgen B.B.

    2014-01-01

    ). At the Universitätsklinikum Gießen und Marburg, Germany, a new second generation RiFi has been developed with two-dimensional groove structures. In this work we evaluate this new RiFi design. Methods: The Monte Carlo (MC) code SHIELD-HIT12A is used to determine the RiFi- induced inhomogeneities in the dose distribution...... for various ion types, initial particle energies and distances from the RiFi to the phantom surface as well as in the depth of the phantom. The beam delivery and monitor system (BAMS) used at Marburg, the Heidelberg Ionentherapiezentrum (HIT), Universit ̈tsklinikum Heidelberg, Germany and the GSI...... Helmholtzzentrum für Schwerionenforschung, Darmstadt, Germany is modeled and simulated. To evaluate the PTV dose coverage performance of the new RiFi design, the heavy ion treatment planning system TRiP98 is used for dose optimization. SHIELD-HIT12A is used to prepare the facility-specific physical dose kernels...

  7. Coarse-grained stochastic processes and kinetic Monte Carlo simulators for the diffusion of interacting particles

    Science.gov (United States)

    Katsoulakis, Markos A.; Vlachos, Dionisios G.

    2003-11-01

    We derive a hierarchy of successively coarse-grained stochastic processes and associated coarse-grained Monte Carlo (CGMC) algorithms directly from the microscopic processes as approximations in larger length scales for the case of diffusion of interacting particles on a lattice. This hierarchy of models spans length scales between microscopic and mesoscopic, satisfies a detailed balance, and gives self-consistent fluctuation mechanisms whose noise is asymptotically identical to the microscopic MC. Rigorous, detailed asymptotics justify and clarify these connections. Gradient continuous time microscopic MC and CGMC simulations are compared under far from equilibrium conditions to illustrate the validity of our theory and delineate the errors obtained by rigorous asymptotics. Information theory estimates are employed for the first time to provide rigorous error estimates between the solutions of microscopic MC and CGMC, describing the loss of information during the coarse-graining process. Simulations under periodic boundary conditions are used to verify the information theory error estimates. It is shown that coarse-graining in space leads also to coarse-graining in time by q2, where q is the level of coarse-graining, and overcomes in part the hydrodynamic slowdown. Operation counting and CGMC simulations demonstrate significant CPU savings in continuous time MC simulations that vary from q3 for short potentials to q4 for long potentials. Finally, connections of the new coarse-grained stochastic processes to stochastic mesoscopic and Cahn-Hilliard-Cook models are made.

  8. Gyrokinetic theory for particle and energy transport in fusion plasmas

    Science.gov (United States)

    Falessi, Matteo Valerio; Zonca, Fulvio

    2018-03-01

    A set of equations is derived describing the macroscopic transport of particles and energy in a thermonuclear plasma on the energy confinement time. The equations thus derived allow studying collisional and turbulent transport self-consistently, retaining the effect of magnetic field geometry without postulating any scale separation between the reference state and fluctuations. Previously, assuming scale separation, transport equations have been derived from kinetic equations by means of multiple-scale perturbation analysis and spatio-temporal averaging. In this work, the evolution equations for the moments of the distribution function are obtained following the standard approach; meanwhile, gyrokinetic theory has been used to explicitly express the fluctuation induced fluxes. In this way, equations for the transport of particles and energy up to the transport time scale can be derived using standard first order gyrokinetics.

  9. ENERGETIC PARTICLE TRANSPORT ACROSS THE MEAN MAGNETIC FIELD: BEFORE DIFFUSION

    International Nuclear Information System (INIS)

    Laitinen, T.; Dalla, S.

    2017-01-01

    Current particle transport models describe the propagation of charged particles across the mean field direction in turbulent plasmas as diffusion. However, recent studies suggest that at short timescales, such as soon after solar energetic particle (SEP) injection, particles remain on turbulently meandering field lines, which results in nondiffusive initial propagation across the mean magnetic field. In this work, we use a new technique to investigate how the particles are displaced from their original field lines, and we quantify the parameters of the transition from field-aligned particle propagation along meandering field lines to particle diffusion across the mean magnetic field. We show that the initial decoupling of the particles from the field lines is slow, and particles remain within a Larmor radius from their initial meandering field lines for tens to hundreds of Larmor periods, for 0.1–10 MeV protons in turbulence conditions typical of the solar wind at 1 au. Subsequently, particles decouple from their initial field lines and after hundreds to thousands of Larmor periods reach time-asymptotic diffusive behavior consistent with particle diffusion across the mean field caused by the meandering of the field lines. We show that the typical duration of the prediffusive phase, hours to tens of hours for 10 MeV protons in 1 au solar wind turbulence conditions, is significant for SEP propagation to 1 au and must be taken into account when modeling SEP propagation in the interplanetary space.

  10. ENERGETIC PARTICLE TRANSPORT ACROSS THE MEAN MAGNETIC FIELD: BEFORE DIFFUSION

    Energy Technology Data Exchange (ETDEWEB)

    Laitinen, T.; Dalla, S., E-mail: tlmlaitinen@uclan.ac.uk [Jeremiah Horrocks Institute, University of Central Lancashire, Preston (United Kingdom)

    2017-01-10

    Current particle transport models describe the propagation of charged particles across the mean field direction in turbulent plasmas as diffusion. However, recent studies suggest that at short timescales, such as soon after solar energetic particle (SEP) injection, particles remain on turbulently meandering field lines, which results in nondiffusive initial propagation across the mean magnetic field. In this work, we use a new technique to investigate how the particles are displaced from their original field lines, and we quantify the parameters of the transition from field-aligned particle propagation along meandering field lines to particle diffusion across the mean magnetic field. We show that the initial decoupling of the particles from the field lines is slow, and particles remain within a Larmor radius from their initial meandering field lines for tens to hundreds of Larmor periods, for 0.1–10 MeV protons in turbulence conditions typical of the solar wind at 1 au. Subsequently, particles decouple from their initial field lines and after hundreds to thousands of Larmor periods reach time-asymptotic diffusive behavior consistent with particle diffusion across the mean field caused by the meandering of the field lines. We show that the typical duration of the prediffusive phase, hours to tens of hours for 10 MeV protons in 1 au solar wind turbulence conditions, is significant for SEP propagation to 1 au and must be taken into account when modeling SEP propagation in the interplanetary space.

  11. Modeling airflow and particle transport/deposition in pulmonary airways.

    Science.gov (United States)

    Kleinstreuer, Clement; Zhang, Zhe; Li, Zheng

    2008-11-30

    A review of research papers is presented, pertinent to computer modeling of airflow as well as nano- and micron-size particle deposition in pulmonary airway replicas. The key modeling steps are outlined, including construction of suitable airway geometries, mathematical description of the air-particle transport phenomena and computer simulation of micron and nanoparticle depositions. Specifically, diffusion-dominated nanomaterial deposits on airway surfaces much more uniformly than micron particles of the same material. This may imply different toxicity effects. Due to impaction and secondary flows, micron particles tend to accumulate around the carinal ridges and to form "hot spots", i.e., locally high concentrations which may lead to tumor developments. Inhaled particles in the size range of 20nm< or =dp< or =3microm may readily reach the deeper lung region. Concerning inhaled therapeutic particles, optimal parameters for mechanical drug-aerosol targeting of predetermined lung areas can be computed, given representative pulmonary airways.

  12. Transient particle transport studies at the W7-AS stellarator

    International Nuclear Information System (INIS)

    Koponen, J.

    2000-01-01

    One of the crucial problems in fusion research is the understanding of the transport of particles and heat in plasmas relevant for energy production. Extensive experimental transport studies have unraveled many details of heat transport in tokamaks and stellarators. However, due to larger experimental difficulties, the properties of particle transport have remained much less known. In particular, very few particle transport studies have been carried out in stellarators. This thesis summarises the transient particle transport experiments carried out at the Wendelstein 7-Advanced Stellarator (W7-AS). The main diagnostics tool was a 10-channel microwave interferometer. A technique for reconstructing the electron density profiles from the multichannel interferometer data was developed and implemented. The interferometer and the reconstruction software provide high quality electron density measurements with high temporal and sufficient spatial resolution. The density reconstruction is based on regularization methods studied during the development work. An extensive program of transient particle transport studies was carried out with the gas modulation method. The experiments resulted in a scaling expression for the diffusion coefficient. Transient inward convection was found in the edge plasma. The role of convection is minor in the core plasma, except at higher heating power, when an outward directed convective flux is observed. Radially peaked density profiles were found in discharges free of significant central density sources. Such density profiles are usually observed in tokamaks, but never before in W7-AS. Existence of an inward pinch is confirmed with two independent transient transport analysis methods. The density peaking is possible if the plasma is heated with extreme off-axis Electron Cyclotron Heating (ECH), when the temperature gradient vanishes in the core plasma, and if the gas puffing level is relatively low. The transport of plasma particles and heat

  13. Particle Transport in ECRH Plasmas of the TJ-II; Transporte de Particulas en Plasmas ECRH del TJ-II

    Energy Technology Data Exchange (ETDEWEB)

    Vargas, V. I.; Lopez-Bruna, D.; Estrada, T.; Guasp, J.; Reynolds, J. M.; Velasco, J. L.; Herranz, J.

    2007-07-01

    We present a systematic study of particle transport in ECRH plasmas of TJ-II with different densities. The goal is to fi nd particle confinement time and electron diffusivity dependence with line-averaged density. The experimental information consists of electron temperature profiles, T{sub e} (Thomson Scattering TS) and electron density, n{sub e}, (TS and reflectometry) and measured puffing data in stationary discharges. The profile of the electron source, Se, was obtained by the 3D Monte-Carlo code EIRENE. The analysis of particle balance has been done by linking the results of the code EIRENE with the results of a model that reproduces ECRH plasmas in stationary conditions. In the range of densities studied (0.58 {<=}n{sub e}> (10{sup 1}9m{sup -}3) {<=}0.80) there are two regions of confinement separated by a threshold density, {approx}0.65 10{sup 1}9m{sup -}3. Below this threshold density the particle confinement time is low, and vice versa. This is reflected in the effective diffusivity, D{sub e}, which in the range of validity of this study, 0.5 <{rho}<0.9 being {rho} normalized plasma radius, decreased significantly above the threshold density. The profiles of D{sub e} are flat for {>=}0,63(10{sup 1}9m{sup -}3). (Author) 35 refs.

  14. Recommended direct simulation Monte Carlo collision model parameters for modeling ionized air transport processes

    Energy Technology Data Exchange (ETDEWEB)

    Swaminathan-Gopalan, Krishnan; Stephani, Kelly A., E-mail: ksteph@illinois.edu [Department of Mechanical Science and Engineering, University of Illinois at Urbana-Champaign, Urbana, Illinois 61801 (United States)

    2016-02-15

    A systematic approach for calibrating the direct simulation Monte Carlo (DSMC) collision model parameters to achieve consistency in the transport processes is presented. The DSMC collision cross section model parameters are calibrated for high temperature atmospheric conditions by matching the collision integrals from DSMC against ab initio based collision integrals that are currently employed in the Langley Aerothermodynamic Upwind Relaxation Algorithm (LAURA) and Data Parallel Line Relaxation (DPLR) high temperature computational fluid dynamics solvers. The DSMC parameter values are computed for the widely used Variable Hard Sphere (VHS) and the Variable Soft Sphere (VSS) models using the collision-specific pairing approach. The recommended best-fit VHS/VSS parameter values are provided over a temperature range of 1000-20 000 K for a thirteen-species ionized air mixture. Use of the VSS model is necessary to achieve consistency in transport processes of ionized gases. The agreement of the VSS model transport properties with the transport properties as determined by the ab initio collision integral fits was found to be within 6% in the entire temperature range, regardless of the composition of the mixture. The recommended model parameter values can be readily applied to any gas mixture involving binary collisional interactions between the chemical species presented for the specified temperature range.

  15. ICF target 2D modeling using Monte Carlo SNB electron thermal transport in DRACO

    Science.gov (United States)

    Chenhall, Jeffrey; Cao, Duc; Moses, Gregory

    2016-10-01

    The iSNB (implicit Schurtz Nicolai Busquet multigroup diffusion electron thermal transport method is adapted into a Monte Carlo (MC) transport method to better model angular and long mean free path non-local effects. The MC model was first implemented in the 1D LILAC code to verify consistency with the iSNB model. Implementation of the MC SNB model in the 2D DRACO code enables higher fidelity non-local thermal transport modeling in 2D implosions such as polar drive experiments on NIF. The final step is to optimize the MC model by hybridizing it with a MC version of the iSNB diffusion method. The hybrid method will combine the efficiency of a diffusion method in intermediate mean free path regions with the accuracy of a transport method in long mean free path regions allowing for improved computational efficiency while maintaining accuracy. Work to date on the method will be presented. This work was supported by Sandia National Laboratories and the Univ. of Rochester Laboratory for Laser Energetics.

  16. Present status of transport code development based on Monte Carlo method

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki

    1985-01-01

    The present status of development in Monte Carlo code is briefly reviewed. The main items are the followings; Application fields, Methods used in Monte Carlo code (geometry spectification, nuclear data, estimator and variance reduction technique) and unfinished works, Typical Monte Carlo codes and Merits of continuous energy Monte Carlo code. (author)

  17. THREEDANT: A code to perform three-dimensional, neutral particle transport calculations

    International Nuclear Information System (INIS)

    Alcouffe, R.E.

    1994-01-01

    The THREEDANT code solves the three-dimensional neutral particle transport equation in its first order, multigroup, discrate ordinate form. The code allows an unlimited number of groups (depending upon the cross section set), angular quadrature up to S-100, and unlimited Pn order again depending upon the cross section set. The code has three options for spatial differencing, diamond with set-to-zero fixup, adaptive weighted diamond, and linear modal. The geometry options are XYZ and RZΘ with a special XYZ option based upon a volume fraction method. This allows objects or bodies of any shape to be modelled as input which gives the code as much geometric description flexibility as the Monte Carlo code MCNP. The transport equation is solved by source iteration accelerated by the DSA method. Both inner and outer iterations are so accelerated. Some results are presented which demonstrate the effectiveness of these techniques. The code is available on several types of computing platforms

  18. Drift Wave Test Particle Transport in Reversed Shear Profile

    International Nuclear Information System (INIS)

    Horton, W.; Park, H.B.; Kwon, J.M.; Stronzzi, D.; Morrison, P.J.; Choi, D.I.

    1998-01-01

    Drift wave maps, area preserving maps that describe the motion of charged particles in drift waves, are derived. The maps allow the integration of particle orbits on the long time scale needed to describe transport. Calculations using the drift wave maps show that dramatic improvement in the particle confinement, in the presence of a given level and spectrum of E x B turbulence, can occur for q(r)-profiles with reversed shear. A similar reduction in the transport, i.e. one that is independent of the turbulence, is observed in the presence of an equilibrium radial electric field with shear. The transport reduction, caused by the combined effects of radial electric field shear and both monotonic and reversed shear magnetic q-profiles, is also investigated

  19. Optimization of magnetic switches for single particle and cell transport

    Energy Technology Data Exchange (ETDEWEB)

    Abedini-Nassab, Roozbeh; Yellen, Benjamin B., E-mail: yellen@duke.edu [Department of Mechanical Engineering and Materials Science, Duke University, Box 90300 Hudson Hall, Durham, North Carolina 27708 (United States); Joint Institute, University of Michigan—Shanghai Jiao Tong University, Shanghai Jiao Tong University, Shanghai 200240 (China); Murdoch, David M. [Department of Medicine, Duke University, Durham, North Carolina 27708 (United States); Kim, CheolGi [Department of Emerging Materials Science, Daegu Gyeongbuk Institute of Science and Technology (DGIST), Daegu 711-873 (Korea, Republic of)

    2014-06-28

    The ability to manipulate an ensemble of single particles and cells is a key aim of lab-on-a-chip research; however, the control mechanisms must be optimized for minimal power consumption to enable future large-scale implementation. Recently, we demonstrated a matter transport platform, which uses overlaid patterns of magnetic films and metallic current lines to control magnetic particles and magnetic-nanoparticle-labeled cells; however, we have made no prior attempts to optimize the device geometry and power consumption. Here, we provide an optimization analysis of particle-switching devices based on stochastic variation in the particle's size and magnetic content. These results are immediately applicable to the design of robust, multiplexed platforms capable of transporting, sorting, and storing single cells in large arrays with low power and high efficiency.

