Condensation and homogenization of cross sections for the deterministic transport codes with Monte Carlo method: Application to the GEN IV fast neutron reactors

International Nuclear Information System (INIS)

Cai, Li

2014-01-01

calculation solver SNATCH in the PARIS code platform. The latter uses the transport theory which is indispensable for the new generation fast reactors analysis. The principal conclusions are as follows: The Monte-Carlo assembly calculation code is an interesting way (in the sense of avoiding the difficulties in the self-shielding calculation, the limited order development of anisotropy parameters, the exact 3D geometries) to validate the deterministic codes like ECCO or APOLLO3 and to produce the multi-group constants for deterministic or Monte-Carlo multi-group calculation codes. The results obtained for the moment with the multi-group constants calculated by TRIPOLI-4 code are comparable with those produced from ECCO, but did not show remarkable advantages. (author) [fr

Adaptive tree multigrids and simplified spherical harmonics approximation in deterministic neutral and charged particle transport

International Nuclear Information System (INIS)

Kotiluoto, P.

2007-05-01

A new deterministic three-dimensional neutral and charged particle transport code, MultiTrans, has been developed. In the novel approach, the adaptive tree multigrid technique is used in conjunction with simplified spherical harmonics approximation of the Boltzmann transport equation. The development of the new radiation transport code started in the framework of the Finnish boron neutron capture therapy (BNCT) project. Since the application of the MultiTrans code to BNCT dose planning problems, the testing and development of the MultiTrans code has continued in conventional radiotherapy and reactor physics applications. In this thesis, an overview of different numerical radiation transport methods is first given. Special features of the simplified spherical harmonics method and the adaptive tree multigrid technique are then reviewed. The usefulness of the new MultiTrans code has been indicated by verifying and validating the code performance for different types of neutral and charged particle transport problems, reported in separate publications. (orig.)

Review of the Monte Carlo and deterministic codes in radiation protection and dosimetry

Energy Technology Data Exchange (ETDEWEB)

Tagziria, H

2000-02-01

Modelling a physical system can be carried out either stochastically or deterministically. An example of the former method is the Monte Carlo technique, in which statistically approximate methods are applied to exact models. No transport equation is solved as individual particles are simulated and some specific aspect (tally) of their average behaviour is recorded. The average behaviour of the physical system is then inferred using the central limit theorem. In contrast, deterministic codes use mathematically exact methods that are applied to approximate models to solve the transport equation for the average particle behaviour. The physical system is subdivided in boxes in the phase-space system and particles are followed from one box to the next. The smaller the boxes the better the approximations become. Although the Monte Carlo method has been used for centuries, its more recent manifestation has really emerged from the Manhattan project of the Word War II. Its invention is thought to be mainly due to Metropolis, Ulah (through his interest in poker), Fermi, von Neuman andRichtmeyer. Over the last 20 years or so, the Monte Carlo technique has become a powerful tool in radiation transport. This is due to users taking full advantage of richer cross section data, more powerful computers and Monte Carlo techniques for radiation transport, with high quality physics and better known source spectra. This method is a common sense approach to radiation transport and its success and popularity is quite often also due to necessity, because measurements are not always possible or affordable. In the Monte Carlo method, which is inherently realistic because nature is statistical, a more detailed physics is made possible by isolation of events while rather elaborate geometries can be modelled. Provided that the physics is correct, a simulation is exactly analogous to an experimenter counting particles. In contrast to the deterministic approach, however, a disadvantage of the

NASA space radiation transport code development consortium

International Nuclear Information System (INIS)

Townsend, L. W.

2005-01-01

Recently, NASA established a consortium involving the Univ. of Tennessee (lead institution), the Univ. of Houston, Roanoke College and various government and national laboratories, to accelerate the development of a standard set of radiation transport computer codes for NASA human exploration applications. This effort involves further improvements of the Monte Carlo codes HETC and FLUKA and the deterministic code HZETRN, including developing nuclear reaction databases necessary to extend the Monte Carlo codes to carry out heavy ion transport, and extending HZETRN to three dimensions. The improved codes will be validated by comparing predictions with measured laboratory transport data, provided by an experimental measurements consortium, and measurements in the upper atmosphere on the balloon-borne Deep Space Test Bed (DSTB). In this paper, we present an overview of the consortium members and the current status and future plans of consortium efforts to meet the research goals and objectives of this extensive undertaking. (authors)

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.

On the progress towards probabilistic basis for deterministic codes

International Nuclear Information System (INIS)

Ellyin, F.

1975-01-01

Fundamentals arguments for a probabilistic basis of codes are presented. A class of code formats is outlined in which explicit statistical measures of uncertainty of design variables are incorporated. The format looks very much like present codes (deterministic) except for having probabilistic background. An example is provided whereby the design factors are plotted against the safety index, the probability of failure, and the risk of mortality. The safety level of the present codes is also indicated. A decision regarding the new probabilistically based code parameters thus could be made with full knowledge of implied consequences

Applications of the 3-D Deterministic Transport Attila(regsign) for Core Safety Analysis

International Nuclear Information System (INIS)

Lucas, D.S.; Gougar, D.; Roth, P.A.; Wareing, T.; Failla, G.; McGhee, J.; Barnett, A.

2004-01-01

An LDRD (Laboratory Directed Research and Development) project is ongoing at the Idaho National Engineering and Environmental Laboratory (INEEL) for applying the three-dimensional multi-group deterministic neutron transport code (Attila(reg s ign)) to criticality, flux and depletion calculations of the Advanced Test Reactor (ATR). This paper discusses the model development, capabilities of Attila, generation of the cross-section libraries, and comparisons to an ATR MCNP model and future

Minaret, a deterministic neutron transport solver for nuclear core calculations

Energy Technology Data Exchange (ETDEWEB)

Moller, J-Y.; Lautard, J-J., E-mail: jean-yves.moller@cea.fr, E-mail: jean-jacques.lautard@cea.fr [CEA - Centre de Saclay , Gif sur Yvette (France)

2011-07-01

We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

Minaret, a deterministic neutron transport solver for nuclear core calculations

International Nuclear Information System (INIS)

Moller, J-Y.; Lautard, J-J.

2011-01-01

We present here MINARET a deterministic transport solver for nuclear core calculations to solve the steady state Boltzmann equation. The code follows the multi-group formalism to discretize the energy variable. It uses discrete ordinate method to deal with the angular variable and a DGFEM to solve spatially the Boltzmann equation. The mesh is unstructured in 2D and semi-unstructured in 3D (cylindrical). Curved triangles can be used to fit the exact geometry. For the curved elements, two different sets of basis functions can be used. Transport solver is accelerated with a DSA method. Diffusion and SPN calculations are made possible by skipping the transport sweep in the source iteration. The transport calculations are parallelized with respect to the angular directions. Numerical results are presented for simple geometries and for the C5G7 Benchmark, JHR reactor and the ESFR (in 2D and 3D). Straight and curved finite element results are compared. (author)

Deterministic and stochastic transport theories for the analysis of complex nuclear systems

International Nuclear Information System (INIS)

Giffard, F.X.

2000-01-01

In the field of reactor and fuel cycle physics, particle transport plays an important role. Neutronic design, operation and evaluation calculations of nuclear systems make use of large and powerful computer codes. However, current limitations in terms of computer resources make it necessary to introduce simplifications and approximations in order to keep calculation time and cost within reasonable limits. Two different types of methods are available in these codes. The first one is the deterministic method, which is applicable in most practical cases but requires approximations. The other method is the Monte Carlo method, which does not make these approximations but which generally requires exceedingly long running times. The main motivation of this work is to investigate the possibility of a combined use of the two methods in such a way as to retain their advantages while avoiding their drawbacks. Our work has mainly focused on the speed-up of 3-D continuous energy Monte Carlo calculations (TRIPOLI-4 code) by means of an optimized biasing scheme derived from importance maps obtained from the deterministic code ERANOS. The application of this method to two different practical shielding-type problems has demonstrated its efficiency: speed-up factors of 100 have been reached. In addition, the method offers the advantage of being easily implemented as it is not very sensitive to the choice of the importance mesh grid. It has also been demonstrated that significant speed-ups can be achieved by this method in the case of coupled neutron-gamma transport problems, provided that the interdependence of the neutron and photon importance maps is taken into account. Complementary studies are necessary to tackle a problem brought out by this work, namely undesirable jumps in the Monte Carlo variance estimates. (author)

On the implementation of a deterministic secure coding protocol using polarization entangled photons

OpenAIRE

Ostermeyer, Martin; Walenta, Nino

2007-01-01

We demonstrate a prototype-implementation of deterministic information encoding for quantum key distribution (QKD) following the ping-pong coding protocol [K. Bostroem, T. Felbinger, Phys. Rev. Lett. 89 (2002) 187902-1]. Due to the deterministic nature of this protocol the need for post-processing the key is distinctly reduced compared to non-deterministic protocols. In the course of our implementation we analyze the practicability of the protocol and discuss some security aspects of informat...

Development and validation of a criticality calculation scheme based on French deterministic transport codes

International Nuclear Information System (INIS)

Santamarina, A.

1991-01-01

A criticality-safety calculational scheme using the automated deterministic code system, APOLLO-BISTRO, has been developed. The cell/assembly code APOLLO is used mainly in LWR and HCR design calculations, and its validation spans a wide range of moderation ratios, including voided configurations. Its recent 99-group library and self-shielded cross-sections has been extensively qualified through critical experiments and PWR spent fuel analysis. The PIC self-shielding formalism enables a rigorous treatment of the fuel double heterogeneity in dissolver medium calculations. BISTRO is an optimized multidimensional SN code, part of the modular CCRR package used mainly in FBR calculations. The APOLLO-BISTRO scheme was applied to the 18 experimental benchmarks selected by the OECD/NEACRP Criticality Calculation Working Group. The Calculation-Experiment discrepancy was within ± 1% in ΔK/K and always looked consistent with the experimental uncertainty margin. In the critical experiments corresponding to a dissolver type benchmark, our tools computed a satisfactory Keff. In the VALDUC fuel storage experiments, with hafnium plates, the computed Keff ranged between 0.994 and 1.003 for the various watergaps spacing the fuel clusters from the absorber plates. The APOLLO-KENOEUR statistic calculational scheme, based on the same self-shielded multigroup library, supplied consistent results within 0.3% in ΔK/K. (Author)

3D-TRANS-2003, Workshop on Common Tools and Interfaces for Radiation Transport Codes

International Nuclear Information System (INIS)

2004-01-01

Description: Contents proceedings of Workshop on Common Tools and Interfaces for Deterministic Radiation Transport, for Monte Carlo and Hybrid Codes with a proposal to develop the following: GERALD - A General Environment for Radiation Analysis and Design. GERALD intends to create a unifying software environment where the user can define, solve and analyse a nuclear radiation transport problem using available numerical tools seamlessly. This environment will serve many purposes: teaching, research, industrial needs. It will also help to preserve the existing analytical and numerical knowledge base. This could represent a significant step towards solving the legacy problem. This activity should contribute to attracting young engineers to nuclear science and engineering and contribute to competence and knowledge preservation and management. This proposal was made at the on Workshop on C ommon Tools and Interfaces for Deterministic Radiation Transport, for Monte Carlo and Hybrid Codes , held from 25-26 September 2003 in connection with the conference SNA-2003. A first success with the development of such tools was achieved with the BOT3P2.0 and 3.0 codes providing an easy procedure and mechanism for defining and displaying 3D geometries and materials both in the form of refineable meshes for deterministic codes or Monte Carlo geometries consistent with deterministic models. Advanced SUSD: Improved tools for Sensitivity/Uncertainty Analysis. The development of tools for the analysis and estimation of sensitivities and uncertainties in calculations, or their propagation through complex computational schemes, in the field of neutronics, thermal hydraulics and also thermo-mechanics is of increasing importance for research and engineering applications. These tools allow establishing better margins for engineering designs and for the safe operation of nuclear facilities. Such tools are not sufficiently developed, but their need is increasingly evident in many activities

Coupling Legacy and Contemporary Deterministic Codes to Goldsim for Probabilistic Assessments of Potential Low-Level Waste Repository Sites

Science.gov (United States)

Mattie, P. D.; Knowlton, R. G.; Arnold, B. W.; Tien, N.; Kuo, M.

2006-12-01

Sandia National Laboratories (Sandia), a U.S. Department of Energy National Laboratory, has over 30 years experience in radioactive waste disposal and is providing assistance internationally in a number of areas relevant to the safety assessment of radioactive waste disposal systems. International technology transfer efforts are often hampered by small budgets, time schedule constraints, and a lack of experienced personnel in countries with small radioactive waste disposal programs. In an effort to surmount these difficulties, Sandia has developed a system that utilizes a combination of commercially available codes and existing legacy codes for probabilistic safety assessment modeling that facilitates the technology transfer and maximizes limited available funding. Numerous codes developed and endorsed by the United States Nuclear Regulatory Commission and codes developed and maintained by United States Department of Energy are generally available to foreign countries after addressing import/export control and copyright requirements. From a programmatic view, it is easier to utilize existing codes than to develop new codes. From an economic perspective, it is not possible for most countries with small radioactive waste disposal programs to maintain complex software, which meets the rigors of both domestic regulatory requirements and international peer review. Therefore, re-vitalization of deterministic legacy codes, as well as an adaptation of contemporary deterministic codes, provides a creditable and solid computational platform for constructing probabilistic safety assessment models. External model linkage capabilities in Goldsim and the techniques applied to facilitate this process will be presented using example applications, including Breach, Leach, and Transport-Multiple Species (BLT-MS), a U.S. NRC sponsored code simulating release and transport of contaminants from a subsurface low-level waste disposal facility used in a cooperative technology transfer

BOT3P: a mesh generation software package for the transport analysis codes Dort, Tort, Twodant, Threedant and MCNP

International Nuclear Information System (INIS)

Orsi, R.

2003-01-01

Bot3p consists of a set of standard Fortran 77 language programs that gives the users of the deterministic transport codes Dort and Tort some useful diagnostic tools to prepare and check the geometry of their input data files for both Cartesian and cylindrical geometries including graphical display modules. Bot3p produces at the same time the geometrical and material distribution data for the deterministic transport codes Twodant and Threedant and, only in three-dimensional (3D) Cartesian geometry, for the Monte Carlo Transport Code MCNP. This makes it possible to compare directly for the same geometry the effects stemming from the use of different data libraries and solution approaches on transport analysis results. Through the use of Bot3p, radiation transport problems with complex 3D geometrical structures can be modelled easily, as a relatively small amount of engineer-time is required and refinement is achieved by changing few parameters. This tool is useful for solving very large challenging problems. (author)

A new modelling of the multigroup scattering cross section in deterministic codes for neutron transport

International Nuclear Information System (INIS)

Calloo, A.A.

2012-01-01

In reactor physics, calculation schemes with deterministic codes are validated with respect to a reference Monte Carlo code. The remaining biases are attributed to the approximations and models induced by the multigroup theory (self-shielding models and expansion of the scattering law using Legendre polynomials) to represent physical phenomena (resonant absorption and scattering anisotropy respectively). This work focuses on the relevance of a polynomial expansion to model the scattering law. Since the outset of reactor physics, the latter has been expanded on a truncated Legendre polynomial basis. However, the transfer cross sections are highly anisotropic, with non-zero values for a very small range of the cosine of the scattering angle. Besides, the finer the energy mesh and the lighter the scattering nucleus, the more exacerbated is the peaked shape of this cross section. As such, the Legendre expansion is less suited to represent the scattering law. Furthermore, this model induces negative values which are non-physical. In this work, various scattering laws are briefly described and the limitations of the existing model are pointed out. Hence, piecewise-constant functions have been used to represent the multigroup scattering cross section. This representation requires a different model for the diffusion source. The discrete ordinates method which is widely employed to solve the transport equation has been adapted. Thus, the finite volume method for angular discretization has been developed and implemented in Paris environment which hosts the S n solver, Snatch. The angular finite volume method has been compared to the collocation method with Legendre moments to ensure its proper performance. Moreover, unlike the latter, this method is adapted for both the Legendre moments and the piecewise-constant functions representations of the scattering cross section. This hybrid-source method has been validated for different cases: fuel cell in infinite lattice

Deterministic dense coding and faithful teleportation with multipartite graph states

International Nuclear Information System (INIS)

Huang, C.-Y.; Yu, I-C.; Lin, F.-L.; Hsu, L.-Y.

2009-01-01

We propose schemes to perform the deterministic dense coding and faithful teleportation with multipartite graph states. We also find the sufficient and necessary condition of a viable graph state for the proposed schemes. That is, for the associated graph, the reduced adjacency matrix of the Tanner-type subgraph between senders and receivers should be invertible.

ZZ BOREHOLE-EB6.8-MG, multi group cross-section library for deterministic and Monte Carlo codes

International Nuclear Information System (INIS)

Kodeli, Ivo; Aldama, Daniel L.; Leege, Piet F.A. de; Legrady, David; Hoogenboom, J. Eduard

2007-01-01

1 - Description: Format: MATXS and ACE; Number of groups: 175 neutron, 45 gamma-ray; Nuclides: H-1, C-12, O-16, Na-23, Mg-nat, Al-27, Si-28, -29, -30, S-nat, Cl-35, -37, K-nat, Ca-nat, Mn-55, Fe-54, -56, -57, -58, I-127, W-nat. Origin: ENDF/B-VI.8; Weighting spectrum: Fission and fusion peak at high energies and a 1/E + thermal Maxwellian extension at low energies. The following materials/nuclides are included in the library: H-1, C-12, O-16, Na-23, Mg-nat, Al-27, Si-28, -29, -30, S-nat, Cl-35, -37, K-nat, Ca-nat, Fe-54, -56, -57, -58, Mn-55, I-127, W-nat. ZZ-BOREHOLE-EB6.8-MG is a multigroup cross section library for deterministic (DOORS, DANTSYS) and Monte Carlo (MCNP) transport codes developed for the oil well logging applications. The library is based on the ENDF/B-VI.8 evaluation and was processed by the NJOY-99 code. The cross sections are given in the 175 neutron and 45 gamma ray group structure. The MATXS format library can be directly used in TRANSX code to prepare the multigroup self-shielded cross sections for deterministic discrete ordinates codes like DOORS and DANTSYS. The data provided in the GROUPR and GAMINR format were converted to the MCNP ACE format by the NSLINK, SCALE and CRSRD codes. IAEA1398/03: Multigroup cross section data for Mn-55 were added in TRANSX format

Probabilistic evaluation of fuel element performance by the combined use of a fast running simplistic and a detailed deterministic fuel performance code

International Nuclear Information System (INIS)

Misfeldt, I.

1980-01-01

A comprehensive evaluation of fuel element performance requires a probabilistic fuel code supported by a well bench-marked deterministic code. This paper presents an analysis of a SGHWR ramp experiment, where the probabilistic fuel code FRP is utilized in combination with the deterministic fuel models FFRS and SLEUTH/SEER. The statistical methods employed in FRP are Monte Carlo simulation or a low-order Taylor approximation. The fast-running simplistic fuel code FFRS is used for the deterministic simulations, whereas simulations with SLEUTH/SEER are used to verify the predictions of FFRS. The ramp test was performed with a SGHWR fuel element, where 9 of the 36 fuel pins failed. There seemed to be good agreement between the deterministic simulations and the experiment, but the statistical evaluation shows that the uncertainty on the important performance parameters is too large for this ''nice'' result. The analysis does therefore indicate a discrepancy between the experiment and the deterministic code predictions. Possible explanations for this disagreement are discussed. (author)

OCA-P, a deterministic and probabilistic fracture-mechanics code for application to pressure vessels

International Nuclear Information System (INIS)

Cheverton, R.D.; Ball, D.G.

1984-05-01

The OCA-P code is a probabilistic fracture-mechanics code that was prepared specifically for evaluating the integrity of pressurized-water reactor vessels when subjected to overcooling-accident loading conditions. The code has two-dimensional- and some three-dimensional-flaw capability; it is based on linear-elastic fracture mechanics; and it can treat cladding as a discrete region. Both deterministic and probabilistic analyses can be performed. For the former analysis, it is possible to conduct a search for critical values of the fluence and the nil-ductility reference temperature corresponding to incipient initiation of the initial flaw. The probabilistic portion of OCA-P is based on Monte Carlo techniques, and simulated parameters include fluence, flaw depth, fracture toughness, nil-ductility reference temperature, and concentrations of copper, nickel, and phosphorous. Plotting capabilities include the construction of critical-crack-depth diagrams (deterministic analysis) and various histograms (probabilistic analysis)

Motivation for Using Generalized Geometry in the Time Dependent Transport Code TDKENO

Energy Technology Data Exchange (ETDEWEB)

Dustin Popp; Zander Mausolff; Sedat Goluoglu

2016-04-01

We are proposing to use the code, TDKENO, to model TREAT. TDKENO solves the time dependent, three dimensional Boltzmann transport equation with explicit representation of delayed neutrons. Instead of directly integrating this equation, the neutron flux is factored into two components – a rapidly varying amplitude equation and a slowly varying shape equation and each is solved separately on different time scales. The shape equation is solved using the 3D Monte Carlo transport code KENO, from Oak Ridge National Laboratory’s SCALE code package. Using the Monte Carlo method to solve the shape equation is still computationally intensive, but the operation is only performed when needed. The amplitude equation is solved deterministically and frequently, so the solution gives an accurate time-dependent solution without having to repeatedly We have modified TDKENO to incorporate KENO-VI so that we may accurately represent the geometries within TREAT. This paper explains the motivation behind using generalized geometry, and provides the results of our modifications. TDKENO uses the Improved Quasi-Static method to accomplish this. In this method, the neutron flux is factored into two components. One component is a purely time-dependent and rapidly varying amplitude function, which is solved deterministically and very frequently (small time steps). The other is a slowly varying flux shape function that weakly depends on time and is only solved when needed (significantly larger time steps).

Deterministic sensitivity analysis for the numerical simulation of contaminants transport

International Nuclear Information System (INIS)

Marchand, E.

2007-12-01

The questions of safety and uncertainty are central to feasibility studies for an underground nuclear waste storage site, in particular the evaluation of uncertainties about safety indicators which are due to uncertainties concerning properties of the subsoil or of the contaminants. The global approach through probabilistic Monte Carlo methods gives good results, but it requires a large number of simulations. The deterministic method investigated here is complementary. Based on the Singular Value Decomposition of the derivative of the model, it gives only local information, but it is much less demanding in computing time. The flow model follows Darcy's law and the transport of radionuclides around the storage site follows a linear convection-diffusion equation. Manual and automatic differentiation are compared for these models using direct and adjoint modes. A comparative study of both probabilistic and deterministic approaches for the sensitivity analysis of fluxes of contaminants through outlet channels with respect to variations of input parameters is carried out with realistic data provided by ANDRA. Generic tools for sensitivity analysis and code coupling are developed in the Caml language. The user of these generic platforms has only to provide the specific part of the application in any language of his choice. We also present a study about two-phase air/water partially saturated flows in hydrogeology concerning the limitations of the Richards approximation and of the global pressure formulation used in petroleum engineering. (author)

Computation of a Canadian SCWR unit cell with deterministic and Monte Carlo codes

International Nuclear Information System (INIS)

Harrisson, G.; Marleau, G.

2012-01-01

The Canadian SCWR has the potential to achieve the goals that the generation IV nuclear reactors must meet. As part of the optimization process for this design concept, lattice cell calculations are routinely performed using deterministic codes. In this study, the first step (self-shielding treatment) of the computation scheme developed with the deterministic code DRAGON for the Canadian SCWR has been validated. Some options available in the module responsible for the resonance self-shielding calculation in DRAGON 3.06 and different microscopic cross section libraries based on the ENDF/B-VII.0 evaluated nuclear data file have been tested and compared to a reference calculation performed with the Monte Carlo code SERPENT under the same conditions. Compared to SERPENT, DRAGON underestimates the infinite multiplication factor in all cases. In general, the original Stammler model with the Livolant-Jeanpierre approximations are the most appropriate self-shielding options to use in this case of study. In addition, the 89 groups WIMS-AECL library for slight enriched uranium and the 172 groups WLUP library for a mixture of plutonium and thorium give the most consistent results with those of SERPENT. (authors)

Depletion methodology in the 3-D whole core transport code DeCART

Energy Technology Data Exchange (ETDEWEB)

Kim, Kang Seog; Cho, Jin Young; Zee, Sung Quun

2005-02-01

Three dimensional whole-core transport code DeCART has been developed to include a characteristics of the numerical reactor to replace partly the experiment. This code adopts the deterministic method in simulating the neutron behavior with the least assumption and approximation. This neutronic code is also coupled with the thermal hydraulic code CFD and the thermo mechanical code to simulate the combined effects. Depletion module has been implemented in DeCART code to predict the depleted composition in the fuel. The exponential matrix method of ORIGEN-2 has been used for the depletion calculation. The library of including decay constants, yield matrix and others has been used and greatly simplified for the calculation efficiency. This report summarizes the theoretical backgrounds and includes the verification of the depletion module in DeCART by performing the benchmark calculations.

A Deterministic Electron, Photon, Proton and Heavy Ion Radiation Transport Suite for the Study of the Jovian System

Science.gov (United States)

Norman, Ryan B.; Badavi, Francis F.; Blattnig, Steve R.; Atwell, William

2011-01-01

A deterministic suite of radiation transport codes, developed at NASA Langley Research Center (LaRC), which describe the transport of electrons, photons, protons, and heavy ions in condensed media is used to simulate exposures from spectral distributions typical of electrons, protons and carbon-oxygen-sulfur (C-O-S) trapped heavy ions in the Jovian radiation environment. The particle transport suite consists of a coupled electron and photon deterministic transport algorithm (CEPTRN) and a coupled light particle and heavy ion deterministic transport algorithm (HZETRN). The primary purpose for the development of the transport suite is to provide a means for the spacecraft design community to rapidly perform numerous repetitive calculations essential for electron, proton and heavy ion radiation exposure assessments in complex space structures. In this paper, the radiation environment of the Galilean satellite Europa is used as a representative boundary condition to show the capabilities of the transport suite. While the transport suite can directly access the output electron spectra of the Jovian environment as generated by the Jet Propulsion Laboratory (JPL) Galileo Interim Radiation Electron (GIRE) model of 2003; for the sake of relevance to the upcoming Europa Jupiter System Mission (EJSM), the 105 days at Europa mission fluence energy spectra provided by JPL is used to produce the corresponding dose-depth curve in silicon behind an aluminum shield of 100 mils ( 0.7 g/sq cm). The transport suite can also accept ray-traced thickness files from a computer-aided design (CAD) package and calculate the total ionizing dose (TID) at a specific target point. In that regard, using a low-fidelity CAD model of the Galileo probe, the transport suite was verified by comparing with Monte Carlo (MC) simulations for orbits JOI--J35 of the Galileo extended mission (1996-2001). For the upcoming EJSM mission with a potential launch date of 2020, the transport suite is used to compute

Deterministic methods in radiation transport. A compilation of papers presented February 4--5, 1992

Energy Technology Data Exchange (ETDEWEB)

Rice, A.F.; Roussin, R.W. [eds.

1992-06-01

The Seminar on Deterministic Methods in Radiation Transport was held February 4--5, 1992, in Oak Ridge, Tennessee. Eleven presentations were made and the full papers are published in this report, along with three that were submitted but not given orally. These papers represent a good overview of the state of the art in the deterministic solution of radiation transport problems for a variety of applications of current interest to the Radiation Shielding Information Center user community.

Deterministic methods in radiation transport. A compilation of papers presented February 4-5, 1992

Energy Technology Data Exchange (ETDEWEB)

Rice, A. F.; Roussin, R. W. [eds.

1992-06-01

The Seminar on Deterministic Methods in Radiation Transport was held February 4--5, 1992, in Oak Ridge, Tennessee. Eleven presentations were made and the full papers are published in this report, along with three that were submitted but not given orally. These papers represent a good overview of the state of the art in the deterministic solution of radiation transport problems for a variety of applications of current interest to the Radiation Shielding Information Center user community.

Deterministic and unambiguous dense coding

International Nuclear Information System (INIS)

Wu Shengjun; Cohen, Scott M.; Sun Yuqing; Griffiths, Robert B.

2006-01-01

Optimal dense coding using a partially-entangled pure state of Schmidt rank D and a noiseless quantum channel of dimension D is studied both in the deterministic case where at most L d messages can be transmitted with perfect fidelity, and in the unambiguous case where when the protocol succeeds (probability τ x ) Bob knows for sure that Alice sent message x, and when it fails (probability 1-τ x ) he knows it has failed. Alice is allowed any single-shot (one use) encoding procedure, and Bob any single-shot measurement. For D≤D a bound is obtained for L d in terms of the largest Schmidt coefficient of the entangled state, and is compared with published results by Mozes et al. [Phys. Rev. A71, 012311 (2005)]. For D>D it is shown that L d is strictly less than D 2 unless D is an integer multiple of D, in which case uniform (maximal) entanglement is not needed to achieve the optimal protocol. The unambiguous case is studied for D≤D, assuming τ x >0 for a set of DD messages, and a bound is obtained for the average . A bound on the average requires an additional assumption of encoding by isometries (unitaries when D=D) that are orthogonal for different messages. Both bounds are saturated when τ x is a constant independent of x, by a protocol based on one-shot entanglement concentration. For D>D it is shown that (at least) D 2 messages can be sent unambiguously. Whether unitary (isometric) encoding suffices for optimal protocols remains a major unanswered question, both for our work and for previous studies of dense coding using partially-entangled states, including noisy (mixed) states

GPU accelerated simulations of 3D deterministic particle transport using discrete ordinates method

International Nuclear Information System (INIS)

Gong Chunye; Liu Jie; Chi Lihua; Huang Haowei; Fang Jingyue; Gong Zhenghu

2011-01-01

Graphics Processing Unit (GPU), originally developed for real-time, high-definition 3D graphics in computer games, now provides great faculty in solving scientific applications. The basis of particle transport simulation is the time-dependent, multi-group, inhomogeneous Boltzmann transport equation. The numerical solution to the Boltzmann equation involves the discrete ordinates (S n ) method and the procedure of source iteration. In this paper, we present a GPU accelerated simulation of one energy group time-independent deterministic discrete ordinates particle transport in 3D Cartesian geometry (Sweep3D). The performance of the GPU simulations are reported with the simulations of vacuum boundary condition. The discussion of the relative advantages and disadvantages of the GPU implementation, the simulation on multi GPUs, the programming effort and code portability are also reported. The results show that the overall performance speedup of one NVIDIA Tesla M2050 GPU ranges from 2.56 compared with one Intel Xeon X5670 chip to 8.14 compared with one Intel Core Q6600 chip for no flux fixup. The simulation with flux fixup on one M2050 is 1.23 times faster than on one X5670.

GPU accelerated simulations of 3D deterministic particle transport using discrete ordinates method

Science.gov (United States)

Gong, Chunye; Liu, Jie; Chi, Lihua; Huang, Haowei; Fang, Jingyue; Gong, Zhenghu

2011-07-01

Graphics Processing Unit (GPU), originally developed for real-time, high-definition 3D graphics in computer games, now provides great faculty in solving scientific applications. The basis of particle transport simulation is the time-dependent, multi-group, inhomogeneous Boltzmann transport equation. The numerical solution to the Boltzmann equation involves the discrete ordinates ( Sn) method and the procedure of source iteration. In this paper, we present a GPU accelerated simulation of one energy group time-independent deterministic discrete ordinates particle transport in 3D Cartesian geometry (Sweep3D). The performance of the GPU simulations are reported with the simulations of vacuum boundary condition. The discussion of the relative advantages and disadvantages of the GPU implementation, the simulation on multi GPUs, the programming effort and code portability are also reported. The results show that the overall performance speedup of one NVIDIA Tesla M2050 GPU ranges from 2.56 compared with one Intel Xeon X5670 chip to 8.14 compared with one Intel Core Q6600 chip for no flux fixup. The simulation with flux fixup on one M2050 is 1.23 times faster than on one X5670.

Heavy-ion transport codes for radiotherapy and radioprotection in space

International Nuclear Information System (INIS)

Mancusi, Davide

2006-06-01

Simulation of the transport of heavy ions in matter is a field of nuclear science that has recently received attention in view of its importance for some relevant applications. Accelerated heavy ions can, for example, be used to treat cancers (heavy-ion radiotherapy) and show some superior qualities with respect to more conventional treatment systems, like photons (x-rays) or protons. Furthermore, long-term manned space missions (like a possible future mission to Mars) pose the challenge to protect astronauts and equipment on board against the harmful space radiation environment, where heavy ions can be responsible for a significant share of the exposure risk. The high accuracy expected from a transport algorithm (especially in the case of radiotherapy) and the large amount of semi-empirical knowledge necessary to even state the transport problem properly rule out any analytical approach; the alternative is to resort to numerical simulations in order to build treatment-planning systems for cancer or to aid space engineers in shielding design. This thesis is focused on the description of HIBRAC, a one-dimensional deterministic code optimised for radiotherapy, and PHITS (Particle and Heavy- Ion Transport System), a general-purpose three-dimensional Monte-Carlo code. The structure of both codes is outlined and some relevant results are presented. In the case of PHITS, we also report the first results of an ongoing comprehensive benchmarking program for the main components of the code; we present the comparison of partial charge-changing cross sections for a 400 MeV/n 40 Ar beam impinging on carbon, polyethylene, aluminium, copper, tin and lead targets

Validation of the 3D finite element transport theory code EVENT for shielding applications

International Nuclear Information System (INIS)

Warner, Paul; Oliveira, R.E. de

2000-01-01

This paper is concerned with the validation of the 3D deterministic neutral-particle transport theory code EVENT for shielding applications. The code is based on the finite element-spherical harmonics (FE-P N ) method which has been extensively developed over the last decade. A general multi-group, anisotropic scattering formalism enables the code to address realistic steady state and time dependent, multi-dimensional coupled neutron/gamma radiation transport problems involving high scattering and deep penetration alike. The powerful geometrical flexibility and competitive computational effort makes the code an attractive tool for shielding applications. In recognition of this, EVENT is currently in the process of being adopted by the UK nuclear industry. The theory behind EVENT is described and its numerical implementation is outlined. Numerical results obtained by the code are compared with predictions of the Monte Carlo code MCBEND and also with the results from benchmark shielding experiments. In particular, results are presented for the ASPIS experimental configuration for both neutron and gamma ray calculations using the BUGLE 96 nuclear data library. (author)

The concerted calculation of the BN-600 reactor for the deterministic and stochastic codes

Science.gov (United States)

Bogdanova, E. V.; Kuznetsov, A. N.

2017-01-01

The solution of the problem of increasing the safety of nuclear power plants implies the existence of complete and reliable information about the processes occurring in the core of a working reactor. Nowadays the Monte-Carlo method is the most general-purpose method used to calculate the neutron-physical characteristic of the reactor. But it is characterized by large time of calculation. Therefore, it may be useful to carry out coupled calculations with stochastic and deterministic codes. This article presents the results of research for possibility of combining stochastic and deterministic algorithms in calculation the reactor BN-600. This is only one part of the work, which was carried out in the framework of the graduation project at the NRC “Kurchatov Institute” in cooperation with S. S. Gorodkov and M. A. Kalugin. It is considering the 2-D layer of the BN-600 reactor core from the international benchmark test, published in the report IAEA-TECDOC-1623. Calculations of the reactor were performed with MCU code and then with a standard operative diffusion algorithm with constants taken from the Monte - Carlo computation. Macro cross-section, diffusion coefficients, the effective multiplication factor and the distribution of neutron flux and power were obtained in 15 energy groups. The reasonable agreement between stochastic and deterministic calculations of the BN-600 is observed.

Heavy-ion transport codes for radiotherapy and radioprotection in space

Energy Technology Data Exchange (ETDEWEB)

Mancusi, Davide

2006-06-15

Simulation of the transport of heavy ions in matter is a field of nuclear science that has recently received attention in view of its importance for some relevant applications. Accelerated heavy ions can, for example, be used to treat cancers (heavy-ion radiotherapy) and show some superior qualities with respect to more conventional treatment systems, like photons (x-rays) or protons. Furthermore, long-term manned space missions (like a possible future mission to Mars) pose the challenge to protect astronauts and equipment on board against the harmful space radiation environment, where heavy ions can be responsible for a significant share of the exposure risk. The high accuracy expected from a transport algorithm (especially in the case of radiotherapy) and the large amount of semi-empirical knowledge necessary to even state the transport problem properly rule out any analytical approach; the alternative is to resort to numerical simulations in order to build treatment-planning systems for cancer or to aid space engineers in shielding design. This thesis is focused on the description of HIBRAC, a one-dimensional deterministic code optimised for radiotherapy, and PHITS (Particle and Heavy- Ion Transport System), a general-purpose three-dimensional Monte-Carlo code. The structure of both codes is outlined and some relevant results are presented. In the case of PHITS, we also report the first results of an ongoing comprehensive benchmarking program for the main components of the code; we present the comparison of partial charge-changing cross sections for a 400 MeV/n {sup 40}Ar beam impinging on carbon, polyethylene, aluminium, copper, tin and lead targets.

SERPENT Monte Carlo reactor physics code

International Nuclear Information System (INIS)

Leppaenen, J.

2010-01-01

SERPENT is a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, developed at VTT Technical Research Centre of Finland since 2004. The code is specialized in lattice physics applications, but the universe-based geometry description allows transport simulation to be carried out in complicated three-dimensional geometries as well. The suggested applications of SERPENT include generation of homogenized multi-group constants for deterministic reactor simulator calculations, fuel cycle studies involving detailed assembly-level burnup calculations, validation of deterministic lattice transport codes, research reactor applications, educational purposes and demonstration of reactor physics phenomena. The Serpent code has been publicly distributed by the OECD/NEA Data Bank since May 2009 and RSICC in the U. S. since March 2010. The code is being used in some 35 organizations in 20 countries around the world. This paper presents an overview of the methods and capabilities of the Serpent code, with examples in the modelling of WWER-440 reactor physics. (Author)

Overcoming a limitation of deterministic dense coding with a nonmaximally entangled initial state

International Nuclear Information System (INIS)

Bourdon, P. S.; Gerjuoy, E.

2010-01-01

Under two-party deterministic dense coding, Alice communicates (perfectly distinguishable) messages to Bob via a qudit from a pair of entangled qudits in pure state |Ψ>. If |Ψ> represents a maximally entangled state (i.e., each of its Schmidt coefficients is √(1/d)), then Alice can convey to Bob one of d 2 distinct messages. If |Ψ> is not maximally entangled, then Ji et al. [Phys. Rev. A 73, 034307 (2006)] have shown that under the original deterministic dense-coding protocol, in which messages are encoded by unitary operations performed on Alice's qudit, it is impossible to encode d 2 -1 messages. Encoding d 2 -2 messages is possible; see, for example, the numerical studies by Mozes et al. [Phys. Rev. A 71, 012311 (2005)]. Answering a question raised by Wu et al. [Phys. Rev. A 73, 042311 (2006)], we show that when |Ψ> is not maximally entangled, the communications limit of d 2 -2 messages persists even when the requirement that Alice encode by unitary operations on her qudit is weakened to allow encoding by more general quantum operators. We then describe a dense-coding protocol that can overcome this limitation with high probability, assuming the largest Schmidt coefficient of |Ψ> is sufficiently close to √(1/d). In this protocol, d 2 -2 of the messages are encoded via unitary operations on Alice's qudit, and the final (d 2 -1)-th message is encoded via a non-trace-preserving quantum operation.

2D deterministic radiation transport with the discontinuous finite element method

International Nuclear Information System (INIS)

Kershaw, D.; Harte, J.

1993-01-01

This report provides a complete description of the analytic and discretized equations for 2D deterministic radiation transport. This computational model has been checked against a wide variety of analytic test problems and found to give excellent results. We make extensive use of the discontinuous finite element method

Deterministic sensitivity analysis for the numerical simulation of contaminants transport; Analyse de sensibilite deterministe pour la simulation numerique du transfert de contaminants

Energy Technology Data Exchange (ETDEWEB)

Marchand, E

2007-12-15

The questions of safety and uncertainty are central to feasibility studies for an underground nuclear waste storage site, in particular the evaluation of uncertainties about safety indicators which are due to uncertainties concerning properties of the subsoil or of the contaminants. The global approach through probabilistic Monte Carlo methods gives good results, but it requires a large number of simulations. The deterministic method investigated here is complementary. Based on the Singular Value Decomposition of the derivative of the model, it gives only local information, but it is much less demanding in computing time. The flow model follows Darcy's law and the transport of radionuclides around the storage site follows a linear convection-diffusion equation. Manual and automatic differentiation are compared for these models using direct and adjoint modes. A comparative study of both probabilistic and deterministic approaches for the sensitivity analysis of fluxes of contaminants through outlet channels with respect to variations of input parameters is carried out with realistic data provided by ANDRA. Generic tools for sensitivity analysis and code coupling are developed in the Caml language. The user of these generic platforms has only to provide the specific part of the application in any language of his choice. We also present a study about two-phase air/water partially saturated flows in hydrogeology concerning the limitations of the Richards approximation and of the global pressure formulation used in petroleum engineering. (author)

Verification & Validation of High-Order Short-Characteristics-Based Deterministic Transport Methodology on Unstructured Grids

International Nuclear Information System (INIS)

Azmy, Yousry; Wang, Yaqi

2013-01-01

The research team has developed a practical, high-order, discrete-ordinates, short characteristics neutron transport code for three-dimensional configurations represented on unstructured tetrahedral grids that can be used for realistic reactor physics applications at both the assembly and core levels. This project will perform a comprehensive verification and validation of this new computational tool against both a continuous-energy Monte Carlo simulation (e.g. MCNP) and experimentally measured data, an essential prerequisite for its deployment in reactor core modeling. Verification is divided into three phases. The team will first conduct spatial mesh and expansion order refinement studies to monitor convergence of the numerical solution to reference solutions. This is quantified by convergence rates that are based on integral error norms computed from the cell-by-cell difference between the code's numerical solution and its reference counterpart. The latter is either analytic or very fine- mesh numerical solutions from independent computational tools. For the second phase, the team will create a suite of code-independent benchmark configurations to enable testing the theoretical order of accuracy of any particular discretization of the discrete ordinates approximation of the transport equation. For each tested case (i.e. mesh and spatial approximation order), researchers will execute the code and compare the resulting numerical solution to the exact solution on a per cell basis to determine the distribution of the numerical error. The final activity comprises a comparison to continuous-energy Monte Carlo solutions for zero-power critical configuration measurements at Idaho National Laboratory's Advanced Test Reactor (ATR). Results of this comparison will allow the investigators to distinguish between modeling errors and the above-listed discretization errors introduced by the deterministic method, and to separate the sources of uncertainty.