  20. Computational methods for two-phase flow and particle transport

    CERN Document Server

    Lee, Wen Ho

    2013-01-01

    This book describes mathematical formulations and computational methods for solving two-phase flow problems with a computer code that calculates thermal hydraulic problems related to light water and fast breeder reactors. The physical model also handles the particle and gas flow problems that arise from coal gasification and fluidized beds. The second part of this book deals with the computational methods for particle transport.

  1. Solitary Model of the Charge Particle Transport in Collisionless Plasma

    International Nuclear Information System (INIS)

    Simonchik, L.V.; Trukhachev, F.M.

    2006-01-01

    The one-dimensional MHD solitary model of charged particle transport in plasma is developed. It is shown that self-consistent electric field of ion-acoustic solitons can displace charged particles in space, which can be a reason of local electric current generation. The displacement amount is order of a few Debye lengths. It is shown that the current associated with soliton cascade has pulsating nature with DC component. Methods of built theory verification in dusty plasma are proposed

  2. Particle Acceleration and Fractional Transport in Turbulent Reconnection

    Science.gov (United States)

    Isliker, Heinz; Pisokas, Theophilos; Vlahos, Loukas; Anastasiadis, Anastasios

    2017-11-01

    We consider a large-scale environment of turbulent reconnection that is fragmented into a number of randomly distributed unstable current sheets (UCSs), and we statistically analyze the acceleration of particles within this environment. We address two important cases of acceleration mechanisms when particles interact with the UCS: (a) electric field acceleration and (b) acceleration by reflection at contracting islands. Electrons and ions are accelerated very efficiently, attaining an energy distribution of power-law shape with an index 1-2, depending on the acceleration mechanism. The transport coefficients in energy space are estimated from test-particle simulation data, and we show that the classical Fokker-Planck (FP) equation fails to reproduce the simulation results when the transport coefficients are inserted into it and it is solved numerically. The cause for this failure is that the particles perform Levy flights in energy space, while the distributions of the energy increments exhibit power-law tails. We then use the fractional transport equation (FTE) derived by Isliker et al., whose parameters and the order of the fractional derivatives are inferred from the simulation data, and solving the FTE numerically, we show that the FTE successfully reproduces the kinetic energy distribution of the test particles. We discuss in detail the analysis of the simulation data and the criteria that allow one to judge the appropriateness of either an FTE or a classical FP equation as a transport model.

  3. Particle Acceleration and Fractional Transport in Turbulent Reconnection

    Energy Technology Data Exchange (ETDEWEB)

    Isliker, Heinz; Pisokas, Theophilos; Vlahos, Loukas [Department of Physics, Aristotle University of Thessaloniki, GR-52124 Thessaloniki (Greece); Anastasiadis, Anastasios [Institute for Astronomy, Astrophysics, Space Applications and Remote Sensing, National Observatory of Athens, GR-15236 Penteli (Greece)

    2017-11-01

    We consider a large-scale environment of turbulent reconnection that is fragmented into a number of randomly distributed unstable current sheets (UCSs), and we statistically analyze the acceleration of particles within this environment. We address two important cases of acceleration mechanisms when particles interact with the UCS: (a) electric field acceleration and (b) acceleration by reflection at contracting islands. Electrons and ions are accelerated very efficiently, attaining an energy distribution of power-law shape with an index 1–2, depending on the acceleration mechanism. The transport coefficients in energy space are estimated from test-particle simulation data, and we show that the classical Fokker–Planck (FP) equation fails to reproduce the simulation results when the transport coefficients are inserted into it and it is solved numerically. The cause for this failure is that the particles perform Levy flights in energy space, while the distributions of the energy increments exhibit power-law tails. We then use the fractional transport equation (FTE) derived by Isliker et al., whose parameters and the order of the fractional derivatives are inferred from the simulation data, and solving the FTE numerically, we show that the FTE successfully reproduces the kinetic energy distribution of the test particles. We discuss in detail the analysis of the simulation data and the criteria that allow one to judge the appropriateness of either an FTE or a classical FP equation as a transport model.

  4. Optimization of FIBMOS Through 2D Silvaco ATLAS and 2D Monte Carlo Particle-based Device Simulations

    OpenAIRE

    Kang, J.; He, X.; Vasileska, D.; Schroder, D. K.

    2001-01-01

    Focused Ion Beam MOSFETs (FIBMOS) demonstrate large enhancements in core device performance areas such as output resistance, hot electron reliability and voltage stability upon channel length or drain voltage variation. In this work, we describe an optimization technique for FIBMOS threshold voltage characterization using the 2D Silvaco ATLAS simulator. Both ATLAS and 2D Monte Carlo particle-based simulations were used to show that FIBMOS devices exhibit enhanced current drive ...

  5. Particle swarm optimization - Genetic algorithm (PSOGA) on linear transportation problem

    Science.gov (United States)

    Rahmalia, Dinita

    2017-08-01

    Linear Transportation Problem (LTP) is the case of constrained optimization where we want to minimize cost subject to the balance of the number of supply and the number of demand. The exact method such as northwest corner, vogel, russel, minimal cost have been applied at approaching optimal solution. In this paper, we use heurisitic like Particle Swarm Optimization (PSO) for solving linear transportation problem at any size of decision variable. In addition, we combine mutation operator of Genetic Algorithm (GA) at PSO to improve optimal solution. This method is called Particle Swarm Optimization - Genetic Algorithm (PSOGA). The simulations show that PSOGA can improve optimal solution resulted by PSO.

  6. SAM-CE, Time-Dependent 3-D Neutron Transport, Gamma Transport in Complex Geometry by Monte-Carlo

    International Nuclear Information System (INIS)

    2003-01-01

    1 - Nature of physical problem solved: The SAM-CE system comprises two Monte Carlo codes, SAM-F and SAM-A. SAM-F supersedes the forward Monte Carlo code, SAM-C. SAM-A is an adjoint Monte Carlo code designed to calculate the response due to fields of primary and secondary gamma radiation. The SAM-CE system is a FORTRAN Monte Carlo computer code designed to solve the time-dependent neutron and gamma-ray transport equations in complex three-dimensional geometries. SAM-CE is applicable for forward neutron calculations and for forward as well as adjoint primary gamma-ray calculations. In addition, SAM-CE is applicable for the gamma-ray stage of the coupled neutron-secondary gamma ray problem, which may be solved in either the forward or the adjoint mode. Time-dependent fluxes, and flux functionals such as dose, heating, count rates, etc., are calculated as functions of energy, time and position. Multiple scoring regions are permitted and these may be either finite volume regions or point detectors or both. Other scores of interest, e.g., collision and absorption densities, etc., are also made. 2 - Method of solution: A special feature of SAM-CE is its use of the 'combinatorial geometry' technique which affords the user geometric capabilities exceeding those available with other commonly used geometric packages. All nuclear interaction cross section data (derived from the ENDF for neutrons and from the UNC-format library for gamma-rays) are tabulated in point energy meshes. The energy meshes for neutrons are internally derived, based on built-in convergence criteria and user- supplied tolerances. Tabulated neutron data for each distinct nuclide are in unique and appropriate energy meshes. Both resolved and unresolved resonance parameters from ENDF data files are treated automatically, and extremely precise and detailed descriptions of cross section behaviour is permitted. Such treatment avoids the ambiguities usually associated with multi-group codes, which use flux

  7. Collective transport of Lennard–Jones particles through one-dimensional periodic potentials

    International Nuclear Information System (INIS)

    He Jian-hui; Wen Jia-le; Chen Pei-rong; Zheng Dong-qin; Zhong Wei-rong

    2017-01-01

    The surrounding media in which transport occurs contains various kinds of fields, such as particle potentials and external potentials. One of the important questions is how elements work and how position and momentum are redistributed in the diffusion under these conditions. For enriching Fick’s law, ordinary non-equilibrium statistical physics can be used to understand the complex process. This study attempts to discuss particle transport in the one-dimensional channel under external potential fields. Two kinds of potentials—the potential well and barrier—which do not change the potential in total, are built during the diffusion process. There are quite distinct phenomena because of the different one-dimensional periodic potentials. By the combination of a Monte Carlo method and molecular dynamics, we meticulously explore why an external potential field impacts transport by the subsection and statistical method. Besides, one piece of evidence of the Maxwell velocity distribution is confirmed under the assumption of local equilibrium. The simple model is based on the key concept that relates the flux to sectional statistics of position and momentum and could be referenced in similar transport problems. (rapid communication)

  8. Simulation of charge generation and transport in semi-conductors under energetic-particle bombardment

    International Nuclear Information System (INIS)

    Martin, R.C.

    1990-01-01

    The passage of energetic ions through semiconductor devices generates excess charge which can produce logic upset, memory change, and device damage. This single event upset (SEU) phenomenon is increasingly important for satellite communications. Experimental and numerical simulation of SEUs is difficult because of the subnanosecond times and large charge densities within the ion track. The objective of this work is twofold: (1) the determination of the track structure and electron-hole pair generation profiles following the passage of an energetic ion; (2) the development and application of a new numerical method for transient charge transport in semiconductor devices. A secondary electron generation and transport model, based on the Monte Carlo method, is developed and coupled to an ion transport code to simulate ion track formation in silicon. A new numerical method is developed for the study of transient charge transport. The numerical method combines an axisymmetric quadratic finite-element formulation for the solution of the potential with particle simulation methods for electron and hole transport. Carrier transport, recombination, and thermal generation of both majority and minority carriers are included. To assess the method, transient one-dimensional solutions for silicon diodes are compared to a fully iterative finite-element method. Simulations of charge collection from ion tracks in three-dimensional axisymmetric devices are presented and compared to previous work. The results of this work for transient current pulses following charged ion passage are in agreement with recent experimental data

  9. Vectorized Monte Carlo

    International Nuclear Information System (INIS)

    Brown, F.B.

    1981-01-01

    Examination of the global algorithms and local kernels of conventional general-purpose Monte Carlo codes shows that multigroup Monte Carlo methods have sufficient structure to permit efficient vectorization. A structured multigroup Monte Carlo algorithm for vector computers is developed in which many particle events are treated at once on a cell-by-cell basis. Vectorization of kernels for tracking and variance reduction is described, and a new method for discrete sampling is developed to facilitate the vectorization of collision analysis. To demonstrate the potential of the new method, a vectorized Monte Carlo code for multigroup radiation transport analysis was developed. This code incorporates many features of conventional general-purpose production codes, including general geometry, splitting and Russian roulette, survival biasing, variance estimation via batching, a number of cutoffs, and generalized tallies of collision, tracklength, and surface crossing estimators with response functions. Predictions of vectorized performance characteristics for the CYBER-205 were made using emulated coding and a dynamic model of vector instruction timing. Computation rates were examined for a variety of test problems to determine sensitivities to batch size and vector lengths. Significant speedups are predicted for even a few hundred particles per batch, and asymptotic speedups by about 40 over equivalent Amdahl 470V/8 scalar codes arepredicted for a few thousand particles per batch. The principal conclusion is that vectorization of a general-purpose multigroup Monte Carlo code is well worth the significant effort required for stylized coding and major algorithmic changes

  10. Measured performances on vectorization and multitasking with a Monte Carlo code for neutron transport problems

    International Nuclear Information System (INIS)

    Chauvet, Y.

    1985-01-01

    This paper summarized two improvements of a real production code by using vectorization and multitasking techniques. After a short description of Monte Carlo algorithms employed in neutron transport problems, the authors briefly describe the work done in order to get a vector code. Vectorization principles are presented and measured performances on the CRAY 1S, CYBER 205 and CRAY X-MP compared in terms of vector lengths. The second part of this work is an adaptation to multitasking on the CRAY X-MP using exclusively standard multitasking tools available with FORTRAN under the COS 1.13 system. Two examples are presented. The goal of the first one is to measure the overhead inherent to multitasking when tasks become too small and to define a granularity threshold, that is to say a minimum size for a task. With the second example they propose a method that is very X-MP oriented in order to get the best speedup factor on such a computer. In conclusion they prove that Monte Carlo algorithms are very well suited to future vector and parallel computers

  11. Measured performances on vectorization and multitasking with a Monte Carlo code for neutron transport problems

    International Nuclear Information System (INIS)

    Chauvet, Y.

    1985-01-01

    This paper summarized two improvements of a real production code by using vectorization and multitasking techniques. After a short description of Monte Carlo algorithms employed in our neutron transport problems, we briefly describe the work we have done in order to get a vector code. Vectorization principles will be presented and measured performances on the CRAY 1S, CYBER 205 and CRAY X-MP compared in terms of vector lengths. The second part of this work is an adaptation to multitasking on the CRAY X-MP using exclusively standard multitasking tools available with FORTRAN under the COS 1.13 system. Two examples will be presented. The goal of the first one is to measure the overhead inherent to multitasking when tasks become too small and to define a granularity threshold, that is to say a minimum size for a task. With the second example we propose a method that is very X-MP oriented in order to get the best speedup factor on such a computer. In conclusion we prove that Monte Carlo algorithms are very well suited to future vector and parallel computers. (orig.)

  12. Monte Carlo modelling of impurity ion transport for a limiter source/sink

    International Nuclear Information System (INIS)

    Stangeby, P.C.; Farrell, C.; Hoskins, S.; Wood, L.

    1988-01-01

    In relating the impurity influx Φ I (0) (atoms per second) into a plasma from the edge to the central impurity ion density n I (0) (ions·m -3 ), it is necessary to know the value of τ I SOL , the average dwell time of impurity ions in the scrape-off layer. It is usually assumed that τ I SOL =L c /c s , the hydrogenic dwell time, where L c is the limiter connection length and c s is the hydrogenic ion acoustic speed. Monte Carlo ion transport results are reported here which show that, for a wall (uniform) influx, τ I SOL is longer than L c /c s , while for a limiter influx it is shorter. Thus for a limiter influx n I (0) is predicted to be smaller than the reference value. Impurities released from the limiter form ever large 'clouds' of successively higher ionization stages. These are reproduced by the Monte Carlo code as are the cloud shapes for a localized impurity injection far from the limiter. (author). 23 refs, 18 figs, 6 tabs

  13. Particle Tracking Model and Abstraction of Transport Processes

    International Nuclear Information System (INIS)

    Robinson, B.

    2000-01-01

    The purpose of the transport methodology and component analysis is to provide the numerical methods for simulating radionuclide transport and model setup for transport in the unsaturated zone (UZ) site-scale model. The particle-tracking method of simulating radionuclide transport is incorporated into the FEHM computer code and the resulting changes in the FEHM code are to be submitted to the software configuration management system. This Analysis and Model Report (AMR) outlines the assumptions, design, and testing of a model for calculating radionuclide transport in the unsaturated zone at Yucca Mountain. In addition, methods for determining colloid-facilitated transport parameters are outlined for use in the Total System Performance Assessment (TSPA) analyses. Concurrently, process-level flow model calculations are being carrier out in a PMR for the unsaturated zone. The computer code TOUGH2 is being used to generate three-dimensional, dual-permeability flow fields, that are supplied to the Performance Assessment group for subsequent transport simulations. These flow fields are converted to input files compatible with the FEHM code, which for this application simulates radionuclide transport using the particle-tracking algorithm outlined in this AMR. Therefore, this AMR establishes the numerical method and demonstrates the use of the model, but the specific breakthrough curves presented do not necessarily represent the behavior of the Yucca Mountain unsaturated zone

  14. Parallelizing an electron transport Monte Carlo simulator (MOCASIN 2.0)

    International Nuclear Information System (INIS)

    Schwetman, H.; Burdick, S.

    1988-01-01

    Electron transport simulators are tools for studying electrical properties of semiconducting materials and devices. As demands for modeling more complex devices and new materials have emerged, so have demands for more processing power. This paper documents a project to convert an electron transport simulator (MOCASIN 2.0) to a parallel processing environment. In addition to describing the conversion, the paper presents PPL, a parallel programming version of C running on a Sequent multiprocessor system. In timing tests, models that simulated the movement of 2,000 particles for 100 time steps were executed on ten processors, with a parallel efficiency of over 97%

  15. Quasilinear Line Broadened Model for Energetic Particle Transport

    Science.gov (United States)

    Ghantous, Katy; Gorelenkov, Nikolai; Berk, Herbert

    2011-10-01

    We present a self-consistent quasi-linear model that describes wave-particle interaction in toroidal geometry and computes fast ion transport during TAE mode evolution. The model bridges the gap between single mode resonances, where it predicts the analytically expected saturation levels, and the case of multiple modes overlapping, where particles diffuse across phase space. Results are presented in the large aspect ratio limit where analytic expressions are used for Fourier harmonics of the power exchange between waves and particles, . Implemention of a more realistic mode structure calculated by NOVAK code are also presented. This work is funded by DOE contract DE-AC02-09CH11466.

  16. Automating methods to improve precision in Monte-Carlo event generation for particle colliders

    International Nuclear Information System (INIS)

    Gleisberg, Tanju

    2008-01-01

    The subject of this thesis was the development of tools for the automated calculation of exact matrix elements, which are a key for the systematic improvement of precision and confidence for theoretical predictions. Part I of this thesis concentrates on the calculations of cross sections at tree level. A number of extensions have been implemented in the matrix element generator AMEGIC++, namely new interaction models such as effective loop-induced couplings of the Higgs boson with massless gauge bosons, required for a number of channels for the Higgs boson search at LHC and anomalous gauge couplings, parameterizing a number of models beyond th SM. Further a special treatment to deal with complicated decay chains of heavy particles has been constructed. A significant effort went into the implementation of methods to push the limits on particle multiplicities. Two recursive methods have been implemented, the Cachazo-Svrcek-Witten recursion and the colour dressed Berends-Giele recursion. For the latter the new module COMIX has been added to the SHERPA framework. The Monte-Carlo phase space integration techniques have been completely revised, which led to significantly reduced statistical error estimates when calculating cross sections and a greatly improved unweighting efficiency for the event generation. Special integration methods have been developed to cope with the newly accessible final states. The event generation framework SHERPA directly benefits from those new developments, improving the precision and the efficiency. Part II was addressed to the automation of QCD calculations at next-to-leading order. A code has been developed, that, for the first time fully automates the real correction part of a NLO calculation. To calculate the correction for a m-parton process obeying the Catani-Seymour dipole subtraction method the following components are provided: 1. the corresponding m+1-parton tree level matrix elements, 2. a number dipole subtraction terms to remove

  17. Automating methods to improve precision in Monte-Carlo event generation for particle colliders

    Energy Technology Data Exchange (ETDEWEB)

    Gleisberg, Tanju

    2008-07-01

    The subject of this thesis was the development of tools for the automated calculation of exact matrix elements, which are a key for the systematic improvement of precision and confidence for theoretical predictions. Part I of this thesis concentrates on the calculations of cross sections at tree level. A number of extensions have been implemented in the matrix element generator AMEGIC++, namely new interaction models such as effective loop-induced couplings of the Higgs boson with massless gauge bosons, required for a number of channels for the Higgs boson search at LHC and anomalous gauge couplings, parameterizing a number of models beyond th SM. Further a special treatment to deal with complicated decay chains of heavy particles has been constructed. A significant effort went into the implementation of methods to push the limits on particle multiplicities. Two recursive methods have been implemented, the Cachazo-Svrcek-Witten recursion and the colour dressed Berends-Giele recursion. For the latter the new module COMIX has been added to the SHERPA framework. The Monte-Carlo phase space integration techniques have been completely revised, which led to significantly reduced statistical error estimates when calculating cross sections and a greatly improved unweighting efficiency for the event generation. Special integration methods have been developed to cope with the newly accessible final states. The event generation framework SHERPA directly benefits from those new developments, improving the precision and the efficiency. Part II was addressed to the automation of QCD calculations at next-to-leading order. A code has been developed, that, for the first time fully automates the real correction part of a NLO calculation. To calculate the correction for a m-parton process obeying the Catani-Seymour dipole subtraction method the following components are provided: 1. the corresponding m+1-parton tree level matrix elements, 2. a number dipole subtraction terms to remove

  18. penORNL: a parallel Monte Carlo photon and electron transport package using PENELOPE

    International Nuclear Information System (INIS)

    Bekar, Kursat B.; Miller, Thomas Martin; Patton, Bruce W.; Weber, Charles F.

    2015-01-01

    The parallel Monte Carlo photon and electron transport code package penORNL was developed at Oak Ridge National Laboratory to enable advanced scanning electron microscope (SEM) simulations on high-performance computing systems. This paper discusses the implementations, capabilities and parallel performance of the new code package. penORNL uses PENELOPE for its physics calculations and provides all available PENELOPE features to the users, as well as some new features including source definitions specifically developed for SEM simulations, a pulse-height tally capability for detailed simulations of gamma and x-ray detectors, and a modified interaction forcing mechanism to enable accurate energy deposition calculations. The parallel performance of penORNL was extensively tested with several model problems, and very good linear parallel scaling was observed with up to 512 processors. penORNL, along with its new features, will be available for SEM simulations upon completion of the new pulse-height tally implementation.

  19. Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes

    International Nuclear Information System (INIS)

    Smith, L.M.; Hochstedler, R.D.

    1997-01-01

    Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code)

  20. TOPIC: a debugging code for torus geometry input data of Monte Carlo transport code

    International Nuclear Information System (INIS)

    Iida, Hiromasa; Kawasaki, Hiromitsu.

    1979-06-01

    TOPIC has been developed for debugging geometry input data of the Monte Carlo transport code. the code has the following features: (1) It debugs the geometry input data of not only MORSE-GG but also MORSE-I capable of treating torus geometry. (2) Its calculation results are shown in figures drawn by Plotter or COM, and the regions not defined or doubly defined are easily detected. (3) It finds a multitude of input data errors in a single run. (4) The input data required in this code are few, so that it is readily usable in a time sharing system of FACOM 230-60/75 computer. Example TOPIC calculations in design study of tokamak fusion reactors (JXFR, INTOR-J) are presented. (author)

  1. Accelerating execution of the integrated TIGER series Monte Carlo radiation transport codes

    Science.gov (United States)

    Smith, L. M.; Hochstedler, R. D.

    1997-02-01

    Execution of the integrated TIGER series (ITS) of coupled electron/photon Monte Carlo radiation transport codes has been accelerated by modifying the FORTRAN source code for more efficient computation. Each member code of ITS was benchmarked and profiled with a specific test case that directed the acceleration effort toward the most computationally intensive subroutines. Techniques for accelerating these subroutines included replacing linear search algorithms with binary versions, replacing the pseudo-random number generator, reducing program memory allocation, and proofing the input files for geometrical redundancies. All techniques produced identical or statistically similar results to the original code. Final benchmark timing of the accelerated code resulted in speed-up factors of 2.00 for TIGER (the one-dimensional slab geometry code), 1.74 for CYLTRAN (the two-dimensional cylindrical geometry code), and 1.90 for ACCEPT (the arbitrary three-dimensional geometry code).

  2. Space applications of the MITS electron-photon Monte Carlo transport code system

    International Nuclear Information System (INIS)

    Kensek, R.P.; Lorence, L.J.; Halbleib, J.A.; Morel, J.E.

    1996-01-01

    The MITS multigroup/continuous-energy electron-photon Monte Carlo transport code system has matured to the point that it is capable of addressing more realistic three-dimensional adjoint applications. It is first employed to efficiently predict point doses as a function of source energy for simple three-dimensional experimental geometries exposed to simulated uniform isotropic planar sources of monoenergetic electrons up to 4.0 MeV. Results are in very good agreement with experimental data. It is then used to efficiently simulate dose to a detector in a subsystem of a GPS satellite due to its natural electron environment, employing a relatively complex model of the satellite. The capability for survivability analysis of space systems is demonstrated, and results are obtained with and without variance reduction

  3. Monte Carlo photon transport on shared memory and distributed memory parallel processors

    International Nuclear Information System (INIS)

    Martin, W.R.; Wan, T.C.; Abdel-Rahman, T.S.; Mudge, T.N.; Miura, K.

    1987-01-01

    Parallelized Monte Carlo algorithms for analyzing photon transport in an inertially confined fusion (ICF) plasma are considered. Algorithms were developed for shared memory (vector and scalar) and distributed memory (scalar) parallel processors. The shared memory algorithm was implemented on the IBM 3090/400, and timing results are presented for dedicated runs with two, three, and four processors. Two alternative distributed memory algorithms (replication and dispatching) were implemented on a hypercube parallel processor (1 through 64 nodes). The replication algorithm yields essentially full efficiency for all cube sizes; with the 64-node configuration, the absolute performance is nearly the same as with the CRAY X-MP. The dispatching algorithm also yields efficiencies above 80% in a large simulation for the 64-processor configuration

  4. Monte Carlo electron-transport calculations for clinical beams using energy grouping

    Energy Technology Data Exchange (ETDEWEB)

    Teng, S P; Anderson, D W; Lindstrom, D G

    1986-01-01

    A Monte Carlo program has been utilized to study the penetration of broad electron beams into a water phantom. The MORSE-E code, originally developed for neutron and photon transport, was chosen for adaptation to electrons because of its versatility. The electron energy degradation model employed logarithmic spacing of electron energy groups and included effects of elastic scattering, inelastic-moderate-energy-loss-processes and inelastic-large-energy-loss-processes (catastrophic). Energy straggling and angular deflections were modeled from group to group, using the Moeller cross section for energy loss, and Goudsmit-Saunderson theory to describe angular deflections. The resulting energy- and electron-deposition distributions in depth were obtained at 10 and 20 MeV and are compared with ETRAN results and broad beam experimental data from clinical accelerators.

  5. GPU-based high performance Monte Carlo simulation in neutron transport

    Energy Technology Data Exchange (ETDEWEB)

    Heimlich, Adino; Mol, Antonio C.A.; Pereira, Claudio M.N.A. [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Inteligencia Artificial Aplicada], e-mail: cmnap@ien.gov.br

    2009-07-01

    Graphics Processing Units (GPU) are high performance co-processors intended, originally, to improve the use and quality of computer graphics applications. Since researchers and practitioners realized the potential of using GPU for general purpose, their application has been extended to other fields out of computer graphics scope. The main objective of this work is to evaluate the impact of using GPU in neutron transport simulation by Monte Carlo method. To accomplish that, GPU- and CPU-based (single and multicore) approaches were developed and applied to a simple, but time-consuming problem. Comparisons demonstrated that the GPU-based approach is about 15 times faster than a parallel 8-core CPU-based approach also developed in this work. (author)

  6. GPU-based high performance Monte Carlo simulation in neutron transport

    International Nuclear Information System (INIS)

    Heimlich, Adino; Mol, Antonio C.A.; Pereira, Claudio M.N.A.

    2009-01-01

    Graphics Processing Units (GPU) are high performance co-processors intended, originally, to improve the use and quality of computer graphics applications. Since researchers and practitioners realized the potential of using GPU for general purpose, their application has been extended to other fields out of computer graphics scope. The main objective of this work is to evaluate the impact of using GPU in neutron transport simulation by Monte Carlo method. To accomplish that, GPU- and CPU-based (single and multicore) approaches were developed and applied to a simple, but time-consuming problem. Comparisons demonstrated that the GPU-based approach is about 15 times faster than a parallel 8-core CPU-based approach also developed in this work. (author)

  7. Scalable and massively parallel Monte Carlo photon transport simulations for heterogeneous computing platforms.

    Science.gov (United States)

    Yu, Leiming; Nina-Paravecino, Fanny; Kaeli, David; Fang, Qianqian

    2018-01-01

    We present a highly scalable Monte Carlo (MC) three-dimensional photon transport simulation platform designed for heterogeneous computing systems. Through the development of a massively parallel MC algorithm using the Open Computing Language framework, this research extends our existing graphics processing unit (GPU)-accelerated MC technique to a highly scalable vendor-independent heterogeneous computing environment, achieving significantly improved performance and software portability. A number of parallel computing techniques are investigated to achieve portable performance over a wide range of computing hardware. Furthermore, multiple thread-level and device-level load-balancing strategies are developed to obtain efficient simulations using multiple central processing units and GPUs. (2018) COPYRIGHT Society of Photo-Optical Instrumentation Engineers (SPIE).

  8. Application of the direct simulation Monte Carlo method to nanoscale heat transfer between a soot particle and the surrounding gas

    International Nuclear Information System (INIS)

    Yang, M.; Liu, F.; Smallwood, G.J.

    2004-01-01

    Laser-Induced Incandescence (LII) technique has been widely used to measure soot volume fraction and primary particle size in flames and engine exhaust. Currently there is lack of quantitative understanding of the shielding effect of aggregated soot particles on its conduction heat loss rate to the surrounding gas. The conventional approach for this problem would be the application of the Monte Carlo (MC) method. This method is based on simulation of the trajectories of individual molecules and calculation of the heat transfer at each of the molecule/molecule collisions and the molecule/particle collisions. As the first step toward calculating the heat transfer between a soot aggregate and the surrounding gas, the Direct Simulation Monte Carlo (DSMC) method was used in this study to calculate the heat transfer rate between a single spherical aerosol particle and its cooler surrounding gas under different conditions of temperature, pressure, and the accommodation coefficient. A well-defined and simple hard sphere model was adopted to describe molecule/molecule elastic collisions. A combination of the specular reflection and completely diffuse reflection model was used to consider molecule/particle collisions. The results obtained by DSMC are in good agreement with the known analytical solution of heat transfer rate for an isolated, motionless sphere in the free-molecular regime. Further the DSMC method was applied to calculate the heat transfer in the transition regime. Our present DSMC results agree very well with published DSMC data. (author)

  9. ITS, TIGER System of Coupled Electron Photon Transport by Monte-Carlo

    International Nuclear Information System (INIS)

    Halbleib, J.A.; Mehlhorn, T.A.; Young, M.F.