A Case for Dynamic Reverse-code Generation to Debug Non-deterministic Programs

Directory of Open Access Journals (Sweden)

Jooyong Yi

2013-09-01

Full Text Available Backtracking (i.e., reverse execution helps the user of a debugger to naturally think backwards along the execution path of a program, and thinking backwards makes it easy to locate the origin of a bug. So far backtracking has been implemented mostly by state saving or by checkpointing. These implementations, however, inherently do not scale. Meanwhile, a more recent backtracking method based on reverse-code generation seems promising because executing reverse code can restore the previous states of a program without state saving. In the literature, there can be found two methods that generate reverse code: (a static reverse-code generation that pre-generates reverse code through static analysis before starting a debugging session, and (b dynamic reverse-code generation that generates reverse code by applying dynamic analysis on the fly during a debugging session. In particular, we espoused the latter one in our previous work to accommodate non-determinism of a program caused by e.g., multi-threading. To demonstrate the usefulness of our dynamic reverse-code generation, this article presents a case study of various backtracking methods including ours. We compare the memory usage of various backtracking methods in a simple but nontrivial example, a bounded-buffer program. In the case of non-deterministic programs such as this bounded-buffer program, our dynamic reverse-code generation outperforms the existing backtracking methods in terms of memory efficiency.

Sub-Transport Layer Coding

DEFF Research Database (Denmark)

Hansen, Jonas; Krigslund, Jeppe; Roetter, Daniel Enrique Lucani

2014-01-01

Packet losses in wireless networks dramatically curbs the performance of TCP. This paper introduces a simple coding shim that aids IP-layer traffic in lossy environments while being transparent to transport layer protocols. The proposed coding approach enables erasure correction while being...... oblivious to the congestion control algorithms of the utilised transport layer protocol. Although our coding shim is indifferent towards the transport layer protocol, we focus on the performance of TCP when ran on top of our proposed coding mechanism due to its widespread use. The coding shim provides gains...

Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system

International Nuclear Information System (INIS)

Yang, W.S.; Lee, C.H.

2008-01-01

Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC 2 -2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC 2 -2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC 2 -2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC 2 -2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC 2 -2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC 2 -2, VIM, and NJOY. For almost all nuclides considered, MC 2 -2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC 2 -2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC 2 -2/TWODANT calculations were in good agreement with MCNP solutions within ∼0.25% Δρ, except a few small LANL fast assemblies. Relative to the MCNP solution, the MC 2 -2/TWODANT

Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

Energy Technology Data Exchange (ETDEWEB)

Yang, W. S.; Lee, C. H. (Nuclear Engineering Division)

2008-05-16

Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC{sup 2}-2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC{sup 2}-2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC{sup 2}-2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC{sup 2}-2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC{sup 2}-2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC{sup 2}-2, VIM, and NJOY. For almost all nuclides considered, MC{sup 2}-2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC{sup 2}-2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC{sup 2}-2/TWODANT calculations were in good agreement with MCNP solutions within {approx}0.25% {Delta}{rho}, except a few small LANL fast assemblies

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)

Deterministic dense coding and entanglement entropy

International Nuclear Information System (INIS)

Bourdon, P. S.; Gerjuoy, E.; McDonald, J. P.; Williams, H. T.

2008-01-01

We present an analytical study of the standard two-party deterministic dense-coding protocol, under which communication of perfectly distinguishable messages takes place via a qudit from a pair of nonmaximally entangled qudits in a pure state |ψ>. Our results include the following: (i) We prove that it is possible for a state |ψ> with lower entanglement entropy to support the sending of a greater number of perfectly distinguishable messages than one with higher entanglement entropy, confirming a result suggested via numerical analysis in Mozes et al. [Phys. Rev. A 71, 012311 (2005)]. (ii) By explicit construction of families of local unitary operators, we verify, for dimensions d=3 and d=4, a conjecture of Mozes et al. about the minimum entanglement entropy that supports the sending of d+j messages, 2≤j≤d-1; moreover, we show that the j=2 and j=d-1 cases of the conjecture are valid in all dimensions. (iii) Given that |ψ> allows the sending of K messages and has √(λ 0 ) as its largest Schmidt coefficient, we show that the inequality λ 0 ≤d/K, established by Wu et al. [Phys. Rev. A 73, 042311 (2006)], must actually take the form λ 0 < d/K if K=d+1, while our constructions of local unitaries show that equality can be realized if K=d+2 or K=2d-1

A Deep Penetration Problem Calculation Using AETIUS:An Easy Modeling Discrete Ordinates Transport Code UsIng Unstructured Tetrahedral Mesh, Shared Memory Parallel

Science.gov (United States)

KIM, Jong Woon; LEE, Young-Ouk

2017-09-01

As computing power gets better and better, computer codes that use a deterministic method seem to be less useful than those using the Monte Carlo method. In addition, users do not like to think about space, angles, and energy discretization for deterministic codes. However, a deterministic method is still powerful in that we can obtain a solution of the flux throughout the problem, particularly as when particles can barely penetrate, such as in a deep penetration problem with small detection volumes. Recently, a new state-of-the-art discrete-ordinates code, ATTILA, was developed and has been widely used in several applications. ATTILA provides the capabilities to solve geometrically complex 3-D transport problems by using an unstructured tetrahedral mesh. Since 2009, we have been developing our own code by benchmarking ATTILA. AETIUS is a discrete ordinates code that uses an unstructured tetrahedral mesh such as ATTILA. For pre- and post- processing, Gmsh is used to generate an unstructured tetrahedral mesh by importing a CAD file (*.step) and visualizing the calculation results of AETIUS. Using a CAD tool, the geometry can be modeled very easily. In this paper, we describe a brief overview of AETIUS and provide numerical results from both AETIUS and a Monte Carlo code, MCNP5, in a deep penetration problem with small detection volumes. The results demonstrate the effectiveness and efficiency of AETIUS for such calculations.

Performance, Accuracy and Efficiency Evaluation of a Three-Dimensional Whole-Core Neutron Transport Code AGENT

International Nuclear Information System (INIS)

Jevremovic, Tatjana; Hursin, Mathieu; Satvat, Nader; Hopkins, John; Xiao, Shanjie; Gert, Godfree

2006-01-01

The AGENT (Arbitrary Geometry Neutron Transport) an open-architecture reactor modeling tool is deterministic neutron transport code for two or three-dimensional heterogeneous neutronic design and analysis of the whole reactor cores regardless of geometry types and material configurations. The AGENT neutron transport methodology is applicable to all generations of nuclear power and research reactors. It combines three theories: (1) the theory of R-functions used to generate real three-dimensional whole-cores of square, hexagonal or triangular cross sections, (2) the planar method of characteristics used to solve isotropic neutron transport in non-homogenized 2D) reactor slices, and (3) the one-dimensional diffusion theory used to couple the planar and axial neutron tracks through the transverse leakage and angular mesh-wise flux values. The R-function-geometrical module allows a sequential building of the layers of geometry and automatic sub-meshing based on the network of domain functions. The simplicity of geometry description and selection of parameters for accurate treatment of neutron propagation is achieved through the Boolean algebraic hierarchically organized simple primitives into complex domains (both being represented with corresponding domain functions). The accuracy is comparable to Monte Carlo codes and is obtained by following neutron propagation through real geometrical domains that does not require homogenization or simplifications. The efficiency is maintained through a set of acceleration techniques introduced at all important calculation levels. The flux solution incorporates power iteration with two different acceleration techniques: Coarse Mesh Re-balancing (CMR) and Coarse Mesh Finite Difference (CMFD). The stand-alone originally developed graphical user interface of the AGENT code design environment allows the user to view and verify input data by displaying the geometry and material distribution. The user can also view the output data such

Convergence studies of deterministic methods for LWR explicit reflector methodology

International Nuclear Information System (INIS)

Canepa, S.; Hursin, M.; Ferroukhi, H.; Pautz, A.

2013-01-01

The standard approach in modem 3-D core simulators, employed either for steady-state or transient simulations, is to use Albedo coefficients or explicit reflectors at the core axial and radial boundaries. In the latter approach, few-group homogenized nuclear data are a priori produced with lattice transport codes using 2-D reflector models. Recently, the explicit reflector methodology of the deterministic CASMO-4/SIMULATE-3 code system was identified to potentially constitute one of the main sources of errors for core analyses of the Swiss operating LWRs, which are all belonging to GII design. Considering that some of the new GIII designs will rely on very different reflector concepts, a review and assessment of the reflector methodology for various LWR designs appeared as relevant. Therefore, the purpose of this paper is to first recall the concepts of the explicit reflector modelling approach as employed by CASMO/SIMULATE. Then, for selected reflector configurations representative of both GII and GUI designs, a benchmarking of the few-group nuclear data produced with the deterministic lattice code CASMO-4 and its successor CASMO-5, is conducted. On this basis, a convergence study with regards to geometrical requirements when using deterministic methods with 2-D homogenous models is conducted and the effect on the downstream 3-D core analysis accuracy is evaluated for a typical GII deflector design in order to assess the results against available plant measurements. (authors)

Use of deterministic methods in survey calculations for criticality problems

International Nuclear Information System (INIS)

Hutton, J.L.; Phenix, J.; Course, A.F.

1991-01-01

A code package using deterministic methods for solving the Boltzmann Transport equation is the WIMS suite. This has been very successful for a range of situations. In particular it has been used with great success to analyse trends in reactivity with a range of changes in state. The WIMS suite of codes have a range of methods and are very flexible in the way they can be combined. A wide variety of situations can be modelled ranging through all the current Thermal Reactor variants to storage systems and items of chemical plant. These methods have recently been enhanced by the introduction of the CACTUS method. This is based on a characteristics technique for solving the Transport equation and has the advantage that complex geometrical situations can be treated. In this paper the basis of the method is outlined and examples of its use are illustrated. In parallel with these developments the validation for out of pile situations has been extended to include experiments with relevance to criticality situations. The paper will summarise this evidence and show how these results point to a partial re-adoption of deterministic methods for some areas of criticality. The paper also presents results to illustrate the use of WIMS in criticality situations and in particular show how it can complement codes such as MONK when used for surveying the reactivity effect due to changes in geometry or materials. (Author)

Deterministic Quantum Secure Direct Communication with Dense Coding and Continuous Variable Operations

International Nuclear Information System (INIS)

Han Lianfang; Chen Yueming; Yuan Hao

2009-01-01

We propose a deterministic quantum secure direct communication protocol by using dense coding. The two check photon sequences are used to check the securities of the channels between the message sender and the receiver. The continuous variable operations instead of the usual discrete unitary operations are performed on the travel photons so that the security of the present protocol can be enhanced. Therefore some specific attacks such as denial-of-service attack, intercept-measure-resend attack and invisible photon attack can be prevented in ideal quantum channel. In addition, the scheme is still secure in noise channel. Furthermore, this protocol has the advantage of high capacity and can be realized in the experiment. (general)

In-facility transport code review

International Nuclear Information System (INIS)

Spore, J.W.; Boyack, B.E.; Bohl, W.R.

1996-07-01

The following computer codes were reviewed by the In-Facility Transport Working Group for application to the in-facility transport of radioactive aerosols, flammable gases, and/or toxic gases: (1) CONTAIN, (2) FIRAC, (3) GASFLOW, (4) KBERT, and (5) MELCOR. Based on the review criteria as described in this report and the versions of each code available at the time of the review, MELCOR is the best code for the analysis of in-facility transport when multidimensional effects are not significant. When multi-dimensional effects are significant, GASFLOW should be used

Solution to the transport equation with anisotropic dispersion in a BWR type assembly using the AZTRAN code

International Nuclear Information System (INIS)

Chepe P, M.; Xolocostli M, J. V.; Gomez T, A. M.; Del Valle G, E.

2016-09-01

Due to the current computing power, the deterministic codes for analyzing nuclear reactors that have been used for several years are becoming more relevant, since much more precise solution techniques can be used; the last century would have been very difficult, since memory and processor capacities were very limited or had high prices on the components. In this work we analyze the effect of the anisotropic dispersion of the effective dispersion section, compared to the isotropic dispersion. The anisotropy implementation was carried out in the AZTRAN transport code, which is part of the AZTLAN platform for nuclear reactors analysis (in development). The AZTRAN code solves the Boltzmann transport equation in one, two and three dimensions at steady state, using the multi-group technique for energy discretization, the RTN-0 nodal method in spatial discretization and for angular discretization the discrete ordinates without considering anisotropy originally. The effect of the anisotropy dispersion on the effective multiplication factor and the axial and radial power on a fuel assembly BWR type are analyzed. (Author)

Particle-tracking code (track3d) for convective solute transport modelling in the geosphere: Description and user`s manual; Programme de reperage de particules (track3d) pour la modelisation du transport par convection des solutes dans la geosphere: description et manuel de l`utilisateur

Energy Technology Data Exchange (ETDEWEB)

Nakka, B W; Chan, T

1994-12-01

A deterministic particle-tracking code (TRACK3D) has been developed to compute convective flow paths of conservative (nonreactive) contaminants through porous geological media. TRACK3D requires the groundwater velocity distribution, which, in our applications, results from flow simulations using AECL`s MOTIF code. The MOTIF finite-element code solves the transient and steady-state coupled equations of groundwater flow, solute transport and heat transport in fractured/porous media. With few modifications, TRACK3D can be used to analyse the velocity distributions calculated by other finite-element or finite-difference flow codes. This report describes the assumptions, limitations, organization, operation and applications of the TRACK3D code, and provides a comprehensive user`s manual.

Analysis of the pool critical assembly benchmark using raptor-M3G, a parallel deterministic radiation transport code - 289

International Nuclear Information System (INIS)

Fischer, G.A.

2010-01-01

The PCA Benchmark is analyzed using RAPTOR-M3G, a parallel SN radiation transport code. A variety of mesh structures, angular quadrature sets, cross section treatments, and reactor dosimetry cross sections are presented. The results show that RAPTOR-M3G is generally suitable for PWR neutron dosimetry applications. (authors)

A deterministic electron, photon, proton and heavy ion transport suite for the study of the Jovian moon Europa

International Nuclear Information System (INIS)

Badavi, Francis F.; Blattnig, Steve R.; Atwell, William; Nealy, John E.; Norman, Ryan B.

2011-01-01

A Langley research center (LaRC) developed deterministic suite of radiation transport codes describing the propagation of electron, photon, proton and heavy ion in condensed media is used to simulate the exposure from the spectral distribution of the aforementioned particles in the Jovian radiation environment. Based on the measurements by the Galileo probe (1995-2003) heavy ion counter (HIC), the choice of trapped heavy ions is limited to carbon, oxygen and sulfur (COS). The deterministic particle transport suite consists of a coupled electron photon algorithm (CEPTRN) and a coupled light heavy ion algorithm (HZETRN). The primary purpose for the development of the transport suite is to provide a means to the spacecraft design community to rapidly perform numerous repetitive calculations essential for electron, photon, proton and heavy ion exposure assessment in a complex space structure. In this paper, the reference radiation environment of the Galilean satellite Europa is used as a representative boundary condition to show the capabilities of the transport suite. While the transport suite can directly access the output electron and proton spectra of the Jovian environment as generated by the jet propulsion laboratory (JPL) Galileo interim radiation electron (GIRE) model of 2003; for the sake of relevance to the upcoming Europa Jupiter system mission (EJSM), the JPL provided Europa mission fluence spectrum, is used to produce the corresponding depth dose curve in silicon behind a default aluminum shield of 100 mils (∼0.7 g/cm 2 ). The transport suite can also accept a geometry describing ray traced thickness file from a computer aided design (CAD) package and calculate the total ionizing dose (TID) at a specific target point within the interior of the vehicle. In that regard, using a low fidelity CAD model of the Galileo probe generated by the authors, the transport suite was verified versus Monte Carlo (MC) simulation for orbits JOI-J35 of the Galileo probe

Prospects in deterministic three dimensional whole-core transport calculations

International Nuclear Information System (INIS)

Sanchez, Richard

2012-01-01

The point we made in this paper is that, although detailed and precise three-dimensional (3D) whole-core transport calculations may be obtained in the future with massively parallel computers, they would have an application to only some of the problems of the nuclear industry, more precisely those regarding multiphysics or for methodology validation or nuclear safety calculations. On the other hand, typical design reactor cycle calculations comprising many one-point core calculations can have very strict constraints in computing time and will not directly benefit from the advances in computations in large scale computers. Consequently, in this paper we review some of the deterministic 3D transport methods which in the very near future may have potential for industrial applications and, even with low-order approximations such as a low resolution in energy, might represent an advantage as compared with present industrial methodology, for which one of the main approximations is due to power reconstruction. These methods comprise the response-matrix method and methods based on the two-dimensional (2D) method of characteristics, such as the fusion method.

Fracture flow code

International Nuclear Information System (INIS)

Dershowitz, W; Herbert, A.; Long, J.

1989-03-01

The hydrology of the SCV site will be modelled utilizing discrete fracture flow models. These models are complex, and can not be fully cerified by comparison to analytical solutions. The best approach for verification of these codes is therefore cross-verification between different codes. This is complicated by the variation in assumptions and solution techniques utilized in different codes. Cross-verification procedures are defined which allow comparison of the codes developed by Harwell Laboratory, Lawrence Berkeley Laboratory, and Golder Associates Inc. Six cross-verification datasets are defined for deterministic and stochastic verification of geometric and flow features of the codes. Additional datasets for verification of transport features will be documented in a future report. (13 figs., 7 tabs., 10 refs.) (authors)

Comparison of deterministic and Monte Carlo methods in shielding design.

Science.gov (United States)

Oliveira, A D; Oliveira, C

2005-01-01

In shielding calculation, deterministic methods have some advantages and also some disadvantages relative to other kind of codes, such as Monte Carlo. The main advantage is the short computer time needed to find solutions while the disadvantages are related to the often-used build-up factor that is extrapolated from high to low energies or with unknown geometrical conditions, which can lead to significant errors in shielding results. The aim of this work is to investigate how good are some deterministic methods to calculating low-energy shielding, using attenuation coefficients and build-up factor corrections. Commercial software MicroShield 5.05 has been used as the deterministic code while MCNP has been used as the Monte Carlo code. Point and cylindrical sources with slab shield have been defined allowing comparison between the capability of both Monte Carlo and deterministic methods in a day-by-day shielding calculation using sensitivity analysis of significant parameters, such as energy and geometrical conditions.

Comparison of deterministic and Monte Carlo methods in shielding design

International Nuclear Information System (INIS)

Oliveira, A. D.; Oliveira, C.

2005-01-01

In shielding calculation, deterministic methods have some advantages and also some disadvantages relative to other kind of codes, such as Monte Carlo. The main advantage is the short computer time needed to find solutions while the disadvantages are related to the often-used build-up factor that is extrapolated from high to low energies or with unknown geometrical conditions, which can lead to significant errors in shielding results. The aim of this work is to investigate how good are some deterministic methods to calculating low-energy shielding, using attenuation coefficients and build-up factor corrections. Commercial software MicroShield 5.05 has been used as the deterministic code while MCNP has been used as the Monte Carlo code. Point and cylindrical sources with slab shield have been defined allowing comparison between the capability of both Monte Carlo and deterministic methods in a day-by-day shielding calculation using sensitivity analysis of significant parameters, such as energy and geometrical conditions. (authors)

Development and implementation of a set of numerical quadratures SQN and EQN type in the transport code AZTRAN

International Nuclear Information System (INIS)

Chepe P, M.; Xolocostli M, J. V.; Gomez T, A. M.; Del Valle G, E.

2015-09-01

The deterministic transport codes for analysis of nuclear reactors have been used for several years already, these codes have evolved in terms of the methodology used and the degree of accuracy, because at the present time has more computer power. In this paper, the transport code used considers the classical technique of multi-group for discretization energy, for space discretization uses the nodal methods, while for the angular discretization the discrete ordinates method is used; so that presents the development and implementation of a set of numerical quadratures of SQ N type symmetrical with the same weight for each angular direction and these are compared with the quadratures of EQ N type. The two sets of numerical quadratures were implemented in the program AZTRAN to a problem with isotropic medium in XYZ geometry, in steady state using the nodal method RTN-0 (Raviart-Thomas-Nedelec). The analyzed results correspond to the effective multiplication factor k eff and neutron angular flux with approximations from S 4 to S 16 . (Author)

Deterministic 3D transport, sensitivity and uncertainty analysis of TPR and reaction rate measurements in HCPB Breeder Blanket mock-up benchmark

International Nuclear Information System (INIS)

Kodeli, I.

2006-01-01

The Helium-Cooled Pebble Bed (HCPB) Breeder Blanket mock-up benchmark experiment was analysed using the deterministic transport, sensitivity and uncertainty code system in order to determine the Tritium Production Rate (TPR) in the ceramic breeder and the neutron reaction rates in beryllium, both nominal values and the corresponding uncertainties. The experiment, performed in 2005 to validate the HCPB concept, consists of a metallic beryllium set-up with two double layers of breeder material (Li 2 CO 3 powder). The reaction rate measurements include the Li 2 CO 3 pellets for the tritium breeding monitoring and activation foils, inserted at several axial and lateral locations in the block. In addition to the well established and validated procedure based on the 2-dimensional (2D) code DORT, a new approach for the 3D modelling was validated based on the TORT/GRTUNCL3D transport codes. The SUSD3D code, also in 3D geometry, was used for the cross-section sensitivity and uncertainty calculations. These studies are useful for the interpretation of the experimental measurements, in particular to assess the uncertainties linked to the basic nuclear data. The TPR, the neutron activation rates and the associated uncertainties were determined using the EFF-3.0 9 Be nuclear cross section and covariance data, and compared with those from other evaluations, like FENDL-2.1. Sensitivity profiles and nuclear data uncertainties of the TPR and detector reaction rates with respect to the cross-sections of 9 Be, 6 Li, 7 Li, O and C were determined at different positions in the experimental block. (author)

Developments based on stochastic and determinist methods for studying complex nuclear systems; Developpements utilisant des methodes stochastiques et deterministes pour l'analyse de systemes nucleaires complexes

Energy Technology Data Exchange (ETDEWEB)

Giffard, F X

2000-05-19

In the field of reactor and fuel cycle physics, particle transport plays and important role. Neutronic design, operation and evaluation calculations of nuclear system make use of large and powerful computer codes. However, current limitations in terms of computer resources make it necessary to introduce simplifications and approximations in order to keep calculation time and cost within reasonable limits. Two different types of methods are available in these codes. The first one is the deterministic method, which is applicable in most practical cases but requires approximations. The other method is the Monte Carlo method, which does not make these approximations but which generally requires exceedingly long running times. The main motivation of this work is to investigate the possibility of a combined use of the two methods in such a way as to retain their advantages while avoiding their drawbacks. Our work has mainly focused on the speed-up of 3-D continuous energy Monte Carlo calculations (TRIPOLI-4 code) by means of an optimized biasing scheme derived from importance maps obtained from the deterministic code ERANOS. The application of this method to two different practical shielding-type problems has demonstrated its efficiency: speed-up factors of 100 have been reached. In addition, the method offers the advantage of being easily implemented as it is not very to the choice of the importance mesh grid. It has also been demonstrated that significant speed-ups can be achieved by this method in the case of coupled neutron-gamma transport problems, provided that the interdependence of the neutron and photon importance maps is taken into account. Complementary studies are necessary to tackle a problem brought out by this work, namely undesirable jumps in the Monte Carlo variance estimates. (author)

A Comparison of Monte Carlo and Deterministic Solvers for keff and Sensitivity Calculations

Energy Technology Data Exchange (ETDEWEB)

Haeck, Wim [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parsons, Donald Kent [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); White, Morgan Curtis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Saller, Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Favorite, Jeffrey A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-12-12

Verification and validation of our solutions for calculating the neutron reactivity for nuclear materials is a key issue to address for many applications, including criticality safety, research reactors, power reactors, and nuclear security. Neutronics codes solve variations of the Boltzmann transport equation. The two main variants are Monte Carlo versus deterministic solutions, e.g. the MCNP [1] versus PARTISN [2] codes, respectively. There have been many studies over the decades that examined the accuracy of such solvers and the general conclusion is that when the problems are well-posed, either solver can produce accurate results. However, the devil is always in the details. The current study examines the issue of self-shielding and the stress it puts on deterministic solvers. Most Monte Carlo neutronics codes use continuous-energy descriptions of the neutron interaction data that are not subject to this effect. The issue of self-shielding occurs because of the discretisation of data used by the deterministic solutions. Multigroup data used in these solvers are the average cross section and scattering parameters over an energy range. Resonances in cross sections can occur that change the likelihood of interaction by one to three orders of magnitude over a small energy range. Self-shielding is the numerical effect that the average cross section in groups with strong resonances can be strongly affected as neutrons within that material are preferentially absorbed or scattered out of the resonance energies. This affects both the average cross section and the scattering matrix.

Experimental aspects of deterministic secure quantum key distribution

Energy Technology Data Exchange (ETDEWEB)

Walenta, Nino; Korn, Dietmar; Puhlmann, Dirk; Felbinger, Timo; Hoffmann, Holger; Ostermeyer, Martin [Universitaet Potsdam (Germany). Institut fuer Physik; Bostroem, Kim [Universitaet Muenster (Germany)

2008-07-01

Most common protocols for quantum key distribution (QKD) use non-deterministic algorithms to establish a shared key. But deterministic implementations can allow for higher net key transfer rates and eavesdropping detection rates. The Ping-Pong coding scheme by Bostroem and Felbinger[1] employs deterministic information encoding in entangled states with its characteristic quantum channel from Bob to Alice and back to Bob. Based on a table-top implementation of this protocol with polarization-entangled photons fundamental advantages as well as practical issues like transmission losses, photon storage and requirements for progress towards longer transmission distances are discussed and compared to non-deterministic protocols. Modifications of common protocols towards a deterministic quantum key distribution are addressed.

Analytical three-dimensional neutron transport benchmarks for verification of nuclear engineering codes. Final report

International Nuclear Information System (INIS)

Ganapol, B.D.; Kornreich, D.E.

1997-01-01

Because of the requirement of accountability and quality control in the scientific world, a demand for high-quality analytical benchmark calculations has arisen in the neutron transport community. The intent of these benchmarks is to provide a numerical standard to which production neutron transport codes may be compared in order to verify proper operation. The overall investigation as modified in the second year renewal application includes the following three primary tasks. Task 1 on two dimensional neutron transport is divided into (a) single medium searchlight problem (SLP) and (b) two-adjacent half-space SLP. Task 2 on three-dimensional neutron transport covers (a) point source in arbitrary geometry, (b) single medium SLP, and (c) two-adjacent half-space SLP. Task 3 on code verification, includes deterministic and probabilistic codes. The primary aim of the proposed investigation was to provide a suite of comprehensive two- and three-dimensional analytical benchmarks for neutron transport theory applications. This objective has been achieved. The suite of benchmarks in infinite media and the three-dimensional SLP are a relatively comprehensive set of one-group benchmarks for isotropically scattering media. Because of time and resource limitations, the extensions of the benchmarks to include multi-group and anisotropic scattering are not included here. Presently, however, enormous advances in the solution for the planar Green's function in an anisotropically scattering medium have been made and will eventually be implemented in the two- and three-dimensional solutions considered under this grant. Of particular note in this work are the numerical results for the three-dimensional SLP, which have never before been presented. The results presented were made possible only because of the tremendous advances in computing power that have occurred during the past decade

A study of physics of sub-critical multiplicative systems driven by sources and the utilization of deterministic codes in calculation of this systems

International Nuclear Information System (INIS)

Antunes, Alberi

2008-01-01

This work presents the Physics of Source Driven Systems (ADS). It shows some statics and K i netics parameters of the reactor Physics and when it is sub critical, that are important in evaluation and definition of these systems. The objective is to demonstrate that there are differences in parameters when the reactor is critical. Moreover, the work shows the differences observed in the parameters for different calculation models. Two calculation methodologies are shown In this dissertation: Gandini and Salvatores and Dulla, and some parameters are calculated. The ANISN deterministic transport code is used in calculation in order to compare these parameters. In a subcritical configuration of IPEN-MB-01 Reactor driven by an external source some parameters are calculated. The conclusions about calculation realized are presented in end of work. (author)

Developments based on stochastic and determinist methods for studying complex nuclear systems; Developpements utilisant des methodes stochastiques et deterministes pour l'analyse de systemes nucleaires complexes

Energy Technology Data Exchange (ETDEWEB)

Giffard, F.X

2000-05-19

In the field of reactor and fuel cycle physics, particle transport plays and important role. Neutronic design, operation and evaluation calculations of nuclear system make use of large and powerful computer codes. However, current limitations in terms of computer resources make it necessary to introduce simplifications and approximations in order to keep calculation time and cost within reasonable limits. Two different types of methods are available in these codes. The first one is the deterministic method, which is applicable in most practical cases but requires approximations. The other method is the Monte Carlo method, which does not make these approximations but which generally requires exceedingly long running times. The main motivation of this work is to investigate the possibility of a combined use of the two methods in such a way as to retain their advantages while avoiding their drawbacks. Our work has mainly focused on the speed-up of 3-D continuous energy Monte Carlo calculations (TRIPOLI-4 code) by means of an optimized biasing scheme derived from importance maps obtained from the deterministic code ERANOS. The application of this method to two different practical shielding-type problems has demonstrated its efficiency: speed-up factors of 100 have been reached. In addition, the method offers the advantage of being easily implemented as it is not very to the choice of the importance mesh grid. It has also been demonstrated that significant speed-ups can be achieved by this method in the case of coupled neutron-gamma transport problems, provided that the interdependence of the neutron and photon importance maps is taken into account. Complementary studies are necessary to tackle a problem brought out by this work, namely undesirable jumps in the Monte Carlo variance estimates. (author)

DANTSYS: a system for deterministic, neutral particle transport calculations

Energy Technology Data Exchange (ETDEWEB)

Alcouffe, R.E.; Baker, R.S.

1996-12-31

The THREEDANT code is the latest addition to our system of codes, DANTSYS, which perform neutral particle transport computations on a given system of interest. The system of codes is distinguished by geometrical or symmetry considerations. For example, ONEDANT and TWODANT are designed for one and two dimensional geometries respectively. We have TWOHEX for hexagonal geometries, TWODANT/GQ for arbitrary quadrilaterals in XY and RZ geometry, and THREEDANT for three-dimensional geometries. The design of this system of codes is such that they share the same input and edit module and hence the input and output is uniform for all the codes (with the obvious additions needed to specify each type of geometry). The codes in this system are also designed to be general purpose solving both eigenvalue and source driven problems. In this paper we concentrate on the THREEDANT module since there are special considerations that need to be taken into account when designing such a module. The main issues that need to be addressed in a three-dimensional transport solver are those of the computational time needed to solve a problem and the amount of storage needed to accomplish that solution. Of course both these issues are directly related to the number of spatial mesh cells required to obtain a solution to a specified accuracy, but is also related to the spatial discretization method chosen and the requirements of the iteration acceleration scheme employed as will be noted below. Another related consideration is the robustness of the resulting algorithms as implemented; because insistence on complete robustness has a significant impact upon the computation time. We address each of these issues in the following through which we give reasons for the choices we have made in our approach to this code. And this is useful in outlining how the code is evolving to better address the shortcomings that presently exist.

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

Parallelization of a three-dimensional whole core transport code DeCART

Energy Technology Data Exchange (ETDEWEB)

Jin Young, Cho; Han Gyu, Joo; Ha Yong, Kim; Moon-Hee, Chang [Korea Atomic Energy Research Institute, Yuseong-gu, Daejon (Korea, Republic of)

2003-07-01

Parallelization of the DeCART (deterministic core analysis based on ray tracing) code is presented that reduces the computational burden of the tremendous computing time and memory required in three-dimensional whole core transport calculations. The parallelization employs the concept of MPI grouping and the MPI/OpenMP mixed scheme as well. Since most of the computing time and memory are used in MOC (method of characteristics) and the multi-group CMFD (coarse mesh finite difference) calculation in DeCART, variables and subroutines related to these two modules are the primary targets for parallelization. Specifically, the ray tracing module was parallelized using a planar domain decomposition scheme and an angular domain decomposition scheme. The parallel performance of the DeCART code is evaluated by solving a rodded variation of the C5G7MOX three dimensional benchmark problem and a simplified three-dimensional SMART PWR core problem. In C5G7MOX problem with 24 CPUs, a speedup of maximum 21 is obtained on an IBM Regatta machine and 22 on a LINUX Cluster in the MOC kernel, which indicates good parallel performance of the DeCART code. In the simplified SMART problem, the memory requirement of about 11 GBytes in the single processor cases reduces to 940 Mbytes with 24 processors, which means that the DeCART code can now solve large core problems with affordable LINUX clusters. (authors)

On the use of SERPENT Monte Carlo code to generate few group diffusion constants

Energy Technology Data Exchange (ETDEWEB)

Piovezan, Pamela, E-mail: pamela.piovezan@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Carluccio, Thiago; Domingos, Douglas Borges; Rossi, Pedro Russo; Mura, Luiz Felipe, E-mail: fermium@cietec.org.b, E-mail: thiagoc@ipen.b [Fermium Tecnologia Nuclear, Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

The accuracy of diffusion reactor codes strongly depends on the quality of the groups constants processing. For many years, the generation of such constants was based on 1-D infinity cell transport calculations. Some developments using collision probability or the method of characteristics allow, nowadays, 2-D assembly group constants calculations. However, these 1-D and 2-D codes how some limitations as , for example, on complex geometries and in the neighborhood of heavy absorbers. On the other hand, since Monte Carlos (MC) codes provide accurate neutro flux distributions, the possibility of using these solutions to provide group constants to full-core reactor diffusion simulators has been recently investigated, especially for the cases in which the geometry and reactor types are beyond the capability of the conventional deterministic lattice codes. The two greatest difficulties on the use of MC codes to group constant generation are the computational costs and the methodological incompatibility between analog MC particle transport simulation and deterministic transport methods based in several approximations. The SERPENT code is a 3-D continuous energy MC transport code with built-in burnup capability that was specially optimized to generate these group constants. In this work, we present the preliminary results of using the SERPENT MC code to generate 3-D two-group diffusion constants for a PWR like assembly. These constants were used in the CITATION diffusion code to investigate the effects of the MC group constants determination on the neutron multiplication factor diffusion estimate. (author)

Developments based on stochastic and determinist methods for studying complex nuclear systems

International Nuclear Information System (INIS)

Giffard, F.X.

2000-01-01

In the field of reactor and fuel cycle physics, particle transport plays and important role. Neutronic design, operation and evaluation calculations of nuclear system make use of large and powerful computer codes. However, current limitations in terms of computer resources make it necessary to introduce simplifications and approximations in order to keep calculation time and cost within reasonable limits. Two different types of methods are available in these codes. The first one is the deterministic method, which is applicable in most practical cases but requires approximations. The other method is the Monte Carlo method, which does not make these approximations but which generally requires exceedingly long running times. The main motivation of this work is to investigate the possibility of a combined use of the two methods in such a way as to retain their advantages while avoiding their drawbacks. Our work has mainly focused on the speed-up of 3-D continuous energy Monte Carlo calculations (TRIPOLI-4 code) by means of an optimized biasing scheme derived from importance maps obtained from the deterministic code ERANOS. The application of this method to two different practical shielding-type problems has demonstrated its efficiency: speed-up factors of 100 have been reached. In addition, the method offers the advantage of being easily implemented as it is not very to the choice of the importance mesh grid. It has also been demonstrated that significant speed-ups can be achieved by this method in the case of coupled neutron-gamma transport problems, provided that the interdependence of the neutron and photon importance maps is taken into account. Complementary studies are necessary to tackle a problem brought out by this work, namely undesirable jumps in the Monte Carlo variance estimates. (author)

Radiation transport calculation methods in BNCT

International Nuclear Information System (INIS)

Koivunoro, H.; Seppaelae, T.; Savolainen, S.

2000-01-01

Boron neutron capture therapy (BNCT) is used as a radiotherapy for malignant brain tumours. Radiation dose distribution is necessary to determine individually for each patient. Radiation transport and dose distribution calculations in BNCT are more complicated than in conventional radiotherapy. Total dose in BNCT consists of several different dose components. The most important dose component for tumour control is therapeutic boron dose D B . The other dose components are gamma dose D g , incident fast neutron dose D f ast n and nitrogen dose D N . Total dose is a weighted sum of the dose components. Calculation of neutron and photon flux is a complex problem and requires numerical methods, i.e. deterministic or stochastic simulation methods. Deterministic methods are based on the numerical solution of Boltzmann transport equation. Such are discrete ordinates (SN) and spherical harmonics (PN) methods. The stochastic simulation method for calculation of radiation transport is known as Monte Carlo method. In the deterministic methods the spatial geometry is partitioned into mesh elements. In SN method angular integrals of the transport equation are replaced with weighted sums over a set of discrete angular directions. Flux is calculated iteratively for all these mesh elements and for each discrete direction. Discrete ordinates transport codes used in the dosimetric calculations are ANISN, DORT and TORT. In PN method a Legendre expansion for angular flux is used instead of discrete direction fluxes, land the angular dependency comes a property of vector function space itself. Thus, only spatial iterations are required for resulting equations. A novel radiation transport code based on PN method and tree-multigrid technique (TMG) has been developed at VTT (Technical Research Centre of Finland). Monte Carlo method solves the radiation transport by randomly selecting neutrons and photons from a prespecified boundary source and following the histories of selected particles

High Energy Transport Code HETC

International Nuclear Information System (INIS)

Gabriel, T.A.

1985-09-01

The physics contained in the High Energy Transport Code (HETC), in particular the collision models, are discussed. An application using HETC as part of the CALOR code system is also given. 19 refs., 5 figs., 3 tabs

Recent developments in discrete ordinates electron transport

International Nuclear Information System (INIS)

Morel, J.E.; Lorence, L.J. Jr.

1986-01-01

The discrete ordinates method is a deterministic method for numerically solving the Boltzmann equation. It was originally developed for neutron transport calculations, but is routinely used for photon and coupled neutron-photon transport calculations as well. The computational state of the art for coupled electron-photon transport (CEPT) calculations is not as developed as that for neutron transport calculations. The only production codes currently available for CEPT calculations are condensed-history Monte Carlo codes such as the ETRAN and ITS codes. A deterministic capability for production calculations is clearly needed. In response to this need, we have begun the development of a production discrete ordinates code for CEPT calculations. The purpose of this paper is to describe the basic approach we are taking, discuss the current status of the project, and present some new computational results. Although further characterization of the coupled electron-photon discrete ordinates method remains to be done, the results to date indicate that the discrete ordinates method can be just as accurate and from 10 to 100 times faster than the Monte Carlo method for a wide variety of problems. We stress that these results are obtained with standard discrete ordinates codes such as ONETRAN. It is clear that even greater efficiency can be obtained by developing a new generation of production discrete ordinates codes specifically designed to solve the Boltzmann-Fokker-Planck equation. However, the prospects for such development in the near future appear to be remote

Deterministic models for energy-loss straggling

International Nuclear Information System (INIS)

Prinja, A.K.; Gleicher, F.; Dunham, G.; Morel, J.E.

1999-01-01

Inelastic ion interactions with target electrons are dominated by extremely small energy transfers that are difficult to resolve numerically. The continuous-slowing-down (CSD) approximation is then commonly employed, which, however, only preserves the mean energy loss per collision through the stopping power, S(E) = ∫ 0 ∞ dEprime (E minus Eprime) σ s (E → Eprime). To accommodate energy loss straggling, a Gaussian distribution with the correct mean-squared energy loss (akin to a Fokker-Planck approximation in energy) is commonly used in continuous-energy Monte Carlo codes. Although this model has the unphysical feature that ions can be upscattered, it nevertheless yields accurate results. A multigroup model for energy loss straggling was recently presented for use in multigroup Monte Carlo codes or in deterministic codes that use multigroup data. The method has the advantage that the mean and mean-squared energy loss are preserved without unphysical upscatter and hence is computationally efficient. Results for energy spectra compared extremely well with Gaussian distributions under the idealized conditions for which the Gaussian may be considered to be exact. Here, the authors present more consistent comparisons by extending the method to accommodate upscatter and, further, compare both methods with exact solutions obtained from an analog Monte Carlo simulation, for a straight-ahead transport problem

Comparison of Non-overlapping and Overlapping Local/Global Iteration Schemes for Whole-Core Deterministic Transport Calculation

International Nuclear Information System (INIS)

Yuk, Seung Su; Cho, Bumhee; Cho, Nam Zin

2013-01-01

In the case of deterministic transport model, fixed-k problem formulation is necessary and the overlapping local domain is chosen. However, as mentioned in, the partial current-based Coarse Mesh Finite Difference (p-CMFD) procedure enables also non-overlapping local/global (NLG) iteration. In this paper, NLG iteration is combined with p-CMFD and with CMFD (augmented with a concept of p-CMFD), respectively, and compared to OLG iteration on a 2-D test problem. Non-overlapping local/global iteration with p-CMFD and CMFD global calculation is introduced and tested on a 2-D deterministic transport problem. The modified C5G7 problem is analyzed with both NLG and OLG methods and the solutions converge to the reference solution except for some cases of NLG with CMFD. NLG with CMFD gives the best performance if the solution converges. But if fission-source iteration in local calculation is not enough, it is prone to diverge. The p-CMFD global solver gives unconditional convergence (for both OLG and NLG). A study of switching scheme is in progress, where NLG/p-CMFD is used as 'starter' and then switched to NLG/CMFD to render the whole-core transport calculation more efficient and robust. Parallel computation is another obvious future work

Transport theory and codes

International Nuclear Information System (INIS)

Clancy, B.E.