    1996-01-01

    1 - Description of program or function: ITS permits a state-of-the-art Monte Carlo solution of linear time-integrated coupled electron/ photon radiation transport problems with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. 2 - Method of solution: Through a machine-portable utility that emulates the basic features of the CDC UPDATE processor, the user selects one of eight codes for running on a machine of one of four (at least) major vendors. With the ITS-3.0 release the PSR-0245/UPEML package is included to perform these functions. The ease with which this utility is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is maximized by employing the best available cross sections and sampling distributions, and the most complete physical model for describing the production and transport of the electron/ photon cascade from 1.0 GeV down to 1.0 keV. Flexibility of construction permits the codes to be tailored to specific applications and the capabilities of the codes to be extended to more complex applications through update procedures. 3 - Restrictions on the complexity of the problem: - Restrictions and/or limitations for ITS depend upon the local operating system

  10. 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...

  11. Applications Of Monte Carlo Radiation Transport Simulation Techniques For Predicting Single Event Effects In Microelectronics

    International Nuclear Information System (INIS)

    Warren, Kevin; Reed, Robert; Weller, Robert; Mendenhall, Marcus; Sierawski, Brian; Schrimpf, Ronald

    2011-01-01

    MRED (Monte Carlo Radiative Energy Deposition) is Vanderbilt University's Geant4 application for simulating radiation events in semiconductors. Geant4 is comprised of the best available computational physics models for the transport of radiation through matter. In addition to basic radiation transport physics contained in the Geant4 core, MRED has the capability to track energy loss in tetrahedral geometric objects, includes a cross section biasing and track weighting technique for variance reduction, and additional features relevant to semiconductor device applications. The crucial element of predicting Single Event Upset (SEU) parameters using radiation transport software is the creation of a dosimetry model that accurately approximates the net collected charge at transistor contacts as a function of deposited energy. The dosimetry technique described here is the multiple sensitive volume (MSV) model. It is shown to be a reasonable approximation of the charge collection process and its parameters can be calibrated to experimental measurements of SEU cross sections. The MSV model, within the framework of MRED, is examined for heavy ion and high-energy proton SEU measurements of a static random access memory.

  12. Interfacial Phonon Transport Through Si/Ge Multilayer Film Using Monte Carlo Scheme With Spectral Transmissivity

    Directory of Open Access Journals (Sweden)

    Xin Ran

    2018-05-01

    Full Text Available The knowledge of interfacial phonon transport accounting for detailed phonon spectral properties is desired because of its importance for design of nanoscale energy systems. In this work, we investigate the interfacial phonon transport through Si/Ge multilayer films using an efficient Monte Carlo scheme with spectral transmissivity, which is validated for cross-plane phonon transport through both Si/Ge single-layer and Si/Ge bi-layer thin films by comparing with the discrete-ordinates solution. Different thermal boundary conductances between even the same material pair are declared at different interfaces within the multilayer system. Furthermore, the thermal boundary conductances at different interfaces show different trends with varying total system size, with the variation slope, very different as well. The results are much different from those in the bi-layer thin film or periodic superlattice. These unusual behaviors can be attributed to the combined interfacial local non-equilibrium effect and constraint effect from other interfaces.

  13. ITS Version 6 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes.

    Energy Technology Data Exchange (ETDEWEB)

    Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William

    2008-04-01

    ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of lineartime-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 90. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.

  14. Monte Carlo simulation of ballistic transport in high-mobility channels

    Energy Technology Data Exchange (ETDEWEB)

    Sabatini, G; Marinchio, H; Palermo, C; Varani, L; Daoud, T; Teissier, R [Institut d' Electronique du Sud (CNRS UMR 5214) - Universite Montpellier II (France); Rodilla, H; Gonzalez, T; Mateos, J, E-mail: sabatini@ies.univ-montp2.f [Departamento de Fisica Aplicada - Universidad de Salamanca (Spain)

    2009-11-15

    By means of Monte Carlo simulations coupled with a two-dimensional Poisson solver, we evaluate directly the possibility to use high mobility materials in ultra fast devices exploiting ballistic transport. To this purpose, we have calculated specific physical quantities such as the transit time, the transit velocity, the free flight time and the mean free path as functions of applied voltage in InAs channels with different lengths, from 2000 nm down to 50 nm. In this way the transition from diffusive to ballistic transport is carefully described. We remark a high value of the mean transit velocity with a maximum of 14x10{sup 5} m/s for a 50 nm-long channel and a transit time shorter than 0.1 ps, corresponding to a cutoff frequency in the terahertz domain. The percentage of ballistic electrons and the number of scatterings as functions of distance are also reported, showing the strong influence of quasi-ballistic transport in the shorter channels.

  15. Monte Carlo simulation of ballistic transport in high-mobility channels

    International Nuclear Information System (INIS)

    Sabatini, G; Marinchio, H; Palermo, C; Varani, L; Daoud, T; Teissier, R; Rodilla, H; Gonzalez, T; Mateos, J

    2009-01-01

    By means of Monte Carlo simulations coupled with a two-dimensional Poisson solver, we evaluate directly the possibility to use high mobility materials in ultra fast devices exploiting ballistic transport. To this purpose, we have calculated specific physical quantities such as the transit time, the transit velocity, the free flight time and the mean free path as functions of applied voltage in InAs channels with different lengths, from 2000 nm down to 50 nm. In this way the transition from diffusive to ballistic transport is carefully described. We remark a high value of the mean transit velocity with a maximum of 14x10 5 m/s for a 50 nm-long channel and a transit time shorter than 0.1 ps, corresponding to a cutoff frequency in the terahertz domain. The percentage of ballistic electrons and the number of scatterings as functions of distance are also reported, showing the strong influence of quasi-ballistic transport in the shorter channels.

  16. Toolkit for high performance Monte Carlo radiation transport and activation calculations for shielding applications in ITER

    International Nuclear Information System (INIS)

    Serikov, A.; Fischer, U.; Grosse, D.; Leichtle, D.; Majerle, M.

    2011-01-01

    The Monte Carlo (MC) method is the most suitable computational technique of radiation transport for shielding applications in fusion neutronics. This paper is intended for sharing the results of long term experience of the fusion neutronics group at Karlsruhe Institute of Technology (KIT) in radiation shielding calculations with the MCNP5 code for the ITER fusion reactor with emphasizing on the use of several ITER project-driven computer programs developed at KIT. Two of them, McCad and R2S, seem to be the most useful in radiation shielding analyses. The McCad computer graphical tool allows to perform automatic conversion of the MCNP models from the underlying CAD (CATIA) data files, while the R2S activation interface couples the MCNP radiation transport with the FISPACT activation allowing to estimate nuclear responses such as dose rate and nuclear heating after the ITER reactor shutdown. The cell-based R2S scheme was applied in shutdown photon dose analysis for the designing of the In-Vessel Viewing System (IVVS) and the Glow Discharge Cleaning (GDC) unit in ITER. Newly developed at KIT mesh-based R2S feature was successfully tested on the shutdown dose rate calculations for the upper port in the Neutral Beam (NB) cell of ITER. The merits of McCad graphical program were broadly acknowledged by the neutronic analysts and its continuous improvement at KIT has introduced its stable and more convenient run with its Graphical User Interface. Detailed 3D ITER neutronic modeling with the MCNP Monte Carlo method requires a lot of computation resources, inevitably leading to parallel calculations on clusters. Performance assessments of the MCNP5 parallel runs on the JUROPA/HPC-FF supercomputer cluster permitted to find the optimal number of processors for ITER-type runs. (author)

  17. An investigation of the adjoint method for external beam radiation therapy treatment planning using Monte Carlo transport

    International Nuclear Information System (INIS)

    Kowalok, M.; Mackie, T.R.

    2001-01-01

    A relatively new technique for achieving the right dose to the right tissue, is intensity modulated radiation therapy (IMRT). In this technique, a megavoltage x-ray beam is rotated around a patient, and the intensity and shape of the beam is modulated as a function of source position and patient anatomy. The relationship between beam-let intensity and patient dose can be expressed under a matrix form where the matrix D ij represents the dose delivered to voxel i by beam-let j per unit fluence. The D ij influence matrix is the key element that enables this approach. In this regard, sensitivity theory lends itself in a natural way to the process of computing beam weights for treatment planning. The solution of the adjoint form of the Boltzmann equation is an adjoint function that describes the importance of particles throughout the system in contributing to the detector response. In this case, adjoint methods can provide the sensitivity of the dose at a single point in the patient with respect to all points in the source field. The purpose of this study is to investigate the feasibility of using the adjoint method and Monte Carlo transport for radiation therapy treatment planning

  18. A graphics-card implementation of Monte-Carlo simulations for cosmic-ray transport

    Science.gov (United States)

    Tautz, R. C.

    2016-05-01

    A graphics card implementation of a test-particle simulation code is presented that is based on the CUDA extension of the C/C++ programming language. The original CPU version has been developed for the calculation of cosmic-ray diffusion coefficients in artificial Kolmogorov-type turbulence. In the new implementation, the magnetic turbulence generation, which is the most time-consuming part, is separated from the particle transport and is performed on a graphics card. In this article, the modification of the basic approach of integrating test particle trajectories to employ the SIMD (single instruction, multiple data) model is presented and verified. The efficiency of the new code is tested and several language-specific accelerating factors are discussed. For the example of isotropic magnetostatic turbulence, sample results are shown and a comparison to the results of the CPU implementation is performed.

  19. FLUKA A multi-particle transport code (program version 2005)

    CERN Document Server

    Ferrari, A; Fassò, A; Ranft, Johannes

    2005-01-01

    This report describes the 2005 version of the Fluka particle transport code. The first part introduces the basic notions, describes the modular structure of the system, and contains an installation and beginner’s guide. The second part complements this initial information with details about the various components of Fluka and how to use them. It concludes with a detailed history and bibliography.

  20. Linear kinetic theory and particle transport in stochastic mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Pomraning, G.C. [Univ. of California, Los Angeles, CA (United States)

    1995-12-31

    We consider the formulation of linear transport and kinetic theory describing energy and particle flow in a random mixture of two or more immiscible materials. Following an introduction, we summarize early and fundamental work in this area, and we conclude with a brief discussion of recent results.

  1. Transient fluctuation relations for time-dependent particle transport

    Science.gov (United States)

    Altland, Alexander; de Martino, Alessandro; Egger, Reinhold; Narozhny, Boris

    2010-09-01

    We consider particle transport under the influence of time-varying driving forces, where fluctuation relations connect the statistics of pairs of time-reversed evolutions of physical observables. In many “mesoscopic” transport processes, the effective many-particle dynamics is dominantly classical while the microscopic rates governing particle motion are of quantum-mechanical origin. We here employ the stochastic path-integral approach as an optimal tool to probe the fluctuation statistics in such applications. Describing the classical limit of the Keldysh quantum nonequilibrium field theory, the stochastic path integral encapsulates the quantum origin of microscopic particle exchange rates. Dynamically, it is equivalent to a transport master equation which is a formalism general enough to describe many applications of practical interest. We apply the stochastic path integral to derive general functional fluctuation relations for current flow induced by time-varying forces. We show that the successive measurement processes implied by this setup do not put the derivation of quantum fluctuation relations in jeopardy. While in many cases the fluctuation relation for a full time-dependent current profile may contain excessive information, we formulate a number of reduced relations, and demonstrate their application to mesoscopic transport. Examples include the distribution of transmitted charge, where we show that the derivation of a fluctuation relation requires the combined monitoring of the statistics of charge and work.

  2. Los Alamos neutral particle transport codes: New and enhanced capabilities

    International Nuclear Information System (INIS)

    Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Clark, B.A.; Koch, K.R.; Marr, D.R.

    1992-01-01

    We present new developments in Los Alamos discrete-ordinates transport codes and introduce THREEDANT, the latest in the series of Los Alamos discrete ordinates transport codes. THREEDANT solves the multigroup, neutral-particle transport equation in X-Y-Z and R-Θ-Z geometries. THREEDANT uses computationally efficient algorithms: Diffusion Synthetic Acceleration (DSA) is used to accelerate the convergence of transport iterations, the DSA solution is accelerated using the multigrid technique. THREEDANT runs on a wide range of computers, from scientific workstations to CRAY supercomputers. The algorithms are highly vectorized on CRAY computers. Recently, the THREEDANT transport algorithm was implemented on the massively parallel CM-2 computer, with performance that is comparable to a single-processor CRAY-YMP We present the results of THREEDANT analysis of test problems

  3. A generalized transport-velocity formulation for smoothed particle hydrodynamics

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Chi; Hu, Xiangyu Y., E-mail: xiangyu.hu@tum.de; Adams, Nikolaus A.

    2017-05-15

    The standard smoothed particle hydrodynamics (SPH) method suffers from tensile instability. In fluid-dynamics simulations this instability leads to particle clumping and void regions when negative pressure occurs. In solid-dynamics simulations, it results in unphysical structure fragmentation. In this work the transport-velocity formulation of Adami et al. (2013) is generalized for providing a solution of this long-standing problem. Other than imposing a global background pressure, a variable background pressure is used to modify the particle transport velocity and eliminate the tensile instability completely. Furthermore, such a modification is localized by defining a shortened smoothing length. The generalized formulation is suitable for fluid and solid materials with and without free surfaces. The results of extensive numerical tests on both fluid and solid dynamics problems indicate that the new method provides a unified approach for multi-physics SPH simulations.

  4. Directed transport of confined Brownian particles with torque

    Science.gov (United States)

    Radtke, Paul K.; Schimansky-Geier, Lutz

    2012-05-01

    We investigate the influence of an additional torque on the motion of Brownian particles confined in a channel geometry with varying width. The particles are driven by random fluctuations modeled by an Ornstein-Uhlenbeck process with given correlation time τc. The latter causes persistent motion and is implemented as (i) thermal noise in equilibrium and (ii) noisy propulsion in nonequilibrium. In the nonthermal process a directed transport emerges; its properties are studied in detail with respect to the correlation time, the torque, and the channel geometry. Eventually, the transport mechanism is traced back to a persistent sliding of particles along the even boundaries in contrast to scattered motion at uneven or rough ones.

  5. Estimates of Lagrangian particle transport by wave groups: forward transport by Stokes drift and backward transport by the return flow

    Science.gov (United States)

    van den Bremer, Ton S.; Taylor, Paul H.

    2014-11-01

    Although the literature has examined Stokes drift, the net Lagrangian transport by particles due to of surface gravity waves, in great detail, the motion of fluid particles transported by surface gravity wave groups has received considerably less attention. In practice nevertheless, the wave field on the open sea often has a group-like structure. The motion of particles is different, as particles at sufficient depth are transported backwards by the Eulerian return current that was first described by Longuet-Higgins & Stewart (1962) and forms an inseparable counterpart of Stokes drift for wave groups ensuring the (irrotational) mass balance holds. We use WKB theory to study the variation of the Lagrangian transport by the return current with depth distinguishing two-dimensional seas, three-dimensional seas, infinite depth and finite depth. We then provide dimensional estimates of the net horizontal Lagrangian transport by the Stokes drift on the one hand and the return flow on the other hand for realistic sea states in all four cases. Finally we propose a simple scaling relationship for the transition depth: the depth above which Lagrangian particles are transported forwards by the Stokes drift and below which such particles are transported backwards by the return current.

  6. Particle transport methods for LWR dosimetry developed by the Penn State transport theory group

    International Nuclear Information System (INIS)

    Haghighat, A.; Petrovic, B.