1986-01-01

This chapter begins with a neutron transport equation which includes the one dimensional plane geometry problems, the one dimensional spherical geometry problems, and numerical solutions. The section on the ANISN code and its look-alikes covers problems which can be solved; eigenvalue problems; outer iteration loop; inner iteration loop; and finite difference solution procedures. The input and output data for ANISN is also discussed. Two dimensional problems such as the DOT code are given. Finally, an overview of the Monte-Carlo methods and codes are elaborated on

Conception and development of an adaptive energy mesher for multigroup library generation of the transport codes

International Nuclear Information System (INIS)

Mosca, P.

2009-12-01

The deterministic transport codes solve the stationary Boltzmann equation in a discretized energy formalism called multigroup. The transformation of continuous data in a multigroup form is obtained by averaging the highly variable cross sections of the resonant isotopes with the solution of the self-shielding models and the remaining ones with the coarse energy spectrum of the reactor type. So far the error of such an approach could only be evaluated retrospectively. To remedy this, we studied in this thesis a set of methods to control a priori the accuracy and the cost of the multigroup transport computation. The energy mesh optimisation is achieved using a two step process: the creation of a reference mesh and its optimized condensation. In the first stage, by refining locally and globally the energy mesh, we seek, on a fine energy mesh with subgroup self-shielding, a solution equivalent to a reference solver (Monte Carlo or pointwise deterministic solver). In the second step, once fixed the number of groups, depending on the acceptable computational cost, and chosen the most appropriate self-shielding models to the reactor type, we look for the best bounds of the reference mesh minimizing reaction rate errors by the particle swarm optimization algorithm. This new approach allows us to define new meshes for fast reactors as accurate as the currently used ones, but with fewer groups. (author)

Feasibility Study of Core Design with a Monte Carlo Code for APR1400 Initial core

Energy Technology Data Exchange (ETDEWEB)

Kim, Jinsun; Chang, Do Ik; Seong, Kibong [KEPCO NF, Daejeon (Korea, Republic of)

2014-10-15

The Monte Carlo calculation becomes more popular and useful nowadays due to the rapid progress in computing power and parallel calculation techniques. There have been many attempts to analyze a commercial core by Monte Carlo transport code using the enhanced computer capability, recently. In this paper, Monte Carlo calculation of APR1400 initial core has been performed and the results are compared with the calculation results of conventional deterministic code to find out the feasibility of core design using Monte Carlo code. SERPENT, a 3D continuous-energy Monte Carlo reactor physics burnup calculation code is used for this purpose and the KARMA-ASTRA code system, which is used for a deterministic code of comparison. The preliminary investigation for the feasibility of commercial core design with Monte Carlo code was performed in this study. Simplified core geometry modeling was performed for the reactor core surroundings and reactor coolant model is based on two region model. The reactivity difference at HZP ARO condition between Monte Carlo code and the deterministic code is consistent with each other and the reactivity difference during the depletion could be reduced by adopting the realistic moderator temperature. The reactivity difference calculated at HFP, BOC, ARO equilibrium condition was 180 ±9 pcm, with axial moderator temperature of a deterministic code. The computing time will be a significant burden at this time for the application of Monte Carlo code to the commercial core design even with the application of parallel computing because numerous core simulations are required for actual loading pattern search. One of the remedy will be a combination of Monte Carlo code and the deterministic code to generate the physics data. The comparison of physics parameters with sophisticated moderator temperature modeling and depletion will be performed for a further study.

Spent Fuel Pool Dose Rate Calculations Using Point Kernel and Hybrid Deterministic-Stochastic Shielding Methods

International Nuclear Information System (INIS)

Matijevic, M.; Grgic, D.; Jecmenica, R.

2016-01-01

This paper presents comparison of the Krsko Power Plant simplified Spent Fuel Pool (SFP) dose rates using different computational shielding methodologies. The analysis was performed to estimate limiting gamma dose rates on wall mounted level instrumentation in case of significant loss of cooling water. The SFP was represented with simple homogenized cylinders (point kernel and Monte Carlo (MC)) or cuboids (MC) using uranium, iron, water, and dry-air as bulk region materials. The pool is divided on the old and new section where the old one has three additional subsections representing fuel assemblies (FAs) with different burnup/cooling time (60 days, 1 year and 5 years). The new section represents the FAs with the cooling time of 10 years. The time dependent fuel assembly isotopic composition was calculated using ORIGEN2 code applied to the depletion of one of the fuel assemblies present in the pool (AC-29). The source used in Microshield calculation is based on imported isotopic activities. The time dependent photon spectra with total source intensity from Microshield multigroup point kernel calculations was then prepared for two hybrid deterministic-stochastic sequences. One is based on SCALE/MAVRIC (Monaco and Denovo) methodology and another uses Monte Carlo code MCNP6.1.1b and ADVANTG3.0.1. code. Even though this model is a fairly simple one, the layers of shielding materials are thick enough to pose a significant shielding problem for MC method without the use of effective variance reduction (VR) technique. For that purpose the ADVANTG code was used to generate VR parameters (SB cards in SDEF and WWINP file) for MCNP fixed-source calculation using continuous energy transport. ADVATNG employs a deterministic forward-adjoint transport solver Denovo which implements CADIS/FW-CADIS methodology. Denovo implements a structured, Cartesian-grid SN solver based on the Koch-Baker-Alcouffe parallel transport sweep algorithm across x-y domain blocks. This was first

Efficiency of transport in periodic potentials: dichotomous noise contra deterministic force

Science.gov (United States)

Spiechowicz, J.; Łuczka, J.; Machura, L.

2016-05-01

We study the transport of an inertial Brownian particle moving in a symmetric and periodic one-dimensional potential, and subjected to both a symmetric, unbiased external harmonic force as well as biased dichotomic noise η (t) also known as a random telegraph signal or a two state continuous-time Markov process. In doing so, we concentrate on the previously reported regime (Spiechowicz et al 2014 Phys. Rev. E 90 032104) for which non-negative biased noise η (t) in the form of generalized white Poissonian noise can induce anomalous transport processes similar to those generated by a deterministic constant force F= but significantly more effective than F, i.e. the particle moves much faster, the velocity fluctuations are noticeably reduced and the transport efficiency is enhanced several times. Here, we confirm this result for the case of dichotomous fluctuations which, in contrast to white Poissonian noise, can assume positive as well as negative values and examine the role of thermal noise in the observed phenomenon. We focus our attention on the impact of bidirectionality of dichotomous fluctuations and reveal that the effect of nonequilibrium noise enhanced efficiency is still detectable. This result may explain transport phenomena occurring in strongly fluctuating environments of both physical and biological origin. Our predictions can be corroborated experimentally by use of a setup that consists of a resistively and capacitively shunted Josephson junction.

CTCN: Colloid transport code -- nuclear

International Nuclear Information System (INIS)

Jain, R.

1993-01-01

This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential-algebraic equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential-algebraic systems

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

Coupled geochemical and solute transport code development

International Nuclear Information System (INIS)

Morrey, J.R.; Hostetler, C.J.

1985-01-01

A number of coupled geochemical hydrologic codes have been reported in the literature. Some of these codes have directly coupled the source-sink term to the solute transport equation. The current consensus seems to be that directly coupling hydrologic transport and chemical models through a series of interdependent differential equations is not feasible for multicomponent problems with complex geochemical processes (e.g., precipitation/dissolution reactions). A two-step process appears to be the required method of coupling codes for problems where a large suite of chemical reactions must be monitored. Two-step structure requires that the source-sink term in the transport equation is supplied by a geochemical code rather than by an analytical expression. We have developed a one-dimensional two-step coupled model designed to calculate relatively complex geochemical equilibria (CTM1D). Our geochemical module implements a Newton-Raphson algorithm to solve heterogeneous geochemical equilibria, involving up to 40 chemical components and 400 aqueous species. The geochemical module was designed to be efficient and compact. A revised version of the MINTEQ Code is used as a parent geochemical code

Faster Heavy Ion Transport for HZETRN

Science.gov (United States)

Slaba, Tony C.

2013-01-01

The deterministic particle transport code HZETRN was developed to enable fast and accurate space radiation transport through materials. As more complex transport solutions are implemented for neutrons, light ions (Z heavy ion (Z > 2) transport algorithm in HZETRN is reviewed, and a simple modification is shown to provide an approximate 5x decrease in execution time for galactic cosmic ray transport. Convergence tests and other comparisons are carried out to verify that numerical accuracy is maintained in the new algorithm.

Biomedical applications of two- and three-dimensional deterministic radiation transport methods

International Nuclear Information System (INIS)

Nigg, D.W.

1992-01-01

Multidimensional deterministic radiation transport methods are routinely used in support of the Boron Neutron Capture Therapy (BNCT) Program at the Idaho National Engineering Laboratory (INEL). Typical applications of two-dimensional discrete-ordinates methods include neutron filter design, as well as phantom dosimetry. The epithermal-neutron filter for BNCT that is currently available at the Brookhaven Medical Research Reactor (BMRR) was designed using such methods. Good agreement between calculated and measured neutron fluxes was observed for this filter. Three-dimensional discrete-ordinates calculations are used routinely for dose-distribution calculations in three-dimensional phantoms placed in the BMRR beam, as well as for treatment planning verification for live canine subjects. Again, good agreement between calculated and measured neutron fluxes and dose levels is obtained

Current status of high energy nucleon-meson transport code

Energy Technology Data Exchange (ETDEWEB)

Takada, Hiroshi; Sasa, Toshinobu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

Current status of design code of accelerator (NMTC/JAERI code), outline of physical model and evaluation of accuracy of code were reported. To evaluate the nuclear performance of accelerator and strong spallation neutron origin, the nuclear reaction between high energy proton and target nuclide and behaviors of various produced particles are necessary. The nuclear design of spallation neutron system used a calculation code system connected the high energy nucleon{center_dot}meson transport code and the neutron{center_dot}photon transport code. NMTC/JAERI is described by the particle evaporation process under consideration of competition reaction of intranuclear cascade and fission process. Particle transport calculation was carried out for proton, neutron, {pi}- and {mu}-meson. To verify and improve accuracy of high energy nucleon-meson transport code, data of spallation and spallation neutron fragment by the integral experiment were collected. (S.Y.)

RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

International Nuclear Information System (INIS)

Salim, Mohd Faiz; Roslan, Ridha; Ibrahim, Mohd Rizal Mamat

2014-01-01

Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges

RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

Energy Technology Data Exchange (ETDEWEB)

Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my [Nuclear Energy Department, Tenaga Nasional Berhad, Level 32, Dua Sentral, 50470 Kuala Lumpur (Malaysia); Roslan, Ridha [Nuclear Installation Division, Atomic Energy Licensing Board, Batu 24, Jalan Dengkil, 43800 Dengkil, Selangor (Malaysia); Ibrahim, Mohd Rizal Mamat [Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

2014-02-12

Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges.

Experimental transport analysis code system in JT-60

International Nuclear Information System (INIS)

Hirayama, Toshio; Shimizu, Katsuhiro; Tani, Keiji; Shirai, Hiroshi; Kikuchi, Mitsuru

1988-03-01

Transport analysis codes have been developed in order to study confinement properties related to particle and energy balance in ohmically and neutral beam heated plasmas of JT-60. The analysis procedure is divided into three steps as follows: 1) LOOK ; The shape of the plasma boundary is identified with a fast boundary identification code of FBI by using magnetic data, and flux surfaces are calculated with a MHD equilibrium code of SELENE. The diagnostic data are mapped to flux surfaces for neutral beam heating calculation and/or for radial transport analysis. 2) OFMC ; On the basis of transformed data, an orbit following Monte Carlo code of OFMC calculates both profiles of power deposition and particle source of neutral beam injected into a plasma. 3) SCOOP ; In the last stage, a one dimensional transport code of SCOOP solves particle and energy balance for electron and ion, in order to evaluate transport coefficients as well as global parameters such as energy confinement time and the stored energy. The analysis results are provided to a data bank of DARTS that is used to find an overview of important consideration on confinement with a regression analysis code of RAC. (author)

MVP utilization for PWR design code

International Nuclear Information System (INIS)

Matsumoto, Hideki; Tahara, Yoshihisa

2001-01-01

MHI studies the method of the spatially dependent resonance cross sections so as to predict the power distribution in a fuel pellet accurately. For this purpose, the multiband method and the Stoker/Weiss method were implemented to the 2 dimensional transport code PHOENIX-P, and the methods were validated by comparing them with MVP code. Although the appropriate reference was not obtain from the deterministic codes on the resonance cross section study, now the Monte Carlo code MVP result is available and useful as reference. It is shown here how MVP is used to develop the multiband method and the Stoker/Weiss method, and how effective the result of MVP is on the study of the resonance cross sections. (author)

Los Alamos radiation transport code system on desktop computing platforms

International Nuclear Information System (INIS)

Briesmeister, J.F.; Brinkley, F.W.; Clark, B.A.; West, J.T.

1990-01-01

The Los Alamos Radiation Transport Code System (LARTCS) consists of state-of-the-art Monte Carlo and discrete ordinates transport codes and data libraries. These codes were originally developed many years ago and have undergone continual improvement. With a large initial effort and continued vigilance, the codes are easily portable from one type of hardware to another. The performance of scientific work-stations (SWS) has evolved to the point that such platforms can be used routinely to perform sophisticated radiation transport calculations. As the personal computer (PC) performance approaches that of the SWS, the hardware options for desk-top radiation transport calculations expands considerably. The current status of the radiation transport codes within the LARTCS is described: MCNP, SABRINA, LAHET, ONEDANT, TWODANT, TWOHEX, and ONELD. Specifically, the authors discuss hardware systems on which the codes run and present code performance comparisons for various machines

Solution to the transport equation with anisotropic dispersion in a BWR type assembly using the AZTRAN code; Solucion de la ecuacion de transporte con dispersion anisotropica en un ensamble tipo BWR usando el codigo AZTRAN

Energy Technology Data Exchange (ETDEWEB)

Chepe P, M. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: liaison.web@gmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. San Pedro Zacatenco, 07730 Ciudad de Mexico (Mexico)

2016-09-15

Due to the current computing power, the deterministic codes for analyzing nuclear reactors that have been used for several years are becoming more relevant, since much more precise solution techniques can be used; the last century would have been very difficult, since memory and processor capacities were very limited or had high prices on the components. In this work we analyze the effect of the anisotropic dispersion of the effective dispersion section, compared to the isotropic dispersion. The anisotropy implementation was carried out in the AZTRAN transport code, which is part of the AZTLAN platform for nuclear reactors analysis (in development). The AZTRAN code solves the Boltzmann transport equation in one, two and three dimensions at steady state, using the multi-group technique for energy discretization, the RTN-0 nodal method in spatial discretization and for angular discretization the discrete ordinates without considering anisotropy originally. The effect of the anisotropy dispersion on the effective multiplication factor and the axial and radial power on a fuel assembly BWR type are analyzed. (Author)

RDS; A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

International Nuclear Information System (INIS)

Mohd Faiz Salim; Ridha Roslan; Mohd Rizal Mamat

2013-01-01

Full-text: Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBIMOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges. (author)

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

Transmutation approximations for the application of hybrid Monte Carlo/deterministic neutron transport to shutdown dose rate analysis

International Nuclear Information System (INIS)

Biondo, Elliott D.; Wilson, Paul P. H.

2017-01-01

In fusion energy systems (FES) neutrons born from burning plasma activate system components. The photon dose rate after shutdown from resulting radionuclides must be quantified. This shutdown dose rate (SDR) is calculated by coupling neutron transport, activation analysis, and photon transport. The size, complexity, and attenuating configuration of FES motivate the use of hybrid Monte Carlo (MC)/deterministic neutron transport. The Multi-Step Consistent Adjoint Driven Importance Sampling (MS-CADIS) method can be used to optimize MC neutron transport for coupled multiphysics problems, including SDR analysis, using deterministic estimates of adjoint flux distributions. When used for SDR analysis, MS-CADIS requires the formulation of an adjoint neutron source that approximates the transmutation process. In this work, transmutation approximations are used to derive a solution for this adjoint neutron source. It is shown that these approximations are reasonably met for typical FES neutron spectra and materials over a range of irradiation scenarios. When these approximations are met, the Groupwise Transmutation (GT)-CADIS method, proposed here, can be used effectively. GT-CADIS is an implementation of the MS-CADIS method for SDR analysis that uses a series of single-energy-group irradiations to calculate the adjoint neutron source. For a simple SDR problem, GT-CADIS provides speedups of 200 100 relative to global variance reduction with the Forward-Weighted (FW)-CADIS method and 9 _± 5 • _1_0_"_4 relative to analog. As a result, this work shows that GT-CADIS is broadly applicable to FES problems and will significantly reduce the computational resources necessary for SDR analysis.

Radiation transport phenomena and modeling - part A: Codes

International Nuclear Information System (INIS)

Lorence, L.J.

1997-01-01

The need to understand how particle radiation (high-energy photons and electrons) from a variety of sources affects materials and electronics has motivated the development of sophisticated computer codes that describe how radiation with energies from 1.0 keV to 100.0 GeV propagates through matter. Predicting radiation transport is the necessary first step in predicting radiation effects. The radiation transport codes that are described here are general-purpose codes capable of analyzing a variety of radiation environments including those produced by nuclear weapons (x-rays, gamma rays, and neutrons), by sources in space (electrons and ions) and by accelerators (x-rays, gamma rays, and electrons). Applications of these codes include the study of radiation effects on electronics, nuclear medicine (imaging and cancer treatment), and industrial processes (food disinfestation, waste sterilization, manufacturing.) The primary focus will be on coupled electron-photon transport codes, with some brief discussion of proton transport. These codes model a radiation cascade in which electrons produce photons and vice versa. This coupling between particles of different types is important for radiation effects. For instance, in an x-ray environment, electrons are produced that drive the response in electronics. In an electron environment, dose due to bremsstrahlung photons can be significant once the source electrons have been stopped

Colloid transport code-nuclear user's manual

International Nuclear Information System (INIS)

Jain, R.

1992-01-01

This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential systems

Benchmarking NNWSI flow and transport codes: COVE 1 results

International Nuclear Information System (INIS)

Hayden, N.K.

1985-06-01

The code verification (COVE) activity of the Nevada Nuclear Waste Storage Investigations (NNWSI) Project is the first step in certification of flow and transport codes used for NNWSI performance assessments of a geologic repository for disposing of high-level radioactive wastes. The goals of the COVE activity are (1) to demonstrate and compare the numerical accuracy and sensitivity of certain codes, (2) to identify and resolve problems in running typical NNWSI performance assessment calculations, and (3) to evaluate computer requirements for running the codes. This report describes the work done for COVE 1, the first step in benchmarking some of the codes. Isothermal calculations for the COVE 1 benchmarking have been completed using the hydrologic flow codes SAGUARO, TRUST, and GWVIP; the radionuclide transport codes FEMTRAN and TRUMP; and the coupled flow and transport code TRACR3D. This report presents the results of three cases of the benchmarking problem solved for COVE 1, a comparison of the results, questions raised regarding sensitivities to modeling techniques, and conclusions drawn regarding the status and numerical sensitivities of the codes. 30 refs

High performance 3D neutron transport on peta scale and hybrid architectures within APOLLO3 code

International Nuclear Information System (INIS)

Jamelot, E.; Dubois, J.; Lautard, J-J.; Calvin, C.; Baudron, A-M.

2011-01-01

APOLLO3 code is a common project of CEA, AREVA and EDF for the development of a new generation system for core physics analysis. We present here the parallelization of two deterministic transport solvers of APOLLO3: MINOS, a simplified 3D transport solver on structured Cartesian and hexagonal grids, and MINARET, a transport solver based on triangular meshes on 2D and prismatic ones in 3D. We used two different techniques to accelerate MINOS: a domain decomposition method, combined with an accelerated algorithm using GPU. The domain decomposition is based on the Schwarz iterative algorithm, with Robin boundary conditions to exchange information. The Robin parameters influence the convergence and we detail how we optimized the choice of these parameters. MINARET parallelization is based on angular directions calculation using explicit message passing. Fine grain parallelization is also available for each angular direction using shared memory multithreaded acceleration. Many performance results are presented on massively parallel architectures using more than 103 cores and on hybrid architectures using some tens of GPUs. This work contributes to the HPC development in reactor physics at the CEA Nuclear Energy Division. (author)

Deterministic Safety Analysis for Nuclear Power Plants. Specific Safety Guide (Russian Edition)

International Nuclear Information System (INIS)

2014-01-01

The objective of this Safety Guide is to provide harmonized guidance to designers, operators, regulators and providers of technical support on deterministic safety analysis for nuclear power plants. It provides information on the utilization of the results of such analysis for safety and reliability improvements. The Safety Guide addresses conservative, best estimate and uncertainty evaluation approaches to deterministic safety analysis and is applicable to current and future designs. Contents: 1. Introduction; 2. Grouping of initiating events and associated transients relating to plant states; 3. Deterministic safety analysis and acceptance criteria; 4. Conservative deterministic safety analysis; 5. Best estimate plus uncertainty analysis; 6. Verification and validation of computer codes; 7. Relation of deterministic safety analysis to engineering aspects of safety and probabilistic safety analysis; 8. Application of deterministic safety analysis; 9. Source term evaluation for operational states and accident conditions; References

Using MCBEND for neutron or gamma-ray deterministic calculations

Science.gov (United States)

Geoff, Dobson; Adam, Bird; Brendan, Tollit; Paul, Smith

2017-09-01

MCBEND 11 is the latest version of the general radiation transport Monte Carlo code from AMEC Foster Wheeler's ANSWERS® Software Service. MCBEND is well established in the UK shielding community for radiation shielding and dosimetry assessments. MCBEND supports a number of acceleration techniques, for example the use of an importance map in conjunction with Splitting/Russian Roulette. MCBEND has a well established automated tool to generate this importance map, commonly referred to as the MAGIC module using a diffusion adjoint solution. This method is fully integrated with the MCBEND geometry and material specification, and can easily be run as part of a normal MCBEND calculation. An often overlooked feature of MCBEND is the ability to use this method for forward scoping calculations, which can be run as a very quick deterministic method. Additionally, the development of the Visual Workshop environment for results display provides new capabilities for the use of the forward calculation as a productivity tool. In this paper, we illustrate the use of the combination of the old and new in order to provide an enhanced analysis capability. We also explore the use of more advanced deterministic methods for scoping calculations used in conjunction with MCBEND, with a view to providing a suite of methods to accompany the main Monte Carlo solver.

Interfacial and Wall Transport Models for SPACE-CAP Code

International Nuclear Information System (INIS)

Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul; Choi, Hoon; Ha, Sang Jun

2009-01-01

The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code

Interfacial and Wall Transport Models for SPACE-CAP Code

Energy Technology Data Exchange (ETDEWEB)

Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech., Seoul (Korea, Republic of); Choi, Hoon; Ha, Sang Jun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2009-10-15

The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code.

SU-G-TeP1-15: Toward a Novel GPU Accelerated Deterministic Solution to the Linear Boltzmann Transport Equation

Energy Technology Data Exchange (ETDEWEB)

Yang, R [University of Alberta, Edmonton, AB (Canada); Fallone, B [University of Alberta, Edmonton, AB (Canada); Cross Cancer Institute, Edmonton, AB (Canada); MagnetTx Oncology Solutions, Edmonton, AB (Canada); St Aubin, J [University of Alberta, Edmonton, AB (Canada); Cross Cancer Institute, Edmonton, AB (Canada)

2016-06-15

Purpose: To develop a Graphic Processor Unit (GPU) accelerated deterministic solution to the Linear Boltzmann Transport Equation (LBTE) for accurate dose calculations in radiotherapy (RT). A deterministic solution yields the potential for major speed improvements due to the sparse matrix-vector and vector-vector multiplications and would thus be of benefit to RT. Methods: In order to leverage the massively parallel architecture of GPUs, the first order LBTE was reformulated as a second order self-adjoint equation using the Least Squares Finite Element Method (LSFEM). This produces a symmetric positive-definite matrix which is efficiently solved using a parallelized conjugate gradient (CG) solver. The LSFEM formalism is applied in space, discrete ordinates is applied in angle, and the Multigroup method is applied in energy. The final linear system of equations produced is tightly coupled in space and angle. Our code written in CUDA-C was benchmarked on an Nvidia GeForce TITAN-X GPU against an Intel i7-6700K CPU. A spatial mesh of 30,950 tetrahedral elements was used with an S4 angular approximation. Results: To avoid repeating a full computationally intensive finite element matrix assembly at each Multigroup energy, a novel mapping algorithm was developed which minimized the operations required at each energy. Additionally, a parallelized memory mapping for the kronecker product between the sparse spatial and angular matrices, including Dirichlet boundary conditions, was created. Atomicity is preserved by graph-coloring overlapping nodes into separate kernel launches. The one-time mapping calculations for matrix assembly, kronecker product, and boundary condition application took 452±1ms on GPU. Matrix assembly for 16 energy groups took 556±3s on CPU, and 358±2ms on GPU using the mappings developed. The CG solver took 93±1s on CPU, and 468±2ms on GPU. Conclusion: Three computationally intensive subroutines in deterministically solving the LBTE have been

A massively parallel method of characteristic neutral particle transport code for GPUs

International Nuclear Information System (INIS)

Boyd, W. R.; Smith, K.; Forget, B.

2013-01-01

Over the past 20 years, parallel computing has enabled computers to grow ever larger and more powerful while scientific applications have advanced in sophistication and resolution. This trend is being challenged, however, as the power consumption for conventional parallel computing architectures has risen to unsustainable levels and memory limitations have come to dominate compute performance. Heterogeneous computing platforms, such as Graphics Processing Units (GPUs), are an increasingly popular paradigm for solving these issues. This paper explores the applicability of GPUs for deterministic neutron transport. A 2D method of characteristics (MOC) code - OpenMOC - has been developed with solvers for both shared memory multi-core platforms as well as GPUs. The multi-threading and memory locality methodologies for the GPU solver are presented. Performance results for the 2D C5G7 benchmark demonstrate 25-35 x speedup for MOC on the GPU. The lessons learned from this case study will provide the basis for further exploration of MOC on GPUs as well as design decisions for hardware vendors exploring technologies for the next generation of machines for scientific computing. (authors)

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.

Coupled electron-photon radiation transport

International Nuclear Information System (INIS)

Lorence, L.; Kensek, R.P.; Valdez, G.D.; Drumm, C.R.; Fan, W.C.; Powell, J.L.

2000-01-01

Massively-parallel computers allow detailed 3D radiation transport simulations to be performed to analyze the response of complex systems to radiation. This has been recently been demonstrated with the coupled electron-photon Monte Carlo code, ITS. To enable such calculations, the combinatorial geometry capability of ITS was improved. For greater geometrical flexibility, a version of ITS is under development that can track particles in CAD geometries. Deterministic radiation transport codes that utilize an unstructured spatial mesh are also being devised. For electron transport, the authors are investigating second-order forms of the transport equations which, when discretized, yield symmetric positive definite matrices. A novel parallelization strategy, simultaneously solving for spatial and angular unknowns, has been applied to the even- and odd-parity forms of the transport equation on a 2D unstructured spatial mesh. Another second-order form, the self-adjoint angular flux transport equation, also shows promise for electron transport

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)

Impact of mesh points number on the accuracy of deterministic calculations of control rods worth for Tehran research reactor

International Nuclear Information System (INIS)

Boustani, Ehsan; Amirkabir University of Technology, Tehran; Khakshournia, Samad

2016-01-01

In this paper two different computational approaches, a deterministic and a stochastic one, were used for calculation of the control rods worth of the Tehran research reactor. For the deterministic approach the MTRPC package composed of the WIMS code and diffusion code CITVAP was used, while for the stochastic one the Monte Carlo code MCNPX was applied. On comparing our results obtained by the Monte Carlo approach and those previously reported in the Safety Analysis Report (SAR) of Tehran research reactor produced by the deterministic approach large discrepancies were seen. To uncover the root cause of these discrepancies, some efforts were made and finally was discerned that the number of spatial mesh points in the deterministic approach was the critical cause of these discrepancies. Therefore, the mesh optimization was performed for different regions of the core such that the results of deterministic approach based on the optimized mesh points have a good agreement with those obtained by the Monte Carlo approach.

Impact of mesh points number on the accuracy of deterministic calculations of control rods worth for Tehran research reactor

Energy Technology Data Exchange (ETDEWEB)

Boustani, Ehsan [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Amirkabir University of Technology, Tehran (Iran, Islamic Republic of). Energy Engineering and Physics Dept.; Khakshournia, Samad [Amirkabir University of Technology, Tehran (Iran, Islamic Republic of). Energy Engineering and Physics Dept.

2016-12-15

In this paper two different computational approaches, a deterministic and a stochastic one, were used for calculation of the control rods worth of the Tehran research reactor. For the deterministic approach the MTRPC package composed of the WIMS code and diffusion code CITVAP was used, while for the stochastic one the Monte Carlo code MCNPX was applied. On comparing our results obtained by the Monte Carlo approach and those previously reported in the Safety Analysis Report (SAR) of Tehran research reactor produced by the deterministic approach large discrepancies were seen. To uncover the root cause of these discrepancies, some efforts were made and finally was discerned that the number of spatial mesh points in the deterministic approach was the critical cause of these discrepancies. Therefore, the mesh optimization was performed for different regions of the core such that the results of deterministic approach based on the optimized mesh points have a good agreement with those obtained by the Monte Carlo approach.

Neutron transport

International Nuclear Information System (INIS)

Berthoud, Georges; Ducros, Gerard; Feron, Damien; Guerin, Yannick; Latge, Christian; Limoge, Yves; Santarini, Gerard; Seiler, Jean-Marie; Vernaz, Etienne; Coste-Delclaux, Mireille; M'Backe Diop, Cheikh; Nicolas, Anne; Andrieux, Catherine; Archier, Pascal; Baudron, Anne-Marie; Bernard, David; Biaise, Patrick; Blanc-Tranchant, Patrick; Bonin, Bernard; Bouland, Olivier; Bourganel, Stephane; Calvin, Christophe; Chiron, Maurice; Damian, Frederic; Dumonteil, Eric; Fausser, Clement; Fougeras, Philippe; Gabriel, Franck; Gagnier, Emmanuel; Gallo, Daniele; Hudelot, Jean-Pascal; Hugot, Francois-Xavier; Dat Huynh, Tan; Jouanne, Cedric; Lautard, Jean-Jacques; Laye, Frederic; Lee, Yi-Kang; Lenain, Richard; Leray, Sylvie; Litaize, Olivier; Magnaud, Christine; Malvagi, Fausto; Mijuin, Dominique; Mounier, Claude; Naury, Sylvie; Nicolas, Anne; Noguere, Gilles; Palau, Jean-Marc; Le Pallec, Jean-Charles; Peneliau, Yannick; Petit, Odile; Poinot-Salanon, Christine; Raepsaet, Xavier; Reuss, Paul; Richebois, Edwige; Roque, Benedicte; Royer, Eric; Saint-Jean, Cyrille de; Santamarina, Alain; Serot, Olivier; Soldevila, Michel; Tommasi, Jean; Trama, Jean-Christophe; Tsilanizara, Aime; Behar, Christophe; Provitina, Olivier; Lecomte, Michael; Forestier, Alain; Bender, Alexandra; Parisot, Jean-Francois; Finot, Pierre

2013-10-01

This bibliographical note presents a reference book which addresses the study of neutron transport in matter, the study of conditions for a chain reaction and the study of modifications of matter composition due to nuclear reactions. This book presents the main nuclear data, their measurement, assessment and processing, and the spallation. It proposes an overview of methods applied for the study of neutron transport: basic equations and their derived forms, deterministic methods and Monte Carlo method of resolution of the Boltzmann equation, methods of resolution of generalized Bateman equations, methods of time resolution of space kinetics coupled equations. It presents the main calculation codes, discusses the qualification and experimental aspects, and gives an overview of neutron transport applications: neutron transport calculation of reactors, neutron transport coupled with other disciplines, physics of fuel cycle, criticality

Development of a model for unsteady deterministic stresses adapted to the multi-stages turbomachines simulation; Developpement d'un modele de tensions deterministes instationnaires adapte a la simulation de turbomachines multi-etagees

Energy Technology Data Exchange (ETDEWEB)

Charbonnier, D.

2004-12-15

The physical phenomena observed in turbomachines are generally three-dimensional and unsteady. A recent study revealed that a three-dimensional steady simulation can reproduce the time-averaged unsteady phenomena, since the steady flow field equations integrate deterministic stresses. The objective of this work is thus to develop an unsteady deterministic stresses model. The analogy with turbulence makes it possible to write transport equations for these stresses. The equations are implemented in steady flow solver and e model for the energy deterministic fluxes is also developed and implemented. Finally, this work shows that a three-dimensional steady simulation, by taking into account unsteady effects with transport equations of deterministic stresses, increases the computing time by only approximately 30 %, which remains very interesting compared to an unsteady simulation. (author)

Hydrogen recycle modeling in transport codes

International Nuclear Information System (INIS)

Howe, H.C.

1979-01-01

The hydrogen recycling models now used in Tokamak transport codes are reviewed and the method by which realistic recycling models are being added is discussed. Present models use arbitrary recycle coefficients and therefore do not model the actual recycling processes at the wall. A model for the hydrogen concentration in the wall serves two purposes: (1) it allows a better understanding of the density behavior in present gas puff, pellet, and neutral beam heating experiments; and (2) it allows one to extrapolate to long pulse devices such as EBT, ISX-C and reactors where the walls are observed or expected to saturate. Several wall models are presently being studied for inclusion in transport codes

Deterministic sensitivity and uncertainty methodology for best estimate system codes applied in nuclear technology

International Nuclear Information System (INIS)

Petruzzi, A.; D'Auria, F.; Cacuci, D.G.

2009-01-01

Nuclear Power Plant (NPP) technology has been developed based on the traditional defense in depth philosophy supported by deterministic and overly conservative methods for safety analysis. In the 1970s [1], conservative hypotheses were introduced for safety analyses to address existing uncertainties. Since then, intensive thermal-hydraulic experimental research has resulted in a considerable increase in knowledge and consequently in the development of best-estimate codes able to provide more realistic information about the physical behaviour and to identify the most relevant safety issues allowing the evaluation of the existing actual margins between the results of the calculations and the acceptance criteria. However, the best-estimate calculation results from complex thermal-hydraulic system codes (like Relap5, Cathare, Athlet, Trace, etc..) are affected by unavoidable approximations that are un-predictable without the use of computational tools that account for the various sources of uncertainty. Therefore the use of best-estimate codes (BE) within the reactor technology, either for design or safety purposes, implies understanding and accepting the limitations and the deficiencies of those codes. Taking into consideration the above framework, a comprehensive approach for utilizing quantified uncertainties arising from Integral Test Facilities (ITFs, [2]) and Separate Effect Test Facilities (SETFs, [3]) in the process of calibrating complex computer models for the application to NPP transient scenarios has been developed. The methodology proposed is capable of accommodating multiple SETFs and ITFs to learn as much as possible about uncertain parameters, allowing for the improvement of the computer model predictions based on the available experimental evidences. The proposed methodology constitutes a major step forward with respect to the generally used expert judgment and statistical methods as it permits a) to establish the uncertainties of any parameter

Multiple component codes based generalized LDPC codes for high-speed optical transport.

Science.gov (United States)

Djordjevic, Ivan B; Wang, Ting

2014-07-14

A class of generalized low-density parity-check (GLDPC) codes suitable for optical communications is proposed, which consists of multiple local codes. It is shown that Hamming, BCH, and Reed-Muller codes can be used as local codes, and that the maximum a posteriori probability (MAP) decoding of these local codes by Ashikhmin-Lytsin algorithm is feasible in terms of complexity and performance. We demonstrate that record coding gains can be obtained from properly designed GLDPC codes, derived from multiple component codes. We then show that several recently proposed classes of LDPC codes such as convolutional and spatially-coupled codes can be described using the concept of GLDPC coding, which indicates that the GLDPC coding can be used as a unified platform for advanced FEC enabling ultra-high speed optical transport. The proposed class of GLDPC codes is also suitable for code-rate adaption, to adjust the error correction strength depending on the optical channel conditions.

Colloid transport code-nuclear user`s manual

Energy Technology Data Exchange (ETDEWEB)

Jain, R. [New Mexico Univ., Albuquerque, NM (United States)

1992-04-03

This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential systems.

Benchmark studies of BOUT++ code and TPSMBI code on neutral transport during SMBI

Energy Technology Data Exchange (ETDEWEB)

Wang, Y.H. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Center for Magnetic Fusion Theory, Chinese Academy of Sciences, Hefei 230031 (China); Wang, Z.H., E-mail: zhwang@swip.ac.cn [Southwestern Institute of Physics, Chengdu 610041 (China); Guo, W., E-mail: wfguo@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Center for Magnetic Fusion Theory, Chinese Academy of Sciences, Hefei 230031 (China); Ren, Q.L. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Sun, A.P.; Xu, M.; Wang, A.K. [Southwestern Institute of Physics, Chengdu 610041 (China); Xiang, N. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Center for Magnetic Fusion Theory, Chinese Academy of Sciences, Hefei 230031 (China)

2017-06-09

SMBI (supersonic molecule beam injection) plays an important role in tokamak plasma fuelling, density control and ELM mitigation in magnetic confinement plasma physics, which has been widely used in many tokamaks. The trans-neut module of BOUT++ code is the only large-scale parallel 3D fluid code used to simulate the SMBI fueling process, while the TPSMBI (transport of supersonic molecule beam injection) code is a recent developed 1D fluid code of SMBI. In order to find a method to increase SMBI fueling efficiency in H-mode plasma, especially for ITER, it is significant to first verify the codes. The benchmark study between the trans-neut module of BOUT++ code and the TPSMBI code on radial transport dynamics of neutral during SMBI has been first successfully achieved in both slab and cylindrical coordinates. The simulation results from the trans-neut module of BOUT++ code and TPSMBI code are consistent very well with each other. Different upwind schemes have been compared to deal with the sharp gradient front region during the inward propagation of SMBI for the code stability. The influence of the WENO3 (weighted essentially non-oscillatory) and the third order upwind schemes on the benchmark results has also been discussed. - Highlights: • A 1D model of SMBI has developed. • Benchmarks of BOUT++ and TPSMBI codes have first been finished. • The influence of the WENO3 and the third order upwind schemes on the benchmark results has also been discussed.

RADTRAN: a computer code to analyze transportation of radioactive material

International Nuclear Information System (INIS)

Taylor, J.M.; Daniel, S.L.

1977-04-01

A computer code is presented which predicts the environmental impact of any specific scheme of radioactive material transportation. Results are presented in terms of annual latent cancer fatalities and annual early fatility probability resulting from exposure, during normal transportation or transport accidents. The code is developed in a generalized format to permit wide application including normal transportation analysis; consideration of alternatives; and detailed consideration of specific sectors of industry

Hazardous waste transportation risk assessment: Benefits of a combined deterministic and probabilistic Monte Carlo approach in expressing risk uncertainty

International Nuclear Information System (INIS)

Policastro, A.J.; Lazaro, M.A.; Cowen, M.A.; Hartmann, H.M.; Dunn, W.E.; Brown, D.F.

1995-01-01

This paper presents a combined deterministic and probabilistic methodology for modeling hazardous waste transportation risk and expressing the uncertainty in that risk. Both the deterministic and probabilistic methodologies are aimed at providing tools useful in the evaluation of alternative management scenarios for US Department of Energy (DOE) hazardous waste treatment, storage, and disposal (TSD). The probabilistic methodology can be used to provide perspective on and quantify uncertainties in deterministic predictions. The methodology developed has been applied to 63 DOE shipments made in fiscal year 1992, which contained poison by inhalation chemicals that represent an inhalation risk to the public. Models have been applied to simulate shipment routes, truck accident rates, chemical spill probabilities, spill/release rates, dispersion, population exposure, and health consequences. The simulation presented in this paper is specific to trucks traveling from DOE sites to their commercial TSD facilities, but the methodology is more general. Health consequences are presented as the number of people with potentially life-threatening health effects. Probabilistic distributions were developed (based on actual item data) for accident release amounts, time of day and season of the accident, and meteorological conditions

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

Release of WIMS10: a versatile reactor physics code for thermal and fast systems - 15467

International Nuclear Information System (INIS)

Lindley, B.A.; Newton, T.D.; Hosking, J.G.; Smith, P.N.; Powney, D.J.; Tollit, B.; Smith, P.J.

2015-01-01

the WIMS code provides a versatile software package for neutronic calculations, which can be applied to all thermal reactor types including mixed moderator systems. It can provide lattice cell and supercell calculations using a range of flux solutions methods to produce the neutronic libraries for use in PANTHER or other whole core analysis codes. With the release of WIMS10, the range of problems which WIMS can solve has been greatly extended. A WIMS/PANTHER calculation route has been developed and validated for part MOX-fuelled PWRs, with calculations showing excellent agreement with 2D core deterministic and Monte Carlo transport solutions. A flexible geometry 3D method of characteristics transport solver, CACTUS3D has been added to the code. CACTUS3D has been benchmarked for a 3D BWR assembly model, and was in good agreement with a direct 172-group solution in the Monte Carlo code MONK. Fast reactor calculations using the ECCO deterministic calculation route have been validated using experimental data from the ZEBRA reactor. Power deposition can be treated through following neutrons and/or photons to their point of interaction. The improved methodology is shown to give more accurate calculation of heat deposition and improve agreement between calculated and measured detector responses for part MOX-fuelled cores. (authors)

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

User's manual for the Oak Ridge Tokamak Transport Code

International Nuclear Information System (INIS)

Munro, J.K.; Hogan, J.T.; Howe, H.C.; Arnurius, D.E.

1977-02-01

A one-dimensional tokamak transport code is described which simulates a plasma discharge using a fluid model which includes power balances for electrons and ions, conservation of mass, and Maxwell's equations. The modular structure of the code allows a user to add models of various physical processes which can modify the discharge behavior. Such physical processes treated in the version of the code described here include effects of plasma transport, neutral gas transport, impurity diffusion, and neutral beam injection. Each process can be modeled by a parameterized analytic formula or at least one detailed numerical calculation. The program logic of each module is presented, followed by detailed descriptions of each subroutine used by the module. The physics underlying the models is only briefly summarized. The transport code was written in IBM FORTRAN-IV and implemented on IBM 360/370 series computers at the Oak Ridge National Laboratory and on the CDC 7600 computers of the Magnetic Fusion Energy (MFE) Computing Center of the Lawrence Livermore Laboratory. A listing of the current reference version is provided on accompanying microfiche

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

Using MCBEND for neutron or gamma-ray deterministic calculations

Directory of Open Access Journals (Sweden)

Geoff Dobson

2017-01-01

Full Text Available MCBEND 11 is the latest version of the general radiation transport Monte Carlo code from AMEC Foster Wheeler’s ANSWERS® Software Service. MCBEND is well established in the UK shielding community for radiation shielding and dosimetry assessments. MCBEND supports a number of acceleration techniques, for example the use of an importance map in conjunction with Splitting/Russian Roulette. MCBEND has a well established automated tool to generate this importance map, commonly referred to as the MAGIC module using a diffusion adjoint solution. This method is fully integrated with the MCBEND geometry and material specification, and can easily be run as part of a normal MCBEND calculation. An often overlooked feature of MCBEND is the ability to use this method for forward scoping calculations, which can be run as a very quick deterministic method. Additionally, the development of the Visual Workshop environment for results display provides new capabilities for the use of the forward calculation as a productivity tool. In this paper, we illustrate the use of the combination of the old and new in order to provide an enhanced analysis capability. We also explore the use of more advanced deterministic methods for scoping calculations used in conjunction with MCBEND, with a view to providing a suite of methods to accompany the main Monte Carlo solver.

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

Implementation of the kinetics in the transport code AZTRAN

International Nuclear Information System (INIS)

Duran G, J. A.; Del Valle G, E.; Gomez T, A. M.

2017-09-01

This paper shows the implementation of the time dependence in the three-dimensional transport code AZTRAN (AZtlan TRANsport), which belongs to the AZTLAN platform, for the analysis of nuclear reactors (currently under development). The AZTRAN code with this implementation is able to numerically solve the time-dependent transport equation in XYZ geometry, for several energy groups, using the discrete ordinate method S n for the discretization of the angular variable, the nodal method RTN-0 for spatial discretization and method 0 for discretization in time. Initially, the code only solved the neutrons transport equation in steady state, so the implementation of the temporal part was made integrating the neutrons transport equation with respect to time and balance equations corresponding to the concentrations of delayed neutron precursors, for which method 0 was applied. After having directly implemented code kinetics, the improved quasi-static method was implemented, which is a tool for reducing computation time, where the angular flow is factored by the product of two functions called shape function and amplitude function, where the first is calculated for long time steps, called macro-steps and the second is resolved for small time steps called micro-steps. In the new version of AZTRAN several Benchmark problems that were taken from the literature were simulated, the problems used are of two and three dimensions which allowed corroborating the accuracy and stability of the code, showing in general in the reference tests a good behavior. (Author)

HETFIS: High-Energy Nucleon-Meson Transport Code with Fission

International Nuclear Information System (INIS)

Barish, J.; Gabriel, T.A.; Alsmiller, F.S.; Alsmiller, R.G. Jr.

1981-07-01

A model that includes fission for predicting particle production spectra from medium-energy nucleon and pion collisions with nuclei (Z greater than or equal to 91) has been incorporated into the nucleon-meson transport code, HETC. This report is primarily concerned with the programming aspects of HETFIS (High-Energy Nucleon-Meson Transport Code with Fission). A description of the program data and instructions for operating the code are given. HETFIS is written in FORTRAN IV for the IBM computers and is readily adaptable to other systems

Survey of 1 1/2D transport codes

International Nuclear Information System (INIS)

Grad, H.

1978-10-01

A survey is given of a family of classical transport codes, recently termed ''1 1/2D'', which efficiently and accurately follow the evolution of plasma configurations on a long time scale, following coupled changes in plasma shape and topology with transport (but not wave motion). Codes have been constructed and operated (since 1974) which include various combinations of finite beta, general plasma cross-section and aspect, various topologies (Doublet, tearing, reversed-field mirror) including time dependent transitions in topology resulting from external coil variation and plasma transport, with models including (classical) tensor resistivity and heat flow as well as the adiabatic limiting case

BOT3P5.2, 3D Mesh Generator and Graphical Display of Geometry for Radiation Transport Codes, Display of Results

International Nuclear Information System (INIS)

Orsi, Roberto; Bidaud, Adrien

2007-01-01

1 - Description of program or function: BOT3P was originally conceived as a set of standard FORTRAN 77 language programs in order to give the users of the DORT and TORT deterministic transport codes some useful diagnostic tools to prepare and check their input data files. Later versions extended the possibility to produce the geometrical, material distribution and fixed neutron source data to other deterministic transport codes such as TWODANT/THREEDANT of the DANTSYS system, PARTISN and, potentially, to any transport code through BOT3P binary output files that can be easily interfaced (see, for example, the Russian two-dimensional (2D) and three-dimensional (3D) discrete ordinates neutron, photon and charged particle transport codes KASKAD-S-2.5 and KATRIN-2.0). As from Version 5.1 BOT3P contained important additions specifically addressed to radiation transport analysis for medical applications. BOT3P-5.2 contains new graphics capabilities. Some of them enable users to select space sub-domains of the total mesh grid in order to improve the zoom simulation of the geometry, both in 2D cuts and in 3D. Moreover the new BOT3P module (PDTM) may improve the interface of BOT3P geometrical models to transport analysis codes. The following programs are included in the BOT3P software package: GGDM, DDM, GGTM, DTM2, DTM3, RVARSCL, COMPARE, MKSRC, CATSM, DTET, and PDTM. The main features of these different programs are described. 2 - Methods: GGDM and GGTM work similarly from the logical point of view. Since the 3D case is more general, the following description refers to GGTM. All the co-ordinate values that characterise the geometrical scheme at the basis of the 3D transport code geometrical and material model are read, sorted and all stored if different from the neighbouring ones more than an input tolerance established by the user. These co-ordinates are always present in the fine-mesh boundary arrays independently of the mesh grid refinement options, because they

Recent developments in the Los Alamos radiation transport code system

International Nuclear Information System (INIS)

Forster, R.A.; Parsons, K.

1997-01-01

A brief progress report on updates to the Los Alamos Radiation Transport Code System (LARTCS) for solving criticality and fixed-source problems is provided. LARTCS integrates the Diffusion Accelerated Neutral Transport (DANT) discrete ordinates codes with the Monte Carlo N-Particle (MCNP) code. The LARCTS code is being developed with a graphical user interface for problem setup and analysis. Progress in the DANT system for criticality applications include a two-dimensional module which can be linked to a mesh-generation code and a faster iteration scheme. Updates to MCNP Version 4A allow statistical checks of calculated Monte Carlo results

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.)

Electron transport code theoretical basis

International Nuclear Information System (INIS)

Dubi, A.; Horowitz, Y.S.

1978-04-01

This report mainly describes the physical and mathematical considerations involved in the treatment of the multiple collision processes. A brief description is given of the traditional methods used in electron transport via Monte Carlo, and a somewhat more detailed description, of the approach to be used in the presently developed code

Validation study of the reactor physics lattice transport code WIMSD-5B by TRX and BAPL critical experiments of light water reactors

International Nuclear Information System (INIS)

Khan, M.J.H.; Alam, A.B.M.K.; Ahsan, M.H.; Mamun, K.A.A.; Islam, S.M.A.

2015-01-01

Highlights: • To validate the reactor physics lattice code WIMSD-5B by this analysis. • To model TRX and BAPL critical experiments using WIMSD-5B. • To compare the calculated results with experiment and MCNP results. • To rely on WIMSD-5B code for TRIGA calculations. - Abstract: The aim of this analysis is to validate the reactor physics lattice transport code WIMSD-5B by TRX (thermal reactor-one region lattice) and BAPL (Bettis Atomic Power Laboratory-one region lattice) critical experiments of light water reactors for neutronics analysis of 3 MW TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh. This analysis is achieved through the analysis of integral parameters of five light water reactor critical experiments TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 based on evaluated nuclear data libraries JEFF-3.1 and ENDF/B-VII.1. In integral measurements, these experiments are considered as standard benchmark lattices for validating the reactor physics lattice transport code WIMSD-5B as well as evaluated nuclear data libraries. The integral parameters of the said critical experiments are calculated using the reactor physics lattice transport code WIMSD-5B. The calculated integral parameters are compared to the measured values as well as the earlier published MCNP results based on the Chinese evaluated nuclear data library CENDL-3.0 for assessment of deterministic calculation. It was found that the calculated integral parameters give mostly reasonable and globally consistent results with the experiment and the MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are well consistent with each other. Therefore, this analysis reveals the validation study of the reactor physics lattice transport code WIMSD-5B based on JEFF-3.1 and ENDF/B-VII.1 libraries and can also be essential to

Regional Atmospheric Transport Code for Hanford Emission Tracking, Version 2 (RATCHET2)

International Nuclear Information System (INIS)

Ramsdell, James V.; Rishel, Jeremy P.

2006-01-01

This manual describes the atmospheric model and computer code for the Atmospheric Transport Module within SAC. The Atmospheric Transport Module, called RATCHET2, calculates the time-integrated air concentration and surface deposition of airborne contaminants to the soil. The RATCHET2 code is an adaptation of the Regional Atmospheric Transport Code for Hanford Emissions Tracking (RATCHET). The original RATCHET code was developed to perform the atmospheric transport for the Hanford Environmental Dose Reconstruction Project. Fundamentally, the two sets of codes are identical; no capabilities have been deleted from the original version of RATCHET. Most modifications are generally limited to revision of the run-specification file to streamline the simulation process for SAC.

Regional Atmospheric Transport Code for Hanford Emission Tracking, Version 2(RATCHET2)

Energy Technology Data Exchange (ETDEWEB)

Ramsdell, James V.; Rishel, Jeremy P.

2006-07-01

This manual describes the atmospheric model and computer code for the Atmospheric Transport Module within SAC. The Atmospheric Transport Module, called RATCHET2, calculates the time-integrated air concentration and surface deposition of airborne contaminants to the soil. The RATCHET2 code is an adaptation of the Regional Atmospheric Transport Code for Hanford Emissions Tracking (RATCHET). The original RATCHET code was developed to perform the atmospheric transport for the Hanford Environmental Dose Reconstruction Project. Fundamentally, the two sets of codes are identical; no capabilities have been deleted from the original version of RATCHET. Most modifications are generally limited to revision of the run-specification file to streamline the simulation process for SAC.

Neutronics comparative analysis of plate-type research reactor using deterministic and stochastic methods

International Nuclear Information System (INIS)

Liu, Shichang; Wang, Guanbo; Wu, Gaochen; Wang, Kan

2015-01-01

Highlights: • DRAGON and DONJON are applied and verified in calculations of research reactors. • Continuous-energy Monte Carlo calculations by RMC are chosen as the references. • “ECCO” option of DRAGON is suitable for the calculations of research reactors. • Manual modifications of cross-sections are not necessary with DRAGON and DONJON. • DRAGON and DONJON agree well with RMC if appropriate treatments are applied. - Abstract: Simulation of the behavior of the plate-type research reactors such as JRR-3M and CARR poses a challenge for traditional neutronics calculation tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity and large leakage of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON and DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic approach. The goal of this research is to examine the capability of the deterministic code system DRAGON and DONJON to reliably simulate the research reactors. The results indicate that the DRAGON and DONJON code system agrees well with the continuous-energy Monte Carlo simulation on both k eff and flux distributions if the appropriate treatments (such as the ECCO option) are applied

Modelling plastic scintillator response to gamma rays using light transport incorporated FLUKA code

Energy Technology Data Exchange (ETDEWEB)

Ranjbar Kohan, M. [Physics Department, Tafresh University, Tafresh (Iran, Islamic Republic of); Etaati, G.R. [Department of Nuclear Engineering and Physics, Amir Kabir University of Technology, Tehran (Iran, Islamic Republic of); Ghal-Eh, N., E-mail: ghal-eh@du.ac.ir [School of Physics, Damghan University, Damghan (Iran, Islamic Republic of); Safari, M.J. [Department of Energy Engineering, Sharif University of Technology, Tehran (Iran, Islamic Republic of); Afarideh, H. [Department of Nuclear Engineering and Physics, Amir Kabir University of Technology, Tehran (Iran, Islamic Republic of); Asadi, E. [Department of Physics, Payam-e-Noor University, Tehran (Iran, Islamic Republic of)

2012-05-15

The response function of NE102 plastic scintillator to gamma rays has been simulated using a joint FLUKA+PHOTRACK Monte Carlo code. The multi-purpose particle transport code, FLUKA, has been responsible for gamma transport whilst the light transport code, PHOTRACK, has simulated the transport of scintillation photons through scintillator and lightguide. The simulation results of plastic scintillator with/without light guides of different surface coverings have been successfully verified with experiments. - Highlights: Black-Right-Pointing-Pointer A multi-purpose code (FLUKA) and a light transport code (PHOTRACK) have been linked. Black-Right-Pointing-Pointer The hybrid code has been used to generate the response function of an NE102 scintillator. Black-Right-Pointing-Pointer The simulated response functions exhibit a good agreement with experimental data.

High energy particle transport code NMTC/JAM

International Nuclear Information System (INIS)

Niita, K.; Takada, H.; Meigo, S.; Ikeda, Y.

2001-01-01

We have developed a high energy particle transport code NMTC/JAM, which is an upgrade version of NMTC/JAERI97. The available energy range of NMTC/JAM is, in principle, extended to 200 GeV for nucleons and mesons including the high energy nuclear reaction code JAM for the intra-nuclear cascade part. We compare the calculations by NMTC/JAM code with the experimental data of thin and thick targets for proton induced reactions up to several 10 GeV. The results of NMTC/JAM code show excellent agreement with the experimental data. From these code validation, it is concluded that NMTC/JAM is reliable in neutronics optimization study of the high intense spallation neutron utilization facility. (author)

Burnup-dependent core neutronics analysis of plate-type research reactor using deterministic and stochastic methods

International Nuclear Information System (INIS)

Liu, Shichang; Wang, Guanbo; Liang, Jingang; Wu, Gaochen; Wang, Kan

2015-01-01

Highlights: • DRAGON & DONJON were applied in burnup calculations of plate-type research reactors. • Continuous-energy Monte Carlo burnup calculations by RMC were chosen as references. • Comparisons of keff, isotopic densities and power distribution were performed. • Reasons leading to discrepancies between two different approaches were analyzed. • DRAGON & DONJON is capable of burnup calculations with appropriate treatments. - Abstract: The burnup-dependent core neutronics analysis of the plate-type research reactors such as JRR-3M poses a challenge for traditional neutronics calculational tools and schemes for power reactors, due to the characteristics of complex geometry, highly heterogeneity, large leakage and the particular neutron spectrum of the research reactors. Two different theoretical approaches, the deterministic and the stochastic methods, are used for the burnup-dependent core neutronics analysis of the JRR-3M plate-type research reactor in this paper. For the deterministic method the neutronics codes DRAGON & DONJON are used, while the continuous-energy Monte Carlo code RMC (Reactor Monte Carlo code) is employed for the stochastic one. In the first stage, the homogenizations of few-group cross sections by DRAGON and the full core diffusion calculations by DONJON have been verified by comparing with the detailed Monte Carlo simulations. In the second stage, the burnup-dependent calculations of both assembly level and the full core level were carried out, to examine the capability of the deterministic code system DRAGON & DONJON to reliably simulate the burnup-dependent behavior of research reactors. The results indicate that both RMC and DRAGON & DONJON code system are capable of burnup-dependent neutronics analysis of research reactors, provided that appropriate treatments are applied in both assembly and core levels for the deterministic codes

Energy-pointwise discrete ordinates transport methods

International Nuclear Information System (INIS)

Williams, M.L.; Asgari, M.; Tashakorri, R.

1997-01-01

A very brief description is given of a one-dimensional code, CENTRM, which computes a detailed, space-dependent flux spectrum in a pointwise-energy representation within the resolved resonance range. The code will become a component in the SCALE system to improve computation of self-shielded cross sections, thereby enhancing the accuracy of codes such as KENO. CENTRM uses discrete-ordinates transport theory with an arbitrary angular quadrature order and a Legendre expansion of scattering anisotropy for moderator materials and heavy nuclides. The CENTRM program provides capability to deterministically compute full energy range, space-dependent angular flux spectra, rigorously accounting for resonance fine-structure and scattering anisotropy effects

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

Performance analysis of multidimensional wavefront algorithms with application to deterministic particle transport

International Nuclear Information System (INIS)

Hoisie, A.; Lubeck, O.; Wasserman, H.

1998-01-01

The authors develop a model for the parallel performance of algorithms that consist of concurrent, two-dimensional wavefronts implemented in a message passing environment. The model, based on a LogGP machine parameterization, combines the separate contributions of computation and communication wavefronts. They validate the model on three important supercomputer systems, on up to 500 processors. They use data from a deterministic particle transport application taken from the ASCI workload, although the model is general to any wavefront algorithm implemented on a 2-D processor domain. They also use the validated model to make estimates of performance and scalability of wavefront algorithms on 100-TFLOPS computer systems expected to be in existence within the next decade as part of the ASCI program and elsewhere. In this context, the authors analyze two problem sizes. Their model shows that on the largest such problem (1 billion cells), inter-processor communication performance is not the bottleneck. Single-node efficiency is the dominant factor

DANTSYS: A diffusion accelerated neutral particle transport code system

Energy Technology Data Exchange (ETDEWEB)

Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Marr, D.R.; O`Dell, R.D.; Walters, W.F.

1995-06-01

The DANTSYS code package includes the following transport codes: ONEDANT, TWODANT, TWODANT/GQ, TWOHEX, and THREEDANT. The DANTSYS code package is a modular computer program package designed to solve the time-independent, multigroup discrete ordinates form of the boltzmann transport equation in several different geometries. The modular construction of the package separates the input processing, the transport equation solving, and the post processing (or edit) functions into distinct code modules: the Input Module, one or more Solver Modules, and the Edit Module, respectively. The Input and Edit Modules are very general in nature and are common to all the Solver Modules. The ONEDANT Solver Module contains a one-dimensional (slab, cylinder, and sphere), time-independent transport equation solver using the standard diamond-differencing method for space/angle discretization. Also included in the package are solver Modules named TWODANT, TWODANT/GQ, THREEDANT, and TWOHEX. The TWODANT Solver Module solves the time-independent two-dimensional transport equation using the diamond-differencing method for space/angle discretization. The authors have also introduced an adaptive weighted diamond differencing (AWDD) method for the spatial and angular discretization into TWODANT as an option. The TWOHEX Solver Module solves the time-independent two-dimensional transport equation on an equilateral triangle spatial mesh. The THREEDANT Solver Module solves the time independent, three-dimensional transport equation for XYZ and RZ{Theta} symmetries using both diamond differencing with set-to-zero fixup and the AWDD method. The TWODANT/GQ Solver Module solves the 2-D transport equation in XY and RZ symmetries using a spatial mesh of arbitrary quadrilaterals. The spatial differencing method is based upon the diamond differencing method with set-to-zero fixup with changes to accommodate the generalized spatial meshing.

DANTSYS: A diffusion accelerated neutral particle transport code system

International Nuclear Information System (INIS)

Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Marr, D.R.; O'Dell, R.D.; Walters, W.F.

1995-06-01

The DANTSYS code package includes the following transport codes: ONEDANT, TWODANT, TWODANT/GQ, TWOHEX, and THREEDANT. The DANTSYS code package is a modular computer program package designed to solve the time-independent, multigroup discrete ordinates form of the boltzmann transport equation in several different geometries. The modular construction of the package separates the input processing, the transport equation solving, and the post processing (or edit) functions into distinct code modules: the Input Module, one or more Solver Modules, and the Edit Module, respectively. The Input and Edit Modules are very general in nature and are common to all the Solver Modules. The ONEDANT Solver Module contains a one-dimensional (slab, cylinder, and sphere), time-independent transport equation solver using the standard diamond-differencing method for space/angle discretization. Also included in the package are solver Modules named TWODANT, TWODANT/GQ, THREEDANT, and TWOHEX. The TWODANT Solver Module solves the time-independent two-dimensional transport equation using the diamond-differencing method for space/angle discretization. The authors have also introduced an adaptive weighted diamond differencing (AWDD) method for the spatial and angular discretization into TWODANT as an option. The TWOHEX Solver Module solves the time-independent two-dimensional transport equation on an equilateral triangle spatial mesh. The THREEDANT Solver Module solves the time independent, three-dimensional transport equation for XYZ and RZΘ symmetries using both diamond differencing with set-to-zero fixup and the AWDD method. The TWODANT/GQ Solver Module solves the 2-D transport equation in XY and RZ symmetries using a spatial mesh of arbitrary quadrilaterals. The spatial differencing method is based upon the diamond differencing method with set-to-zero fixup with changes to accommodate the generalized spatial meshing

High energy particle transport code NMTC/JAM

International Nuclear Information System (INIS)

Niita, Koji; Meigo, Shin-ichiro; Takada, Hiroshi; Ikeda, Yujiro

2001-03-01

We have developed a high energy particle transport code NMTC/JAM, which is an upgraded version of NMTC/JAERI97. The applicable energy range of NMTC/JAM is extended in principle up to 200 GeV for nucleons and mesons by introducing the high energy nuclear reaction code JAM for the intra-nuclear cascade part. For the evaporation and fission process, we have also implemented a new model, GEM, by which the light nucleus production from the excited residual nucleus can be described. According to the extension of the applicable energy, we have upgraded the nucleon-nucleus non-elastic, elastic and differential elastic cross section data by employing new systematics. In addition, the particle transport in a magnetic field has been implemented for the beam transport calculations. In this upgrade, some new tally functions are added and the format of input of data has been improved very much in a user friendly manner. Due to the implementation of these new calculation functions and utilities, consequently, NMTC/JAM enables us to carry out reliable neutronics study of a large scale target system with complex geometry more accurately and easily than before. This report serves as a user manual of the code. (author)

Jetto a free boundary plasma transport code

International Nuclear Information System (INIS)

Cenacchi, G.; Taroni, A.

1988-01-01

JETTO is a one-and-a-half-dimensional transport code calculating the evolution of plasma parameters in a time dependent axisymmetric MHD equilibrium configuration. A splitting technique gives a consistent solution of coupled equilibrium and transport equations. The plasma boundary is free and defined either by its contact with a limiter (wall) or by a separatrix or by the toroidal magnetic flux. The Grad's approach to the equilibrium problem with adiabatic (or similar) constraints is adopted. This method consists of iterating by alternately solving the Grad-Schluter-Shafranov equation (PDE) and the ODE obtained by averaging the PDE over the magnetic surfaces. The bidimensional equation of the poloidal flux is solved by a finite difference scheme, whereas a Runge-Kutta method is chosen for the averaged equilibrium equation. The 1D transport equations (averaged over the magnetic surfaces) for the electron and ion densities and energies and for the rotational transform are written in terms of a coordinate (ρ) related to the toroidal flux. Impurity transport is also considered, under the hypothesis of coronal equilibrium. The transport equations are solved by an implicit scheme in time and by a finite difference scheme in space. The centering of the source terms and transport coefficients is performed using a Predictor-Corrector scheme. The basic version of the code is described here in detail; input and output parameters are also listed

The RADionuclide Transport, Removal, and Dose (RADTRAD) code

International Nuclear Information System (INIS)

Miller, L.A.; Chanin, D.I.; Lee, J.

1993-01-01

The RADionuclide Transport, Removal, And Dose (RADTRAD) code is designed for US Nuclear Regulatory Commission (USNRC) use to calculate the radiological consequences to the offsite population and to control room operators following a design-basis accident at Light Water Reactor (LWR) power plants. This code utilizes updated reactor accident source terms published in draft NUREG-1465, ''Accident Source Terms for Light-Water Nuclear Power Plants.'' The code will track the transport of radionuclides as they are released from the reactor pressure vessel, travel through the primary containment and other buildings, and are released to the environment. As the radioactive material is transported through the primary containment and other buildings, credit for several removal mechanisms may be taken including sprays, suppression pools, overlying pools, filters, and natural deposition. Simple models are available for these different removal mechanisms that use, as input, information about the conditions in the plant and predict either a removal coefficient (λ) or decontamination factor. The user may elect to use these models or input a single value for a removal coefficient or decontamination factor

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.)

Development of Deterministic and Probabilistic Fracture Mechanics Analysis Code PROFAS-RV for Reactor Pressure Vessel - Progress of the Work

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong Min; Lee, Bong Sang [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In this study, a deterministic/probabilistic fracture mechanics analysis program for reactor pressure vessel, PROFAS-RV, is developed. This program can evaluate failure probability of RPV using recent radiation embrittlement model of 10CFR50.61a and stress intensity factor calculation method of RCC-MRx code as well as the required basic functions of PFM program. Applications of some new radiation embrittlement model, material database, calculation method of stress intensity factors, and others which can improve fracture mechanics assessment of RPV are introduced. The purpose of this study is to develop a probabilistic fracture mechanics (PFM) analysis program for RPV considering above modification and application of newly developed models and calculation methods. In this paper, it deals with the development progress of the PFM analysis program for RPV, PROFAS-RV. The PROFAS-RV is being tested with other codes, and it is expected to revise and upgrade by reflecting the latest model and calculation method continuously. These efforts can minimize the uncertainty of the integrity evaluation for the reactor pressure vessel.

Time-dependent deterministic transport on parallel architectures using PARTISN

International Nuclear Information System (INIS)

Alcouffe, R.E.; Baker, R.S.

1998-01-01

In addition to the ability to solve the static transport equation, the authors have also incorporated time dependence into the parallel S N code PARTISN. Using a semi-implicit scheme, PARTISN is capable of performing time-dependent calculations for both fissioning and pure source driven problems. They have applied this to various types of problems such as shielding and prompt fission experiments. This paper describes the form of the time-dependent equations implemented, their solution strategies in PARTISN including iteration acceleration, and the strategies used for time-step control. Results are presented for a iron-water shielding calculation and a criticality excursion in a uranium solution configuration

Comparative static simulations of a CANDU6 cell using different transport codes

Energy Technology Data Exchange (ETDEWEB)

Mahjoub, M.; Koclas, J., E-mail: mehdi.mahjoub@polymtl.ca [Ecole Polytechnique de Montreal, QC (Canada)

2015-07-01

The solution of the time dependent Boltzmann equation remains quite a challenge. We are in the process of developing such a method using the stochastic Monte Carlo approach for two reasons: First, at the cell level, we will be able to obtain time dependent homogenized cross sections for use in full core diffusion calculations. Second, the Monte Carlo methods are scalable to perform full core if and when appropriate computer resources become available. The Time dependent approach will be concretized a new module that will be added to an existing Monte Carlo code. As a first step towards this goal, we need to choose the initial Monte Carlo code to be used as start point. For this reason, we have compared results concerning the void reactivity of a fresh fuel CANDU6 cell, using two Monte Carlo codes, namely VTT developed SERPENT and MIT developed OpenMC together with the deterministic DRAGON code. Several libraries are used in this comparison. We conclude that OpenMC is a good candidate for implementation of a time dependent simulation. (author)

Deterministic sensitivity and uncertainty analysis for large-scale computer models

International Nuclear Information System (INIS)

Worley, B.A.; Pin, F.G.; Oblow, E.M.; Maerker, R.E.; Horwedel, J.E.; Wright, R.Q.

1988-01-01

The fields of sensitivity and uncertainty analysis have traditionally been dominated by statistical techniques when large-scale modeling codes are being analyzed. These methods are able to estimate sensitivities, generate response surfaces, and estimate response probability distributions given the input parameter probability distributions. Because the statistical methods are computationally costly, they are usually applied only to problems with relatively small parameter sets. Deterministic methods, on the other hand, are very efficient and can handle large data sets, but generally require simpler models because of the considerable programming effort required for their implementation. The first part of this paper reports on the development and availability of two systems, GRESS and ADGEN, that make use of computer calculus compilers to automate the implementation of deterministic sensitivity analysis capability into existing computer models. This automation removes the traditional limitation of deterministic sensitivity methods. This second part of the paper describes a deterministic uncertainty analysis method (DUA) that uses derivative information as a basis to propagate parameter probability distributions to obtain result probability distributions. This paper is applicable to low-level radioactive waste disposal system performance assessment

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)

Development and implementation of a set of numerical quadratures SQ{sub N} and EQ{sub N} type in the transport code AZTRAN; Desarrollo e implementacion de un conjunto de cuadraturas numericas de tipo SQ{sub N} y EQ{sub N} en el codigo de transporte AZTRAN

Energy Technology Data Exchange (ETDEWEB)

Chepe P, M. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Del Valle G, E., E-mail: liaison.web@gmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, 07738 Ciudad de Mexico (Mexico)

2015-09-15

The deterministic transport codes for analysis of nuclear reactors have been used for several years already, these codes have evolved in terms of the methodology used and the degree of accuracy, because at the present time has more computer power. In this paper, the transport code used considers the classical technique of multi-group for discretization energy, for space discretization uses the nodal methods, while for the angular discretization the discrete ordinates method is used; so that presents the development and implementation of a set of numerical quadratures of SQ{sub N} type symmetrical with the same weight for each angular direction and these are compared with the quadratures of EQ{sub N} type. The two sets of numerical quadratures were implemented in the program AZTRAN to a problem with isotropic medium in XYZ geometry, in steady state using the nodal method RTN-0 (Raviart-Thomas-Nedelec). The analyzed results correspond to the effective multiplication factor k{sub eff} and neutron angular flux with approximations from S{sub 4} to S{sub 16}. (Author)

Line and lattice networks under deterministic interference models

NARCIS (Netherlands)

Goseling, Jasper; Gastpar, Michael; Weber, Jos H.

Capacity bounds are compared for four different deterministic models of wireless networks, representing four different ways of handling broadcast and superposition in the physical layer. In particular, the transport capacity under a multiple unicast traffic pattern is studied for a 1-D network of

Discrete-ordinates electron transport calculations using standard neutron transport codes

International Nuclear Information System (INIS)

Morel, J.E.

1979-01-01

The primary purpose of this work was to develop a method for using standard neutron transport codes to perform electron transport calculations. The method is to develop approximate electron cross sections which are sufficiently well-behaved to be treated with standard S/sub n/ methods, but which nonetheless yield flux solutions which are very similar to the exact solutions. The main advantage of this approach is that, once the approximate cross sections are constructed, their multigroup Legendre expansion coefficients can be calculated and input to any standard S/sub n/ code. Discrete-ordinates calculations were performed to determine the accuracy of the flux solutions for problems corresponding to 1.0-MeV electrons incident upon slabs of aluminum and gold. All S/sub n/ calculations were compared with similar calculations performed with an electron Monte Carlo code, considered to be exact. In all cases, the discrete-ordinates solutions for integral flux quantities (i.e., scalar flux, energy deposition profiles, etc.) are generally in agreement with the Monte Carlo solutions to within approximately 5% or less. The central conclusion is that integral electron flux quantities can be efficiently and accurately calculated using standard S/sub n/ codes in conjunction with approximate cross sections. Furthermore, if group structures and approximate cross section construction are optimized, accurate differential flux energy spectra may also be obtainable without having to use an inordinately large number of energy groups. 1 figure

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Intracoin - International Nuclide Transport Code Intercomparison Study

International Nuclear Information System (INIS)

1984-09-01

The purpose of the project is to obtain improved knowledge of the influence of various strategies for radionuclide transport modelling for the safety assessment of final repositories for nuclear waste. This is a report of the first phase of the project which was devoted to a comparison of the numerical accuracy of the computer codes used in the study. The codes can be divided into five groups, namely advection-dispersion models, models including matrix diffusion and chemical effects and finally combined models. The results are presented as comparisons of calculations since the objective of level 1 was code verification. (G.B.)

Parallel thermal radiation transport in two dimensions

International Nuclear Information System (INIS)

Smedley-Stevenson, R.P.; Ball, S.R.

2003-01-01

This paper describes the distributed memory parallel implementation of a deterministic thermal radiation transport algorithm in a 2-dimensional ALE hydrodynamics code. The parallel algorithm consists of a variety of components which are combined in order to produce a state of the art computational capability, capable of solving large thermal radiation transport problems using Blue-Oak, the 3 Tera-Flop MPP (massive parallel processors) computing facility at AWE (United Kingdom). Particular aspects of the parallel algorithm are described together with examples of the performance on some challenging applications. (author)

Parallel thermal radiation transport in two dimensions

Energy Technology Data Exchange (ETDEWEB)

Smedley-Stevenson, R.P.; Ball, S.R. [AWE Aldermaston (United Kingdom)

2003-07-01

This paper describes the distributed memory parallel implementation of a deterministic thermal radiation transport algorithm in a 2-dimensional ALE hydrodynamics code. The parallel algorithm consists of a variety of components which are combined in order to produce a state of the art computational capability, capable of solving large thermal radiation transport problems using Blue-Oak, the 3 Tera-Flop MPP (massive parallel processors) computing facility at AWE (United Kingdom). Particular aspects of the parallel algorithm are described together with examples of the performance on some challenging applications. (author)

Development of TIGER code for radionuclide transport in a geochemically evolving region

International Nuclear Information System (INIS)

Mihara, Morihiro; Ooi, Takao

2004-01-01

In a transuranic (TRU) waste geological disposal facility, using cementitious materials is being considered. Cementitious materials will gradually dissolve in groundwater over the long-term. In the performance assessment report of a TRU waste repository in Japan already published, the most conservative radionuclide migration parameter set was selected considering the evolving cementitious material. Therefore, a tool to perform the calculation of radionuclide transport considering long-term geochemically evolving cementitious materials, named the TIGER code, Transport In Geochemically Evolving Region was developed to calculate a more realistic performance assessment. It can calculate radionuclide transport in engineered and natural barrier systems. In this report, mathematical equations of this code are described and validated with analytical solutions and results of other codes for radionuclide transport. The more realistic calculation of radionuclide transport for a TRU waste geological disposal system using the TIGER code could be performed. (author)

Diffusion in Deterministic Interacting Lattice Systems

Science.gov (United States)

Medenjak, Marko; Klobas, Katja; Prosen, Tomaž

2017-09-01

We study reversible deterministic dynamics of classical charged particles on a lattice with hard-core interaction. It is rigorously shown that the system exhibits three types of transport phenomena, ranging from ballistic, through diffusive to insulating. By obtaining an exact expressions for the current time-autocorrelation function we are able to calculate the linear response transport coefficients, such as the diffusion constant and the Drude weight. Additionally, we calculate the long-time charge profile after an inhomogeneous quench and obtain diffusive profilewith the Green-Kubo diffusion constant. Exact analytical results are corroborated by Monte Carlo simulations.

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

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

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

Documentation for TRACE: an interactive beam-transport code

International Nuclear Information System (INIS)

Crandall, K.R.; Rusthoi, D.P.

1985-01-01

TRACE is an interactive, first-order, beam-dynamics computer program. TRACE includes space-charge forces and mathematical models for a number of beamline elements not commonly found in beam-transport codes, such as permanent-magnet quadrupoles, rf quadrupoles, rf gaps, accelerator columns, and accelerator tanks. TRACE provides an immediate graphic display of calculative results, has a powerful and easy-to-use command procedure, includes eight different types of beam-matching or -fitting capabilities, and contains its own internal HELP package. This report describes the models and equations used for each of the transport elements, the fitting procedures, and the space-charge/emittance calculations, and provides detailed instruction for using the code

Deterministic geologic processes and stochastic modeling

International Nuclear Information System (INIS)

Rautman, C.A.; Flint, A.L.

1992-01-01

This paper reports that recent outcrop sampling at Yucca Mountain, Nevada, has produced significant new information regarding the distribution of physical properties at the site of a potential high-level nuclear waste repository. consideration of the spatial variability indicates that her are a number of widespread deterministic geologic features at the site that have important implications for numerical modeling of such performance aspects as ground water flow and radionuclide transport. Because the geologic processes responsible for formation of Yucca Mountain are relatively well understood and operate on a more-or-less regional scale, understanding of these processes can be used in modeling the physical properties and performance of the site. Information reflecting these deterministic geologic processes may be incorporated into the modeling program explicitly using geostatistical concepts such as soft information, or implicitly, through the adoption of a particular approach to modeling

SCALE6 Hybrid Deterministic-Stochastic Shielding Methodology for PWR Containment Calculations

International Nuclear Information System (INIS)

Matijevic, Mario; Pevec, Dubravko; Trontl, Kresimir

2014-01-01

The capabilities and limitations of SCALE6/MAVRIC hybrid deterministic-stochastic shielding methodology (CADIS and FW-CADIS) are demonstrated when applied to a realistic deep penetration Monte Carlo (MC) shielding problem of full-scale PWR containment model. The ultimate goal of such automatic variance reduction (VR) techniques is to achieve acceptable precision for the MC simulation in reasonable time by preparation of phase-space VR parameters via deterministic transport theory methods (discrete ordinates SN) by generating space-energy mesh-based adjoint function distribution. The hybrid methodology generates VR parameters that work in tandem (biased source distribution and importance map) in automated fashion which is paramount step for MC simulation of complex models with fairly uniform mesh tally uncertainties. The aim in this paper was determination of neutron-gamma dose rate distribution (radiation field) over large portions of PWR containment phase-space with uniform MC uncertainties. The sources of ionizing radiation included fission neutrons and gammas (reactor core) and gammas from activated two-loop coolant. Special attention was given to focused adjoint source definition which gave improved MC statistics in selected materials and/or regions of complex model. We investigated benefits and differences of FW-CADIS over CADIS and manual (i.e. analog) MC simulation of particle transport. Computer memory consumption by deterministic part of hybrid methodology represents main obstacle when using meshes with millions of cells together with high SN/PN parameters, so optimization of control and numerical parameters of deterministic module plays important role for computer memory management. We investigated the possibility of using deterministic module (memory intense) with broad group library v7 2 7n19g opposed to fine group library v7 2 00n47g used with MC module to fully take effect of low energy particle transport and secondary gamma emission. Compared with

Deterministic methods for sensitivity and uncertainty analysis in large-scale computer models

International Nuclear Information System (INIS)

Worley, B.A.; Oblow, E.M.; Pin, F.G.; Maerker, R.E.; Horwedel, J.E.; Wright, R.Q.; Lucius, J.L.

1987-01-01

The fields of sensitivity and uncertainty analysis are dominated by statistical techniques when large-scale modeling codes are being analyzed. This paper reports on the development and availability of two systems, GRESS and ADGEN, that make use of computer calculus compilers to automate the implementation of deterministic sensitivity analysis capability into existing computer models. This automation removes the traditional limitation of deterministic sensitivity methods. The paper describes a deterministic uncertainty analysis method (DUA) that uses derivative information as a basis to propagate parameter probability distributions to obtain result probability distributions. The paper demonstrates the deterministic approach to sensitivity and uncertainty analysis as applied to a sample problem that models the flow of water through a borehole. The sample problem is used as a basis to compare the cumulative distribution function of the flow rate as calculated by the standard statistical methods and the DUA method. The DUA method gives a more accurate result based upon only two model executions compared to fifty executions in the statistical case

A benchmark comparison of the Canadian Supercritical Water-Cooled Reactor (SCWR) 64-element fuel lattice cell parameters using various computer codes

Energy Technology Data Exchange (ETDEWEB)

Sharpe, J.; Salaun, F.; Hummel, D.; Moghrabi, A., E-mail: sharpejr@mcmaster.ca [McMaster University, Hamilton, ON (Canada); Nowak, M. [McMaster University, Hamilton, ON (Canada); Institut National Polytechnique de Grenoble, Phelma, Grenoble (France); Pencer, J. [McMaster University, Hamilton, ON (Canada); Canadian Nuclear Laboratories, Chalk River, ON, (Canada); Novog, D.; Buijs, A. [McMaster University, Hamilton, ON (Canada)

2015-07-01

Discrepancies in key lattice physics parameters have been observed between various deterministic (e.g. DRAGON and WIMS-AECL) and stochastic (MCNP, KENO) neutron transport codes in modeling previous versions of the Canadian SCWR lattice cell. Further, inconsistencies in these parameters have also been observed when using different nuclear data libraries. In this work, the predictions of k∞, various reactivity coefficients, and relative ring-averaged pin powers have been re-evaluated using these codes and libraries with the most recent 64-element fuel assembly geometry. A benchmark problem has been defined to quantify the dissimilarities between code results for a number of responses along the fuel channel under prescribed hot full power (HFP), hot zero power (HZP) and cold zero power (CZP) conditions and at several fuel burnups (0, 25 and 50 MW·d·kg{sup -1} [HM]). Results from deterministic (TRITON, DRAGON) and stochastic codes (MCNP6, KENO V.a and KENO-VI) are presented. (author)

A benchmark comparison of the Canadian Supercritical Water-Cooled Reactor (SCWR) 64-element fuel lattice cell parameters using various computer codes

International Nuclear Information System (INIS)

Sharpe, J.; Salaun, F.; Hummel, D.; Moghrabi, A.; Nowak, M.; Pencer, J.; Novog, D.; Buijs, A.