    1997-01-01

    This paper reviews advanced particle transport theory methods developed by the Penn State Transport Theory Group (PSTTG) over the past several years. These methods have been developed in response to increasing needs for accuracy of results and for three-dimensional modeling of nuclear systems

  7. Fundamentals of Monte Carlo

    International Nuclear Information System (INIS)

    Wollaber, Allan Benton

    2016-01-01

    This is a powerpoint presentation which serves as lecture material for the Parallel Computing summer school. It goes over the fundamentals of the Monte Carlo calculation method. The material is presented according to the following outline: Introduction (background, a simple example: estimating @@), Why does this even work? (The Law of Large Numbers, The Central Limit Theorem), How to sample (inverse transform sampling, rejection), and An example from particle transport.

  8. Fundamentals of Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Wollaber, Allan Benton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-06-16

    This is a powerpoint presentation which serves as lecture material for the Parallel Computing summer school. It goes over the fundamentals of the Monte Carlo calculation method. The material is presented according to the following outline: Introduction (background, a simple example: estimating π), Why does this even work? (The Law of Large Numbers, The Central Limit Theorem), How to sample (inverse transform sampling, rejection), and An example from particle transport.

  9. Consistency evaluation between EGSnrc and Geant4 charged particle transport in an equilibrium magnetic field.

    Science.gov (United States)

    Yang, Y M; Bednarz, B

    2013-02-21

    Following the proposal by several groups to integrate magnetic resonance imaging (MRI) with radiation therapy, much attention has been afforded to examining the impact of strong (on the order of a Tesla) transverse magnetic fields on photon dose distributions. The effect of the magnetic field on dose distributions must be considered in order to take full advantage of the benefits of real-time intra-fraction imaging. In this investigation, we compared the handling of particle transport in magnetic fields between two Monte Carlo codes, EGSnrc and Geant4, to analyze various aspects of their electromagnetic transport algorithms; both codes are well-benchmarked for medical physics applications in the absence of magnetic fields. A water-air-water slab phantom and a water-lung-water slab phantom were used to highlight dose perturbations near high- and low-density interfaces. We have implemented a method of calculating the Lorentz force in EGSnrc based on theoretical models in literature, and show very good consistency between the two Monte Carlo codes. This investigation further demonstrates the importance of accurate dosimetry for MRI-guided radiation therapy (MRIgRT), and facilitates the integration of a ViewRay MRIgRT system in the University of Wisconsin-Madison's Radiation Oncology Department.

  10. Consistency evaluation between EGSnrc and Geant4 charged particle transport in an equilibrium magnetic field

    International Nuclear Information System (INIS)

    Yang, Y M; Bednarz, B

    2013-01-01

    Following the proposal by several groups to integrate magnetic resonance imaging (MRI) with radiation therapy, much attention has been afforded to examining the impact of strong (on the order of a Tesla) transverse magnetic fields on photon dose distributions. The effect of the magnetic field on dose distributions must be considered in order to take full advantage of the benefits of real-time intra-fraction imaging. In this investigation, we compared the handling of particle transport in magnetic fields between two Monte Carlo codes, EGSnrc and Geant4, to analyze various aspects of their electromagnetic transport algorithms; both codes are well-benchmarked for medical physics applications in the absence of magnetic fields. A water–air–water slab phantom and a water–lung–water slab phantom were used to highlight dose perturbations near high- and low-density interfaces. We have implemented a method of calculating the Lorentz force in EGSnrc based on theoretical models in literature, and show very good consistency between the two Monte Carlo codes. This investigation further demonstrates the importance of accurate dosimetry for MRI-guided radiation therapy (MRIgRT), and facilitates the integration of a ViewRay MRIgRT system in the University of Wisconsin-Madison's Radiation Oncology Department. (note)

  11. SU-E-T-590: Optimizing Magnetic Field Strengths with Matlab for An Ion-Optic System in Particle Therapy Consisting of Two Quadrupole Magnets for Subsequent Simulations with the Monte-Carlo Code FLUKA

    International Nuclear Information System (INIS)

    Baumann, K; Weber, U; Simeonov, Y; Zink, K

    2015-01-01

    Purpose: Aim of this study was to optimize the magnetic field strengths of two quadrupole magnets in a particle therapy facility in order to obtain a beam quality suitable for spot beam scanning. Methods: The particle transport through an ion-optic system of a particle therapy facility consisting of the beam tube, two quadrupole magnets and a beam monitor system was calculated with the help of Matlab by using matrices that solve the equation of motion of a charged particle in a magnetic field and field-free region, respectively. The magnetic field strengths were optimized in order to obtain a circular and thin beam spot at the iso-center of the therapy facility. These optimized field strengths were subsequently transferred to the Monte-Carlo code FLUKA and the transport of 80 MeV/u C12-ions through this ion-optic system was calculated by using a user-routine to implement magnetic fields. The fluence along the beam-axis and at the iso-center was evaluated. Results: The magnetic field strengths could be optimized by using Matlab and transferred to the Monte-Carlo code FLUKA. The implementation via a user-routine was successful. Analyzing the fluence-pattern along the beam-axis the characteristic focusing and de-focusing effects of the quadrupole magnets could be reproduced. Furthermore the beam spot at the iso-center was circular and significantly thinner compared to an unfocused beam. Conclusion: In this study a Matlab tool was developed to optimize magnetic field strengths for an ion-optic system consisting of two quadrupole magnets as part of a particle therapy facility. These magnetic field strengths could subsequently be transferred to and implemented in the Monte-Carlo code FLUKA to simulate the particle transport through this optimized ion-optic system

  12. Applying Dispersive Changes to Lagrangian Particles in Groundwater Transport Models

    Science.gov (United States)

    Konikow, Leonard F.

    2010-01-01

    Method-of-characteristics groundwater transport models require that changes in concentrations computed within an Eulerian framework to account for dispersion be transferred to moving particles used to simulate advective transport. A new algorithm was developed to accomplish this transfer between nodal values and advecting particles more precisely and realistically compared to currently used methods. The new method scales the changes and adjustments of particle concentrations relative to limiting bounds of concentration values determined from the population of adjacent nodal values. The method precludes unrealistic undershoot or overshoot for concentrations of individual particles. In the new method, if dispersion causes cell concentrations to decrease during a time step, those particles in the cell having the highest concentration will decrease the most, and those with the lowest concentration will decrease the least. The converse is true if dispersion is causing concentrations to increase. Furthermore, if the initial concentration on a particle is outside the range of the adjacent nodal values, it will automatically be adjusted in the direction of the acceptable range of values. The new method is inherently mass conservative. ?? US Government 2010.

  13. Monte Carlo study of electron-plasmon scattering effects on hot electron transport in GaAs

    International Nuclear Information System (INIS)

    Popov, V.V.; Bagaeva, T.Yu.; Solodkaya, T.I.

    1994-07-01

    It is shown using Monte Carlo simulation that electron-plasmon scattering affects substantially the hot-electron energy distribution function and transport properties in bulk GaAs. However, this effect is found to be much less than that predicted in earlier paper of other authors. (author). 5 refs, 7 figs

  14. Parallelization of MCNP Monte Carlo neutron and photon transport code in parallel virtual machine and message passing interface

    International Nuclear Information System (INIS)

    Deng Li; Xie Zhongsheng

    1999-01-01

    The coupled neutron and photon transport Monte Carlo code MCNP (version 3B) has been parallelized in parallel virtual machine (PVM) and message passing interface (MPI) by modifying a previous serial code. The new code has been verified by solving sample problems. The speedup increases linearly with the number of processors and the average efficiency is up to 99% for 12-processor. (author)

  15. Monte Carlo method implemented in a finite element code with application to dynamic vacuum in particle accelerators

    CERN Document Server

    Garion, C

    2009-01-01

    Modern particle accelerators require UHV conditions during their operation. In the accelerating cavities, breakdowns can occur, releasing large amount of gas into the vacuum chamber. To determine the pressure profile along the cavity as a function of time, the time-dependent behaviour of the gas has to be simulated. To do that, it is useful to apply accurate three-dimensional method, such as Test Particles Monte Carlo. In this paper, a time-dependent Test Particles Monte Carlo is used. It has been implemented in a Finite Element code, CASTEM. The principle is to track a sample of molecules during time. The complex geometry of the cavities can be created either in the FE code or in a CAD software (CATIA in our case). The interface between the two softwares to export the geometry from CATIA to CASTEM is given. The algorithm of particle tracking for collisionless flow in the FE code is shown. Thermal outgassing, pumping surfaces and electron and/or ion stimulated desorption can all be generated as well as differ...

  16. Modeling reactive transport with particle tracking and kernel estimators

    Science.gov (United States)

    Rahbaralam, Maryam; Fernandez-Garcia, Daniel; Sanchez-Vila, Xavier

    2015-04-01

    Groundwater reactive transport models are useful to assess and quantify the fate and transport of contaminants in subsurface media and are an essential tool for the analysis of coupled physical, chemical, and biological processes in Earth Systems. Particle Tracking Method (PTM) provides a computationally efficient and adaptable approach to solve the solute transport partial differential equation. On a molecular level, chemical reactions are the result of collisions, combinations, and/or decay of different species. For a well-mixed system, the chem- ical reactions are controlled by the classical thermodynamic rate coefficient. Each of these actions occurs with some probability that is a function of solute concentrations. PTM is based on considering that each particle actually represents a group of molecules. To properly simulate this system, an infinite number of particles is required, which is computationally unfeasible. On the other hand, a finite number of particles lead to a poor-mixed system which is limited by diffusion. Recent works have used this effect to actually model incomplete mix- ing in naturally occurring porous media. In this work, we demonstrate that this effect in most cases should be attributed to a defficient estimation of the concentrations and not to the occurrence of true incomplete mixing processes in porous media. To illustrate this, we show that a Kernel Density Estimation (KDE) of the concentrations can approach the well-mixed solution with a limited number of particles. KDEs provide weighting functions of each particle mass that expands its region of influence, hence providing a wider region for chemical reactions with time. Simulation results show that KDEs are powerful tools to improve state-of-the-art simulations of chemical reactions and indicates that incomplete mixing in diluted systems should be modeled based on alternative conceptual models and not on a limited number of particles.

  17. Monte Carlo simulation of neutron transport in a homogeneous reactor with a partially inserted control rod

    International Nuclear Information System (INIS)

    Karlsson, J.K.H.; Linden, P.

    1997-01-01

    The neutron transport in a bare, cylindrical and homogeneous reactor, with and without the presence of a central partially inserted control rod, has been simulated by using a Monte Carlo transport code. The behaviour of both the flux and current in this system have been investigated. We have found that the flux and especially the current are strongly affected by the presence of the control rod in its close vicinity. The results indicate the feasibility to identify the position and especially the tip of the rod from the flux and current. Further, the direction to the rod can be found from the current vector. The information content regarding the position of the rod, in both the neutron flux and the current, decays strongly as a function of distance and it is dependent on the size of the rod. In our model, the practical range over which the flux or current can be a useful indicator of the position of the tip of the rod is about 10-15 cm for a rod with a diameter of 2 cm. The practical range for identification of the position of the rod is greater for a rod of larger diameter

  18. Oxygen transport properties estimation by classical trajectory–direct simulation Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Bruno, Domenico, E-mail: domenico.bruno@cnr.it [Istituto di Metodologie Inorganiche e dei Plasmi, Consiglio Nazionale delle Ricerche– Via G. Amendola 122, 70125 Bari (Italy); Frezzotti, Aldo, E-mail: aldo.frezzotti@polimi.it; Ghiroldi, Gian Pietro, E-mail: gpghiro@gmail.com [Dipartimento di Scienze e Tecnologie Aerospaziali, Politecnico di Milano–Via La Masa 34, 20156 Milano (Italy)

    2015-05-15

    Coupling direct simulation Monte Carlo (DSMC) simulations with classical trajectory calculations is a powerful tool to improve predictive capabilities of computational dilute gas dynamics. The considerable increase in computational effort outlined in early applications of the method can be compensated by running simulations on massively parallel computers. In particular, Graphics Processing Unit acceleration has been found quite effective in reducing computing time of classical trajectory (CT)-DSMC simulations. The aim of the present work is to study dilute molecular oxygen flows by modeling binary collisions, in the rigid rotor approximation, through an accurate Potential Energy Surface (PES), obtained by molecular beams scattering. The PES accuracy is assessed by calculating molecular oxygen transport properties by different equilibrium and non-equilibrium CT-DSMC based simulations that provide close values of the transport properties. Comparisons with available experimental data are presented and discussed in the temperature range 300–900 K, where vibrational degrees of freedom are expected to play a limited (but not always negligible) role.

  19. Effect of interface roughness on the carrier transport in germanium MOSFETs investigated by Monte Carlo method

    International Nuclear Information System (INIS)

    Gang, Du; Xiao-Yan, Liu; Zhi-Liang, Xia; Jing-Feng, Yang; Ru-Qi, Han

    2010-01-01

    Interface roughness strongly influences the performance of germanium metal–organic–semiconductor field effect transistors (MOSFETs). In this paper, a 2D full-band Monte Carlo simulator is used to study the impact of interface roughness scattering on electron and hole transport properties in long- and short- channel Ge MOSFETs inversion layers. The carrier effective mobility in the channel of Ge MOSFETs and the in non-equilibrium transport properties are investigated. Results show that both electron and hole mobility are strongly influenced by interface roughness scattering. The output curves for 50 nm channel-length double gate n and p Ge MOSFET show that the drive currents of n- and p-Ge MOSFETs have significant improvement compared with that of Si n- and p-MOSFETs with smooth interface between channel and gate dielectric. The 82% and 96% drive current enhancement are obtained for the n- and p-MOSFETs with the completely smooth interface. However, the enhancement decreases sharply with the increase of interface roughness. With the very rough interface, the drive currents of Ge MOSFETs are even less than that of Si MOSFETs. Moreover, the significant velocity overshoot also has been found in Ge MOSFETs. (condensed matter: electronic structure, electrical, magnetic, and optical properties)

  20. 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)

  1. Co-combustion of peanut hull and coal blends: Artificial neural networks modeling, particle swarm optimization and Monte Carlo simulation.

    Science.gov (United States)

    Buyukada, Musa

    2016-09-01

    Co-combustion of coal and peanut hull (PH) were investigated using artificial neural networks (ANN), particle swarm optimization, and Monte Carlo simulation as a function of blend ratio, heating rate, and temperature. The best prediction was reached by ANN61 multi-layer perception model with a R(2) of 0.99994. Blend ratio of 90 to 10 (PH to coal, wt%), temperature of 305°C, and heating rate of 49°Cmin(-1) were determined as the optimum input values and yield of 87.4% was obtained under PSO optimized conditions. The validation experiments resulted in yields of 87.5%±0.2 after three replications. Monte Carlo simulations were used for the probabilistic assessments of stochastic variability and uncertainty associated with explanatory variables of co-combustion process. Copyright © 2016 Elsevier Ltd. All rights reserved.