2015-01-01

Discrepancies in key lattice physics parameters have been observed between various deterministic (e.g. DRAGON and WIMS-AECL) and stochastic (MCNP, KENO) neutron transport codes in modeling previous versions of the Canadian SCWR lattice cell. Further, inconsistencies in these parameters have also been observed when using different nuclear data libraries. In this work, the predictions of k∞, various reactivity coefficients, and relative ring-averaged pin powers have been re-evaluated using these codes and libraries with the most recent 64-element fuel assembly geometry. A benchmark problem has been defined to quantify the dissimilarities between code results for a number of responses along the fuel channel under prescribed hot full power (HFP), hot zero power (HZP) and cold zero power (CZP) conditions and at several fuel burnups (0, 25 and 50 MW·d·kg"-"1 [HM]). Results from deterministic (TRITON, DRAGON) and stochastic codes (MCNP6, KENO V.a and KENO-VI) are presented. (author)

Code of Practice for the safe transport of radioactive substances 1990

International Nuclear Information System (INIS)

1990-01-01

This Federal Code revises an earlier Code on the same subject issued in 1982 and was formulated under the Environment Protection (Nuclear Codes) Act 1978. The purpose of the Code is to establish uniform safety standards, applicable throughout the Commonwealth of Australia, to provide for the protection of persons and the environment, against any dangers associated with the transport of radioactive substances. The Code uses as a basis the IAEA Regulations for the Safe Transport of Radioactive Materials. This new edition takes into account the 1985 Edition of the Regulations incorporating the 1988 Supplement and provides, furthermore, that radiation protection standards will also be subject to recommendations of the Australian National Health and Medical Research Council [fr

HYDRASTAR - a code for stochastic simulation of groundwater flow

International Nuclear Information System (INIS)

Norman, S.

1992-05-01

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

Development of a tracer transport option for the NAPSAC fracture network computer code

International Nuclear Information System (INIS)

Herbert, A.W.

1990-06-01

The Napsac computer code predicts groundwater flow through fractured rock using a direct fracture network approach. This paper describes the development of a tracer transport algorithm for the NAPSAC code. A very efficient particle-following approach is used enabling tracer transport to be predicted through large fracture networks. The new algorithm is tested against three test examples. These demonstrations confirm the accuracy of the code for simple networks, where there is an analytical solution to the transport problem, and illustrates the use of the computer code on a more realistic problem. (author)

Multidimensional electron-photon transport with standard discrete ordinates codes

International Nuclear Information System (INIS)

Drumm, C.R.

1995-01-01

A method is described for generating electron cross sections that are compatible with standard discrete ordinates codes without modification. There are many advantages of using an established discrete ordinates solver, e.g. immediately available adjoint capability. Coupled electron-photon transport capability is needed for many applications, including the modeling of the response of electronics components to space and man-made radiation environments. The cross sections have been successfully used in the DORT, TWODANT and TORT discrete ordinates codes. The cross sections are shown to provide accurate and efficient solutions to certain multidimensional electronphoton transport problems

Probabilistic evaluations for CANTUP computer code analysis improvement

International Nuclear Information System (INIS)

Florea, S.; Pavelescu, M.

2004-01-01

Structural analysis with finite element method is today an usual way to evaluate and predict the behavior of structural assemblies subject to hard conditions in order to ensure their safety and reliability during their operation. A CANDU 600 fuel channel is an example of an assembly working in hard conditions, in which, except the corrosive and thermal aggression, long time irradiation, with implicit consequences on material properties evolution, interferes. That leads inevitably to material time-dependent properties scattering, their dynamic evolution being subject to a great degree of uncertainness. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods in order to predict the structural component response. This work initiates the possibility to extend the deterministic thermomechanical evaluation on fuel channel components to probabilistic structural mechanics approach starting with deterministic analysis performed with CANTUP computer code which is a code developed to predict the long term mechanical behavior of the pressure tube - calandria tube assembly. To this purpose the structure of deterministic calculus CANTUP computer code has been reviewed. The code has been adapted from LAHEY 77 platform to Microsoft Developer Studio - Fortran Power Station platform. In order to perform probabilistic evaluations, it was added a part to the deterministic code which, using a subroutine from IMSL library from Microsoft Developer Studio - Fortran Power Station platform, generates pseudo-random values of a specified value. It was simulated a normal distribution around the deterministic value and 5% standard deviation for Young modulus material property in order to verify the statistical calculus of the creep behavior. The tube deflection and effective stresses were the properties subject to probabilistic evaluation. All the values of these properties obtained for all the values for

User's manual for the Oak Ridge Tokamak Transport Code

Energy Technology Data Exchange (ETDEWEB)

Munro, J.K.; Hogan, J.T.; Howe, H.C.; Arnurius, D.E.

1977-02-01

A one-dimensional tokamak transport code is described which simulates a plasma discharge using a fluid model which includes power balances for electrons and ions, conservation of mass, and Maxwell's equations. The modular structure of the code allows a user to add models of various physical processes which can modify the discharge behavior. Such physical processes treated in the version of the code described here include effects of plasma transport, neutral gas transport, impurity diffusion, and neutral beam injection. Each process can be modeled by a parameterized analytic formula or at least one detailed numerical calculation. The program logic of each module is presented, followed by detailed descriptions of each subroutine used by the module. The physics underlying the models is only briefly summarized. The transport code was written in IBM FORTRAN-IV and implemented on IBM 360/370 series computers at the Oak Ridge National Laboratory and on the CDC 7600 computers of the Magnetic Fusion Energy (MFE) Computing Center of the Lawrence Livermore Laboratory. A listing of the current reference version is provided on accompanying microfiche.

Application of the three-dimensional Oak Ridge transport code

International Nuclear Information System (INIS)

Rhoades, W.A.; Childs, R.L.; Emmett, M.B.; Cramer, S.N.

1984-01-01

TORT, a 3-d extension of the DOT discrete ordinates transport code is now in production use for studies of radiation penetration into large concrete and masonry structures. This paper discusses certain features of the new code and shows representative results, including comparisons with Monte Carlo calculations

One-, two- and three-dimensional transport codes using multi-group double-differential form cross sections

International Nuclear Information System (INIS)

Mori, Takamasa; Nakagawa, Masayuki; Sasaki, Makoto.

1988-11-01

We have developed a group of computer codes to realize the accurate transport calculation by using the multi-group double-differential form cross section. This type of cross section can correctly take account of the energy-angle correlated reaction kinematics. Accordingly, the transport phenomena in materials with highly anisotropic scattering are accurately calculated by using this cross section. They include the following four codes or code systems: PROF-DD : a code system to generate the multi-group double-differential form cross section library by processing basic nuclear data file compiled in the ENDF / B-IV or -V format, ANISN-DD : a one-dimensional transport code based on the discrete ordinate method, DOT-DD : a two-dimensional transport code based on the discrete ordinate method, MORSE-DD : a three-dimensional transport code based on the Monte Carlo method. In addition to these codes, several auxiliary codes have been developed to process calculated results. This report describes the calculation algorithm employed in these codes and how to use them. (author)

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

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.)

Transport code and nuclear data in intermediate energy region

Energy Technology Data Exchange (ETDEWEB)

Hasegawa, Akira; Odama, Naomitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Maekawa, F.; Ueki, K.; Kosaka, K.; Oyama, Y.

1998-11-01

We briefly reviewed the problems of intermediate energy nuclear data file and transport codes in connection with processing of the data. This is a summary of our group in the task force on JENDL High Energy File Integral Evaluation (JHEFIE). In this article we stress the necessity of the production of intermediate evaluated nuclear data file up to 3 GeV for the application of accelerator driven transmutation (ADT) system. And also we state the necessity of having our own transport code system to calculate the radiation fields using these evaluated files from the strategic points of view to keep our development of the ADT technology completely free from other conditions outside of our own such as imported codes and data with poor maintenance or unknown accuracy. (author)

Transport code and nuclear data in intermediate energy region

International Nuclear Information System (INIS)

Hasegawa, Akira; Odama, Naomitsu; Maekawa, F.; Ueki, K.; Kosaka, K.; Oyama, Y.

1998-01-01

We briefly reviewed the problems of intermediate energy nuclear data file and transport codes in connection with processing of the data. This is a summary of our group in the task force on JENDL High Energy File Integral Evaluation (JHEFIE). In this article we stress the necessity of the production of intermediate evaluated nuclear data file up to 3 GeV for the application of accelerator driven transmutation (ADT) system. And also we state the necessity of having our own transport code system to calculate the radiation fields using these evaluated files from the strategic points of view to keep our development of the ADT technology completely free from other conditions outside of our own such as imported codes and data with poor maintenance or unknown accuracy. (author)

Numerical model for two-dimensional hydrodynamics and energy transport. [VECTRA code

Energy Technology Data Exchange (ETDEWEB)

Trent, D.S.

1973-06-01

The theoretical basis and computational procedure of the VECTRA computer program are presented. VECTRA (Vorticity-Energy Code for TRansport Analysis) is designed for applying numerical simulation to a broad range of intake/discharge flows in conjunction with power plant hydrological evaluation. The code computational procedure is based on finite-difference approximation of the vorticity-stream function partial differential equations which govern steady flow momentum transport of two-dimensional, incompressible, viscous fluids in conjunction with the transport of heat and other constituents.

Regional Atmospheric Transport Code for Hanford Emission Tracking (RATCHET)

International Nuclear Information System (INIS)

Ramsdell, J.V. Jr.; Simonen, C.A.; Burk, K.W.

1994-02-01

The purpose of the Hanford Environmental Dose Reconstruction (HEDR) Project is to estimate radiation doses that individuals may have received from operations at the Hanford Site since 1944. This report deals specifically with the atmospheric transport model, Regional Atmospheric Transport Code for Hanford Emission Tracking (RATCHET). RATCHET is a major rework of the MESOILT2 model used in the first phase of the HEDR Project; only the bookkeeping framework escaped major changes. Changes to the code include (1) significant changes in the representation of atmospheric processes and (2) incorporation of Monte Carlo methods for representing uncertainty in input data, model parameters, and coefficients. To a large extent, the revisions to the model are based on recommendations of a peer working group that met in March 1991. Technical bases for other portions of the atmospheric transport model are addressed in two other documents. This report has three major sections: a description of the model, a user's guide, and a programmer's guide. These sections discuss RATCHET from three different perspectives. The first provides a technical description of the code with emphasis on details such as the representation of the model domain, the data required by the model, and the equations used to make the model calculations. The technical description is followed by a user's guide to the model with emphasis on running the code. The user's guide contains information about the model input and output. The third section is a programmer's guide to the code. It discusses the hardware and software required to run the code. The programmer's guide also discusses program structure and each of the program elements

The KFA-Version of the high-energy transport code HETC and the generalized evaluation code SIMPEL

International Nuclear Information System (INIS)

Cloth, P.; Filges, D.; Sterzenbach, G.; Armstrong, T.W.; Colborn, B.L.

1983-03-01

This document describes the updates that have been made to the high-energy transport code HETC for use in the German spallation-neutron source project SNQ. Performance and purpose of the subsidiary code SIMPEL that has been written for general analysis of the HETC output are also described. In addition means of coupling to low energy transport programs, such as the Monte-Carlo code MORSE is provided. As complete input descriptions for HETC and SIMPEL are given together with a sample problem, this document can serve as a user's manual for these two codes. The document is also an answer to the demand that has been issued by a greater community of HETC users on the ICANS-IV meeting, Oct 20-24 1980, Tsukuba-gun, Japan for a complete description of at least one single version of HETC among the many different versions that exist. (orig.)

A Deterministic Safety Assessment of a Pyro-processed Waste Repository

International Nuclear Information System (INIS)

Lee, Youn Myoung; Jeong, Jong Tae; Choi, Jong Won

2012-01-01

A GoldSim template program for a safety assessment of a hybrid-typed repository system, called 'A-KRS', in which two kinds of pyro-processed radioactive wastes, low-level metal wastes and ceramic high-level wastes that arise from the pyro-processing of PWR nuclear spent fuels are disposed of, has been developed. This program is ready both for a deterministic and probabilistic total system performance assessment which is able to evaluate nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural and manmade events, and scenarios. The A-KRS has been deterministically assessed with 5 various normal and abnormal scenarios associated with nuclide release and transport in and around the repository. Dose exposure rates to the farming exposure group have been evaluated in accordance with all the scenarios and then compared among other.

Simulation of the burnup in cell calculation using the WIMSD-5B Code considering different nuclear data libraries

International Nuclear Information System (INIS)

Tavares, Desirée Yael de Sena; Silva, Adilson Costa da; Lima, Zelmo Rodrigues de

2017-01-01

This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows to determine the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups. (author)

Simulation of the burnup in cell calculation using the WIMSD-5B Code considering different nuclear data libraries

Energy Technology Data Exchange (ETDEWEB)

Tavares, Desirée Yael de Sena; Silva, Adilson Costa da; Lima, Zelmo Rodrigues de, E-mail: zelmolima@yahoo.com.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2017-07-01

This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows to determine the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups. (author)

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.)

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.

Plasmator. A numerical code for simulation of plasma transport in Tokamaks

International Nuclear Information System (INIS)

Guasp, J.

1979-01-01

Plasmator is a flexible monodimensional numerical code for plasma transport in Tokamaks of circular cross-section, it allows neutral particle transport and impurity effects. The code leaves a total freedom in the analytical form of transport coefficients. It has been writen in Fortran-V for the UNIVAC-1100/80 from JEN and allows for the possibility of graphics for radial profiles and temporal evolution of the main plasma magnitudes, as well in three-dimensional as in two-dimensional representation either on a Calcomp plotter or in the printer. (author)

Aerosol sampling and Transport Efficiency Calculation (ASTEC) and application to surtsey/DCH aerosol sampling system: Code version 1.0: Code description and user's manual

International Nuclear Information System (INIS)

Yamano, N.; Brockmann, J.E.

1989-05-01

This report describes the features and use of the Aerosol Sampling and Transport Efficiency Calculation (ASTEC) Code. The ASTEC code has been developed to assess aerosol transport efficiency source term experiments at Sandia National Laboratories. This code also has broad application for aerosol sampling and transport efficiency calculations in general as well as for aerosol transport considerations in nuclear reactor safety issues. 32 refs., 31 figs., 7 tabs

Nonterminals, homomorphisms and codings in different variations of OL-systems. I. Deterministic systems

DEFF Research Database (Denmark)

Nielsen, Mogens; Rozenberg, Grzegorz; Salomaa, Arto

1974-01-01

The use of nonterminals versus the use of homomorphisms of different kinds in the basic types of deterministic OL-systems is studied. A rather surprising result is that in some cases the use of nonterminals produces a comparatively low generative capacity, whereas in some other cases the use of n...

Differential Space-Time Block Code Modulation for DS-CDMA Systems

Directory of Open Access Journals (Sweden)

Liu Jianhua

2002-01-01

Full Text Available A differential space-time block code (DSTBC modulation scheme is used to improve the performance of DS-CDMA systems in fast time-dispersive fading channels. The resulting scheme is referred to as the differential space-time block code modulation for DS-CDMA (DSTBC-CDMA systems. The new modulation and demodulation schemes are especially studied for the down-link transmission of DS-CDMA systems. We present three demodulation schemes, referred to as the differential space-time block code Rake (D-Rake receiver, differential space-time block code deterministic (D-Det receiver, and differential space-time block code deterministic de-prefix (D-Det-DP receiver, respectively. The D-Det receiver exploits the known information of the spreading sequences and their delayed paths deterministically besides the Rake type combination; consequently, it can outperform the D-Rake receiver, which employs the Rake type combination only. The D-Det-DP receiver avoids the effect of intersymbol interference and hence can offer better performance than the D-Det receiver.

BERMUDA-1DG: a one-dimensional photon transport code

International Nuclear Information System (INIS)

Suzuki, Tomoo; Hasegawa, Akira; Nakashima, Hiroshi; Kaneko, Kunio.

1984-10-01

A one-dimensional photon transport code BERMUDA-1DG has been developed for spherical and infinite slab geometries. The purpose of development is to equip the function of gamma rays calculation for the BERMUDA code system, which was developed by 1983 only for neutron transport calculation as a preliminary version. A group constants library has been prepared for 30 nuclides, and it now consists of the 36-group total cross sections and secondary gamma ray yields by the 120-group neutron flux. For the Compton scattering, group-angle transfer matrices are accurately obtained by integrating the Klein-Nishina formula taking into account the energy and scattering angle correlation. The pair production cross sections are now calculated in the code from atomic number and midenergy of each group. To obtain angular flux distribution, the transport equation is solved in the same way as in case of neutron, using the direct integration method in a multigroup model. Both of an independent gamma ray source problem and a neutron-gamma source problem are possible to be solved. This report is written as a user's manual with a brief description of the calculational method. (author)

FTR power-to-melt study. Phase I. Evaluation of deterministic models

International Nuclear Information System (INIS)

1978-09-01

SIEX is an HEDL fuel thermal performance code. It is designed to be a fast running code for steady state thermal performance analysis of coolant, cladding and fuel temperature throughout the history of a fuel element. The need to have a good predictive model coupled with short running time in the probabilistic analysis has made SIEX one of the potential deterministic models to be adopted. The probabilistic code to be developed will be a general thermal performance code acceptable on a national basis. It is, therefore, necessary to ensure that the physical model meets the requirements of an analytical tool. Since SIEX incorporates some physically-based correlated models, its general validity and limitations should be evaluated before being adopted

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

RADTRAN 5: A computer code for transportation risk analysis

International Nuclear Information System (INIS)

Neuhauser, K.S.; Kanipe, F.L.

1991-01-01

RADTRAN 5 is a computer code developed at Sandia National Laboratories (SNL) in Albuquerque, NM, to estimate radiological and nonradiological risks of radioactive materials transportation. RADTRAN 5 is written in ANSI Standard FORTRAN 77 and contains significant advances in the methodology for route-specific analysis first developed by SNL for RADTRAN 4 (Neuhauser and Kanipe, 1992). Like the previous RADTRAN codes, RADTRAN 5 contains two major modules for incident-free and accident risk amlysis, respectively. All commercially important transportation modes may be analyzed with RADTRAN 5: highway by combination truck; highway by light-duty vehicle; rail; barge; ocean-going ship; cargo air; and passenger air

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)

Applications of the Los Alamos High Energy Transport code

International Nuclear Information System (INIS)

Waters, L.; Gavron, A.; Prael, R.E.

1992-01-01

Simulation codes reliable through a large range of energies are essential to analyze the environment of vehicles and habitats proposed for space exploration. The LAHET monte carlo code has recently been expanded to track high energy hadrons with FLUKA, while retaining the original Los Alamos version of HETC at lower energies. Electrons and photons are transported with EGS4, and an interface to the MCNP monte carlo code is provided to analyze neutrons with kinetic energies less than 20 MeV. These codes are benchmarked by comparison of LAHET/MCNP calculations to data from the Brookhaven experiment E814 participant calorimeter

Library system for a one dimensional tokamak transport code: (LIBJT60), 1

International Nuclear Information System (INIS)

Hirayama, Toshio

1982-12-01

A library system is developed to control and manage huge programs in terms of FORTRAN source. It is applied to widely used one dimensional tokamak transport codes (LIBJT60), which have been developed in the Division of Large Tokamak Development. The structure of data and program in the transport code turn out to be flexible enough to respond to various demands and this gigantic code frame work can be decomposed into groups of a compact code with a specific function. Some editing support tools for programming and debugging are also developed to save programming work. By applying this library system, users can obtain a code whose functions can be efficiently developed. (author)

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

The Initial Atmospheric Transport (IAT) Code: Description and Validation

Energy Technology Data Exchange (ETDEWEB)

Morrow, Charles W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bartel, Timothy James [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-10-01

The Initial Atmospheric Transport (IAT) computer code was developed at Sandia National Laboratories as part of their nuclear launch accident consequences analysis suite of computer codes. The purpose of IAT is to predict the initial puff/plume rise resulting from either a solid rocket propellant or liquid rocket fuel fire. The code generates initial conditions for subsequent atmospheric transport calculations. The Initial Atmospheric Transfer (IAT) code has been compared to two data sets which are appropriate to the design space of space launch accident analyses. The primary model uncertainties are the entrainment coefficients for the extended Taylor model. The Titan 34D accident (1986) was used to calibrate these entrainment settings for a prototypic liquid propellant accident while the recent Johns Hopkins University Applied Physics Laboratory (JHU/APL, or simply APL) large propellant block tests (2012) were used to calibrate the entrainment settings for prototypic solid propellant accidents. North American Meteorology (NAM )formatted weather data profiles are used by IAT to determine the local buoyancy force balance. The IAT comparisons for the APL solid propellant tests illustrate the sensitivity of the plume elevation to the weather profiles; that is, the weather profile is a dominant factor in determining the plume elevation. The IAT code performed remarkably well and is considered validated for neutral weather conditions.

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...

On the use of the Serpent Monte Carlo code for few-group cross section generation

International Nuclear Information System (INIS)

Fridman, E.; Leppaenen, J.

2011-01-01

Research highlights: → B1 methodology was used for generation of leakage-corrected few-group cross sections in the Serpent Monte-Carlo code. → Few-group constants generated by Serpent were compared with those calculated by Helios deterministic lattice transport code. → 3D analysis of a PWR core was performed by a nodal diffusion code DYN3D employing two-group cross section sets generated by Serpent and Helios. → An excellent agreement in the results of 3D core calculations obtained with Helios and Serpent generated cross-section libraries was observed. - Abstract: Serpent is a recently developed 3D continuous-energy Monte Carlo (MC) reactor physics burnup calculation code. Serpent is specifically designed for lattice physics applications including generation of homogenized few-group constants for full-core core simulators. Currently in Serpent, the few-group constants are obtained from the infinite-lattice calculations with zero neutron current at the outer boundary. In this study, in order to account for the non-physical infinite-lattice approximation, B1 methodology, routinely used by deterministic lattice transport codes, was considered for generation of leakage-corrected few-group cross sections in the Serpent code. A preliminary assessment of the applicability of the B1 methodology for generation of few-group constants in the Serpent code was carried out according to the following steps. Initially, the two-group constants generated by Serpent were compared with those calculated by Helios deterministic lattice transport code. Then, a 3D analysis of a Pressurized Water Reactor (PWR) core was performed by the nodal diffusion code DYN3D employing two-group cross section sets generated by Serpent and Helios. At this stage thermal-hydraulic (T-H) feedback was neglected. The DYN3D results were compared with those obtained from the 3D full core Serpent MC calculations. Finally, the full core DYN3D calculations were repeated taking into account T-H feedback and

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

International Nuclear Information System (INIS)

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

2009-09-01

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

MIGFRAC - a code for modelling of radionuclide transport in fracture media

International Nuclear Information System (INIS)

Satyanarayana, S.V.M.; Mohankumar, N.; Sasidhar, P.

2002-05-01

Radionuclides migrate through diffusion process from radioactive waste disposal facilities into fractures present in the host rock. The transport phenomenon is aided by the circulating ground waters. To model the transport of radionuclides in the charnockite rock formations present at Kalpakkam, a numerical code - MIGFRAC has been developed at SHINE Group, IGCAR. The code has been subjected to rigorous tests and the results of the build up of radionuclide concentrations are validated with a test case up to a distance of 100 meter along the fracture. The report discusses the model, code features and the results obtained up to a distance of 400 meter are presented. (author)

Application of Inverse Gamma Transport to Material Thickness Identification with SGRD Code

Directory of Open Access Journals (Sweden)

Humbert Philippe

2017-01-01

Full Text Available SGRD (Spectroscopy, Gamma rays, Rapid, Deterministic code is used to infer the dimensions of a one dimensional model of a shielded gamma ray source. The method is based on the simulation of the uncollided leakage current of discrete gamma lines that are produced by nuclear decay. Experimentally, the unscattered gamma lines leakage current is obtained by processing high precision gamma spectroscopy measurements. The material thicknesses are computed with SGRD using a fast ray-tracing algorithm embedded in a non-linear multidimensional iterative optimization procedure that minimizes the error metric between calculated and measured signatures. For verification, numerical results on a test problem are presented.

Integrated transport code system for a multicomponent plasma in a gas dynamic trap

International Nuclear Information System (INIS)

Anikeev, A.V.; Karpushov, A.N.; Noak, K.; Strogalova, S.L.

2000-01-01

This report is focused on the development of the theoretical and numerical models of multicomponent high-β plasma confinement and transport in the gas-dynamic trap (GDT). In order to simulate the plasma behavior in the GDT as well as that in the GDT-based neutron source the Integrated Transport Code System is developed from existing stand-alone codes calculating the target plasma, the fast ions and the neutral gas in the GDT. The code system considers the full dependence of the transport phenomena on space, time, energy and angle variables as well as the interactions between the particle fields [ru

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...

Fast Computation of Pulse Height Spectra Using SGRD Code

Directory of Open Access Journals (Sweden)

Humbert Philippe

2017-01-01

Full Text Available SGRD (Spectroscopy, Gamma rays, Rapid, Deterministic code is used for fast calculation of the gamma ray spectrum produced by a spherical shielded source and measured by a detector. The photon source lines originate from the radioactive decay of the unstable isotopes. The emission rate and spectrum of these primary sources are calculated using the DARWIN code. The leakage spectrum is separated in two parts, the uncollided component is transported by ray-tracing and the scattered component is calculated using a multigroup discrete ordinates method. The pulsed height spectrum is then simulated by folding the leakage spectrum with the detector response functions which are pre-calculated using MCNP5 code for each considered detector type. An application to the simulation of the gamma spectrum produced by a natural uranium ball coated with plexiglass and measured using a NaI detector is presented.

Coupling of system thermal–hydraulics and Monte-Carlo code: Convergence criteria and quantification of correlation between statistical uncertainty and coupled error

International Nuclear Information System (INIS)

Wu, Xu; Kozlowski, Tomasz

2015-01-01

Highlights: • Coupling of Monte Carlo code Serpent and thermal–hydraulics code RELAP5. • A convergence criterion is developed based on the statistical uncertainty of power. • Correlation between MC statistical uncertainty and coupled error is quantified. • Both UO 2 and MOX single assembly models are used in the coupled simulation. • Validation of coupling results with a multi-group transport code DeCART. - Abstract: Coupled multi-physics approach plays an important role in improving computational accuracy. Compared with deterministic neutronics codes, Monte Carlo codes have the advantage of a higher resolution level. In the present paper, a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, Serpent, is coupled with a thermal–hydraulics safety analysis code, RELAP5. The coupled Serpent/RELAP5 code capability is demonstrated by the improved axial power distribution of UO 2 and MOX single assembly models, based on the OECD-NEA/NRC PWR MOX/UO 2 Core Transient Benchmark. Comparisons of calculation results using the coupled code with those from the deterministic methods, specifically heterogeneous multi-group transport code DeCART, show that the coupling produces more precise results. A new convergence criterion for the coupled simulation is developed based on the statistical uncertainty in power distribution in the Monte Carlo code, rather than ad-hoc criteria used in previous research. The new convergence criterion is shown to be more rigorous, equally convenient to use but requiring a few more coupling steps to converge. Finally, the influence of Monte Carlo statistical uncertainty on the coupled error of power and thermal–hydraulics parameters is quantified. The results are presented such that they can be used to find the statistical uncertainty to use in Monte Carlo in order to achieve a desired precision in coupled simulation

Code-To-Code Benchmarking Of The Porflow And GoldSim Contaminant Transport Models Using A Simple 1-D Domain - 11191

International Nuclear Information System (INIS)

Hiergesell, R.; Taylor, G.

2010-01-01

An investigation was conducted to compare and evaluate contaminant transport results of two model codes, GoldSim and Porflow, using a simple 1-D string of elements in each code. Model domains were constructed to be identical with respect to cell numbers and dimensions, matrix material, flow boundary and saturation conditions. One of the codes, GoldSim, does not simulate advective movement of water; therefore the water flux term was specified as a boundary condition. In the other code, Porflow, a steady-state flow field was computed and contaminant transport was simulated within that flow-field. The comparisons were made solely in terms of the ability of each code to perform contaminant transport. The purpose of the investigation was to establish a basis for, and to validate follow-on work that was conducted in which a 1-D GoldSim model developed by abstracting information from Porflow 2-D and 3-D unsaturated and saturated zone models and then benchmarked to produce equivalent contaminant transport results. A handful of contaminants were selected for the code-to-code comparison simulations, including a non-sorbing tracer and several long- and short-lived radionuclides exhibiting both non-sorbing to strongly-sorbing characteristics with respect to the matrix material, including several requiring the simulation of in-growth of daughter radionuclides. The same diffusion and partitioning coefficients associated with each contaminant and the half-lives associated with each radionuclide were incorporated into each model. A string of 10-elements, having identical spatial dimensions and properties, were constructed within each code. GoldSim's basic contaminant transport elements, Mixing cells, were utilized in this construction. Sand was established as the matrix material and was assigned identical properties (e.g. bulk density, porosity, saturated hydraulic conductivity) in both codes. Boundary conditions applied included an influx of water at the rate of 40 cm/yr at one

Development and validation of ALEPH Monte Carlo burn-up code

International Nuclear Information System (INIS)

Stankovskiy, A.; Van den Eynde, G.; Vidmar, T.

2011-01-01

The Monte-Carlo burn-up code ALEPH is being developed in SCK-CEN since 2004. Belonging to the category of shells coupling Monte Carlo transport (MCNP or MCNPX) and 'deterministic' depletion codes (ORIGEN-2.2), ALEPH possess some unique features that distinguish it from other codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. Recent improvements of ALEPH concern full implementation of general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII, JENDL-3.3). The upgraded version of the code is capable to treat isomeric branching ratios, neutron induced fission product yields, spontaneous fission yields and energy release per fission recorded in ENDF-formatted data files. The alternative algorithm for time evolution of nuclide concentrations is added. A predictor-corrector mechanism and the calculation of nuclear heating are available as well. The validation of the code on REBUS experimental programme results has been performed. The upgraded version of ALEPH has shown better agreement with measured data than other codes, including previous version of ALEPH. (authors)

RADTRAN II: revised computer code to analyze transportation of radioactive material

International Nuclear Information System (INIS)

Taylor, J.M.; Daniel, S.L.

1982-10-01

A revised and updated version of the RADTRAN computer code is presented. This code has the capability to predict the radiological impacts associated with specific schemes of radioactive material shipments and mode specific transport variables

Validation of comprehensive space radiation transport code

International Nuclear Information System (INIS)

Shinn, J.L.; Simonsen, L.C.; Cucinotta, F.A.

1998-01-01

The HZETRN code has been developed over the past decade to evaluate the local radiation fields within sensitive materials on spacecraft in the space environment. Most of the more important nuclear and atomic processes are now modeled and evaluation within a complex spacecraft geometry with differing material components, including transition effects across boundaries of dissimilar materials, are included. The atomic/nuclear database and transport procedures have received limited validation in laboratory testing with high energy ion beams. The codes have been applied in design of the SAGE-III instrument resulting in material changes to control injurious neutron production, in the study of the Space Shuttle single event upsets, and in validation with space measurements (particle telescopes, tissue equivalent proportional counters, CR-39) on Shuttle and Mir. The present paper reviews the code development and presents recent results in laboratory and space flight validation

Aqueous Transport Code Revisions Using Geographic Information Systems

International Nuclear Information System (INIS)

Chen, K.F.

2003-01-01

STREAM II, developed at the Savannah River Site (SRS) for execution on a personal computer, is an emergency response code that predicts downstream pollutant concentrations for releases from the SRS area to the Savannah River for emergency response management decision making. The STREAM II code consists of pre-processor, calculation, and post-processor modules. The pre-processor module provides a graphical user interface (GUI) for inputting the initial release data. The GUI passes the user specified data to the calculation module that calculates the pollutant concentrations at downstream locations and the transport times. The calculation module of the STREAM II adopts the transport module of the WASP5 code. WASP5 is a US Environmental Protection Agency water quality analysis program that simulates pollutant transport and fate through surface water using a finite difference method to solve the transport equation. The calculated downstream pollutant concentrations and travel times a re passed to the post-processor for display on the computer screen in graphical and tabular forms. To minimize the user's effort in the emergency situation, the required input parameters are limited to the time and date of release, type of release, location of release, amount and duration of release, and the calculation units. The user, however, could only select one of the seventeen predetermined locations. Hence, STREAM II could not be used for situations in which release locations differ from the seventeen predetermined locations. To eliminate this limitation, STREAM II has been revised to allow users to select the release location anywhere along the specified SRS main streams or the Savannah River by mouse-selection from a map displayed on the computer monitor. The required modifications to STREAM II using geographic information systems (GIS) software is discussed in this paper

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.

Self-Shielding Treatment to Perform Cell Calculation for Seed Furl In Th/U Pwr Using Dragon Code

Directory of Open Access Journals (Sweden)

Ahmed Amin El Said Abd El Hameed

2015-08-01

Full Text Available Time and precision of the results are the most important factors in any code used for nuclear calculations. Despite of the high accuracy of Monte Carlo codes, MCNP and Serpent, in many cases their relatively long computational time leads to difficulties in using any of them as the main calculation code. Usually, Monte Carlo codes are used only to benchmark the results. The deterministic codes, which are usually used in nuclear reactor’s calculations, have limited precision, due to the approximations in the methods used to solve the multi-group transport equation. Self- Shielding treatment, an algorithm that produces an average cross-section defined over the complete energy domain of the neutrons in a nuclear reactor, is responsible for the biggest error in any deterministic codes. There are mainly two resonance self-shielding models commonly applied: models based on equivalence and dilution and models based on subgroup approach. The fundamental problem with any self-shielding method is that it treats any isotope as there are no other isotopes with resonance present in the reactor. The most practical way to solve this problem is to use multi-energy groups (50-200 that are chosen in a way that allows us to use all major resonances without self-shielding. In this paper, we perform cell calculations, for a fresh seed fuel pin which is used in thorium/uranium reactors, by solving 172 energy group transport equation using the deterministic DRAGON code, for the two types of self-shielding models (equivalence and dilution models and subgroup models Using WIMS-D5 and DRAGON data libraries. The results are then tested by comparing it with the stochastic MCNP5 code.  We also tested the sensitivity of the results to a specific change in self-shielding method implemented, for example the effect of applying Livolant-Jeanpierre Normalization scheme and Rimman Integration improvement on the equivalence and dilution method, and the effect of using Ribbon

FLAME: A finite element computer code for contaminant transport n variably-saturated media

International Nuclear Information System (INIS)

Baca, R.G.; Magnuson, S.O.

1992-06-01

A numerical model was developed for use in performance assessment studies at the INEL. The numerical model referred to as the FLAME computer code, is designed to simulate subsurface contaminant transport in a variably-saturated media. The code can be applied to model two-dimensional contaminant transport in an and site vadose zone or in an unconfined aquifer. In addition, the code has the capability to describe transport processes in a porous media with discrete fractures. This report presents the following: description of the conceptual framework and mathematical theory, derivations of the finite element techniques and algorithms, computational examples that illustrate the capability of the code, and input instructions for the general use of the code. The development of the FLAME computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of energy Order 5820.2A

FLAME: A finite element computer code for contaminant transport n variably-saturated media

Energy Technology Data Exchange (ETDEWEB)

Baca, R.G.; Magnuson, S.O.

1992-06-01

A numerical model was developed for use in performance assessment studies at the INEL. The numerical model referred to as the FLAME computer code, is designed to simulate subsurface contaminant transport in a variably-saturated media. The code can be applied to model two-dimensional contaminant transport in an and site vadose zone or in an unconfined aquifer. In addition, the code has the capability to describe transport processes in a porous media with discrete fractures. This report presents the following: description of the conceptual framework and mathematical theory, derivations of the finite element techniques and algorithms, computational examples that illustrate the capability of the code, and input instructions for the general use of the code. The development of the FLAME computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of energy Order 5820.2A.

Validation of the transportation computer codes HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND

International Nuclear Information System (INIS)

Maheras, S.J.; Pippen, H.K.

1995-05-01

The computer codes HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND were used to estimate radiation doses from the transportation of radioactive material in the Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Environmental Impact Statement. HIGHWAY and INTERLINE were used to estimate transportation routes for truck and rail shipments, respectively. RADTRAN 4 was used to estimate collective doses from incident-free transportation and the risk (probability x consequence) from transportation accidents. RISKIND was used to estimate incident-free radiation doses for maximally exposed individuals and the consequences from reasonably foreseeable transportation accidents. The purpose of this analysis is to validate the estimates made by these computer codes; critiques of the conceptual models used in RADTRAN 4 are also discussed. Validation is defined as ''the test and evaluation of the completed software to ensure compliance with software requirements.'' In this analysis, validation means that the differences between the estimates generated by these codes and independent observations are small (i.e., within the acceptance criterion established for the validation analysis). In some cases, the independent observations used in the validation were measurements; in other cases, the independent observations used in the validation analysis were generated using hand calculations. The results of the validation analyses performed for HIGHWAY, INTERLINE, RADTRAN 4, and RISKIND show that the differences between the estimates generated using the computer codes and independent observations were small. Based on the acceptance criterion established for the validation analyses, the codes yielded acceptable results; in all cases the estimates met the requirements for successful validation

Towards a heavy-ion transport capability in the MARS15 Code

International Nuclear Information System (INIS)

Mokhov, N.V.; Gudima, K.K.; Mashnik, S.G.; Rakhno, I.L.; Striganov, S.

2004-01-01

In order to meet the challenges of new accelerator and space projects and further improve modelling of radiation effects in microscopic objects, heavy-ion interaction and transport physics have been recently incorporated into the MARS15 Monte Carlo code. A brief description of new modules is given in comparison with experimental data. The MARS Monte Carlo code is widely used in numerous accelerator, detector, shielding and cosmic ray applications. The needs of the Relativistic Heavy-Ion Collider, Large Hadron Collider, Rare Isotope Accelerator and NASA projects have recently induced adding heavy-ion interaction and transport physics to the MARS15 code. The key modules of the new implementation are described below along with their comparisons to experimental data.

Implementation of the kinetics in the transport code AZTRAN; Implementacion de la cinetica en el codigo de transporte AZTRAN

Energy Technology Data Exchange (ETDEWEB)

Duran G, J. A.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: redfield1290@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2017-09-15

This paper shows the implementation of the time dependence in the three-dimensional transport code AZTRAN (AZtlan TRANsport), which belongs to the AZTLAN platform, for the analysis of nuclear reactors (currently under development). The AZTRAN code with this implementation is able to numerically solve the time-dependent transport equation in XYZ geometry, for several energy groups, using the discrete ordinate method S{sub n} for the discretization of the angular variable, the nodal method RTN-0 for spatial discretization and method 0 for discretization in time. Initially, the code only solved the neutrons transport equation in steady state, so the implementation of the temporal part was made integrating the neutrons transport equation with respect to time and balance equations corresponding to the concentrations of delayed neutron precursors, for which method 0 was applied. After having directly implemented code kinetics, the improved quasi-static method was implemented, which is a tool for reducing computation time, where the angular flow is factored by the product of two functions called shape function and amplitude function, where the first is calculated for long time steps, called macro-steps and the second is resolved for small time steps called micro-steps. In the new version of AZTRAN several Benchmark problems that were taken from the literature were simulated, the problems used are of two and three dimensions which allowed corroborating the accuracy and stability of the code, showing in general in the reference tests a good behavior. (Author)

Present status of nucleon-meson transport code NMTC/JAERI

International Nuclear Information System (INIS)

Takada, H.; Meigo, S.

2001-01-01

The nucleon-meson transport code NMTC/JAM has been developed for the neutronics design study of the joint project for high-intensity proton accelerators with a power of mega-watts. The applicable energy range is extended by the inclusion of the jet AA microscopic transport model (JAM). The nucleon-nucleus cross sections are also updated for accurate transport calculation. The applicability of NMTC/JAM is studied through the analyses of thick target experiments such as neutron transmission through shield and activation reaction rate measurements. (orig.)

Development of a fuel depletion sensitivity calculation module for multi-cell problems in a deterministic reactor physics code system CBZ

International Nuclear Information System (INIS)

Chiba, Go; Kawamoto, Yosuke; Narabayashi, Tadashi

2016-01-01

Highlights: • A new functionality of fuel depletion sensitivity calculations is developed in a code system CBZ. • This is based on the generalized perturbation theory for fuel depletion problems. • The theory with a multi-layer depletion step division scheme is described. • Numerical techniques employed in actual implementation are also provided. - Abstract: A new functionality of fuel depletion sensitivity calculations is developed as one module in a deterministic reactor physics code system CBZ. This is based on the generalized perturbation theory for fuel depletion problems. The theory for fuel depletion problems with a multi-layer depletion step division scheme is described in detail. Numerical techniques employed in actual implementation are also provided. Verification calculations are carried out for a 3 × 3 multi-cell problem consisting of two different types of fuel pins. It is shown that the sensitivities of nuclide number densities after fuel depletion with respect to the nuclear data calculated by the new module agree well with reference sensitivities calculated by direct numerical differentiation. To demonstrate the usefulness of the new module, fuel depletion sensitivities in different multi-cell arrangements are compared and non-negligible differences are observed. Nuclear data-induced uncertainties of nuclide number densities obtained with the calculated sensitivities are also compared.

Transoptr-a second order beam transport design code with automatic internal optimization and general constraints

International Nuclear Information System (INIS)

Heighway, E.A.

1980-07-01

A second order beam transport design code with parametric optimization is described. The code analyzes the transport of charged particle beams through a user defined magnet system. The magnet system parameters are varied (within user defined limits) until the properties of the transported beam and/or the system transport matrix match those properties requested by the user. The code uses matrix formalism to represent the transport elements and optimization is achieved using the variable metric method. Any constraints that can be expressed algebraically may be included by the user as part of his design. Instruction in the use of the program is given. (auth)

Theory of contributon transport

International Nuclear Information System (INIS)

Painter, J.W.; Gerstl, S.A.W.; Pomraning, G.C.

1980-10-01

A general discussion of the physics of contributon transport is presented. To facilitate this discussion, a Boltzmann-like transport equation for contributons is obtained, and special contributon cross sections are defined. However, the main goal of this study is to identify contributon transport equations and investigate possible deterministic solution techniques. Four approaches to the deterministic solution of the contributon transport problem are investigated. These approaches are an attempt to exploit certain attractive properties of the contributon flux, psi = phi phi + , where phi and phi + are the solutions to the forward and adjoint Boltzmann transport equations

Deterministic Graphical Games Revisited

DEFF Research Database (Denmark)

Andersson, Daniel; Hansen, Kristoffer Arnsfelt; Miltersen, Peter Bro

2008-01-01

We revisit the deterministic graphical games of Washburn. A deterministic graphical game can be described as a simple stochastic game (a notion due to Anne Condon), except that we allow arbitrary real payoffs but disallow moves of chance. We study the complexity of solving deterministic graphical...... games and obtain an almost-linear time comparison-based algorithm for computing an equilibrium of such a game. The existence of a linear time comparison-based algorithm remains an open problem....

Exponential power spectra, deterministic chaos and Lorentzian pulses in plasma edge dynamics

International Nuclear Information System (INIS)

Maggs, J E; Morales, G J

2012-01-01

Exponential spectra have been observed in the edges of tokamaks, stellarators, helical devices and linear machines. The observation of exponential power spectra is significant because such a spectral character has been closely associated with the phenomenon of deterministic chaos by the nonlinear dynamics community. The proximate cause of exponential power spectra in both magnetized plasma edges and nonlinear dynamics models is the occurrence of Lorentzian pulses in the time signals of fluctuations. Lorentzian pulses are produced by chaotic behavior in the separatrix regions of plasma E × B flow fields or the limit cycle regions of nonlinear models. Chaotic advection, driven by the potential fields of drift waves in plasmas, results in transport. The observation of exponential power spectra and Lorentzian pulses suggests that fluctuations and transport at the edge of magnetized plasmas arise from deterministic, rather than stochastic, dynamics. (paper)

Control-oriented approaches to anticipating synchronization of chaotic deterministic ratchets

Energy Technology Data Exchange (ETDEWEB)

Xu Shiyun [State Key Lab for Turbulence and Complex Systems, Department of Mechanics and Aerospace Engineering, College of Engineering, Peking University, Beijing 100871 (China)], E-mail: xushiyun@pku.edu.cn; Yang Ying [State Key Lab for Turbulence and Complex Systems, Department of Mechanics and Aerospace Engineering, College of Engineering, Peking University, Beijing 100871 (China)], E-mail: yy@mech.pku.edu.cn; Song Lei [State Key Lab for Turbulence and Complex Systems, Department of Mechanics and Aerospace Engineering, College of Engineering, Peking University, Beijing 100871 (China)

2009-06-15

In virtue of techniques derived from nonlinear control system theory, we establish conditions under which one could obtain anticipating synchronization between two periodically driven deterministic ratchets that are able to exhibit directed transport with a finite velocity. Criteria are established in order to guarantee the anticipating synchronization property of such systems as well as characterize phase space dynamics of the ratchet transporting behaviors. These results allow one to predict the chaotic direct transport features of particles on a ratchet potential using a copy of the same system that performs as a slave, which are verified through numerical simulation.

Deterministic Earthquake Hazard Assessment by Public Agencies in California

Science.gov (United States)

Mualchin, L.

2005-12-01

Even in its short recorded history, California has experienced a number of damaging earthquakes that have resulted in new codes and other legislation for public safety. In particular, the 1971 San Fernando earthquake produced some of the most lasting results such as the Hospital Safety Act, the Strong Motion Instrumentation Program, the Alquist-Priolo Special Studies Zone Act, and the California Department of Transportation (Caltrans') fault-based deterministic seismic hazard (DSH) map. The latter product provides values for earthquake ground motions based on Maximum Credible Earthquakes (MCEs), defined as the largest earthquakes that can reasonably be expected on faults in the current tectonic regime. For surface fault rupture displacement hazards, detailed study of the same faults apply. Originally, hospital, dam, and other critical facilities used seismic design criteria based on deterministic seismic hazard analyses (DSHA). However, probabilistic methods grew and took hold by introducing earthquake design criteria based on time factors and quantifying "uncertainties", by procedures such as logic trees. These probabilistic seismic hazard analyses (PSHA) ignored the DSH approach. Some agencies were influenced to adopt only the PSHA method. However, deficiencies in the PSHA method are becoming recognized, and the use of the method is now becoming a focus of strong debate. Caltrans is in the process of producing the fourth edition of its DSH map. The reason for preferring the DSH method is that Caltrans believes it is more realistic than the probabilistic method for assessing earthquake hazards that may affect critical facilities, and is the best available method for insuring public safety. Its time-invariant values help to produce robust design criteria that are soundly based on physical evidence. And it is the method for which there is the least opportunity for unwelcome surprises.

Steady-State Gyrokinetics Transport Code (SSGKT), A Scientific Application Partnership with the Framework Application for Core-Edge Transport Simulations, Final Report

Energy Technology Data Exchange (ETDEWEB)

Fahey, Mark R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Candy, Jeff [General Atomics, San Diego, CA (United States)

2013-11-07

This project initiated the development of TGYRO - a steady-state Gyrokinetic transport code (SSGKT) that integrates micro-scale GYRO turbulence simulations into a framework for practical multi-scale simulation of conventional tokamaks as well as future reactors. Using a lightweight master transport code, multiple independent (each massively parallel) gyrokinetic simulations are coordinated. The capability to evolve profiles using the TGLF model was also added to TGYRO and represents a more typical use-case for TGYRO. The goal of the project was to develop a steady-state Gyrokinetic transport code (SSGKT) that integrates micro-scale gyrokinetic turbulence simulations into a framework for practical multi-scale simulation of a burning plasma core ? the International Thermonuclear Experimental Reactor (ITER) in particular. This multi-scale simulation capability will be used to predict the performance (the fusion energy gain, Q) given the H-mode pedestal temperature and density. At present, projections of this type rely on transport models like GLF23, which are based on rather approximate fits to the results of linear and nonlinear simulations. Our goal is to make these performance projections with precise nonlinear gyrokinetic simulations. The method of approach is to use a lightweight master transport code to coordinate multiple independent (each massively parallel) gyrokinetic simulations using the GYRO code. This project targets the practical multi-scale simulation of a reactor core plasma in order to predict the core temperature and density profiles given the H-mode pedestal temperature and density. A master transport code will provide feedback to O(16) independent gyrokinetic simulations (each massively parallel). A successful feedback scheme offers a novel approach to predictive modeling of an important national and international problem. Success in this area of fusion simulations will allow US scientists to direct the research path of ITER over the next two

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)

Height-Deterministic Pushdown Automata

DEFF Research Database (Denmark)

Nowotka, Dirk; Srba, Jiri

2007-01-01

We define the notion of height-deterministic pushdown automata, a model where for any given input string the stack heights during any (nondeterministic) computation on the input are a priori fixed. Different subclasses of height-deterministic pushdown automata, strictly containing the class...... of regular languages and still closed under boolean language operations, are considered. Several of such language classes have been described in the literature. Here, we suggest a natural and intuitive model that subsumes all the formalisms proposed so far by employing height-deterministic pushdown automata...

Improved core-edge tokamak transport simulations with the CORSICA 2 code

International Nuclear Information System (INIS)

Tarditi, A.; Cohen, R.H.; Crotinger, J.A.

1996-01-01

The CORSICA 2 code models the nonlinear transport between the core and the edge of a tokamak plasma. The code couples a 2D axisymmetric edge/SOL model (UEDGE) to a 1D model for the radial core transport in toroidal flux coordinates (the transport module from the CORSICA 1 code). The core density and temperature profiles are joined to the flux-surface average profiles from the 2D code sufficiently inside the magnetic separatrix, at a flux surface on which the edge profiles are approximately constant. In the present version of the code, the deuterium density and electron and ion temperatures are coupled. The electron density is determined by imposing quasi-neutrality, both in the core and in the edge. The model allows the core-edge coupling of multiple ion densities while retaining a single temperature (corresponding to the equilibration value) for the all ion species. Applications of CORSICA 2 to modeling the DIII-D tokamak are discussed. This work will focus on the simulation of the L-H transition, coupling a single ion species (deuterium) and the two (electron and ion) temperatures. These simulations will employ a new self-consistent model for the L-H transition that is being implemented in the UEDGE code. Applications to the modeling of ITER ignition scenarios are also discussed. This will involve coupling a second density species (the thermal alphas), bringing the total number of coupled variables up to four. Finally, the progress in evolving the magnetic geometry is discussed. Currently, this geometry is calculated by CORSICA's MHD equilibrium module (TEQ) at the beginning of the run and fixed thereafter. However, CORSICA 1 can evolve this geometry quasistatically, and this quasistatic treatment is being extended to include the edge/SOL geometry. Recent improvements for code speed-up are also presented

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...

BOT3P - Bologna Transport Analysis Pre-Post-Processors Version 3.0

International Nuclear Information System (INIS)

Orsi, Roberto

2004-01-01

BOT3P is a set of standard FORTRAN 77 language programs developed at the ENEA-Bologna Nuclear Data Centre. BOT3P Version 1.0 was originally conceived to give the users of the DORT and TORT deterministic transport codes some useful diagnostic tools to prepare and to check their input data files. BOT3P Version 3.0 contains some important additions in the input geometrical model description, such as 'rod' and 'hexagonal' geometrical objects, respecting the exact cross-sectional area value and very suitable to describe a reactor lattice in detail. Moreover, it has extended the possibility to produce the geometrical, material distribution, and fixed neutron source data for the deterministic transport codes TWODANT and THREEDANT of the DANTSYS system and for the PARTISN code too, starting from the same input to BOT3P. When users require X-Y-Z TORT/THREEDANT/PARTISN mesh grids to be generated, BOT3P Version 3.0 produces a geometrical input for the MCNP Monte Carlo transport code also, where the MCNP cells correspond to the X-Y-Z bodies created for TORT.BOT3P Version 3.0 lets users specify areas/volumes of the model where the zone/material distribution can be defined not only by a combinatorial geometry but also by an external source, such as one originated from computerized tomography scan data (only for three-dimensional applications) and from one or more external DORT/TORT input files. BOT3P was developed on a DIGITAL UNIX ALPHA 500/333 workstation and successfully used in some complex neutron shielding and criticality benchmarks. It was also tested on Red Hat Linux 7.1 and is designed to run on most UNIX platforms. All BOT3P versions are publicly available from the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Data Bank (NEA-1627, NEA-1678)

Deterministic behavioural models for concurrency

DEFF Research Database (Denmark)

Sassone, Vladimiro; Nielsen, Mogens; Winskel, Glynn

1993-01-01

This paper offers three candidates for a deterministic, noninterleaving, behaviour model which generalizes Hoare traces to the noninterleaving situation. The three models are all proved equivalent in the rather strong sense of being equivalent as categories. The models are: deterministic labelled...... event structures, generalized trace languages in which the independence relation is context-dependent, and deterministic languages of pomsets....

Deterministic thermostats, theories of nonequilibrium systems and parallels with the ergodic condition

International Nuclear Information System (INIS)

Jepps, Owen G; Rondoni, Lamberto

2010-01-01

Deterministic 'thermostats' are mathematical tools used to model nonequilibrium steady states of fluids. The resulting dynamical systems correctly represent the transport properties of these fluids and are easily simulated on modern computers. More recently, the connection between such thermostats and entropy production has been exploited in the development of nonequilibrium fluid theories. The purpose and limitations of deterministic thermostats are discussed in the context of irreversible thermodynamics and the development of theories of nonequilibrium phenomena. We draw parallels between the development of such nonequilibrium theories and the development of notions of ergodicity in equilibrium theories. (topical review)

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)

Contaminant transport in fracture networks with heterogeneous rock matrices. The Picnic code

International Nuclear Information System (INIS)

Barten, Werner; Robinson, Peter C.

2001-02-01

In the context of safety assessment of radioactive waste repositories, complex radionuclide transport models covering key safety-relevant processes play a major role. In recent Swiss safety assessments, such as Kristallin-I, an important drawback was the limitation in geosphere modelling capability to account for geosphere heterogeneities. In marked contrast to this limitation in modelling capabilities, great effort has been put into investigating the heterogeneity of the geosphere as it impacts on hydrology. Structural geological methods have been used to look at the geometry of the flow paths on a small scale and the diffusion and sorption properties of different rock materials have been investigated. This huge amount of information could however be only partially applied in geosphere transport modelling. To make use of these investigations the 'PICNIC project' was established as a joint cooperation of PSI/Nagra and QuantiSci to provide a new geosphere transport model for Swiss safety assessment of radioactive waste repositories. The new transport code, PICNIC, can treat all processes considered in the older geosphere model RANCH MD generally used in the Kristallin-I study and, in addition, explicitly accounts for the heterogeneity of the geosphere on different spatial scales. The effects and transport phenomena that can be accounted for by PICNIC are a combination of (advective) macro-dispersion due to transport in a network of conduits (legs), micro-dispersion in single legs, one-dimensional or two-dimensional matrix diffusion into a wide range of homogeneous and heterogeneous rock matrix geometries, linear sorption of nuclides in the flow path and the rock matrix and radioactive decay and ingrowth in the case of nuclide chains. Analytical and numerical Laplace transformation methods are integrated in a newly developed hierarchical linear response concept to efficiently account for the transport mechanisms considered which typically act on extremely different

Contaminant transport in fracture networks with heterogeneous rock matrices. The Picnic code

Energy Technology Data Exchange (ETDEWEB)

Barten, Werner [Paul Scherrer Inst., CH-5232 Villigen PSI (Switzerland); Robinson, Peter C. [QuantiSci Limited, Henley-on-Thames (United Kingdom)

2001-02-01

In the context of safety assessment of radioactive waste repositories, complex radionuclide transport models covering key safety-relevant processes play a major role. In recent Swiss safety assessments, such as Kristallin-I, an important drawback was the limitation in geosphere modelling capability to account for geosphere heterogeneities. In marked contrast to this limitation in modelling capabilities, great effort has been put into investigating the heterogeneity of the geosphere as it impacts on hydrology. Structural geological methods have been used to look at the geometry of the flow paths on a small scale and the diffusion and sorption properties of different rock materials have been investigated. This huge amount of information could however be only partially applied in geosphere transport modelling. To make use of these investigations the 'PICNIC project' was established as a joint cooperation of PSI/Nagra and QuantiSci to provide a new geosphere transport model for Swiss safety assessment of radioactive waste repositories. The new transport code, PICNIC, can treat all processes considered in the older geosphere model RANCH MD generally used in the Kristallin-I study and, in addition, explicitly accounts for the heterogeneity of the geosphere on different spatial scales. The effects and transport phenomena that can be accounted for by PICNIC are a combination of (advective) macro-dispersion due to transport in a network of conduits (legs), micro-dispersion in single legs, one-dimensional or two-dimensional matrix diffusion into a wide range of homogeneous and heterogeneous rock matrix geometries, linear sorption of nuclides in the flow path and the rock matrix and radioactive decay and ingrowth in the case of nuclide chains. Analytical and numerical Laplace transformation methods are integrated in a newly developed hierarchical linear response concept to efficiently account for the transport mechanisms considered which typically act on extremely

Application of neutron/gamma transport codes for the design of explosive detection systems

International Nuclear Information System (INIS)

Elias, E.; Shayer, Z.

1994-01-01

Applications of neutron and gamma transport codes to the design of nuclear techniques for detecting concealed explosives material are discussed. The methodology of integrating radiation transport computations in the development, optimization and analysis phases of these new technologies is discussed. Transport and Monte Carlo codes are used for proof of concepts, guide the system integration, reduce the extend of experimental program and provide insight into the physical problem involved. The paper concentrates on detection techniques based on thermal and fast neutron interactions in the interrogated object. (authors). 6 refs., 1 tab., 5 figs

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.

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)

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)

Deterministic calculation of the effective delayed neutron fraction without using the adjoint neutron flux - 299

International Nuclear Information System (INIS)

Talamo, A.; Gohar, Y.; Aliberti, G.; Zhong, Z.; Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C.; Serafimovich, I.

2010-01-01

In 1997, Bretscher calculated the effective delayed neutron fraction by the k-ratio method. The Bretscher's approach is based on calculating the multiplication factor of a nuclear reactor core with and without the contribution of delayed neutrons. The multiplication factor set by the delayed neutrons (the delayed multiplication factor) is obtained as the difference between the total and the prompt multiplication factors. Bretscher evaluated the effective delayed neutron fraction as the ratio between the delayed and total multiplication factors (therefore the method is often referred to as k-ratio method). In the present work, the k-ratio method is applied by deterministic nuclear codes. The ENDF/B nuclear data library of the fuel isotopes ( 238 U and 238 U) have been processed by the NJOY code with and without the delayed neutron data to prepare multigroup WIMSD nuclear data libraries for the DRAGON code. The DRAGON code has been used for preparing the PARTISN macroscopic cross sections. This calculation methodology has been applied to the YALINA-Thermal assembly of Belarus. The assembly has been modeled and analyzed using PARTISN code with 69 energy groups and 60 different material zones. The deterministic and Monte Carlo results for the effective delayed neutron fraction obtained by the k-ratio method agree very well. The results also agree with the values obtained by using the adjoint flux. (authors)

Risk-based and deterministic regulation

International Nuclear Information System (INIS)

Fischer, L.E.; Brown, N.W.

1995-07-01

Both risk-based and deterministic methods are used for regulating the nuclear industry to protect the public safety and health from undue risk. The deterministic method is one where performance standards are specified for each kind of nuclear system or facility. The deterministic performance standards address normal operations and design basis events which include transient and accident conditions. The risk-based method uses probabilistic risk assessment methods to supplement the deterministic one by (1) addressing all possible events (including those beyond the design basis events), (2) using a systematic, logical process for identifying and evaluating accidents, and (3) considering alternative means to reduce accident frequency and/or consequences. Although both deterministic and risk-based methods have been successfully applied, there is need for a better understanding of their applications and supportive roles. This paper describes the relationship between the two methods and how they are used to develop and assess regulations in the nuclear industry. Preliminary guidance is suggested for determining the need for using risk based methods to supplement deterministic ones. However, it is recommended that more detailed guidance and criteria be developed for this purpose

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

Verification of the network flow and transport/distributed velocity (NWFT/DVM) computer code

International Nuclear Information System (INIS)

Duda, L.E.

1984-05-01

The Network Flow and Transport/Distributed Velocity Method (NWFT/DVM) computer code was developed primarily to fulfill a need for a computationally efficient ground-water flow and contaminant transport capability for use in risk analyses where, quite frequently, large numbers of calculations are required. It is a semi-analytic, quasi-two-dimensional network code that simulates ground-water flow and the transport of dissolved species (radionuclides) in a saturated porous medium. The development of this code was carried out under a program funded by the US Nuclear Regulatory Commission (NRC) to develop a methodology for assessing the risk from disposal of radioactive wastes in deep geologic formations (FIN: A-1192 and A-1266). In support to the methodology development program, the NRC has funded a separate Maintenance of Computer Programs Project (FIN: A-1166) to ensure that the codes developed under A-1192 or A-1266 remain consistent with current operating systems, are as error-free as possible, and have up-to-date documentations for reference by the NRC staff. Part of this effort would include verification and validation tests to assure that a code correctly performs the operations specified and/or is representing the processes or system for which it is intended. This document contains four verification problems for the NWFT/DVM computer code. Two of these problems are analytical verifications of NWFT/DVM where results are compared to analytical solutions. The other two are code-to-code verifications where results from NWFT/DVM are compared to those of another computer code. In all cases NWFT/DVM showed good agreement with both the analytical solutions and the results from the other code

Recommendations for computer code selection of a flow and transport code to be used in undisturbed vadose zone calculations for TWRS immobilized wastes environmental analyses

International Nuclear Information System (INIS)

VOOGD, J.A.

1999-01-01

An analysis of three software proposals is performed to recommend a computer code for immobilized low activity waste flow and transport modeling. The document uses criteria restablished in HNF-1839, ''Computer Code Selection Criteria for Flow and Transport Codes to be Used in Undisturbed Vadose Zone Calculation for TWRS Environmental Analyses'' as the basis for this analysis

The one-dimensional transport codes MAKOKOT. Presentation and directions for use

International Nuclear Information System (INIS)

Capes, H.; Mercier, C.; Morera, J.P.

1986-06-01

In this note are presented the different one-dimensional evolution codes available to date under the generic name MAKOKOT. They are six principal codes: - TRANS: for ion and electron transport; -NEUTRE: for neutrals; -IMPUR: for impurities; -ECRH: for electron cyclotron resonance; -DENT: for sawtooth modelling and analysis; -BILAN: for global verification of conservation. One supplementary code is added which is an impurity evolution code; it takes in account, in 1-D geometry, the buffer zone generated by the limiter between the hot plasma and the wall. An abundant bibliography is given. A comprehensive manner of using is given which underlines the use versatility of these codes [fr

Monte Carlo burnup codes acceleration using the correlated sampling method

International Nuclear Information System (INIS)

Dieudonne, C.

2013-01-01

For several years, Monte Carlo burnup/depletion codes have appeared, which couple Monte Carlo codes to simulate the neutron transport to deterministic methods, which handle the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3-dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the Monte Carlo solver called at each time step. In this document we present an original methodology to avoid the repetitive and time-expensive Monte Carlo simulations, and to replace them by perturbation calculations: indeed the different burnup steps may be seen as perturbations of the isotopic concentration of an initial Monte Carlo simulation. In a first time we will present this method, and provide details on the perturbative technique used, namely the correlated sampling. In a second time we develop a theoretical model to study the features of the correlated sampling method to understand its effects on depletion calculations. In a third time the implementation of this method in the TRIPOLI-4 code will be discussed, as well as the precise calculation scheme used to bring important speed-up of the depletion calculation. We will begin to validate and optimize the perturbed depletion scheme with the calculation of a REP-like fuel cell depletion. Then this technique will be used to calculate the depletion of a REP-like assembly, studied at beginning of its cycle. After having validated the method with a reference calculation we will show that it can speed-up by nearly an order of magnitude standard Monte-Carlo depletion codes. (author) [fr

Electronic spectrum of a deterministic single-donor device in silicon

International Nuclear Information System (INIS)

Fuechsle, Martin; Miwa, Jill A.; Mahapatra, Suddhasatta; Simmons, Michelle Y.; Hollenberg, Lloyd C. L.

2013-01-01

We report the fabrication of a single-electron transistor (SET) based on an individual phosphorus dopant that is deterministically positioned between the dopant-based electrodes of a transport device in silicon. Electronic characterization at mK-temperatures reveals a charging energy that is very similar to the value expected for isolated P donors in a bulk Si environment. Furthermore, we find indications for bulk-like one-electron excited states in the co-tunneling spectrum of the device, in sharp contrast to previous reports on transport through single dopants

Pseudo-deterministic Algorithms

OpenAIRE

Goldwasser , Shafi

2012-01-01

International audience; In this talk we describe a new type of probabilistic algorithm which we call Bellagio Algorithms: a randomized algorithm which is guaranteed to run in expected polynomial time, and to produce a correct and unique solution with high probability. These algorithms are pseudo-deterministic: they can not be distinguished from deterministic algorithms in polynomial time by a probabilistic polynomial time observer with black box access to the algorithm. We show a necessary an...

Detector placement optimization for cargo containers using deterministic adjoint transport examination for SNM detection

International Nuclear Information System (INIS)

McLaughlin, Trevor D.; Sjoden, Glenn E.; Manalo, Kevin L.

2011-01-01

With growing concerns over port security and the potential for illicit trafficking of SNM through portable cargo shipping containers, efforts are ongoing to reduce the threat via container monitoring. This paper focuses on answering an important question of how many detectors are necessary for adequate coverage of a cargo container considering the detection of neutrons and gamma rays. Deterministic adjoint transport calculations are performed with compressed helium- 3 polyethylene moderated neutron detectors and sodium activated cesium-iodide gamma-ray scintillation detectors on partial and full container models. Results indicate that the detector capability is dependent on source strength and potential shielding. Using a surrogate weapons grade plutonium leakage source, it was determined that for a 20 foot ISO container, five neutron detectors and three gamma detectors are necessary for adequate coverage. While a large CsI(Na) gamma detector has the potential to monitor the entire height of the container for SNM, the He-3 neutron detector is limited to roughly 1.25 m in depth. Detector blind spots are unavoidable inside the container volume unless additional measures are taken for adequate coverage. (author)

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.

Available computer codes and data for radiation transport analysis

International Nuclear Information System (INIS)

Trubey, D.K.; Maskewitz, B.F.; Roussin, R.W.

1975-01-01

The Radiation Shielding Information Center (RSIC), sponsored and supported by the Energy Research and Development Administration (ERDA) and the Defense Nuclear Agency (DNA), is a technical institute serving the radiation transport and shielding community. It acquires, selects, stores, retrieves, evaluates, analyzes, synthesizes, and disseminates information on shielding and ionizing radiation transport. The major activities include: (1) operating a computer-based information system and answering inquiries on radiation analysis, (2) collecting, checking out, packaging, and distributing large computer codes, and evaluated and processed data libraries. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results

Development of a CAD-based neutron transport code with the method of characteristics

International Nuclear Information System (INIS)

Chen Zhenping; Wang Dianxi; He Tao; Wang Guozhong; Zheng Huaqing

2012-01-01

The main problem determining whether the method of characteristics (MOC) can be used in complicated and highly heterogeneous geometry is how to combine an effective geometry processing method with MOC. In this study, a new idea making use of MCAM, which is a Mutlti-Calculation Automatic Modeling for Neutronics and Radiation Transport program developed by FDS Team, for geometry description and ray tracing of particle transport was brought forward to solve the geometry problem mentioned above. Based on the theory and approach as the foregoing statement, a two dimensional neutron transport code was developed which had been integrated into VisualBUS, developed by FDS Team. Several benchmarks were used to verify the validity of the code and the numerical results were coincident with the reference values very well, which indicated the accuracy and feasibility of the method and the MOC code. (authors)

Verification of the shift Monte Carlo code with the C5G7 reactor benchmark

International Nuclear Information System (INIS)

Sly, N. C.; Mervin, B. T.; Mosher, S. W.; Evans, T. M.; Wagner, J. C.; Maldonado, G. I.

2012-01-01

Shift is a new hybrid Monte Carlo/deterministic radiation transport code being developed at Oak Ridge National Laboratory. At its current stage of development, Shift includes a parallel Monte Carlo capability for simulating eigenvalue and fixed-source multigroup transport problems. This paper focuses on recent efforts to verify Shift's Monte Carlo component using the two-dimensional and three-dimensional C5G7 NEA benchmark problems. Comparisons were made between the benchmark eigenvalues and those output by the Shift code. In addition, mesh-based scalar flux tally results generated by Shift were compared to those obtained using MCNP5 on an identical model and tally grid. The Shift-generated eigenvalues were within three standard deviations of the benchmark and MCNP5-1.60 values in all cases. The flux tallies generated by Shift were found to be in very good agreement with those from MCNP. (authors)

Linking the plasma code EDGE2D to the neutral code NIMBUS for a self consistent transport model of the boundary

International Nuclear Information System (INIS)

De Matteis, A.

1987-01-01

This report describes the fully automatic linkage between the finite difference, two-dimensional code EDGE2D, based on the classical Braginskii partial differential equations of ion transport, and the Monte Carlo code NIMBUS, which solves the integral form of the stationary, linear Boltzmann equation for neutral transport in a plasma. The coupling has been performed for the real poloidal geometry of JET with two belt-limiters and real magnetic configurations with or without a single-null point. The new integrated system starts from the magnetic geometry computed by predictive or interpretative equilibrium codes and yields the plasma and neutrals characteristics in the edge

Deterministic calculations of radiation doses from brachytherapy seeds

International Nuclear Information System (INIS)

Reis, Sergio Carneiro dos; Vasconcelos, Vanderley de; Santos, Ana Maria Matildes dos

2009-01-01

Brachytherapy is used for treating certain types of cancer by inserting radioactive sources into tumours. CDTN/CNEN is developing brachytherapy seeds to be used mainly in prostate cancer treatment. Dose calculations play a very significant role in the characterization of the developed seeds. The current state-of-the-art of computation dosimetry relies on Monte Carlo methods using, for instance, MCNP codes. However, deterministic calculations have some advantages, as, for example, short computer time to find solutions. This paper presents a software developed to calculate doses in a two-dimensional space surrounding the seed, using a deterministic algorithm. The analysed seeds consist of capsules similar to IMC6711 (OncoSeed), that are commercially available. The exposure rates and absorbed doses are computed using the Sievert integral and the Meisberger third order polynomial, respectively. The software also allows the isodose visualization at the surface plan. The user can choose between four different radionuclides ( 192 Ir, 198 Au, 137 Cs and 60 Co). He also have to enter as input data: the exposure rate constant; the source activity; the active length of the source; the number of segments in which the source will be divided; the total source length; the source diameter; and the actual and effective source thickness. The computed results were benchmarked against results from literature and developed software will be used to support the characterization process of the source that is being developed at CDTN. The software was implemented using Borland Delphi in Windows environment and is an alternative to Monte Carlo based codes. (author)

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.

Sensitivity analysis of a low-level waste environmental transport code

International Nuclear Information System (INIS)

Hiromoto, G.

1989-01-01

Results are presented from a sensivity analysis of a computer code designed to simulate the environmental transport of radionuclides buried at shallow land waste repositories. A sensitivity analysis methodology, based on the surface response replacement and statistic sensitivity estimators, was developed to address the relative importance of the input parameters on the model output. Response surface replacement for the model was constructed by stepwise regression, after sampling input vectors from range and distribution of the input variables, and running the code to generate the associated output data. Sensitivity estimators were compute using the partial rank correlation coefficients and the standardized rank regression coefficients. The results showed that the tecniques employed in this work provides a feasible means to perform a sensitivity analysis of a general not-linear environmental radionuclides transport models. (author) [pt

Deterministic Modeling of the High Temperature Test Reactor

International Nuclear Information System (INIS)

Ortensi, J.; Cogliati, J.J.; Pope, M.A.; Ferrer, R.M.; Ougouag, A.M.

2010-01-01

Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL's current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green's Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2-3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the control

Multidimensional electron-photon transport with standard discrete ordinates codes

International Nuclear Information System (INIS)

Drumm, C.R.

1997-01-01

A method is described for generating electron cross sections that are comparable with standard discrete ordinates codes without modification. There are many advantages of using an established discrete ordinates solver, e.g. immediately available adjoint capability. Coupled electron-photon transport capability is needed for many applications, including the modeling of the response of electronics components to space and man-made radiation environments. The cross sections have been successfully used in the DORT, TWODANT and TORT discrete ordinates codes. The cross sections are shown to provide accurate and efficient solutions to certain multidimensional electron-photon transport problems. The key to the method is a simultaneous solution of the continuous-slowing-down (CSD) portion and elastic-scattering portion of the scattering source by the Goudsmit-Saunderson theory. The resulting multigroup-Legendre cross sections are much smaller than the true scattering cross sections that they represent. Under certain conditions, the cross sections are guaranteed positive and converge with a low-order Legendre expansion

Multidimensional electron-photon transport with standard discrete ordinates codes

International Nuclear Information System (INIS)

Drumm, C.R.

1997-01-01

A method is described for generating electron cross sections that are compatible with standard discrete ordinates codes without modification. There are many advantages to using an established discrete ordinates solver, e.g., immediately available adjoint capability. Coupled electron-photon transport capability is needed for many applications, including the modeling of the response of electronics components to space and synthetic radiation environments. The cross sections have been successfully used in the DORT, TWODANT, and TORT discrete ordinates codes. The cross sections are shown to provide accurate and efficient solutions to certain multidimensional electron-photon transport problems. The key to the method is a simultaneous solution of the continuous-slowing-down and elastic-scattering portions of the scattering source by the Goudsmit-Saunderson theory. The resulting multigroup-Legendre cross sections are much smaller than the true scattering cross sections that they represent. Under certain conditions, the cross sections are guaranteed positive and converge with a low-order Legendre expansion

Deterministic simulation of first-order scattering in virtual X-ray imaging

Energy Technology Data Exchange (ETDEWEB)

Freud, N. E-mail: nicolas.freud@insa-lyon.fr; Duvauchelle, P.; Pistrui-Maximean, S.A.; Letang, J.-M.; Babot, D

2004-07-01

A deterministic algorithm is proposed to compute the contribution of first-order Compton- and Rayleigh-scattered radiation in X-ray imaging. This algorithm has been implemented in a simulation code named virtual X-ray imaging. The physical models chosen to account for photon scattering are the well-known form factor and incoherent scattering function approximations, which are recalled in this paper and whose limits of validity are briefly discussed. The proposed algorithm, based on a voxel discretization of the inspected object, is presented in detail, as well as its results in simple configurations, which are shown to converge when the sampling steps are chosen sufficiently small. Simple criteria for choosing correct sampling steps (voxel and pixel size) are established. The order of magnitude of the computation time necessary to simulate first-order scattering images amounts to hours with a PC architecture and can even be decreased down to minutes, if only a profile is computed (along a linear detector). Finally, the results obtained with the proposed algorithm are compared to the ones given by the Monte Carlo code Geant4 and found to be in excellent accordance, which constitutes a validation of our algorithm. The advantages and drawbacks of the proposed deterministic method versus the Monte Carlo method are briefly discussed.

Probabilistic versus deterministic hazard assessment in liquefaction susceptible zones

Science.gov (United States)

Daminelli, Rosastella; Gerosa, Daniele; Marcellini, Alberto; Tento, Alberto

2015-04-01

Probabilistic seismic hazard assessment (PSHA), usually adopted in the framework of seismic codes redaction, is based on Poissonian description of the temporal occurrence, negative exponential distribution of magnitude and attenuation relationship with log-normal distribution of PGA or response spectrum. The main positive aspect of this approach stems into the fact that is presently a standard for the majority of countries, but there are weak points in particular regarding the physical description of the earthquake phenomenon. Factors like site effects, source characteristics like duration of the strong motion and directivity that could significantly influence the expected motion at the site are not taken into account by PSHA. Deterministic models can better evaluate the ground motion at a site from a physical point of view, but its prediction reliability depends on the degree of knowledge of the source, wave propagation and soil parameters. We compare these two approaches in selected sites affected by the May 2012 Emilia-Romagna and Lombardia earthquake, that caused widespread liquefaction phenomena unusually for magnitude less than 6. We focus on sites liquefiable because of their soil mechanical parameters and water table level. Our analysis shows that the choice between deterministic and probabilistic hazard analysis is strongly dependent on site conditions. The looser the soil and the higher the liquefaction potential, the more suitable is the deterministic approach. Source characteristics, in particular the duration of strong ground motion, have long since recognized as relevant to induce liquefaction; unfortunately a quantitative prediction of these parameters appears very unlikely, dramatically reducing the possibility of their adoption in hazard assessment. Last but not least, the economic factors are relevant in the choice of the approach. The case history of 2012 Emilia-Romagna and Lombardia earthquake, with an officially estimated cost of 6 billions

3-D extension C5G7 MOX benchmark calculation using threedant code

International Nuclear Information System (INIS)

Kim, H.Ch.; Han, Ch.Y.; Kim, J.K.; Na, B.Ch.

2005-01-01

It pursued the benchmark on deterministic 3-D MOX fuel assembly transport calculations without spatial homogenization (C5G7 MOX Benchmark Extension). The goal of this benchmark is to provide a more through test results for the abilities of current available 3-D methods to handle the spatial heterogeneities of reactor core. The benchmark requires solutions in the form of normalized pin powers as well as the eigenvalue for each of the control rod configurations; without rod, with A rods, and with B rods. In this work, the DANTSYS code package was applied to analyze the 3-D Extension C5G7 MOX Benchmark problems. The THREEDANT code within the DANTSYS code package, which solves the 3-D transport equation in x-y-z, and r-z-theta geometries, was employed to perform the benchmark calculations. To analyze the benchmark with the THREEDANT code, proper spatial and angular approximations were made. Several calculations were performed to investigate the effects of the different spatial approximations on the accuracy. The results from these sensitivity studies were analyzed and discussed. From the results, it is found that the 4*4 grid per pin cell is sufficiently refined so that very little benefit is obtained by increasing the mesh size. (authors)

An overview of the activities of the OECD/NEA Task Force on adapting computer codes in nuclear applications to parallel architectures

Energy Technology Data Exchange (ETDEWEB)

Kirk, B.L. [Oak Ridge National Lab., TN (United States); Sartori, E. [OCDE/OECD NEA Data Bank, Issy-les-Moulineaux (France); Viedma, L.G. de [Consejo de Seguridad Nuclear, Madrid (Spain)

1997-06-01

Subsequent to the introduction of High Performance Computing in the developed countries, the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) created the Task Force on Adapting Computer Codes in Nuclear Applications to Parallel Architectures (under the guidance of the Nuclear Science Committee`s Working Party on Advanced Computing) to study the growth area in supercomputing and its applicability to the nuclear community`s computer codes. The result has been four years of investigation for the Task Force in different subject fields - deterministic and Monte Carlo radiation transport, computational mechanics and fluid dynamics, nuclear safety, atmospheric models and waste management.

An overview of the activities of the OECD/NEA Task Force on adapting computer codes in nuclear applications to parallel architectures

International Nuclear Information System (INIS)

Kirk, B.L.; Sartori, E.; Viedma, L.G. de

1997-01-01

Subsequent to the introduction of High Performance Computing in the developed countries, the Organization for Economic Cooperation and Development/Nuclear Energy Agency (OECD/NEA) created the Task Force on Adapting Computer Codes in Nuclear Applications to Parallel Architectures (under the guidance of the Nuclear Science Committee's Working Party on Advanced Computing) to study the growth area in supercomputing and its applicability to the nuclear community's computer codes. The result has been four years of investigation for the Task Force in different subject fields - deterministic and Monte Carlo radiation transport, computational mechanics and fluid dynamics, nuclear safety, atmospheric models and waste management

APOLLO-2: An advanced transport code for LWRs

International Nuclear Information System (INIS)

Mathonniere, G.

1995-01-01

APOLLO-2 is a fully modular code in which each module corresponds to a specific task: access to the cross-sections libraries, creation of isotopes medium or mixtures, geometry definition, self-shielding calculations, computation of multigroup collision probabilities, flux solver, depletion calculations, transport-transport or transport-diffusion equivalence process, SN calculations, etc... Modules communicate exclusively by ''objects'' containing structured data, these objects are identified and handled by user's given names. Among the major improvements offered by APOLLO-2 the modelization of the self-shielding: it is possible now to deal with a great precision, checked versus Montecarlo calculations, a fuel rod divided into several concentric rings. So the total production of Plutonium is quite better estimated than before and its radial distribution may be predicted also with a good accuracy. Thanks to the versatility of the code some reference calculations and routine ones may be compared easily because only one parameter is changed; for example the self-shielding approximations are modified, the libraries or the flux solver being exactly the same. Other interesting features have been introduced in APOLLO-2: the new isotopes JEF.2 are available in 99 and 172 energy groups libraries, the surface leakage model improves the calculation of the control rod efficiency, the flux-current method allows faster calculations, the possibility of an automatic convergence checking during the depletion calculations coupled with fully automatic corrections, heterogeneous diffusion coefficients used for voiding analysis. 17 refs, 1 tab

Sensitivity/uncertainty analysis of a borehole scenario comparing Latin Hypercube Sampling and deterministic sensitivity approaches

International Nuclear Information System (INIS)

Harper, W.V.; Gupta, S.K.

1983-10-01

A computer code was used to study steady-state flow for a hypothetical borehole scenario. The model consists of three coupled equations with only eight parameters and three dependent variables. This study focused on steady-state flow as the performance measure of interest. Two different approaches to sensitivity/uncertainty analysis were used on this code. One approach, based on Latin Hypercube Sampling (LHS), is a statistical sampling method, whereas, the second approach is based on the deterministic evaluation of sensitivities. The LHS technique is easy to apply and should work well for codes with a moderate number of parameters. Of deterministic techniques, the direct method is preferred when there are many performance measures of interest and a moderate number of parameters. The adjoint method is recommended when there are a limited number of performance measures and an unlimited number of parameters. This unlimited number of parameters capability can be extremely useful for finite element or finite difference codes with a large number of grid blocks. The Office of Nuclear Waste Isolation will use the technique most appropriate for an individual situation. For example, the adjoint method may be used to reduce the scope to a size that can be readily handled by a technique such as LHS. Other techniques for sensitivity/uncertainty analysis, e.g., kriging followed by conditional simulation, will be used also. 15 references, 4 figures, 9 tables

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...

BALDUR: a one-dimensional plasma transport code

International Nuclear Information System (INIS)

Singer, C.E.; Post, D.E.; Mikkelsen, D.R.