  2. Determination of peripheral underdosage at the lung-tumor interface using Monte Carlo radiation transport calculations

    International Nuclear Information System (INIS)

    Taylor, Michael; Dunn, Leon; Kron, Tomas; Height, Felicity; Franich, Rick

    2012-01-01

    Prediction of dose distributions in close proximity to interfaces is difficult. In the context of radiotherapy of lung tumors, this may affect the minimum dose received by lesions and is particularly important when prescribing dose to covering isodoses. The objective of this work is to quantify underdosage in key regions around a hypothetical target using Monte Carlo dose calculation methods, and to develop a factor for clinical estimation of such underdosage. A systematic set of calculations are undertaken using 2 Monte Carlo radiation transport codes (EGSnrc and GEANT4). Discrepancies in dose are determined for a number of parameters, including beam energy, tumor size, field size, and distance from chest wall. Calculations were performed for 1-mm 3 regions at proximal, distal, and lateral aspects of a spherical tumor, determined for a 6-MV and a 15-MV photon beam. The simulations indicate regions of tumor underdose at the tumor-lung interface. Results are presented as ratios of the dose at key peripheral regions to the dose at the center of the tumor, a point at which the treatment planning system (TPS) predicts the dose more reliably. Comparison with TPS data (pencil-beam convolution) indicates such underdosage would not have been predicted accurately in the clinic. We define a dose reduction factor (DRF) as the average of the dose in the periphery in the 6 cardinal directions divided by the central dose in the target, the mean of which is 0.97 and 0.95 for a 6-MV and 15-MV beam, respectively. The DRF can assist clinicians in the estimation of the magnitude of potential discrepancies between prescribed and delivered dose distributions as a function of tumor size and location. Calculation for a systematic set of “generic” tumors allows application to many classes of patient case, and is particularly useful for interpreting clinical trial data.

  3. Determination of peripheral underdosage at the lung-tumor interface using Monte Carlo radiation transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Michael, E-mail: michael.taylor@rmit.edu.au [School of Applied Sciences, College of Science, Engineering and Health, RMIT University, Melbourne, Victoria (Australia); Physical Sciences, Peter MacCallum Cancer Centre, East Melbourne, Victoria (Australia); Dunn, Leon; Kron, Tomas; Height, Felicity; Franich, Rick [School of Applied Sciences, College of Science, Engineering and Health, RMIT University, Melbourne, Victoria (Australia); Physical Sciences, Peter MacCallum Cancer Centre, East Melbourne, Victoria (Australia)

    2012-04-01

    Prediction of dose distributions in close proximity to interfaces is difficult. In the context of radiotherapy of lung tumors, this may affect the minimum dose received by lesions and is particularly important when prescribing dose to covering isodoses. The objective of this work is to quantify underdosage in key regions around a hypothetical target using Monte Carlo dose calculation methods, and to develop a factor for clinical estimation of such underdosage. A systematic set of calculations are undertaken using 2 Monte Carlo radiation transport codes (EGSnrc and GEANT4). Discrepancies in dose are determined for a number of parameters, including beam energy, tumor size, field size, and distance from chest wall. Calculations were performed for 1-mm{sup 3} regions at proximal, distal, and lateral aspects of a spherical tumor, determined for a 6-MV and a 15-MV photon beam. The simulations indicate regions of tumor underdose at the tumor-lung interface. Results are presented as ratios of the dose at key peripheral regions to the dose at the center of the tumor, a point at which the treatment planning system (TPS) predicts the dose more reliably. Comparison with TPS data (pencil-beam convolution) indicates such underdosage would not have been predicted accurately in the clinic. We define a dose reduction factor (DRF) as the average of the dose in the periphery in the 6 cardinal directions divided by the central dose in the target, the mean of which is 0.97 and 0.95 for a 6-MV and 15-MV beam, respectively. The DRF can assist clinicians in the estimation of the magnitude of potential discrepancies between prescribed and delivered dose distributions as a function of tumor size and location. Calculation for a systematic set of 'generic' tumors allows application to many classes of patient case, and is particularly useful for interpreting clinical trial data.

  4. Numerical investigations for insulation particle transport phenomena in water flow

    International Nuclear Information System (INIS)

    Krepper, E.; Grahn, A.; Alt, S.; Kaestner, W.; Kratzsch, A.; Seeliger, A.

    2005-01-01

    The investigation of insulation debris generation, transport and sedimentation gains importance regarding the reactor safety research for PWR and BWR considering the long term behaviour of emergency core coolant systems during all types of LOCA. The insulation debris released near the break during LOCA consists of a mixture of very different particles concerning size, shape, consistence and other properties. Some fraction of the released insulation debris will be transported into the reactor sump where it may affect emergency core cooling. Open questions of generic interest are e.g. the sedimentation of the insulation debris in a water pool, possible re-suspension, transport in the sump water flow, particle load on strainers and corresponding difference pressure. A joint research project in cooperation with Institute of Process Technology, Process Automation and Measuring Technology (IPM) Zittau deals with the experimental investigation and the development of CFD models for the description of particle transport phenomena in coolant flow. While experiments are performed at the IPM-Zittau, theoretical work is concentrated at Forschungszentrum Rossendorf. In the present paper the basic concepts for CFD modelling are described and first results including feasibility studies are shown. During the ongoing work further results are expected. (author)

  5. A study on the particle penetration in RMS Right Single Quotation Marks particle transport system

    International Nuclear Information System (INIS)

    Son, S. M.; Oh, S. H.; Choi, C. R.

    2014-01-01

    In nuclear facilities, a radiation monitoring system (RMS) monitors the exhaust gas containing the radioactive material. Samples of exhaust gas are collected in the downstream region of air cleaning units (ACUs) in order to examine radioactive materials. It is possible to predict an amount of radioactive material by analyzing the corrected samples. Representation of the collected samples should be assured in order to accurately sense and measure of radioactive materials. The radius of curvature is mainly 5 times of tube diameter. Sometimes, a booster fan is additionally added to enhance particle penetration rate... In this study, particle penetrations are calculated to evaluate particle penetration rate with various design parameters (tube lengths, tube declined angles, radius of curvatures, etc). The particle penetration rates have been calculated for several elements in the particle transport system. In general, the horizontal length of tube and the number of bending tube have a big impact on the penetration rate in the particle transport system. If the sampling location is far from the radiation monitoring system, additional installation of booster fans could be considered in case of large diameter tubes, but is not recommended in case of small diameter tube. In order to enhance particle penetration rate, the following works are recommended by priority. 1) to reduce the interval between sampling location and radiation monitoring system 2) to reduce the number of the bending tube

  6. Monte Carlo Methods in ICF

    Science.gov (United States)

    Zimmerman, George B.

    Monte Carlo methods appropriate to simulate the transport of x-rays, neutrons, ions and electrons in Inertial Confinement Fusion targets are described and analyzed. The Implicit Monte Carlo method of x-ray transport handles symmetry within indirect drive ICF hohlraums well, but can be improved 50X in efficiency by angular biasing the x-rays towards the fuel capsule. Accurate simulation of thermonuclear burn and burn diagnostics involves detailed particle source spectra, charged particle ranges, inflight reaction kinematics, corrections for bulk and thermal Doppler effects and variance reduction to obtain adequate statistics for rare events. It is found that the effects of angular Coulomb scattering must be included in models of charged particle transport through heterogeneous materials.

  7. Monte Carlo methods in ICF

    International Nuclear Information System (INIS)

    Zimmerman, G.B.

    1997-01-01

    Monte Carlo methods appropriate to simulate the transport of x-rays, neutrons, ions and electrons in Inertial Confinement Fusion targets are described and analyzed. The Implicit Monte Carlo method of x-ray transport handles symmetry within indirect drive ICF hohlraums well, but can be improved 50X in efficiency by angular biasing the x-rays towards the fuel capsule. Accurate simulation of thermonuclear burn and burn diagnostics involves detailed particle source spectra, charged particle ranges, inflight reaction kinematics, corrections for bulk and thermal Doppler effects and variance reduction to obtain adequate statistics for rare events. It is found that the effects of angular Coulomb scattering must be included in models of charged particle transport through heterogeneous materials. copyright 1997 American Institute of Physics

  8. Monte Carlo methods in ICF

    International Nuclear Information System (INIS)

    Zimmerman, George B.

    1997-01-01

    Monte Carlo methods appropriate to simulate the transport of x-rays, neutrons, ions and electrons in Inertial Confinement Fusion targets are described and analyzed. The Implicit Monte Carlo method of x-ray transport handles symmetry within indirect drive ICF hohlraums well, but can be improved 50X in efficiency by angular biasing the x-rays towards the fuel capsule. Accurate simulation of thermonuclear burn and burn diagnostics involves detailed particle source spectra, charged particle ranges, inflight reaction kinematics, corrections for bulk and thermal Doppler effects and variance reduction to obtain adequate statistics for rare events. It is found that the effects of angular Coulomb scattering must be included in models of charged particle transport through heterogeneous materials

  9. Monte Carlo simulations for plasma physics

    International Nuclear Information System (INIS)

    Okamoto, M.; Murakami, S.; Nakajima, N.; Wang, W.X.

    2000-07-01

    Plasma behaviours are very complicated and the analyses are generally difficult. However, when the collisional processes play an important role in the plasma behaviour, the Monte Carlo method is often employed as a useful tool. For examples, in neutral particle injection heating (NBI heating), electron or ion cyclotron heating, and alpha heating, Coulomb collisions slow down high energetic particles and pitch angle scatter them. These processes are often studied by the Monte Carlo technique and good agreements can be obtained with the experimental results. Recently, Monte Carlo Method has been developed to study fast particle transports associated with heating and generating the radial electric field. Further it is applied to investigating the neoclassical transport in the plasma with steep gradients of density and temperatures which is beyong the conventional neoclassical theory. In this report, we briefly summarize the researches done by the present authors utilizing the Monte Carlo method. (author)

  10. DRIFT-INDUCED PERPENDICULAR TRANSPORT OF SOLAR ENERGETIC PARTICLES

    International Nuclear Information System (INIS)

    Marsh, M. S.; Dalla, S.; Kelly, J.; Laitinen, T.

    2013-01-01

    Drifts are known to play a role in galactic cosmic ray transport within the heliosphere and are a standard component of cosmic ray propagation models. However, the current paradigm of solar energetic particle (SEP) propagation holds the effects of drifts to be negligible, and they are not accounted for in most current SEP modeling efforts. We present full-orbit test particle simulations of SEP propagation in a Parker spiral interplanetary magnetic field (IMF), which demonstrate that high-energy particle drifts cause significant asymmetric propagation perpendicular to the IMF. Thus in many cases the assumption of field-aligned propagation of SEPs may not be valid. We show that SEP drifts have dependencies on energy, heliographic latitude, and charge-to-mass ratio that are capable of transporting energetic particles perpendicular to the field over significant distances within interplanetary space, e.g., protons of initial energy 100 MeV propagate distances across the field on the order of 1 AU, over timescales typical of a gradual SEP event. Our results demonstrate the need for current models of SEP events to include the effects of particle drift. We show that the drift is considerably stronger for heavy ion SEPs due to their larger mass-to-charge ratio. This paradigm shift has important consequences for the modeling of SEP events and is crucial to the understanding and interpretation of in situ observations

  11. A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom

    International Nuclear Information System (INIS)

    Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H.

    2014-08-01

    As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)

  12. A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom

    Energy Technology Data Exchange (ETDEWEB)

    Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H., E-mail: mbellezzo@gmail.br [Instituto de Pesquisas Energeticas e Nucleares / CNEN, Av. Lineu Prestes 2242, Cidade Universitaria, 05508-000 Sao Paulo (Brazil)

    2014-08-15

    As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)

  13. MONTE CARLO NEUTRINO TRANSPORT THROUGH REMNANT DISKS FROM NEUTRON STAR MERGERS

    Energy Technology Data Exchange (ETDEWEB)

    Richers, Sherwood; Ott, Christian D. [TAPIR, Mailcode 350-17, Walter Burke Institute for Theoretical Physics, California Institute of Technology, Pasadena, CA 91125 (United States); Kasen, Daniel; Fernández, Rodrigo [Department of Astronomy and Theoretical Astrophysics Center, University of California, Berkeley, CA 94720 (United States); O’Connor, Evan [Department of Physics, Campus Code 8202, North Carolina State University, Raleigh, NC 27695 (United States)

    2015-11-01

    We present Sedonu, a new open source, steady-state, special relativistic Monte Carlo (MC) neutrino transport code, available at bitbucket.org/srichers/sedonu. The code calculates the energy- and angle-dependent neutrino distribution function on fluid backgrounds of any number of spatial dimensions, calculates the rates of change of fluid internal energy and electron fraction, and solves for the equilibrium fluid temperature and electron fraction. We apply this method to snapshots from two-dimensional simulations of accretion disks left behind by binary neutron star mergers, varying the input physics and comparing to the results obtained with a leakage scheme for the cases of a central black hole and a central hypermassive neutron star. Neutrinos are guided away from the densest regions of the disk and escape preferentially around 45° from the equatorial plane. Neutrino heating is strengthened by MC transport a few scale heights above the disk midplane near the innermost stable circular orbit, potentially leading to a stronger neutrino-driven wind. Neutrino cooling in the dense midplane of the disk is stronger when using MC transport, leading to a globally higher cooling rate by a factor of a few and a larger leptonization rate by an order of magnitude. We calculate neutrino pair annihilation rates and estimate that an energy of 2.8 × 10{sup 46} erg is deposited within 45° of the symmetry axis over 300 ms when a central BH is present. Similarly, 1.9 × 10{sup 48} erg is deposited over 3 s when an HMNS sits at the center, but neither estimate is likely to be sufficient to drive a gamma-ray burst jet.

  14. MONTE CARLO NEUTRINO TRANSPORT THROUGH REMNANT DISKS FROM NEUTRON STAR MERGERS

    International Nuclear Information System (INIS)

    Richers, Sherwood; Ott, Christian D.; Kasen, Daniel; Fernández, Rodrigo; O’Connor, Evan

    2015-01-01

    We present Sedonu, a new open source, steady-state, special relativistic Monte Carlo (MC) neutrino transport code, available at bitbucket.org/srichers/sedonu. The code calculates the energy- and angle-dependent neutrino distribution function on fluid backgrounds of any number of spatial dimensions, calculates the rates of change of fluid internal energy and electron fraction, and solves for the equilibrium fluid temperature and electron fraction. We apply this method to snapshots from two-dimensional simulations of accretion disks left behind by binary neutron star mergers, varying the input physics and comparing to the results obtained with a leakage scheme for the cases of a central black hole and a central hypermassive neutron star. Neutrinos are guided away from the densest regions of the disk and escape preferentially around 45° from the equatorial plane. Neutrino heating is strengthened by MC transport a few scale heights above the disk midplane near the innermost stable circular orbit, potentially leading to a stronger neutrino-driven wind. Neutrino cooling in the dense midplane of the disk is stronger when using MC transport, leading to a globally higher cooling rate by a factor of a few and a larger leptonization rate by an order of magnitude. We calculate neutrino pair annihilation rates and estimate that an energy of 2.8 × 10 46 erg is deposited within 45° of the symmetry axis over 300 ms when a central BH is present. Similarly, 1.9 × 10 48 erg is deposited over 3 s when an HMNS sits at the center, but neither estimate is likely to be sufficient to drive a gamma-ray burst jet

  15. Time-dependent Perpendicular Transport of Energetic Particles for Different Turbulence Configurations and Parallel Transport Models

    Energy Technology Data Exchange (ETDEWEB)

    Lasuik, J.; Shalchi, A., E-mail: andreasm4@yahoo.com [Department of Physics and Astronomy, University of Manitoba, Winnipeg, MB R3T 2N2 (Canada)

    2017-09-20

    Recently, a new theory for the transport of energetic particles across a mean magnetic field was presented. Compared to other nonlinear theories the new approach has the advantage that it provides a full time-dependent description of the transport. Furthermore, a diffusion approximation is no longer part of that theory. The purpose of this paper is to combine this new approach with a time-dependent model for parallel transport and different turbulence configurations in order to explore the parameter regimes for which we get ballistic transport, compound subdiffusion, and normal Markovian diffusion.