1986-07-01

The purpose of BALDUR is to calculate the evolution of plasma parameters in an MHD equilibrium which can be approximated by concentric circular flux surfaces. Transport of up to six species of ionized particles, of electron and ion energy, and of poloidal magnetic flux is computed. A wide variety of source terms are calculated including those due to neutral gas, fusion, and auxiliary heating. The code is primarily designed for modeling tokamak plasmas but could be adapted to other toroidal confinement systems

A Monte Carlo Code for Relativistic Radiation Transport Around Kerr Black Holes

Science.gov (United States)

Schnittman, Jeremy David; Krolik, Julian H.

2013-01-01

We present a new code for radiation transport around Kerr black holes, including arbitrary emission and absorption mechanisms, as well as electron scattering and polarization. The code is particularly useful for analyzing accretion flows made up of optically thick disks and optically thin coronae. We give a detailed description of the methods employed in the code and also present results from a number of numerical tests to assess its accuracy and convergence.

Development and application of a new deterministic method for calculating computer model result uncertainties

International Nuclear Information System (INIS)

Maerker, R.E.; Worley, B.A.

1989-01-01

Interest in research into the field of uncertainty analysis has recently been stimulated as a result of a need in high-level waste repository design assessment for uncertainty information in the form of response complementary cumulative distribution functions (CCDFs) to show compliance with regulatory requirements. The solution to this problem must obviously rely on the analysis of computer code models, which, however, employ parameters that can have large uncertainties. The motivation for the research presented in this paper is a search for a method involving a deterministic uncertainty analysis approach that could serve as an improvement over those methods that make exclusive use of statistical techniques. A deterministic uncertainty analysis (DUA) approach based on the use of first derivative information is the method studied in the present procedure. The present method has been applied to a high-level nuclear waste repository problem involving use of the codes ORIGEN2, SAS, and BRINETEMP in series, and the resulting CDF of a BRINETEMP result of interest is compared with that obtained through a completely statistical analysis

CFRX, a one-and-a-quarter-dimensional transport code for field-reversed configuration studies

International Nuclear Information System (INIS)

Hsiao Mingyuan

1989-01-01

A one-and-a-quarter-dimensional transport code, which includes radial as well as some two-dimensional effects for field-reversed configurations, is described. The set of transport equations is transformed to a set of new independent and dependent variables and is solved as a coupled initial-boundary value problem. The code simulation includes both the closed and open field regions. The axial effects incorporated include global axial force balance, axial losses in the open field region, and flux surface averaging over the closed field region. A typical example of the code results is also given. (orig.)

DRAGON solutions to the 3D transport benchmark over a range in parameter space

International Nuclear Information System (INIS)

Martin, Nicolas; Hebert, Alain; Marleau, Guy

2010-01-01

DRAGON solutions to the 'NEA suite of benchmarks for 3D transport methods and codes over a range in parameter space' are discussed in this paper. A description of the benchmark is first provided, followed by a detailed review of the different computational models used in the lattice code DRAGON. Two numerical methods were selected for generating the required quantities for the 729 configurations of this benchmark. First, S N calculations were performed using fully symmetric angular quadratures and high-order diamond differencing for spatial discretization. To compare S N results with those of another deterministic method, the method of characteristics (MoC) was also considered for this benchmark. Comparisons between reference solutions, S N and MoC results illustrate the advantages and drawbacks of each methods for this 3-D transport problem.

Code manual for MACCS2: Volume 1, user's guide

International Nuclear Information System (INIS)

Chanin, D.I.; Young, M.L.

1997-03-01

This report describes the use of the MACCS2 code. The document is primarily a user's guide, though some model description information is included. MACCS2 represents a major enhancement of its predecessor MACCS, the MELCOR Accident Consequence Code System. MACCS, distributed by government code centers since 1990, was developed to evaluate the impacts of severe accidents at nuclear power plants on the surrounding public. The principal phenomena considered are atmospheric transport and deposition under time-variant meteorology, short- and long-term mitigative actions and exposure pathways, deterministic and stochastic health effects, and economic costs. No other U.S. code that is publicly available at present offers all these capabilities. MACCS2 was developed as a general-purpose tool applicable to diverse reactor and nonreactor facilities licensed by the Nuclear Regulatory Commission or operated by the Department of Energy or the Department of Defense. The MACCS2 package includes three primary enhancements: (1) a more flexible emergency-response model, (2) an expanded library of radionuclides, and (3) a semidynamic food-chain model. Other improvements are in the areas of phenomenological modeling and new output options. Initial installation of the code, written in FORTRAN 77, requires a 486 or higher IBM-compatible PC with 8 MB of RAM

Improvement of Monte Carlo code A3MCNP for large-scale shielding problems

International Nuclear Information System (INIS)

Miyake, Y.; Ohmura, M.; Hasegawa, T.; Ueki, K.; Sato, O.; Haghighat, A.; Sjoden, G.E.

2004-01-01

A 3 MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic 'importance' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3 MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A 3 MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3 MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3 MCNP (referred to as A 3 MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3 MCNPV for a concrete cask streaming problem and a PWR dosimetry problem. (author)

Uncertainty analysis of neutron transport calculation

International Nuclear Information System (INIS)

Oka, Y.; Furuta, K.; Kondo, S.

1987-01-01

A cross section sensitivity-uncertainty analysis code, SUSD was developed. The code calculates sensitivity coefficients for one and two-dimensional transport problems based on the first order perturbation theory. Variance and standard deviation of detector responses or design parameters can be obtained using cross section covariance matrix. The code is able to perform sensitivity-uncertainty analysis for secondary neutron angular distribution(SAD) and secondary neutron energy distribution(SED). Covariances of 6 Li and 7 Li neutron cross sections in JENDL-3PR1 were evaluated including SAD and SED. Covariances of Fe and Be were also evaluated. The uncertainty of tritium breeding ratio, fast neutron leakage flux and neutron heating was analysed on four types of blanket concepts for a commercial tokamak fusion reactor. The uncertainty of tritium breeding ratio was less than 6 percent. Contribution from SAD/SED uncertainties are significant for some parameters. Formulas to estimate the errors of numerical solution of the transport equation were derived based on the perturbation theory. This method enables us to deterministically estimate the numerical errors due to iterative solution, spacial discretization and Legendre polynomial expansion of transfer cross-sections. The calculational errors of the tritium breeding ratio and the fast neutron leakage flux of the fusion blankets were analysed. (author)

Current status of radiation transport tools for proliferation and terrorism prevention

International Nuclear Information System (INIS)

Sale, K.E.

2004-01-01

We present the current status and future plans for the set of calculational tools and data bases developed and maintained at LLNL. The calculational tools include the Monte Carlo codes TART and COG as well as the deterministic code ARDRA. In addition to these codes presently in use there is a major development effort for a new massively parallel transport code. An important part of the capability we're developing is a sophisticated user interface, based on a commercial 3-D modeling product, to improve the model development process. A major part of this user interface tool is being developed by Strela under the Nuclear Cities Initiative. Strela has developed a hub-and-spoke technology for code input interconversions (between COG, TART and MCNP) and will produce the plug-ins that extend the capabilities of the 3-D modeler for use as a radiation transport input generator. The major advantages of this approach are the built-in user interface for 3-D modeling and the ability to read a large variety of CAD-file formats. In addition to supporting our current radiation transport codes and developing new capabilities we are working on some nuclear data needs for homeland security. These projects are carried out and the Lawrence Berkeley National Laboratory 88' cyclotron and at the Institute for Nuclear Research of the Nation Academy of Science of Ukraine under and STCU contract. (author)

Integrated Deterministic-Probabilistic Safety Assessment Methodologies

Energy Technology Data Exchange (ETDEWEB)

Kudinov, P.; Vorobyev, Y.; Sanchez-Perea, M.; Queral, C.; Jimenez Varas, G.; Rebollo, M. J.; Mena, L.; Gomez-Magin, J.

2014-02-01

IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) is a family of methods which use tightly coupled probabilistic and deterministic approaches to address respective sources of uncertainties, enabling Risk informed decision making in a consistent manner. The starting point of the IDPSA framework is that safety justification must be based on the coupling of deterministic (consequences) and probabilistic (frequency) considerations to address the mutual interactions between stochastic disturbances (e.g. failures of the equipment, human actions, stochastic physical phenomena) and deterministic response of the plant (i.e. transients). This paper gives a general overview of some IDPSA methods as well as some possible applications to PWR safety analyses. (Author)

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).

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)

Simplified model for radioactive contaminant transport: the TRANSS code

International Nuclear Information System (INIS)

Simmons, C.S.; Kincaid, C.T.; Reisenauer, A.E.

1986-09-01

A simplified ground-water transport model called TRANSS was devised to estimate the rate of migration of a decaying radionuclide that is subject to sorption governed by a linear isotherm. Transport is modeled as a contaminant mass transmitted along a collection of streamlines constituting a streamtube, which connects a source release zone with an environmental arrival zone. The probability-weighted contaminant arrival distribution along each streamline is represented by an analytical solution of the one-dimensional advection-dispersion equation with constant velocity and dispersion coefficient. The appropriate effective constant velocity for each streamline is based on the exact travel time required to traverse a streamline with a known length. An assumption used in the model to facilitate the mathematical simplification is that transverse dispersion within a streamtube is negligible. Release of contaminant from a source is described in terms of a fraction-remaining curve provided as input information. However, an option included in the code is the calculation of a fraction-remaining curve based on four specialized release models: (1) constant release rate, (2) solubility-controlled release, (3) adsorption-controlled release, and (4) diffusion-controlled release from beneath an infiltration barrier. To apply the code, a user supplies only a certain minimal number of parameters: a probability-weighted list of travel times for streamlines, a local-scale dispersion coefficient, a sorption distribution coefficient, total initial radionuclide inventory, radioactive half-life, a release model choice, and size dimensions of the source. The code is intended to provide scoping estimates of contaminant transport and does not predict the evolution of a concentration distribution in a ground-water flow field. Moreover, the required travel times along streamlines must be obtained from a prior ground-water flow simulation

Review and assessment of thermodynamic and transport properties for the CONTAIN Code

International Nuclear Information System (INIS)

Valdez, G.D.

1988-12-01

A study was carried out to review available data and correlations on the thermodynamic and transport properties of materials applicable to the CONTAIN computer code. CONTAIN is the NRC's best-estimate, mechanistic computer code for modeling containment response to a severe accident. Where appropriate, recommendations have been made for suitable approximations for material properties of interests. Based on a modified Benedict-Webb-Rubin (BWR) equation of state, a procedure is introduced for calculating thermodynamic properties for common gases in the CONTAIN code. These gases are nitrogen, oxygen, hydrogen, carbon dioxide, carbon monoxide, steam, helium, and argon. The thermodynamic equations for density, currently represented in CONTAIN by relatively simple fits, were independently checked and are recommended to be replaced by the Lee-Kesler equation of state which substantially improves accuracy without too much sacrifice in computational efficiency. The accuracy of the calculated values have been found to be generally acceptable. Various correlations and models for single component gas transport properties, viscosity and thermal conductivity, were also assessed with available experimental data. When a suitable correlation or model was not available, transport properties were obtained by performing least-squares fit on experimental data. 50 refs., 126 figs., 3 tabs

RADTRAN 5 - A computer code for transportation risk analysis

International Nuclear Information System (INIS)

Neuhauser, K.S.; Kanipe, F.L.

1993-01-01

The RADTRAN 5 computer code has been developed to estimate radiological and nonradiological risks of radioactive materials transportation. RADTRAN 5 is written in ANSI standard FORTRAN 77; the code contains significant advances in the methodology first pioneered with the LINK option of RADTRAN 4. A major application of the LINK methodology is route-specific analysis. Another application is comparisons of attributes along the same route segments. Nonradiological risk factors have been incorporated to allow users to estimate nonradiological fatalities and injuries that might occur during the transportation event(s) being analyzed. These fatalities include prompt accidental fatalities from mechanical causes. Values of these risk factors for the United States have been made available in the code as optional defaults. Several new health effects models have been published in the wake of the Hiroshima-Nagasaki dosimetry reassessment, and this has emphasized the need for flexibility in the RADTRAN approach to health-effects calculations. Therefore, the basic set of health-effects conversion equations in RADTRAN have been made user-definable. All parameter values can be changed by the user, but a complete set of default values are available for both the new International Commission on Radiation Protection model (ICRP Publication 60) and the recent model of the U.S. National Research Council's Committee on the Biological Effects of Radiation (BEIR V). The meteorological input data tables have been modified to permit optional entry of maximum downwind distances for each dose isopleth. The expected dose to an individual in each isodose area is also calculated and printed automatically. Examples are given that illustrate the power and flexibility of the RADTRAN 5 computer code. (J.P.N.)

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.

THE UNPREDICTABILITY CLAUSE IN TRANSPORT CONTRACTS, ACCORDING TO THE NEW CIVIL CODE

Directory of Open Access Journals (Sweden)

Adriana Elena BELU

2014-05-01

Full Text Available Until the enforcement of the highly controversial transport law, transport companies must already observe the provisions of the new Civil Code1 in their transport business. One of the novelties in the new Civil Code, that came into force on October 1, 2011, refers to the unpredictability clause: recurring to this clause, in certain situations to be precisely analysed by courts, parties may even be exempted from certain contractual obligations, when the court decides to rescind the contract based on objective criteria, not imputable to the party that no longer can properly fulfil the obligations that had been undertaken when the contract had been made. However, this solution only is provided after all means of negotiation and mediation between parties are exhausted. The clause meets current market requirements, under which many companies have to deal with bad paying partners.

Metropol: A computer code for the simulation of transport of contaminants with groundwater

International Nuclear Information System (INIS)

Sauter, F.J.; Hassanizadeh, S.M.; Leijnse, A.; Glasbergen, P.; Slot, A.F.M.

1990-01-01

In this report a description is given of the computer code Metropol. This code simulates the three-dimensional flow of groundwater with varying density and the simultaneous transport of contaminants in low concentration and is based on the finite element method. The basic equations for groundwater flow and transport are described as well as the mathematical techniques used to solve these equations. Pre-processing facilities for mesh generation and post-processing facilities such as particle tracking are also discussed. This work was part of the Community Mirage project Second phase, research area Calculation tools

A one-dimensional transport code for the simulation of D-T burning tokamak plasma

International Nuclear Information System (INIS)

Tone, Tatsuzo; Maki, Koichi; Kasai, Masao; Nishida, Hidetsugu

1980-11-01

A one-dimensional transport code for D-T burning tokamak plasma has been developed, which simulates the spatial behavior of fuel ions(D, T), alpha particles, impurities, temperatures of ions and electrons, plasma current, neutrals, heating of alpha and injected beam particles. The basic transport equations are represented by one generalized equation so that the improvement of models and the addition of new equations may be easily made. A model of burn control using a variable toroidal field ripple is employed. This report describes in detail the simulation model, numerical method and the usage of the code. Some typical examples to which the code has been applied are presented. (author)

The Identity of Information: How Deterministic Dependencies Constrain Information Synergy and Redundancy

Directory of Open Access Journals (Sweden)

Daniel Chicharro

2018-03-01

Full Text Available Understanding how different information sources together transmit information is crucial in many domains. For example, understanding the neural code requires characterizing how different neurons contribute unique, redundant, or synergistic pieces of information about sensory or behavioral variables. Williams and Beer (2010 proposed a partial information decomposition (PID that separates the mutual information that a set of sources contains about a set of targets into nonnegative terms interpretable as these pieces. Quantifying redundancy requires assigning an identity to different information pieces, to assess when information is common across sources. Harder et al. (2013 proposed an identity axiom that imposes necessary conditions to quantify qualitatively common information. However, Bertschinger et al. (2012 showed that, in a counterexample with deterministic target-source dependencies, the identity axiom is incompatible with ensuring PID nonnegativity. Here, we study systematically the consequences of information identity criteria that assign identity based on associations between target and source variables resulting from deterministic dependencies. We show how these criteria are related to the identity axiom and to previously proposed redundancy measures, and we characterize how they lead to negative PID terms. This constitutes a further step to more explicitly address the role of information identity in the quantification of redundancy. The implications for studying neural coding are discussed.

PHITS: Particle and heavy ion transport code system, version 2.23

International Nuclear Information System (INIS)

Niita, Koji; Matsuda, Norihiro; Iwamoto, Yosuke; Sato, Tatsuhiko; Nakashima, Hiroshi; Sakamoto, Yukio; Iwase, Hiroshi; Sihver, Lembit

2010-10-01

A Particle and Heavy-Ion Transport code System PHITS has been developed under the collaboration of JAEA (Japan Atomic Energy Agency), RIST (Research Organization for Information Science and Technology) and KEK (High Energy Accelerator Research Organization). PHITS can deal with the transport of all particles (nucleons, nuclei, mesons, photons, and electrons) over wide energy ranges, using several nuclear reaction models and nuclear data libraries. Geometrical configuration of the simulation can be set with GG (General Geometry) or CG (Combinatorial Geometry). Various quantities such as heat deposition, track length and production yields can be deduced from the simulation, using implemented estimator functions called 'tally'. The code also has a function to draw 2D and 3D figures of the calculated results as well as the setup geometries, using a code ANGEL. Because of these features, PHITS has been widely used for various purposes such as designs of accelerator shielding, radiation therapy and space exploration. Recently PHITS introduces an event generator for particle transport parts in the low energy region. Thus, PHITS was completely rewritten for the introduction of the event generator for neutron-induced reactions in energy region less than 20 MeV. Furthermore, several new tallis were incorporated for estimation of the relative biological effects. This document provides a manual of the new PHITS. (author)

Development and application of neutron transport methods and uncertainty analyses for reactor core calculations. Technical report; Entwicklung und Einsatz von Neutronentransportmethoden und Unsicherheitsanalysen fuer Reaktorkernberechnungen. Technischer Bericht

Energy Technology Data Exchange (ETDEWEB)

Zwermann, W.; Aures, A.; Bernnat, W.; and others

2013-06-15

This report documents the status of the research and development goals reached within the reactor safety research project RS1503 ''Development and Application of Neutron Transport Methods and Uncertainty Analyses for Reactor Core Calculations'' as of the 1{sup st} quarter of 2013. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts. The contributions of this project towards achieving this goal are the further development, validation, and application of deterministic and stochastic calculation programmes and of methods for uncertainty and sensitivity analyses, as well as the assessment of artificial neutral networks, for providing a complete nuclear calculation chain. This comprises processing nuclear basis data, creating multi-group data for diffusion and transport codes, obtaining reference solutions for stationary states with Monte Carlo codes, performing coupled 3D full core analyses in diffusion approximation and with other deterministic and also Monte Carlo transport codes, and implementing uncertainty and sensitivity analyses with the aim of propagating uncertainties through the whole calculation chain from fuel assembly, spectral and depletion calculations to coupled transient analyses. This calculation chain shall be applicable to light water reactors and also to innovative reactor concepts, and therefore has to be extensively validated with the help of benchmarks and critical experiments.

3D Monte-Carlo transport calculations of whole slab reactor cores: validation of deterministic neutronic calculation routes

International Nuclear Information System (INIS)

Palau, J.M.

2005-01-01

This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U 235 , U 238 , Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)

3D Monte-Carlo transport calculations of whole slab reactor cores: validation of deterministic neutronic calculation routes

Energy Technology Data Exchange (ETDEWEB)

Palau, J M [CEA Cadarache, Service de Physique des Reacteurs et du Cycle, Lab. de Projets Nucleaires, 13 - Saint-Paul-lez-Durance (France)

2005-07-01

This paper presents how Monte-Carlo calculations (French TRIPOLI4 poly-kinetic code with an appropriate pre-processing and post-processing software called OVNI) are used in the case of 3-dimensional heterogeneous benchmarks (slab reactor cores) to reduce model biases and enable a thorough and detailed analysis of the performances of deterministic methods and their associated data libraries with respect to key neutron parameters (reactivity, local power). Outstanding examples of application of these tools are presented regarding the new numerical methods implemented in the French lattice code APOLLO2 (advanced self-shielding models, new IDT characteristics method implemented within the discrete-ordinates flux solver model) and the JEFF3.1 nuclear data library (checked against JEF2.2 previous file). In particular we have pointed out, by performing multigroup/point-wise TRIPOLI4 (assembly and core) calculations, the efficiency (in terms of accuracy and computation time) of the new IDT method developed in APOLLO2. In addition, by performing 3-dimensional TRIPOLI4 calculations of the whole slab core (few millions of elementary volumes), the high quality of the new JEFF3.1 nuclear data files and revised evaluations (U{sup 235}, U{sup 238}, Hf) for reactivity prediction of slab cores critical experiments has been stressed. As a feedback of the whole validation process, improvements in terms of nuclear data (mainly Hf capture cross-sections) and numerical methods (advanced quadrature formulas accounting validation results, validation of new self-shielding models, parallelization) are suggested to improve even more the APOLLO2-CRONOS2 standard calculation route. (author)

A toolkit for integrated deterministic and probabilistic assessment for hydrogen infrastructure.

Energy Technology Data Exchange (ETDEWEB)

Groth, Katrina M.; Tchouvelev, Andrei V.

2014-03-01

There has been increasing interest in using Quantitative Risk Assessment [QRA] to help improve the safety of hydrogen infrastructure and applications. Hydrogen infrastructure for transportation (e.g. fueling fuel cell vehicles) or stationary (e.g. back-up power) applications is a relatively new area for application of QRA vs. traditional industrial production and use, and as a result there are few tools designed to enable QRA for this emerging sector. There are few existing QRA tools containing models that have been developed and validated for use in small-scale hydrogen applications. However, in the past several years, there has been significant progress in developing and validating deterministic physical and engineering models for hydrogen dispersion, ignition, and flame behavior. In parallel, there has been progress in developing defensible probabilistic models for the occurrence of events such as hydrogen release and ignition. While models and data are available, using this information is difficult due to a lack of readily available tools for integrating deterministic and probabilistic components into a single analysis framework. This paper discusses the first steps in building an integrated toolkit for performing QRA on hydrogen transportation technologies and suggests directions for extending the toolkit.

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

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)

EBR-II Static Neutronic Calculations by PHISICS / MCNP6 codes

Energy Technology Data Exchange (ETDEWEB)

Paolo Balestra; Carlo Parisi; Andrea Alfonsi

2016-02-01

The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) on the Shutdown Heat Removal Tests (SHRT) performed in the '80s at the Experimental fast Breeder Reactor EBR-II, USA. The scope of the CRP is to improve and validate the simulation tools for the study and the design of the liquid metal cooled fast reactors. Moreover, training of the next generation of fast reactor analysts is being also considered the other scope of the CRP. In this framework, a static neutronic model was developed, using state-of-the art neutron transport codes like SCALE/PHISICS (deterministic solution) and MCNP6 (stochastic solution). Comparison between both solutions is briefly illustrated in this summary.

Development of three-dimensional transport code by the double finite element method

International Nuclear Information System (INIS)

Fujimura, Toichiro

1985-01-01

Development of a three-dimensional neutron transport code by the double finite element method is described. Both of the Galerkin and variational methods are adopted to solve the problem, and then the characteristics of them are compared. Computational results of the collocation method, developed as a technique for the vaviational one, are illustrated in comparison with those of an Ssub(n) code. (author)

Code manual for MACCS2: Volume 1, user`s guide

Energy Technology Data Exchange (ETDEWEB)

Chanin, D.I.; Young, M.L.

1997-03-01

This report describes the use of the MACCS2 code. The document is primarily a user`s guide, though some model description information is included. MACCS2 represents a major enhancement of its predecessor MACCS, the MELCOR Accident Consequence Code System. MACCS, distributed by government code centers since 1990, was developed to evaluate the impacts of severe accidents at nuclear power plants on the surrounding public. The principal phenomena considered are atmospheric transport and deposition under time-variant meteorology, short- and long-term mitigative actions and exposure pathways, deterministic and stochastic health effects, and economic costs. No other U.S. code that is publicly available at present offers all these capabilities. MACCS2 was developed as a general-purpose tool applicable to diverse reactor and nonreactor facilities licensed by the Nuclear Regulatory Commission or operated by the Department of Energy or the Department of Defense. The MACCS2 package includes three primary enhancements: (1) a more flexible emergency-response model, (2) an expanded library of radionuclides, and (3) a semidynamic food-chain model. Other improvements are in the areas of phenomenological modeling and new output options. Initial installation of the code, written in FORTRAN 77, requires a 486 or higher IBM-compatible PC with 8 MB of RAM.

360° deterministic magnetization rotation in a three-ellipse magnetoelectric heterostructure

Science.gov (United States)

Kundu, Auni A.; Chavez, Andres C.; Keller, Scott M.; Carman, Gregory P.; Lynch, Christopher S.

2018-03-01

A magnetic dipole-coupled magnetoelectric heterostructure comprised of three closely spaced ellipse shapes was designed and shown to be capable of achieving deterministic in-plane magnetization rotation. The design approach used a combination of conventional micromagnetic simulations to obtain preliminary configurations followed by simulations using a fully strain-coupled, time domain micromagnetic code for a detailed assessment of performance. The conventional micromagnetic code has short run times and was used to refine the ellipse shape and orientation, but it does not accurately capture the effects of the strain gradients present in the piezoelectric and magnetostrictive layers that contribute to magnetization reorientation. The fully coupled code was used to assess the effects of strain and magnetic field gradients on precessional switching in the side ellipses and on the resulting dipole-field driven magnetization reorientation in the center ellipse. The work led to a geometry with a CoFeB ellipse (125 nm × 95 nm × 4 nm) positioned between two smaller CoFeB ellipses (75 nm × 50 nm × 4 nm) on a 500 nm PZT-5H film substrate clamped at its bottom surface. The smaller ellipses were oriented at 45° and positioned at 70° and 250° about the central ellipse due to the film deposition on a thick substrate. A 7.3 V pulse applied to the PZT for 0.22 ns produced 180° switching of the magnetization in the outer ellipses that then drove switching in the center ellipse through dipole-dipole coupling. Full 360° deterministic rotation was achieved with a second pulse. The temporal response of the resulting design is discussed.

Nonterminals and codings in defining variations of OL-systems

DEFF Research Database (Denmark)

Skyum, Sven

1974-01-01

The use of nonterminals versus the use of codings in variations of OL-systems is studied. It is shown that the use of nonterminals produces a comparatively low generative capacity in deterministic systems while it produces a comparatively high generative capacity in nondeterministic systems. Fina....... Finally it is proved that the family of context-free languages is contained in the family generated by codings on propagating OL-systems with a finite set of axioms, which was one of the open problems in [10]. All the results in this paper can be found in [71] and [72].......The use of nonterminals versus the use of codings in variations of OL-systems is studied. It is shown that the use of nonterminals produces a comparatively low generative capacity in deterministic systems while it produces a comparatively high generative capacity in nondeterministic systems...

The OpenMOC method of characteristics neutral particle transport code

International Nuclear Information System (INIS)

Boyd, William; Shaner, Samuel; Li, Lulu; Forget, Benoit; Smith, Kord

2014-01-01

Highlights: • An open source method of characteristics neutron transport code has been developed. • OpenMOC shows nearly perfect scaling on CPUs and 30× speedup on GPUs. • Nonlinear acceleration techniques demonstrate a 40× reduction in source iterations. • OpenMOC uses modern software design principles within a C++ and Python framework. • Validation with respect to the C5G7 and LRA benchmarks is presented. - Abstract: The method of characteristics (MOC) is a numerical integration technique for partial differential equations, and has seen widespread use for reactor physics lattice calculations. The exponential growth in computing power has finally brought the possibility for high-fidelity full core MOC calculations within reach. The OpenMOC code is being developed at the Massachusetts Institute of Technology to investigate algorithmic acceleration techniques and parallel algorithms for MOC. OpenMOC is a free, open source code written using modern software languages such as C/C++ and CUDA with an emphasis on extensible design principles for code developers and an easy to use Python interface for code users. The present work describes the OpenMOC code and illustrates its ability to model large problems accurately and efficiently

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

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.

Theoretical background and user's manual for the computer code on groundwater flow and radionuclide transport calculation in porous rock

International Nuclear Information System (INIS)

Shirakawa, Toshihiko; Hatanaka, Koichiro

2001-11-01

In order to document a basic manual about input data, output data, execution of computer code on groundwater flow and radionuclide transport calculation in heterogeneous porous rock, we investigated the theoretical background about geostatistical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport which calculates water flow in three dimension, the path of moving radionuclide, and one dimensional radionuclide migration. In this report, based on above investigation we describe the geostatistical background about simulating heterogeneous permeability field. And we describe construction of files, input and output data, a example of calculating of the programs which simulates heterogeneous permeability field, and calculates groundwater flow and radionuclide transport. Therefore, we can document a manual by investigating the theoretical background about geostatistical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport calculation. And we can model heterogeneous porous rock and analyze groundwater flow and radionuclide transport by utilizing the information from this report. (author)

Integral transport computation of gamma detector response with the CPM2 code

International Nuclear Information System (INIS)

Jones, D.B.

1989-12-01

CPM-2 Version 3 is an enhanced version of the CPM-2 lattice physics computer code which supports the capabilities to (1) perform a two-dimensional gamma flux calculation and (2) perform Restart/Data file maintenance operations. The Gamma Calculation Module implemented in CPM-2 was first developed for EPRI in the CASMO-1 computer code by Studsvik Energiteknik under EPRI Agreement RP2352-01. The gamma transport calculation uses the CPM-HET code module to calculate the transport of gamma rays in two dimensions in a mixed cylindrical-rectangular geometry, where the basic fuel assembly and component regions are maintained in a rectangular geometry, but the fuel pins are represented as cylinders within a square pin cell mesh. Such a capability is needed to represent gamma transport in an essentially transparent medium containing spatially distributed ''black'' cylindrical pins. Under a subcontract to RP2352-01, RPI developed the gamma production and gamma interaction library used for gamma calculation. The CPM-2 gamma calculation was verified against reference results generated by Studsvik using the CASMO-1 program. The CPM-2 Restart/Data file maintenance capabilities provide the user with options to copy files between Restart/Data tapes and to purge files from the Restart/Data tapes

Iterative acceleration methods for Monte Carlo and deterministic criticality calculations

Energy Technology Data Exchange (ETDEWEB)

Urbatsch, T.J.

1995-11-01

If you have ever given up on a nuclear criticality calculation and terminated it because it took so long to converge, you might find this thesis of interest. The author develops three methods for improving the fission source convergence in nuclear criticality calculations for physical systems with high dominance ratios for which convergence is slow. The Fission Matrix Acceleration Method and the Fission Diffusion Synthetic Acceleration (FDSA) Method are acceleration methods that speed fission source convergence for both Monte Carlo and deterministic methods. The third method is a hybrid Monte Carlo method that also converges for difficult problems where the unaccelerated Monte Carlo method fails. The author tested the feasibility of all three methods in a test bed consisting of idealized problems. He has successfully accelerated fission source convergence in both deterministic and Monte Carlo criticality calculations. By filtering statistical noise, he has incorporated deterministic attributes into the Monte Carlo calculations in order to speed their source convergence. He has used both the fission matrix and a diffusion approximation to perform unbiased accelerations. The Fission Matrix Acceleration method has been implemented in the production code MCNP and successfully applied to a real problem. When the unaccelerated calculations are unable to converge to the correct solution, they cannot be accelerated in an unbiased fashion. A Hybrid Monte Carlo method weds Monte Carlo and a modified diffusion calculation to overcome these deficiencies. The Hybrid method additionally possesses reduced statistical errors.

Iterative acceleration methods for Monte Carlo and deterministic criticality calculations

International Nuclear Information System (INIS)

Urbatsch, T.J.

1995-11-01

If you have ever given up on a nuclear criticality calculation and terminated it because it took so long to converge, you might find this thesis of interest. The author develops three methods for improving the fission source convergence in nuclear criticality calculations for physical systems with high dominance ratios for which convergence is slow. The Fission Matrix Acceleration Method and the Fission Diffusion Synthetic Acceleration (FDSA) Method are acceleration methods that speed fission source convergence for both Monte Carlo and deterministic methods. The third method is a hybrid Monte Carlo method that also converges for difficult problems where the unaccelerated Monte Carlo method fails. The author tested the feasibility of all three methods in a test bed consisting of idealized problems. He has successfully accelerated fission source convergence in both deterministic and Monte Carlo criticality calculations. By filtering statistical noise, he has incorporated deterministic attributes into the Monte Carlo calculations in order to speed their source convergence. He has used both the fission matrix and a diffusion approximation to perform unbiased accelerations. The Fission Matrix Acceleration method has been implemented in the production code MCNP and successfully applied to a real problem. When the unaccelerated calculations are unable to converge to the correct solution, they cannot be accelerated in an unbiased fashion. A Hybrid Monte Carlo method weds Monte Carlo and a modified diffusion calculation to overcome these deficiencies. The Hybrid method additionally possesses reduced statistical errors

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

Deterministic extraction from weak random sources

CERN Document Server

Gabizon, Ariel

2011-01-01

In this research monograph, the author constructs deterministic extractors for several types of sources, using a methodology of recycling randomness which enables increasing the output length of deterministic extractors to near optimal length.

PFLOTRAN: Reactive Flow & Transport Code for Use on Laptops to Leadership-Class Supercomputers

Energy Technology Data Exchange (ETDEWEB)

Hammond, Glenn E.; Lichtner, Peter C.; Lu, Chuan; Mills, Richard T.

2012-04-18

PFLOTRAN, a next-generation reactive flow and transport code for modeling subsurface processes, has been designed from the ground up to run efficiently on machines ranging from leadership-class supercomputers to laptops. Based on an object-oriented design, the code is easily extensible to incorporate additional processes. It can interface seamlessly with Fortran 9X, C and C++ codes. Domain decomposition parallelism is employed, with the PETSc parallel framework used to manage parallel solvers, data structures and communication. Features of the code include a modular input file, implementation of high-performance I/O using parallel HDF5, ability to perform multiple realization simulations with multiple processors per realization in a seamless manner, and multiple modes for multiphase flow and multicomponent geochemical transport. Chemical reactions currently implemented in the code include homogeneous aqueous complexing reactions and heterogeneous mineral precipitation/dissolution, ion exchange, surface complexation and a multirate kinetic sorption model. PFLOTRAN has demonstrated petascale performance using 2{sup 17} processor cores with over 2 billion degrees of freedom. Accomplishments achieved to date include applications to the Hanford 300 Area and modeling CO{sub 2} sequestration in deep geologic formations.

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)

Non-standard model for electron heat transport for multidimensional hydrodynamic codes

Energy Technology Data Exchange (ETDEWEB)

Nicolai, Ph.; Busquet, M.; Schurtz, G. [CEA/DAM-Ile de France, 91 - Bruyeres Le Chatel (France)

2000-07-01

In simulations of laser-produced plasma, modeling of heat transport requires an artificial limitation of standard Spitzer-Haerm fluxes. To improve heat conduction processing, we have developed a multidimensional model which accounts for non-local features of heat transport and effects of self-generated magnetic fields. This consistent treatment of both mechanisms has been implemented in a two-dimensional radiation-hydrodynamic code. First results indicate good agreements between simulations and experimental data. (authors)

Non-standard model for electron heat transport for multidimensional hydrodynamic codes

International Nuclear Information System (INIS)

Nicolai, Ph.; Busquet, M.; Schurtz, G.

2000-01-01

In simulations of laser-produced plasma, modeling of heat transport requires an artificial limitation of standard Spitzer-Haerm fluxes. To improve heat conduction processing, we have developed a multidimensional model which accounts for non-local features of heat transport and effects of self-generated magnetic fields. This consistent treatment of both mechanisms has been implemented in a two-dimensional radiation-hydrodynamic code. First results indicate good agreements between simulations and experimental data. (authors)

Comparative study of boron transport models in NRC Thermal-Hydraulic Code Trace

Energy Technology Data Exchange (ETDEWEB)

Olmo-Juan, Nicolás; Barrachina, Teresa; Miró, Rafael; Verdú, Gumersindo; Pereira, Claubia, E-mail: nioljua@iqn.upv.es, E-mail: tbarrachina@iqn.upv.es, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es, E-mail: claubia@nuclear.ufmg.br [Institute for Industrial, Radiophysical and Environmental Safety (ISIRYM). Universitat Politècnica de València (Spain); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

Recently, the interest in the study of various types of transients involving changes in the boron concentration inside the reactor, has led to an increase in the interest of developing and studying new models and tools that allow a correct study of boron transport. Therefore, a significant variety of different boron transport models and spatial difference schemes are available in the thermal-hydraulic codes, as TRACE. According to this interest, in this work it will be compared the results obtained using the different boron transport models implemented in the NRC thermal-hydraulic code TRACE. To do this, a set of models have been created using the different options and configurations that could have influence in boron transport. These models allow to reproduce a simple event of filling or emptying the boron concentration in a long pipe. Moreover, with the aim to compare the differences obtained when one-dimensional or three-dimensional components are chosen, it has modeled many different cases using only pipe components or a mix of pipe and vessel components. In addition, the influence of the void fraction in the boron transport has been studied and compared under close conditions to BWR commercial model. A final collection of the different cases and boron transport models are compared between them and those corresponding to the analytical solution provided by the Burgers equation. From this comparison, important conclusions are drawn that will be the basis of modeling the boron transport in TRACE adequately. (author)

Extensions of the 3-dimensional plasma transport code E3D

International Nuclear Information System (INIS)

Runov, A.; Schneider, R.; Kasilov, S.; Reiter, D.

2004-01-01

One important aspect of modern fusion research is plasma edge physics. Fluid transport codes extending beyond the standard 2-D code packages like B2-Eirene or UEDGE are under development. A 3-dimensional plasma fluid code, E3D, based upon the Multiple Coordinate System Approach and a Monte Carlo integration procedure has been developed for general magnetic configurations including ergodic regions. These local magnetic coordinates lead to a full metric tensor which accurately accounts for all transport terms in the equations. Here, we discuss new computational aspects of the realization of the algorithm. The main limitation to the Monte Carlo code efficiency comes from the restriction on the parallel jump of advancing test particles which must be small compared to the gradient length of the diffusion coefficient. In our problems, the parallel diffusion coefficient depends on both plasma and magnetic field parameters. Usually, the second dependence is much more critical. In order to allow long parallel jumps, this dependence can be eliminated in two steps: first, the longitudinal coordinate x 3 of local magnetic coordinates is modified in such a way that in the new coordinate system the metric determinant and contra-variant components of the magnetic field scale along the magnetic field with powers of the magnetic field module (like in Boozer flux coordinates). Second, specific weights of the test particles are introduced. As a result of increased parallel jump length, the efficiency of the code is about two orders of magnitude better. (copyright 2004 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

TRIDENT-CTR: a two-dimensional transport code for CTR applications

International Nuclear Information System (INIS)

Seed, T.J.

1978-01-01

TRIDENT-CTR is a two-dimensional x-y and r-z geometry multigroup neutral transport code developed at Los Alamos for toroidal calculations. The use of triangular finite elements gives it the geometric flexibility to cope with the nonorthogonal shapes of many toroidal designs of current interest in the CTR community

Application of a CADIS-like variance reduction technique to electron transport

International Nuclear Information System (INIS)

Dionne, B.; Haghighat, A.

2004-01-01

This paper studies the use of approximate deterministic importance functions to calculate the lower-weight bounds of the MCNP5 weight-window variance reduction technique when applied to electron transport simulations. This approach follows the CADIS (Consistent Adjoint Driven Importance Sampling) methodology developed for neutral particles shielding calculations. The importance functions are calculated using the one-dimensional CEPXS/ONELD code package. Considering a simple 1-D problem, this paper shows that our methodology can produce speedups up to ∼82 using an approximate electron importance function distributions computed in ∼8 seconds. (author)

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.

Multiple-canister flow and transport code in 2-dimensional space. MCFT2D: user's manual

International Nuclear Information System (INIS)

Lim, Doo-Hyun

2006-03-01

A two-dimensional numerical code, MCFT2D (Multiple-Canister Flow and Transport code in 2-Dimensional space), has been developed for groundwater flow and radionuclide transport analyses in a water-saturated high-level radioactive waste (HLW) repository with multiple canisters. A multiple-canister configuration and a non-uniform flow field of the host rock are incorporated in the MCFT2D code. Effects of heterogeneous flow field of the host rock on migration of nuclides can be investigated using MCFT2D. The MCFT2D enables to take into account the various degrees of the dependency of canister configuration for nuclide migration in a water-saturated HLW repository, while the dependency was assumed to be either independent or perfectly dependent in previous studies. This report presents features of the MCFT2D code, numerical simulation using MCFT2D code, and graphical representation of the numerical results. (author)

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)

Path Toward a Unified Geometry for Radiation Transport

Science.gov (United States)

Lee, Kerry

The Direct Accelerated Geometry for Radiation Analysis and Design (DAGRAD) element of the RadWorks Project under Advanced Exploration Systems (AES) within the Space Technology Mission Directorate (STMD) of NASA will enable new designs and concepts of operation for radiation risk assessment, mitigation and protection. This element is designed to produce a solution that will allow NASA to calculate the transport of space radiation through complex CAD models using the state-of-the-art analytic and Monte Carlo radiation transport codes. Due to the inherent hazard of astronaut and spacecraft exposure to ionizing radiation in low-Earth orbit (LEO) or in deep space, risk analyses must be performed for all crew vehicles and habitats. Incorporating these analyses into the design process can minimize the mass needed solely for radiation protection. Transport of the radiation fields as they pass through shielding and body materials can be simulated using Monte Carlo techniques or described by the Boltzmann equation, which is obtained by balancing changes in particle fluxes as they traverse a small volume of material with the gains and losses caused by atomic and nuclear collisions. Deterministic codes that solve the Boltzmann transport equation, such as HZETRN (high charge and energy transport code developed by NASA LaRC), are generally computationally faster than Monte Carlo codes such as FLUKA, GEANT4, MCNP(X) or PHITS; however, they are currently limited to transport in one dimension, which poorly represents the secondary light ion and neutron radiation fields. NASA currently uses HZETRN space radiation transport software, both because it is computationally efficient and because proven methods have been developed for using this software to analyze complex geometries. Although Monte Carlo codes describe the relevant physics in a fully three-dimensional manner, their computational costs have thus far prevented their widespread use for analysis of complex CAD models, leading

ORNL probabilistic fracture-mechanics code OCA-P

International Nuclear Information System (INIS)

Cheverton, R.D.; Ball, D.G.