  16. Recent advances in neutral particle transport methods and codes

    International Nuclear Information System (INIS)

    Azmy, Y.Y.

    1996-01-01

    An overview of ORNL's three-dimensional neutral particle transport code, TORT, is presented. Special features of the code that make it invaluable for large applications are summarized for the prospective user. Advanced capabilities currently under development and installation in the production release of TORT are discussed; they include: multitasking on Cray platforms running the UNICOS operating system; Adjacent cell Preconditioning acceleration scheme; and graphics codes for displaying computed quantities such as the flux. Further developments for TORT and its companion codes to enhance its present capabilities, as well as expand its range of applications are disucssed. Speculation on the next generation of neutron particle transport codes at ORNL, especially regarding unstructured grids and high order spatial approximations, are also mentioned

  17. Dust particle diffusion in ion beam transport region

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, N.; Okajima, Y.; Romero, C. F.; Kuwata, Y.; Kasuya, T.; Wada, M., E-mail: mwada@mail.doshisha.ac.jp [Graduate school of Science and Engineering, Doshisha University, Kyotanabe, Kyoto 610-0321 (Japan)

    2016-02-15

    Dust particles of μm size produced by a monoplasmatron ion source are observed by a laser light scattering. The scattered light signal from an incident laser at 532 nm wavelength indicates when and where a particle passes through the ion beam transport region. As the result, dusts with the size more than 10 μm are found to be distributed in the center of the ion beam, while dusts with the size less than 10 μm size are distributed along the edge of the ion beam. Floating potential and electron temperature at beam transport region are measured by an electrostatic probe. This observation can be explained by a charge up model of the dust in the plasma boundary region.

  18. Particle transport in JET and TCV-H mode plasmas

    International Nuclear Information System (INIS)

    Maslov, M.

    2009-10-01

    Understanding particle transport physics is of great importance for magnetically confined plasma devices and for the development of thermonuclear fusion power for energy production. From the beginnings of fusion research, more than half a century ago, the problem of heat transport in tokamaks attracted the attention of researchers, but the particle transport phenomena were largely neglected until fairly recently. As tokamak physics advanced to its present level, the physics community realized that there are many hurdles to the development of fusion power beyond the energy confinement. Particle transport is one of the outstanding issues. The aim of this thesis work is to study the anomalous (turbulence driven) particle transport in tokamaks on the basis of experiments on two different devices: JET (Joint European Torus) and TCV (Tokamak à Configuration Variable). In particular the physics of particle inward convection (pinch), which causes formation of peaked density profiles, is addressed in this work. Density profile peaking has a direct, favorable effect on fusion power in a reactor, we therefore also propose an extrapolation to the international experimental reactor ITER, which is currently under construction. To complete the thesis research, a comprehensive experimental database was created on the basis of data collected on JET and TCV during the duration of the thesis. Improvements of the density profile measurements techniques and careful analysis of the experimental data allowed us to derive the dependencies of density profile shape on the relevant plasma parameters. These improved techniques also allowed us to dispel any doubts that had been voiced about previous results. The major conclusions from previous work on JET and other tokamaks were generally confirmed, with some minor supplements. The main novelty of the thesis resides in systematic tests of the predictions of linear gyrokinetic simulations of the ITG (Ion Temperature Gradient) mode against the

  19. Gyrokinetics Simulation of Energetic Particle Turbulence and Transport

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, Patrick H.

    2011-09-21

    Progress in research during this year elucidated the physics of precession resonance and its interaction with radial scattering to form phase space density granulations. Momentum theorems for drift wave-zonal flow systems involving precession resonance were derived. These are directly generalizable to energetic particle modes. A novel nonlinear, subcritical growth mechanism was identified, which has now been verified by simulation. These results strengthen the foundation of our understanding of transport in burning plasmas

  20. Fluid description of particle transport in hf heated magnetized plasma

    International Nuclear Information System (INIS)

    Klima, R.

    1980-01-01

    Particle fluxes averaged over high-frequency oscillations are analyzed. The collisional effects and the kinetic mechanisms of energy absorption are included. Spatial dependences of both the high-frequency and the (quasi-)steady electromagnetic fields are arbitrary. The equations governing the fluxes are deduced from the moments of the averaged kinetic equation. Explicit expressions for steady state fluxes are given in terms of electromagnetic field quantities. The results can also be applied to anomalous transport phenomena in weakly turbulent plasmas. (author)

  1. Gyrokinetics Simulation of Energetic Particle Turbulence and Transport

    International Nuclear Information System (INIS)

    Diamond, Patrick H.

    2011-01-01

    Progress in research during this year elucidated the physics of precession resonance and its interaction with radial scattering to form phase space density granulations. Momentum theorems for drift wave-zonal flow systems involving precession resonance were derived. These are directly generalizable to energetic particle modes. A novel nonlinear, subcritical growth mechanism was identified, which has now been verified by simulation. These results strengthen the foundation of our understanding of transport in burning plasmas

  2. Evaluation of five dry particle deposition parameterizations for incorporation into atmospheric transport models

    Science.gov (United States)

    Khan, Tanvir R.; Perlinger, Judith A.

    2017-10-01

    Despite considerable effort to develop mechanistic dry particle deposition parameterizations for atmospheric transport models, current knowledge has been inadequate to propose quantitative measures of the relative performance of available parameterizations. In this study, we evaluated the performance of five dry particle deposition parameterizations developed by Zhang et al. (2001) (Z01), Petroff and Zhang (2010) (PZ10), Kouznetsov and Sofiev (2012) (KS12), Zhang and He (2014) (ZH14), and Zhang and Shao (2014) (ZS14), respectively. The evaluation was performed in three dimensions: model ability to reproduce observed deposition velocities, Vd (accuracy); the influence of imprecision in input parameter values on the modeled Vd (uncertainty); and identification of the most influential parameter(s) (sensitivity). The accuracy of the modeled Vd was evaluated using observations obtained from five land use categories (LUCs): grass, coniferous and deciduous forests, natural water, and ice/snow. To ascertain the uncertainty in modeled Vd, and quantify the influence of imprecision in key model input parameters, a Monte Carlo uncertainty analysis was performed. The Sobol' sensitivity analysis was conducted with the objective to determine the parameter ranking from the most to the least influential. Comparing the normalized mean bias factors (indicators of accuracy), we find that the ZH14 parameterization is the most accurate for all LUCs except for coniferous forest, for which it is second most accurate. From Monte Carlo simulations, the estimated mean normalized uncertainties in the modeled Vd obtained for seven particle sizes (ranging from 0.005 to 2.5 µm) for the five LUCs are 17, 12, 13, 16, and 27 % for the Z01, PZ10, KS12, ZH14, and ZS14 parameterizations, respectively. From the Sobol' sensitivity results, we suggest that the parameter rankings vary by particle size and LUC for a given parameterization. Overall, for dp = 0.001 to 1.0 µm, friction velocity was one of

  3. Evaluation of five dry particle deposition parameterizations for incorporation into atmospheric transport models

    Directory of Open Access Journals (Sweden)

    T. R. Khan

    2017-10-01

    Full Text Available Despite considerable effort to develop mechanistic dry particle deposition parameterizations for atmospheric transport models, current knowledge has been inadequate to propose quantitative measures of the relative performance of available parameterizations. In this study, we evaluated the performance of five dry particle deposition parameterizations developed by Zhang et al. (2001 (Z01, Petroff and Zhang (2010 (PZ10, Kouznetsov and Sofiev (2012 (KS12, Zhang and He (2014 (ZH14, and Zhang and Shao (2014 (ZS14, respectively. The evaluation was performed in three dimensions: model ability to reproduce observed deposition velocities, Vd (accuracy; the influence of imprecision in input parameter values on the modeled Vd (uncertainty; and identification of the most influential parameter(s (sensitivity. The accuracy of the modeled Vd was evaluated using observations obtained from five land use categories (LUCs: grass, coniferous and deciduous forests, natural water, and ice/snow. To ascertain the uncertainty in modeled Vd, and quantify the influence of imprecision in key model input parameters, a Monte Carlo uncertainty analysis was performed. The Sobol' sensitivity analysis was conducted with the objective to determine the parameter ranking from the most to the least influential. Comparing the normalized mean bias factors (indicators of accuracy, we find that the ZH14 parameterization is the most accurate for all LUCs except for coniferous forest, for which it is second most accurate. From Monte Carlo simulations, the estimated mean normalized uncertainties in the modeled Vd obtained for seven particle sizes (ranging from 0.005 to 2.5 µm for the five LUCs are 17, 12, 13, 16, and 27 % for the Z01, PZ10, KS12, ZH14, and ZS14 parameterizations, respectively. From the Sobol' sensitivity results, we suggest that the parameter rankings vary by particle size and LUC for a given parameterization. Overall, for dp  =  0.001 to 1.0

  4. KIM, Steady-State Transport for Fixed Source in 2-D Thermal Reactor by Monte-Carlo

    International Nuclear Information System (INIS)

    Cupini, E.; De Matteis, A.; Simonini, R.

    1980-01-01

    1 - Description of problem or function: KIM (K-infinite Monte Carlo) is a program which solves the steady-state linear transport equation for a fixed-source problem (or, by successive fixed-source runs, for the eigenvalue problem) in a two-dimensional infinite thermal reactor lattice. The main quantities computed in some broad energy groups are the following: - Fluxes and cross sections averaged over the region (i.e. a space portion that can be unconnected but contains everywhere the same homogeneous material), grouping of regions, the whole element. - Average absorption and fission rates per nuclide. - Average flux, absorption and production distributions versus energy. 2 - Method of solution: Monte Carlo simulation is used by tracing particle histories from fission birth down through the resonance region until absorption in the thermal range. The program is organised in three sections for fast, epithermal and thermal simulation, respectively; each section implements a particular model for both numerical techniques and cross section representation (energy groups in the fast section, groups or resonance parameters in the epithermal section, points in the thermal section). During slowing down (energy above 1 eV) nuclei are considered as stationary, with the exception of some resonance nuclei whose spacing between resonances is much greater than the resonance width. The Doppler broadening of s-wave resonances of these nuclides is taken into account by computing cross sections at the current neutron energy and at the temperature of the nucleus hit. During thermalization (energy below 1 eV) the thermal motion of some nuclides is also considered, by exploiting scattering kernels provided by the library for light water, heavy water and oxygen at several temperatures. KIM includes splitting and Russian roulette. A characteristic feature of the program is its approach to the lattice geometry. In fact, besides the usual continuous treatment of the geometry using the well

  5. Procedure for obtaining neutron diffusion coefficients from neutron transport Monte Carlo calculations (AWBA Development Program)

    International Nuclear Information System (INIS)

    Gast, R.C.

    1981-08-01

    A procedure for defining diffusion coefficients from Monte Carlo calculations that results in suitable ones for use in neutron diffusion theory calculations is not readily obtained. This study provides a survey of the methods used to define diffusion coefficients from deterministic calculations and provides a discussion as to why such traditional methods cannot be used in Monte Carlo. This study further provides the empirical procedure used for defining diffusion coefficients from the RCP01 Monte Carlo program

  6. Transport of Particle Swarms Through Variable Aperture Fractures

    Science.gov (United States)

    Boomsma, E.; Pyrak-Nolte, L. J.

    2012-12-01

    Particle transport through fractured rock is a key concern with the increased use of micro- and nano-size particles in consumer products as well as from other activities in the sub- and near surface (e.g. mining, industrial waste, hydraulic fracturing, etc.). While particle transport is often studied as the transport of emulsions or dispersions, particles may also enter the subsurface from leaks or seepage that lead to particle swarms. Swarms are drop-like collections of millions of colloidal-sized particles that exhibit a number of unique characteristics when compared to dispersions and emulsions. Any contaminant or engineered particle that forms a swarm can be transported farther, faster, and more cohesively in fractures than would be expected from a traditional dispersion model. In this study, the effects of several variable aperture fractures on colloidal swarm cohesiveness and evolution were studied as a swarm fell under gravity and interacted with the fracture walls. Transparent acrylic was used to fabricate synthetic fracture samples with (1) a uniform aperture, (2) a converging region followed by a uniform region (funnel shaped), (3) a uniform region followed by a diverging region (inverted funnel), and (4) a cast of a an induced fracture from a carbonate rock. All of the samples consisted of two blocks that measured 100 x 100 x 50 mm. The minimum separation between these blocks determined the nominal aperture (0.5 mm to 20 mm). During experiments a fracture was fully submerged in water and swarms were released into it. The swarms consisted of a dilute suspension of 3 micron polystyrene fluorescent beads (1% by mass) with an initial volume of 5μL. The swarms were illuminated with a green (525 nm) LED array and imaged optically with a CCD camera. The variation in fracture aperture controlled swarm behavior. Diverging apertures caused a sudden loss of confinement that resulted in a rapid change in the swarm's shape as well as a sharp increase in its velocity

  7. Monte Carlo methods for neutron transport on graphics processing units using Cuda - 015

    International Nuclear Information System (INIS)

    Nelson, A.G.; Ivanov, K.N.

    2010-01-01

    This work examined the feasibility of utilizing Graphics Processing Units (GPUs) to accelerate Monte Carlo neutron transport simulations. First, a clean-sheet MC code was written in C++ for an x86 CPU and later ported to run on GPUs using NVIDIA's CUDA programming language. After further optimization, the GPU ran 21 times faster than the CPU code when using single-precision floating point math. This can be further increased with no additional effort if accuracy is sacrificed for speed: using a compiler flag, the speedup was increased to 22x. Further, if double-precision floating point math is desired for neutron tracking through the geometry, a speedup of 11x was obtained. The GPUs have proven to be useful in this study, but the current generation does have limitations: the maximum memory currently available on a single GPU is only 4 GB; the GPU RAM does not provide error-checking and correction; and the optimization required for large speedups can lead to confusing code. (authors)

  8. 3D Monte Carlo model of optical transport in laser-irradiated cutaneous vascular malformations

    Science.gov (United States)

    Majaron, Boris; Milanič, Matija; Jia, Wangcun; Nelson, J. S.

    2010-11-01

    We have developed a three-dimensional Monte Carlo (MC) model of optical transport in skin and applied it to analysis of port wine stain treatment with sequential laser irradiation and intermittent cryogen spray cooling. Our MC model extends the approaches of the popular multi-layer model by Wang et al.1 to three dimensions, thus allowing treatment of skin inclusions with more complex geometries and arbitrary irradiation patterns. To overcome the obvious drawbacks of either "escape" or "mirror" boundary conditions at the lateral boundaries of the finely discretized volume of interest (VOI), photons exiting the VOI are propagated in laterally infinite tissue layers with appropriate optical properties, until they loose all their energy, escape into the air, or return to the VOI, but the energy deposition outside of the VOI is not computed and recorded. After discussing the selection of tissue parameters, we apply the model to analysis of blood photocoagulation and collateral thermal damage in treatment of port wine stain (PWS) lesions with sequential laser irradiation and intermittent cryogen spray cooling.

  9. Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

    2015-07-01

    Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

  10. Particle Tracking Model and Abstraction of Transport Processes

    Energy Technology Data Exchange (ETDEWEB)

    B. Robinson

    2004-10-21

    The purpose of this report is to document the abstraction model being used in total system performance assessment (TSPA) model calculations for radionuclide transport in the unsaturated zone (UZ). The UZ transport abstraction model uses the particle-tracking method that is incorporated into the finite element heat and mass model (FEHM) computer code (Zyvoloski et al. 1997 [DIRS 100615]) to simulate radionuclide transport in the UZ. This report outlines the assumptions, design, and testing of a model for calculating radionuclide transport in the UZ at Yucca Mountain. In addition, methods for determining and inputting transport parameters are outlined for use in the TSPA for license application (LA) analyses. Process-level transport model calculations are documented in another report for the UZ (BSC 2004 [DIRS 164500]). Three-dimensional, dual-permeability flow fields generated to characterize UZ flow (documented by BSC 2004 [DIRS 169861]; DTN: LB03023DSSCP9I.001 [DIRS 163044]) are converted to make them compatible with the FEHM code for use in this abstraction model. This report establishes the numerical method and demonstrates the use of the model that is intended to represent UZ transport in the TSPA-LA. Capability of the UZ barrier for retarding the transport is demonstrated in this report, and by the underlying process model (BSC 2004 [DIRS 164500]). The technical scope, content, and management of this report are described in the planning document ''Technical Work Plan for: Unsaturated Zone Transport Model Report Integration'' (BSC 2004 [DIRS 171282]). Deviations from the technical work plan (TWP) are noted within the text of this report, as appropriate. The latest version of this document is being prepared principally to correct parameter values found to be in error due to transcription errors, changes in source data that were not captured in the report, calculation errors, and errors in interpretation of source data.

  11. Particle Tracking Model and Abstraction of Transport Processes

    International Nuclear Information System (INIS)

    Robinson, B.

    2004-01-01

    The purpose of this report is to document the abstraction model being used in total system performance assessment (TSPA) model calculations for radionuclide transport in the unsaturated zone (UZ). The UZ transport abstraction model uses the particle-tracking method that is incorporated into the finite element heat and mass model (FEHM) computer code (Zyvoloski et al. 1997 [DIRS 100615]) to simulate radionuclide transport in the UZ. This report outlines the assumptions, design, and testing of a model for calculating radionuclide transport in the UZ at Yucca Mountain. In addition, methods for determining and inputting transport parameters are outlined for use in the TSPA for license application (LA) analyses. Process-level transport model calculations are documented in another report for the UZ (BSC 2004 [DIRS 164500]). Three-dimensional, dual-permeability flow fields generated to characterize UZ flow (documented by BSC 2004 [DIRS 169861]; DTN: LB03023DSSCP9I.001 [DIRS 163044]) are converted to make them compatible with the FEHM code for use in this abstraction model. This report establishes the numerical method and demonstrates the use of the model that is intended to represent UZ transport in the TSPA-LA. Capability of the UZ barrier for retarding the transport is demonstrated in this report, and by the underlying process model (BSC 2004 [DIRS 164500]). The technical scope, content, and management of this report are described in the planning document ''Technical Work Plan for: Unsaturated Zone Transport Model Report Integration'' (BSC 2004 [DIRS 171282]). Deviations from the technical work plan (TWP) are noted within the text of this report, as appropriate. The latest version of this document is being prepared principally to correct parameter values found to be in error due to transcription errors, changes in source data that were not captured in the report, calculation errors, and errors in interpretation of source data

  12. Characterization of molecule and particle transport through nanoscale conduits

    Science.gov (United States)

    Alibakhshi, Mohammad Amin

    Nanofluidic devices have been of great interest due to their applications in variety of fields, including energy conversion and storage, water desalination, biological and chemical separations, and lab-on-a-chip devices. Although these applications cross the boundaries of many different disciplines, they all share the demand for understanding transport in nanoscale conduits. In this thesis, different elusive aspects of molecule and particle transport through nanofluidic conduits are investigated, including liquid and ion transport in nanochannels, diffusion- and reaction-governed enzyme transport in nanofluidic channels, and finally translocation of nanobeads through nanopores. Liquid or solvent transport through nanoconfinements is an essential yet barely characterized component of any nanofluidic systems. In the first chapter, water transport through single hydrophilic nanochannels with heights down to 7 nm is experimentally investigated using a new measurement technique. This technique has been developed based on the capillary flow and a novel hybrid nanochannel design and is capable of characterizing flow in both single nanoconduits as well as nanoporous media. The presence of a 0.7 nm thick hydration layer on hydrophilic surfaces and its effect on increasing the hydraulic resistance of the nanochannels is verified. Next, ion transport in a new class of nanofluidic rectifiers is theoretically and experimentally investigated. These so called nanofluidic diodes are nanochannels with asymmetric geometries which preferentially allow ion transport in one direction. A nondimensional number as a function of electrolyte concentration, nanochannel dimensions, and surface charge is derived that summarizes the rectification behavior of this system. In the fourth chapter, diffusion- and reaction-governed enzyme transport in nanofluidic channels is studied and the theoretical background necessary for understanding enzymatic activity in nanofluidic channels is presented. A

  13. Phase-coexistence simulations of fluid mixtures by the Markov Chain Monte Carlo method using single-particle models

    KAUST Repository

    Li, Jun

    2013-09-01

    We present a single-particle Lennard-Jones (L-J) model for CO2 and N2. Simplified L-J models for other small polyatomic molecules can be obtained following the methodology described herein. The phase-coexistence diagrams of single-component systems computed using the proposed single-particle models for CO2 and N2 agree well with experimental data over a wide range of temperatures. These diagrams are computed using the Markov Chain Monte Carlo method based on the Gibbs-NVT ensemble. This good agreement validates the proposed simplified models. That is, with properly selected parameters, the single-particle models have similar accuracy in predicting gas-phase properties as more complex, state-of-the-art molecular models. To further test these single-particle models, three binary mixtures of CH4, CO2 and N2 are studied using a Gibbs-NPT ensemble. These results are compared against experimental data over a wide range of pressures. The single-particle model has similar accuracy in the gas phase as traditional models although its deviation in the liquid phase is greater. Since the single-particle model reduces the particle number and avoids the time-consuming Ewald summation used to evaluate Coulomb interactions, the proposed model improves the computational efficiency significantly, particularly in the case of high liquid density where the acceptance rate of the particle-swap trial move increases. We compare, at constant temperature and pressure, the Gibbs-NPT and Gibbs-NVT ensembles to analyze their performance differences and results consistency. As theoretically predicted, the agreement between the simulations implies that Gibbs-NVT can be used to validate Gibbs-NPT predictions when experimental data is not available. © 2013 Elsevier Inc.

  14. PENGEOM-A general-purpose geometry package for Monte Carlo simulation of radiation transport in material systems defined by quadric surfaces

    Science.gov (United States)

    Almansa, Julio; Salvat-Pujol, Francesc; Díaz-Londoño, Gloria; Carnicer, Artur; Lallena, Antonio M.; Salvat, Francesc

    2016-02-01

    The Fortran subroutine package PENGEOM provides a complete set of tools to handle quadric geometries in Monte Carlo simulations of radiation transport. The material structure where radiation propagates is assumed to consist of homogeneous bodies limited by quadric surfaces. The PENGEOM subroutines (a subset of the PENELOPE code) track particles through the material structure, independently of the details of the physics models adopted to describe the interactions. Although these subroutines are designed for detailed simulations of photon and electron transport, where all individual interactions are simulated sequentially, they can also be used in mixed (class II) schemes for simulating the transport of high-energy charged particles, where the effect of soft interactions is described by the random-hinge method. The definition of the geometry and the details of the tracking algorithm are tailored to optimize simulation speed. The use of fuzzy quadric surfaces minimizes the impact of round-off errors. The provided software includes a Java graphical user interface for editing and debugging the geometry definition file and for visualizing the material structure. Images of the structure are generated by using the tracking subroutines and, hence, they describe the geometry actually passed to the simulation code.

  15. Measurement of particle transport coefficients on Alcator C-Mod

    International Nuclear Information System (INIS)

    Luke, T.C.T.

    1994-10-01

    The goal of this thesis was to study the behavior of the plasma transport during the divertor detachment in order to explain the central electron density rise. The measurement of particle transport coefficients requires sophisticated diagnostic tools. A two color interferometer system was developed and installed on Alcator C-Mod to measure the electron density with high spatial (∼ 2 cm) and high temporal (≤ 1.0 ms) resolution. The system consists of 10 CO 2 (10.6 μm) and 4 HeNe (.6328 μm) chords that are used to measure the line integrated density to within 0.08 CO 2 degrees or 2.3 x 10 16 m -2 theoretically. Using the two color interferometer, a series of gas puffing experiments were conducted. The density was varied above and below the threshold density for detachment at a constant magnetic field and plasma current. Using a gas modulation technique, the particle diffusion, D, and the convective velocity, V, were determined. Profiles were inverted using a SVD inversion and the transport coefficients were extracted with a time regression analysis and a transport simulation analysis. Results from each analysis were in good agreement. Measured profiles of the coefficients increased with the radius and the values were consistent with measurements from other experiments. The values exceeded neoclassical predictions by a factor of 10. The profiles also exhibited an inverse dependence with plasma density. The scaling of both attached and detached plasmas agreed well with this inverse scaling. This result and the lack of change in the energy and impurity transport indicate that there was no change in the underlying transport processes after detachment

  16. Measurement of particle transport coefficients on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Luke, T.C.T.

    1994-10-01

    The goal of this thesis was to study the behavior of the plasma transport during the divertor detachment in order to explain the central electron density rise. The measurement of particle transport coefficients requires sophisticated diagnostic tools. A two color interferometer system was developed and installed on Alcator C-Mod to measure the electron density with high spatial ({approx} 2 cm) and high temporal ({le} 1.0 ms) resolution. The system consists of 10 CO{sub 2} (10.6 {mu}m) and 4 HeNe (.6328 {mu}m) chords that are used to measure the line integrated density to within 0.08 CO{sub 2} degrees or 2.3 {times} 10{sup 16}m{sup {minus}2} theoretically. Using the two color interferometer, a series of gas puffing experiments were conducted. The density was varied above and below the threshold density for detachment at a constant magnetic field and plasma current. Using a gas modulation technique, the particle diffusion, D, and the convective velocity, V, were determined. Profiles were inverted using a SVD inversion and the transport coefficients were extracted with a time regression analysis and a transport simulation analysis. Results from each analysis were in good agreement. Measured profiles of the coefficients increased with the radius and the values were consistent with measurements from other experiments. The values exceeded neoclassical predictions by a factor of 10. The profiles also exhibited an inverse dependence with plasma density. The scaling of both attached and detached plasmas agreed well with this inverse scaling. This result and the lack of change in the energy and impurity transport indicate that there was no change in the underlying transport processes after detachment.

  17. Particle transport in 3He-rich events: wave-particle interactions and particle anisotropy measurements

    Directory of Open Access Journals (Sweden)

    B. T. Tsurutani

    2002-04-01

    Full Text Available Energetic particles and MHD waves are studied using simultaneous ISEE-3 data to investigate particle propagation and scattering between the source near the Sun and 1 AU. 3 He-rich events are of particular interest because they are typically low intensity "scatter-free" events. The largest solar proton events are of interest because they have been postulated to generate their own waves through beam instabilities. For 3 He-rich events, simultaneous interplanetary magnetic spectra are measured. The intensity of the interplanetary "fossil" turbulence through which the particles have traversed is found to be at the "quiet" to "intermediate" level of IMF activity. Pitch angle scattering rates and the corresponding particle mean free paths lW - P are calculated using the measured wave intensities, polarizations, and k directions. The values of lW - P are found to be ~ 5 times less than the value of lHe , the latter derived from He intensity and anisotropy time profiles. It is demonstrated by computer simulation that scattering rates through a 90° pitch angle are lower than that of other pitch angles, and that this is a possible explanation for the discrepancy between the lW - P and lHe values. At this time the scattering mechanism(s is unknown. We suggest a means where a direct comparison between the two l values could be made. Computer simulations indicate that although scattering through 90° is lower, it still occurs. Possibilities are either large pitch angle scattering through resonant interactions, or particle mirroring off of field compression regions. The largest solar proton events are analyzed to investigate the possibilities of local wave generation at 1 AU. In accordance with the results of a previous calculation (Gary et al., 1985 of beam stability, proton beams at 1 AU are found to be marginally stable. No evidence for substantial wave amplitude was found. Locally generated waves, if present, were less than 10-3 nT 2 Hz-1 at the leading

  18. Particle transport in 3He-rich events: wave-particle interactions and particle anisotropy measurements

    Directory of Open Access Journals (Sweden)

    T. Hada

    Full Text Available Energetic particles and MHD waves are studied using simultaneous ISEE-3 data to investigate particle propagation and scattering between the source near the Sun and 1 AU. 3 He-rich events are of particular interest because they are typically low intensity "scatter-free" events. The largest solar proton events are of interest because they have been postulated to generate their own waves through beam instabilities. For 3 He-rich events, simultaneous interplanetary magnetic spectra are measured. The intensity of the interplanetary "fossil" turbulence through which the particles have traversed is found to be at the "quiet" to "intermediate" level of IMF activity. Pitch angle scattering rates and the corresponding particle mean free paths lW - P are calculated using the measured wave intensities, polarizations, and k directions. The values of lW - P are found to be ~ 5 times less than the value of lHe , the latter derived from He intensity and anisotropy time profiles. It is demonstrated by computer simulation that scattering rates through a 90° pitch angle are lower than that of other pitch angles, and that this is a possible explanation for the discrepancy between the lW - P and lHe values. At this time the scattering mechanism(s is unknown. We suggest a means where a direct comparison between the two l values could be made. Computer simulations indicate that although scattering through 90° is lower, it still occurs. Possibilities are either large pitch angle scattering through resonant interactions, or particle mirroring off of field compression regions. The largest solar proton events are analyzed to investigate the possibilities of local wave generation at 1 AU. In accordance with the results of a previous calculation (Gary et al., 1985 of beam stability, proton beams at 1 AU are found to be marginally stable. No evidence for substantial wave amplitude was found. Locally generated waves, if present, were less than 10-3 nT 2 Hz-1 at the leading

  19. A Proposal of New Spherical Particle Modeling Method Based on Stochastic Sampling of Particle Locations in Monte Carlo Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Song Hyun; Kim, Do Hyun; Kim, Jong Kyung [Hanyang Univ., Seoul (Korea, Republic of); Noh, Jea Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    To the high computational efficiency and user convenience, the implicit method had received attention; however, it is noted that the implicit method in the previous studies has low accuracy at high packing fraction. In this study, a new implicit method, which can be used at any packing fraction with high accuracy, is proposed. In this study, the implicit modeling method in the spherical particle distributed medium for using the MC simulation is proposed. A new concept in the spherical particle sampling was developed to solve the problems in the previous implicit methods. The sampling method was verified by simulating the sampling method in the infinite and finite medium. The results show that the particle implicit modeling with the proposed method was accurately performed in all packing fraction boundaries. It is expected that the proposed method can be efficiently utilized for the spherical particle distributed mediums, which are the fusion reactor blanket, VHTR reactors, and shielding analysis.

  20. Extracting the Single-Particle Gap in Carbon Nanotubes with Lattice Quantum Monte Carlo

    Directory of Open Access Journals (Sweden)

    Berkowitz Evan

    2018-01-01

    Full Text Available We show how lattice Quantum Monte Carlo simulations can be used to calculate electronic properties of carbon nanotubes in the presence of strong electron-electron correlations. We employ the path integral formalism and use methods developed within the lattice QCD community for our numerical work and compare our results to empirical data of the Anti-Ferromagnetic Mott Insulating gap in large diameter tubes.