1984-01-01

The computer code OCA-P was developed at the request of the USNRC for the purpose of helping to evaluate the integrity of PWR pressure vessels during overcooling accidents (OCA's). The code can be used for both deterministic and probabilistic fracture-mechanics calculations, and consists essentially of OCA-II and a Monte Carlo routine similar to that developed by Strosnider et al. In the probabilistic mode OCA-P generates a large number of vessels (10 6 more or less), each with a different combination of the various values of the different parameters involved in the analysis of flaw behavior. For each of these vessels a deterministic fracture-mechanics analysis is performed (calculation of K/sub I/, K/sub Ic/, K/sub Ia/) to determine whether vessel failure takes place. The conditional probability of failure is simply the number of vessels that fail divided by the number of vessels generated. OCA-II is used for the deterministic analysis. Basic input to OCA-II includes, among other things, the primry-system pressure transient and the temperature transient for the coolant in the reactor-vessel downcomer. With this and additional information available OCA-II performs a one-dimensional thermal analysis to obtain the temperature distribution in the wall as a function of time and then a one-dimensional linear-elastic stress analysis. OCA-P has been checked against similar codes and is presently being used in the Integrated Pressurized Thermal Shock Program for specific PWR plants

Evaluation of scattering laws and cross sections for calculation of production and transport of cold and ultracold neutrons

International Nuclear Information System (INIS)

Bernnat, W.; Keinert, J.; Mattes, M.

2004-01-01

For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H 2 O, liquid He, liquid D 2 O, liquid and solid H 2 and D 2 , solid CH 4 and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common S N -transport codes and the Monte Carlo Code MCNP. (orig.)

MATADOR (Methods for the Analysis of Transport And Deposition Of Radionuclides) code description and User's Manual

International Nuclear Information System (INIS)

Avci, H.I.; Raghuram, S.; Baybutt, P.

1985-04-01

A new computer code called MATADOR (Methods for the Analysis of Transport And Deposition Of Radionuclides) has been developed to replace the CORRAL-2 computer code which was written for the Reactor Safety Study (WASH-1400). This report is a User's Manual for MATADOR. MATADOR is intended for use in system risk studies to analyze radionuclide transport and deposition in reactor containments. The principal output of the code is information on the timing and magnitude of radionuclide releases to the environment as a result of severely degraded core accidents. MATADOR considers the transport of radionuclides through the containment and their removal by natural deposition and by engineered safety systems such as sprays. It is capable of analyzing the behavior of radionuclides existing either as vapors or aerosols in the containment. The code requires input data on the source terms into the containment, the geometry of the containment, and thermal-hydraulic conditions in the containment

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

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)

A 3D transport-based core analysis code for research reactors with unstructured geometry

International Nuclear Information System (INIS)

Zhang, Tengfei; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi; Li, Yunzhao

2013-01-01

Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal S N method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of k eff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

EPRI-LATTICE: a multigroup neutron transport code for light water reactor lattice physics calculations

International Nuclear Information System (INIS)

Jones, D.B.

1986-01-01

EPRI-LATTICE is a multigroup neutron transport computer code for the analysis of light water reactor fuel assemblies. It can solve the two-dimensional neutron transport problem by two distinct methods: (a) the method of collision probabilities and (b) the method of discrete ordinates. The code was developed by S. Levy Inc. as an account of work sponsored by the Electric Power Research Institute (EPRI). The collision probabilities calculation in EPRI-LATTICE (L-CP) is based on the same methodology that exists in the lattice codes CPM-2 and EPRI-CPM. Certain extensions have been made to the data representations of the CPM programs to improve the overall accuracy of the calculation. The important extensions include unique representations of scattering matrices and fission fractions (chi) for each composition in the problem. A new capability specifically developed for the EPRI-LATTICE code is a discrete ordinates methodology. The discrete ordinates calculation in EPRI-LATTICE (L-SN) is based on the discrete S/sub n/ methodology that exists in the TWODANT program. In contrast to TWODANT, which utilizes synthetic diffusion acceleration and supports multiple geometries, only the transport equations are solved by L-SN and only the data representations for the two-dimensional geometry are treated

Application of the three-dimensional transport code to analysis of the neutron streaming experiment

International Nuclear Information System (INIS)

Chatani, K.; Slater, C.O.

1990-01-01

The neutron streaming through an experimental mock-up of a Clinch River Breeder Reactor (CRBR) prototypic coolant pipe chaseway was recalculated with a three-dimensional discrete ordinates code. The experiment was conducted at the Tower Shielding Facility at Oak Ridge National Laboratory in 1976 and 1977. The measurement of the neutron flux, using Bonner ball detectors, indicated nine orders of attenuation in the empty pipeway, which contained two 90-deg bends and was surrounded by concrete walls. The measurement data were originally analyzed using the DOT3.5 two-dimensional discrete ordinates radiation transport code. However, the results did not agree with measurement data at the bend because of the difficulties in modeling the three-dimensional configurations using two-dimensional methods. The two-dimensional calculations used a three-step procedure in which each of the three legs making the two 90-deg bends was a separate calculation. The experiment was recently analyzed with the TORT three-dimensional discrete ordinates radiation transport code, not only to compare the calculational results with the experimental results, but also to compare with results obtained from analyses in Japan using DOT3.5, MORSE, and ENSEMBLE, which is a three-dimensional discrete ordinates radiation transport code developed in Japan

The neutron transport code DTF-TRACA. User's manual and input data

International Nuclear Information System (INIS)

Anhert, C.

1979-01-01

A user's manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data description is given. The new options developped at JEN are included too. (author)

Asymptotic diffusion limit of cell temperature discretisation schemes for thermal radiation transport

Energy Technology Data Exchange (ETDEWEB)

Smedley-Stevenson, Richard P., E-mail: richard.smedley-stevenson@awe.co.uk [AWE PLC, Aldermaston, Reading, Berkshire, RG7 4PR (United Kingdom); Department of Earth Science and Engineering, Imperial College London, SW7 2AZ (United Kingdom); McClarren, Ryan G., E-mail: rmcclarren@ne.tamu.edu [Department of Nuclear Engineering, Texas A & M University, College Station, TX 77843-3133 (United States)

2015-04-01

This paper attempts to unify the asymptotic diffusion limit analysis of thermal radiation transport schemes, for a linear-discontinuous representation of the material temperature reconstructed from cell centred temperature unknowns, in a process known as ‘source tilting’. The asymptotic limits of both Monte Carlo (continuous in space) and deterministic approaches (based on linear-discontinuous finite elements) for solving the transport equation are investigated in slab geometry. The resulting discrete diffusion equations are found to have nonphysical terms that are proportional to any cell-edge discontinuity in the temperature representation. Based on this analysis it is possible to design accurate schemes for representing the material temperature, for coupling thermal radiation transport codes to a cell centred representation of internal energy favoured by ALE (arbitrary Lagrange–Eulerian) hydrodynamics schemes.

Asymptotic diffusion limit of cell temperature discretisation schemes for thermal radiation transport

International Nuclear Information System (INIS)

Smedley-Stevenson, Richard P.; McClarren, Ryan G.

2015-01-01

This paper attempts to unify the asymptotic diffusion limit analysis of thermal radiation transport schemes, for a linear-discontinuous representation of the material temperature reconstructed from cell centred temperature unknowns, in a process known as ‘source tilting’. The asymptotic limits of both Monte Carlo (continuous in space) and deterministic approaches (based on linear-discontinuous finite elements) for solving the transport equation are investigated in slab geometry. The resulting discrete diffusion equations are found to have nonphysical terms that are proportional to any cell-edge discontinuity in the temperature representation. Based on this analysis it is possible to design accurate schemes for representing the material temperature, for coupling thermal radiation transport codes to a cell centred representation of internal energy favoured by ALE (arbitrary Lagrange–Eulerian) hydrodynamics schemes

The DANTE Boltzmann transport solver: An unstructured mesh, 3-D, spherical harmonics algorithm compatible with parallel computer architectures

International Nuclear Information System (INIS)

McGhee, J.M.; Roberts, R.M.; Morel, J.E.

1997-01-01

A spherical harmonics research code (DANTE) has been developed which is compatible with parallel computer architectures. DANTE provides 3-D, multi-material, deterministic, transport capabilities using an arbitrary finite element mesh. The linearized Boltzmann transport equation is solved in a second order self-adjoint form utilizing a Galerkin finite element spatial differencing scheme. The core solver utilizes a preconditioned conjugate gradient algorithm. Other distinguishing features of the code include options for discrete-ordinates and simplified spherical harmonics angular differencing, an exact Marshak boundary treatment for arbitrarily oriented boundary faces, in-line matrix construction techniques to minimize memory consumption, and an effective diffusion based preconditioner for scattering dominated problems. Algorithm efficiency is demonstrated for a massively parallel SIMD architecture (CM-5), and compatibility with MPP multiprocessor platforms or workstation clusters is anticipated

Performance of the improved version of Monte Carlo Code A3MCNP for cask shielding design

International Nuclear Information System (INIS)

Hasegawa, T.; Ueki, K.; Sato, O.; Sjoden, G.E.; Miyake, Y.; Ohmura, M.; Haghighat, A.

2004-01-01

A 3 MCNP (Automatic Adjoint Accelerated MCNP) is a revised version of the MCNP Monte Carlo code, that automatically prepares variance reduction parameters for the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. Using a deterministic ''importance'' (or adjoint) function, CADIS performs source and transport biasing within the weight-window technique. The current version of A 3 MCNP uses the 3-D Sn transport TORT code to determine a 3-D importance function distribution. Based on simulation of several real-life problems, it is demonstrated that A3MCNP provides precise calculation results with a remarkably short computation time by using the proper and objective variance reduction parameters. However, since the first version of A 3 MCNP provided only a point source configuration option for large-scale shielding problems, such as spent-fuel transport casks, a large amount of memory may be necessary to store enough points to properly represent the source. Hence, we have developed an improved version of A 3 MCNP (referred to as A 3 MCNPV) which has a volumetric source configuration option. This paper describes the successful use of A 3 MCNPV for cask neutron and gamma-ray shielding problem

76 FR 2744 - Disclosure of Code-Share Service by Air Carriers and Sellers of Air Transportation

Science.gov (United States)

2011-01-14

... DEPARTMENT OF TRANSPORTATION Office of the Secretary Disclosure of Code-Share Service by Air Carriers and Sellers of Air Transportation AGENCY: Office of the Secretary, Department of Transportation..., their agents, and third party sellers of air transportation in view of recent amendments to 49 U.S.C...

Development of Parallel Computing Framework to Enhance Radiation Transport Code Capabilities for Rare Isotope Beam Facility Design

Energy Technology Data Exchange (ETDEWEB)

Kostin, Mikhail [Michigan State Univ., East Lansing, MI (United States); Mokhov, Nikolai [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Niita, Koji [Research Organization for Information Science and Technology, Ibaraki-ken (Japan)

2013-09-25

A parallel computing framework has been developed to use with general-purpose radiation transport codes. The framework was implemented as a C++ module that uses MPI for message passing. It is intended to be used with older radiation transport codes implemented in Fortran77, Fortran 90 or C. The module is significantly independent of radiation transport codes it can be used with, and is connected to the codes by means of a number of interface functions. The framework was developed and tested in conjunction with the MARS15 code. It is possible to use it with other codes such as PHITS, FLUKA and MCNP after certain adjustments. Besides the parallel computing functionality, the framework offers a checkpoint facility that allows restarting calculations with a saved checkpoint file. The checkpoint facility can be used in single process calculations as well as in the parallel regime. The framework corrects some of the known problems with the scheduling and load balancing found in the original implementations of the parallel computing functionality in MARS15 and PHITS. The framework can be used efficiently on homogeneous systems and networks of workstations, where the interference from the other users is possible.

Benchmarking of EPRI-cell epithermal methods with the point-energy discrete-ordinates code (OZMA)

International Nuclear Information System (INIS)

Williams, M.L.; Wright, R.Q.; Barhen, J.; Rothenstein, W.

1982-01-01

The purpose of the present study is to benchmark E-C resonance-shielding and cell-averaging methods against a rigorous deterministic solution on a fine-group level (approx. 30 groups between 1 eV and 5.5 keV). The benchmark code used is OZMA, which solves the space-dependent slowing-down equations using continuous-energy discrete ordinates or integral transport theory to produce fine-group cross sections. Results are given for three water-moderated lattices - a mixed oxide, a uranium method, and a tight-pitch high-conversion uranium oxide configuration. The latter two lattices were chosen because of the strong self shielding of the 238 U resonances

The neutron transport code DTF-Traca users manual and input data

Energy Technology Data Exchange (ETDEWEB)

Ahnert, C

1979-07-01

This is a users manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data descriptions is given. The new options developed at JEN are included too. (Author) 18 refs.

The neutron transport code DTF-Traca users manual and input data

International Nuclear Information System (INIS)

Ahnert, C.

1979-01-01

This is a users manual of the neutron transport code DTF-TRACA, which is a version of the original DTF-IV with some modifications made at JEN. A detailed input data descriptions is given. The new options developed at JEN are included too. (Author) 18 refs

Aspects of cell calculations in deterministic reactor core analysis

International Nuclear Information System (INIS)

Varvayanni, M.; Savva, P.; Catsaros, N.

2011-01-01

Τhe capability of achieving optimum utilization of the deterministic neutronic codes is very important, since, although elaborate tools, they are still widely used for nuclear reactor core analyses, due to specific advantages that they present compared to Monte Carlo codes. The user of a deterministic neutronic code system has to make some significant physical assumptions if correct results are to be obtained. A decisive first step at which such assumptions are required is the one-dimensional cell calculations, which provide the neutronic properties of the homogenized core cells and collapse the cross sections into user-defined energy groups. One of the most crucial determinations required at the above stage and significantly influencing the subsequent three-dimensional calculations of reactivity, concerns the transverse leakages, associated to each one-dimensional, user-defined core cell. For the appropriate definition of the transverse leakages several parameters concerning the core configuration must be taken into account. Moreover, the suitability of the assumptions made for the transverse cell leakages, depends on earlier user decisions, such as those made for the core partition into homogeneous cells. In the present work, the sensitivity of the calculated core reactivity to the determined leakages of the individual cells constituting the core, is studied. Moreover, appropriate assumptions concerning the transverse leakages in the one-dimensional cell calculations are searched out. The study is performed examining also the influence of the core size and the reflector existence, while the effect of the decisions made for the core partition into homogenous cells is investigated. In addition, the effect of broadened moderator channels formed within the core (e.g. by removing fuel plates to create space for control rod hosting) is also examined. Since the study required a large number of conceptual core configurations, experimental data could not be available for

Development of computer code on sodium-water reaction products transport

International Nuclear Information System (INIS)

Arikawa, H.; Yoshioka, N.; Suemori, M.; Nishida, K.

1988-01-01

The LMFBR concept eliminating the secondary sodium system has been considered to be one of the most promissing concepts for offering cost reductions. In this reactor concept, the evaluation of effects on reactor core by the sodium-water reaction products (SWRPs) during sodium-water reaction at primary steam generator becomes one of the major safety issues. In this study, the calculation code was developed as the first step of the processes of establishing the evaluation method for SWRP effects. The calculation code, called SPROUT, simulates the SWRPs transport and distribution in primary sodium system using the system geometry, thermal hydraulic data and sodium-water reacting conditions as input. This code principally models SWRPs behavior. The paper contain the modelings for SWRPs behaviors, with solution, precipation, deposition and so on, and the results and discussions of the demonstration calculation for a typical FBR plant eliminating the secondary sodium system

Deterministic modelling and stochastic simulation of biochemical pathways using MATLAB.

Science.gov (United States)

Ullah, M; Schmidt, H; Cho, K H; Wolkenhauer, O

2006-03-01

The analysis of complex biochemical networks is conducted in two popular conceptual frameworks for modelling. The deterministic approach requires the solution of ordinary differential equations (ODEs, reaction rate equations) with concentrations as continuous state variables. The stochastic approach involves the simulation of differential-difference equations (chemical master equations, CMEs) with probabilities as variables. This is to generate counts of molecules for chemical species as realisations of random variables drawn from the probability distribution described by the CMEs. Although there are numerous tools available, many of them free, the modelling and simulation environment MATLAB is widely used in the physical and engineering sciences. We describe a collection of MATLAB functions to construct and solve ODEs for deterministic simulation and to implement realisations of CMEs for stochastic simulation using advanced MATLAB coding (Release 14). The program was successfully applied to pathway models from the literature for both cases. The results were compared to implementations using alternative tools for dynamic modelling and simulation of biochemical networks. The aim is to provide a concise set of MATLAB functions that encourage the experimentation with systems biology models. All the script files are available from www.sbi.uni-rostock.de/ publications_matlab-paper.html.

Improvements to the National Transport Code Collaboration Data Server

Science.gov (United States)

Alexander, David A.

2001-10-01

The data server of the National Transport Code Colaboration Project provides a universal network interface to interpolated or raw transport data accessible by a universal set of names. Data can be acquired from a local copy of the Iternational Multi-Tokamak (ITER) profile database as well as from TRANSP trees of MDS Plus data systems on the net. Data is provided to the user's network client via a CORBA interface, thus providing stateful data server instances, which have the advantage of remembering the desired interpolation, data set, etc. This paper will review the status and discuss the recent improvements made to the data server, such as the modularization of the data server and the addition of hdf5 and MDS Plus data file writing capability.

Development of three-dimensional neoclassical transport simulation code with high performance Fortran on a vector-parallel computer

International Nuclear Information System (INIS)

Satake, Shinsuke; Okamoto, Masao; Nakajima, Noriyoshi; Takamaru, Hisanori

2005-11-01

A neoclassical transport simulation code (FORTEC-3D) applicable to three-dimensional configurations has been developed using High Performance Fortran (HPF). Adoption of computing techniques for parallelization and a hybrid simulation model to the δf Monte-Carlo method transport simulation, including non-local transport effects in three-dimensional configurations, makes it possible to simulate the dynamism of global, non-local transport phenomena with a self-consistent radial electric field within a reasonable computation time. In this paper, development of the transport code using HPF is reported. Optimization techniques in order to achieve both high vectorization and parallelization efficiency, adoption of a parallel random number generator, and also benchmark results, are shown. (author)

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.

Analysis and development of deterministic and stochastic neutron noise computing techniques with applications to thermal and fast reactors

International Nuclear Information System (INIS)

Rouchon, Amelie

2016-01-01

Neutron noise analysis addresses the description of small time-dependent flux fluctuations induced by small global or local perturbations of the macroscopic cross-sections. These fluctuations may occur in nuclear reactors due to density fluctuations of the coolant, to vibrations of fuel elements, control rods, or any other structures in the core. In power reactors, ex-core and in-core detectors can be used to monitor neutron noise with the aim of detecting possible anomalies and taking the necessary measures for continuous safe power production. The objective of this thesis is to develop techniques for neutron noise analysis and especially to implement a neutron noise solver in the deterministic transport code APOLLO3 developed at CEA. A new Monte Carlo algorithm that solves the transport equations for the neutron noise has been also developed. In addition, a new vibration model has been developed. Moreover, a method based on the determination of a new steady state has been proposed for the linear and the nonlinear full theory so as to improve the traditional neutron noise theory. In order to test these new developments we have performed neutron noise simulations in one-dimensional systems and in a large pressurized water reactor with heavy baffle in two and three dimensions with APOLLO3 in diffusion and transport theories. (author) [fr

The beta equilibrium, stability, and transport codes. Applications to the design of stellarators

International Nuclear Information System (INIS)

Bauer, F.; Garabedian, P.; Betancourt, O.; Wakatani, M.

1987-01-01

This book gives a detailed exposition of the available computational methods, documents the codes, and presents many examples showing how to run them and how to interpret the results. A listing of the recently completed BETA transport code is included. Current stellarator experiments are discussed, and the book contains significant applications to the design of major new stellarator experiments that are now in the planning stage

BLT [Breach, Leach, and Transport]: A source term computer code for low-level waste shallow land burial

International Nuclear Information System (INIS)

Suen, C.J.; Sullivan, T.M.

1990-01-01

This paper discusses the development of a source term model for low-level waste shallow land burial facilities and separates the problem into four individual compartments. These are water flow, corrosion and subsequent breaching of containers, leaching of the waste forms, and solute transport. For the first and the last compartments, we adopted the existing codes, FEMWATER and FEMWASTE, respectively. We wrote two new modules for the other two compartments in the form of two separate Fortran subroutines -- BREACH and LEACH. They were incorporated into a modified version of the transport code FEMWASTE. The resultant code, which contains all three modules of container breaching, waste form leaching, and solute transport, was renamed BLT (for Breach, Leach, and Transport). This paper summarizes the overall program structure and logistics, and presents two examples from the results of verification and sensitivity tests. 6 refs., 7 figs., 1 tab

An Implementation of Interfacial Transport Equation into the CUPID code

Energy Technology Data Exchange (ETDEWEB)

Park, Ik Kyu; Cho, Heong Kyu; Yoon, Han Young; Jeong, Jae Jun

2009-11-15

A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components for a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted a three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAS, semi-implicit ICE, SIMPLE. The governing equations for a 2-phase flow are composed of mass, momentum, and energy conservation equations for each phase. These equation sets are closed by the interfacial transfer rate of mass, momentum, and energy. The interfacial transfer of mass, momentum, and energy occurs through the interfacial area, and this area plays an important role in the transfer rate. The flow regime based correlations are used for calculating the interracial area in the traditional style 2-phase flow model. This is dependent upon the flow regime and is limited to the fully developed 2-phase flow region. Its application to the multi-dimensional 2-phase flow has some limitation because it adopts the measured results of 2-phase flow in the 1-dimensional tube. The interfacial area concentration transport equation had been suggested in order to calculate the interfacial area without the interfacial area correlations. The source terms to close the interfacial area transport equation should be further developed for a wide ranger usage of it. In this study, the one group interfacial area concentration transport equation has been implemented into the CUPID code. This interfacial area concentration transport equation can be used instead of the interfacial area concentration correlations for the bubbly flow region.

An Implementation of Interfacial Transport Equation into the CUPID code

International Nuclear Information System (INIS)

Park, Ik Kyu; Cho, Heong Kyu; Yoon, Han Young; Jeong, Jae Jun

2009-11-01

A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components for a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted a three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAS, semi-implicit ICE, SIMPLE. The governing equations for a 2-phase flow are composed of mass, momentum, and energy conservation equations for each phase. These equation sets are closed by the interfacial transfer rate of mass, momentum, and energy. The interfacial transfer of mass, momentum, and energy occurs through the interfacial area, and this area plays an important role in the transfer rate. The flow regime based correlations are used for calculating the interracial area in the traditional style 2-phase flow model. This is dependent upon the flow regime and is limited to the fully developed 2-phase flow region. Its application to the multi-dimensional 2-phase flow has some limitation because it adopts the measured results of 2-phase flow in the 1-dimensional tube. The interfacial area concentration transport equation had been suggested in order to calculate the interfacial area without the interfacial area correlations. The source terms to close the interfacial area transport equation should be further developed for a wide ranger usage of it. In this study, the one group interfacial area concentration transport equation has been implemented into the CUPID code. This interfacial area concentration transport equation can be used instead of the interfacial area concentration correlations for the bubbly flow region

Present status of reactor physics in the United States and Japan-II. 1. Deterministic Transport Methods for Reactor Analysis

International Nuclear Information System (INIS)

Adams, Marvin L.

2001-01-01

We discuss deterministic transport methods used today in neutronic analysis of nuclear reactors. This discussion is not exhaustive; our goal is to provide an overview of the methods that are most widely used for analyzing light water reactors (LWRs) and that (in our opinion) hold the most promise for the future. The current practice of LWR analysis involves the following steps: 1. Evaluate cross sections from measurements and models. 2. Obtain weighted-average cross sections over dozens to hundreds of energy intervals; the result is a 'fine-group' cross-section set. 3. [Optional] Modify the fine-group set: Further collapse it using information specific to your class of reactors and/or alter parameters so that computations better agree with experiments. The result is a 'many-group library'. 4. Perform pin cell transport calculations (usually one-dimensional cylindrical); use the results to collapse the many-group library to a medium-group set, and/or spatially average the cross sections over the pin cells. 5. Perform assembly-level transport calculations with the medium-group set. It is becoming common practice to use essentially exact geometry (no pin cell homogenization). It may soon become common to skip step 4 and use the many-group library. The output is a library of few-group cross sections, spatially averaged over the assembly, parameterized to cover the full range of operating conditions. 6. Perform full-core calculations with few-group diffusion theory that contains significant homogenizations and limited transport corrections. We discuss steps 4, 5, and 6 and focus mainly on step 5. One cannot review a large topic in a short summary without simplifying reality, omitting important details, and neglecting some methods that deserve attention; for this we apologize in advance. (author)

Extending CANTUP code analysis to probabilistic evaluations

International Nuclear Information System (INIS)

Florea, S.

2001-01-01

The structural analysis with numerical methods based on final element method plays at present a central role in evaluations and predictions of structural systems which require safety and reliable operation in aggressive environmental conditions. This is the case too for the CANDU - 600 fuel channel, where besides the corrosive and thermal aggression upon the Zr97.5Nb2.5 pressure tubes, a lasting irradiation adds which has marked consequences upon the materials properties evolution. This results in an unavoidable spreading in the materials properties in time, affected by high uncertainties. Consequently, the deterministic evaluation with computation codes based on finite element method are supplemented by statistic and probabilistic methods of evaluation of the response of structural components. This paper reports the works on extending the thermo-mechanical evaluation of the fuel channel components in the frame of probabilistic structure mechanics based on statistical methods and developed upon deterministic CANTUP code analyses. CANTUP code was adapted from LAHEY 77 platform onto Microsoft Developer Studio - Fortran Power Station 4.0 platform. To test the statistical evaluation of the creeping behaviour of pressure tube, the value of longitudinal elasticity modulus (Young) was used, as random variable, with a normal distribution around value, as used in deterministic analyses. The influence of the random quantity upon the hog and effective stress developed in the pressure tube for to time values, specific to primary and secondary creep was studied. The results obtained after a five year creep, corresponding to the secondary creep are presented

BLINDAGE: A neutron and gamma-ray transport code for shieldings with the removal-diffusion technique coupled with the point-kernel technique

International Nuclear Information System (INIS)

Fanaro, L.C.C.B.

1984-01-01

It was developed the BLINDAGE computer code for the radiation transport (neutrons and gammas) calculation. The code uses the removal - diffusion method for neutron transport and point-kernel technique with buil-up factors for gamma-rays. The results obtained through BLINDAGE code are compared with those obtained with the ANISN and SABINE computer codes. (Author) [pt

Core 2D. A code for non-isothermal water flow and reactive solute transport. Users manual version 2

Energy Technology Data Exchange (ETDEWEB)

Samper, J.; Juncosa, R.; Delgado, J.; Montenegro, L. [Universidad de A Coruna (Spain)

2000-07-01

Understanding natural groundwater quality patterns, quantifying groundwater pollution and assessing the effects of waste disposal, require modeling tools accounting for water flow, and transport of heat and dissolved species as well as their complex interactions with solid and gases phases. This report contains the users manual of CORE ''2D Version V.2.0, a COde for modeling water flow (saturated and unsaturated), heat transport and multicomponent Reactive solute transport under both local chemical equilibrium and kinetic conditions. it is an updated and improved version of CORE-LE-2D V0 (Samper et al., 1988) which in turns is an extended version of TRANQUI, a previous reactive transport code (ENRESA, 1995). All these codes were developed within the context of Research Projects funded by ENRESA and the European Commission. (Author)

Core2D. A code for non-isothermal water flow and reactive solute transport. Users manual version 2

International Nuclear Information System (INIS)

Samper, J.; Juncosa, R.; Delgado, J.; Montenegro, L.

2000-01-01

Understanding natural groundwater quality patterns, quantifying groundwater pollution and assessing the effects of waste disposal, require modeling tools accounting for water flow, and transport of heat and dissolved species as well as their complex interactions with solid and gases phases. This report contains the users manual of CORE ''2D Version V.2.0, a COde for modeling water flow (saturated and unsaturated), heat transport and multicomponent Reactive solute transport under both local chemical equilibrium and kinetic conditions. it is an updated and improved version of CORE-LE-2D V0 (Samper et al., 1988) which in turns is an extended version of TRANQUI, a previous reactive transport code (ENRESA, 1995). All these codes were developed within the context of Research Projects funded by ENRESA and the European Commission. (Author)

Core 2D. A code for non-isothermal water flow and reactive solute transport. Users manual version 2

Energy Technology Data Exchange (ETDEWEB)

Samper, J; Juncosa, R; Delgado, J; Montenegro, L [Universidad de A Coruna (Spain)

2000-07-01

Understanding natural groundwater quality patterns, quantifying groundwater pollution and assessing the effects of waste disposal, require modeling tools accounting for water flow, and transport of heat and dissolved species as well as their complex interactions with solid and gases phases. This report contains the users manual of CORE ''2D Version V.2.0, a COde for modeling water flow (saturated and unsaturated), heat transport and multicomponent Reactive solute transport under both local chemical equilibrium and kinetic conditions. it is an updated and improved version of CORE-LE-2D V0 (Samper et al., 1988) which in turns is an extended version of TRANQUI, a previous reactive transport code (ENRESA, 1995). All these codes were developed within the context of Research Projects funded by ENRESA and the European Commission. (Author)

Evaluation of scattering laws and cross sections for calculation of production and transport of cold and ultracold neutrons

Energy Technology Data Exchange (ETDEWEB)

Bernnat, W.; Keinert, J.; Mattes, M. [Inst. for Nuclear Energy and Energy Systems, Univ. of Stuttgart, Stuttgart (Germany)

2004-03-01

For the calculation of neutron spectra in cold and super thermal sources scattering laws for a variety of liquid and solid cyrogenic materials were evaluated and prepared for use in deterministic and Monte Carlo transport calculations. For moderator materials like liquid and solid H{sub 2}O, liquid He, liquid D{sub 2}O, liquid and solid H{sub 2} and D{sub 2}, solid CH{sub 4} and structure materials such as Al, Bi, Pb, ZrHx, and graphite scattering law data and cross sections are available. The evaluated data were validated by comparison with measured cross sections and comparison of measured and calculated neutron spectra as far as available. Further applications are the calculation of production and transport and storing of ultra cold neutrons (UCN) in different UCN sources. The data structures of the evaluated data are prepared for the common S{sub N}-transport codes and the Monte Carlo Code MCNP. (orig.)

Current status of radiation transport tools for proliferation and terrorism prevention

International Nuclear Information System (INIS)

Sale, K.E.

2004-01-01

Full text: We will present the current status and future plans for the set of calculational tools and databases developed and maintained at LLNL. The calculational tools include the Monte Carlo codes TART 1) and COG 2) as well as the deterministic code ARDRA 3) . In addition to these codes we use currently there is a major development effort for a new massively parallel transport code. An important part of the capability we're developing is a sophisticated user interface, based on a commercial 3-D modeling product, to improve the model development process. A major part of this user interface tool is being developed by Strela 4) under the Nuclear Cities Initiative. Strela has developed a hub-and-spoke technology for code input interconversions (between COG, TART and MCNP) and will produce the plug-ins that extend the capabilities of the 3-D modeler for use as a radiation transport input generator. The major advantages of this approach are the built-in user interface for 3-D modeling and the ability to read a large variety of CAD-file formats. In addition to supporting our current radiation transport codes and developing new capabilities we are working on some nuclear data needs for homeland security. These projects are carried out and the Lawrence Berkeley National Laboratory 88' cyclotron and at the Institute for Nuclear Research of the Nation Academy of Science of Ukraine under and STCU contract. Reference: 1. http://www.llnl.gov/cullen1/mc/htm; 2. http://www-phys.llnl.gov/N_Div/COG/ETR/ETR_9306.html; 3. http://www.llnl.gov/CASC/asciturb/talks/PPT-HTML.131596/tsld030.htm; 4. http://strela.snz.ru/

Particle and heavy ion transport code system, PHITS, version 2.52

International Nuclear Information System (INIS)

Sato, Tatsuhiko; Matsuda, Norihiro; Hashimoto, Shintaro; Iwamoto, Yosuke; Noda, Shusaku; Ogawa, Tatsuhiko; Nakashima, Hiroshi; Fukahori, Tokio; Okumura, Keisuke; Kai, Tetsuya; Niita, Koji; Iwase, Hiroshi; Chiba, Satoshi; Furuta, Takuya; Sihver, Lembit

2013-01-01

An upgraded version of the Particle and Heavy Ion Transport code System, PHITS2.52, was developed and released to the public. The new version has been greatly improved from the previously released version, PHITS2.24, in terms of not only the code itself but also the contents of its package, such as the attached data libraries. In the new version, a higher accuracy of simulation was achieved by implementing several latest nuclear reaction models. The reliability of the simulation was improved by modifying both the algorithms for the electron-, positron-, and photon-transport simulations and the procedure for calculating the statistical uncertainties of the tally results. Estimation of the time evolution of radioactivity became feasible by incorporating the activation calculation program DCHAIN-SP into the new package. The efficiency of the simulation was also improved as a result of the implementation of shared-memory parallelization and the optimization of several time-consuming algorithms. Furthermore, a number of new user-support tools and functions that help users to intuitively and effectively perform PHITS simulations were developed and incorporated. Due to these improvements, PHITS is now a more powerful tool for particle transport simulation applicable to various research and development fields, such as nuclear technology, accelerator design, medical physics, and cosmic-ray research. (author)

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)

Contribution of the deterministic approach to the characterization of seismic input

International Nuclear Information System (INIS)

Panza, G.F.; Romanelli, F.; Vaccari, F.; Decanini, L.; Mollaioli, F.

1999-10-01

Traditional methods use either a deterministic or a probabilistic approach, based on empirically derived laws for ground motion attenuation. The realistic definition of seismic input can be performed by means of advanced modelling codes based on the modal summation technique. These codes and their extension to laterally heterogeneous structures allow us to accurately calculate synthetic signals, complete of body waves and of surface waves, corresponding to different source and anelastic structural models, taking into account the effect of local geological conditions. This deterministic approach is capable to address some aspects largely overlooked in the probabilistic approach: (a) the effect of crustal properties on attenuation are not neglected; (b) the ground motion parameters are derived from synthetic time histories. and not from overly simplified attenuation functions; (c) the resulting maps are in terms of design parameters directly, and do not require the adaptation of probabilistic maps to design ground motions; and (d) such maps address the issue of the deterministic definition of ground motion in a way which permits the generalization of design parameters to locations where there is little seismic history. The methodology has been applied to a large part of south-eastern Europe, in the framework of the EU-COPERNICUS project 'Quantitative Seismic Zoning of the Circum Pannonian Region'. Maps of various seismic hazard parameters numerically modelled, and whenever possible tested against observations, such as peak ground displacement, velocity and acceleration, of practical use for the design of earthquake-safe structures, have been produced. The results of a standard probabilistic approach are compared with the findings based on the deterministic approach. A good agreement is obtained except for the Vrancea (Romania) zone, where the attenuation relations used in the probabilistic approach seem to underestimate, mainly at large distances, the seismic hazard

Draft ASME code case on ductile cast iron for transport packaging

International Nuclear Information System (INIS)

Saegusa, T.; Arai, T.; Hirose, M.; Kobayashi, T.; Tezuka, Y.; Urabe, N.; Hueggenberg, R.

2004-01-01

The current Rules for Construction of ''Containment Systems for Storage and Transport Packagings of Spent Nuclear Fuel and High Level Radioactive Material and Waste'' of Division 3 in Section III of ASME Code (2001 Edition) does not include ductile cast iron in its list of materials permitted for use. The Rules specify required fracture toughness values of ferritic steel material for nominal wall thickness 5/8 to 12 inches (16 to 305 mm). New rule for ductile cast iron for transport packaging of which wall thickness is greater than 12 inches (305mm) is required

Computer codes for three dimensional mass transport with non-linear sorption

International Nuclear Information System (INIS)

Noy, D.J.

1985-03-01

The report describes the mathematical background and data input to finite element programs for three dimensional mass transport in a porous medium. The transport equations are developed and sorption processes are included in a general way so that non-linear equilibrium relations can be introduced. The programs are described and a guide given to the construction of the required input data sets. Concluding remarks indicate that the calculations require substantial computer resources and suggest that comprehensive preliminary analysis with lower dimensional codes would be important in the assessment of field data. (author)

Development of a computer code for low-and intermediate-level radioactive waste disposal safety assessment

International Nuclear Information System (INIS)

Park, J. W.; Kim, C. L.; Lee, E. Y.; Lee, Y. M.; Kang, C. H.; Zhou, W.; Kozak, M. W.

2002-01-01

A safety assessment code, called SAGE (Safety Assessment Groundwater Evaluation), has been developed to describe post-closure radionuclide releases and potential radiological doses for low- and intermediate-level radioactive waste (LILW) disposal in an engineered vault facility in Korea. The conceptual model implemented in the code is focused on the release of radionuclide from a gradually degrading engineered barrier system to an underlying unsaturated zone, thence to a saturated groundwater zone. The radionuclide transport equations are solved by spatially discretizing the disposal system into a series of compartments. Mass transfer between compartments is by diffusion/dispersion and advection. In all compartments, radionuclides are decayed either as a single-member chain or as multi-member chains. The biosphere is represented as a set of steady-state, radionuclide-specific pathway dose conversion factors that are multiplied by the appropriate release rate from the far field for each pathway. The code has the capability to treat input parameters either deterministically or probabilistically. Parameter input is achieved through a user-friendly Graphical User Interface. An application is presented, which is compared against safety assessment results from the other computer codes, to benchmark the reliability of system-level conceptual modeling of the code

An upgraded version of the nucleon meson transport code: NMTC/JAERI97

Energy Technology Data Exchange (ETDEWEB)

Takada, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yoshizawa, Nobuaki; Kosako, Kazuaki; Ishibashi, Kenji

1998-02-01

The nucleon-meson transport code NMTC/JAERI is upgraded to NMTC/JAERI97 which has new features not only in physics model and nuclear data but also in computational procedure. NMTC/JAERI97 implements the following two new physics models: an intranuclear cascade model taking account of the in-medium nuclear effects and the preequilibrium calculation model based on the exciton one. For treating the nucleon transport process more accurately, the nucleon-nucleus cross sections are revised to those derived by the systematics of Pearlstein. Moreover, the level density parameter derived by Ignatyuk is included as a new option for particle evaporation calculation. Other than those physical aspects, a new geometry package based on the Combinatorial Geometry with multi-array system and the importance sampling technique are implemented in the code. Tally function is also employed for obtaining such physical quantities as neutron energy spectra, heat deposition and nuclide yield without editing a history file. The resultant NMTC/JAERI97 is tuned to be executed on the UNIX system. This paper explains about the function, physics models and geometry model adopted in NMTC/JAERI97 and guides how to use the code. (author)

Savannah River Laboratory DOSTOMAN code: a compartmental pathways computer model of contaminant transport

International Nuclear Information System (INIS)

King, C.M.; Wilhite, E.L.; Root, R.W. Jr.

1985-01-01

The Savannah River Laboratory DOSTOMAN code has been used since 1978 for environmental pathway analysis of potential migration of radionuclides and hazardous chemicals. The DOSTOMAN work is reviewed including a summary of historical use of compartmental models, the mathematical basis for the DOSTOMAN code, examples of exact analytical solutions for simple matrices, methods for numerical solution of complex matrices, and mathematical validation/calibration of the SRL code. The review includes the methodology for application to nuclear and hazardous chemical waste disposal, examples of use of the model in contaminant transport and pathway analysis, a user's guide for computer implementation, peer review of the code, and use of DOSTOMAN at other Department of Energy sites. 22 refs., 3 figs

Analysis of the anisotropy effects with the AZTRAN code

International Nuclear Information System (INIS)

Xolocostli, V.; Vargas, S.; Gomez, A.; Del Valle, E.

2017-09-01

Among the improvements that are made for the deterministic codes with which nuclear reactors are analyzed, is the implementation of the dispersion anisotropic dispersion section, which can obtain better results. With the current computing technology is possible to carry out these implementations, since the computation time is no longer a considerable problem as in the past. In this paper we analyze some effects of anisotropy in the AZTRAN code, a code that solves the Boltzmann transport equation in one, two and three dimensions at steady state, using the multigroup technique, the nodal method RTN-0 and ordered discrete, which is part of the AZTLAN platform for analysis of nuclear reactors, which is currently under development. The implementation of the anisotropy in the AZTRAN code is one of the latest improvements that have been made to the code, leading to different tests and analyzes regarding the anisotropic dispersion, some as a test with homogeneous fuel assemblies. In the case presented here, the benchmark problem of a fuel assembly type BWR is analyzed, which is part of the Benchmark problem suite for reactor physics study of LWR next generation fuels, proposed by the Committee on Reactor Physics organized by the Japan Atomic Energy Research Institute (JAERI). In this problem the behavior of the infinite multiplication factor (k inf ) is analyzed, as well as the behavior of using odd and even anisotropy approximation with respect to the symmetry in the radial power of the assembly. (Author)

Progress in multidimensional neutron transport computation

International Nuclear Information System (INIS)

Lewis, E.E.

1977-01-01

The methods available for solution of the time-independent neutron transport problems arising in the analysis of nuclear systems are examined. The merits of deterministic and Monte Carlo methods are briefly compared. The capabilities of deterministic computational methods derived from the first-order form of the transport equation, from the second-order even-parity form of this equation, and from integral transport formulations are discussed in some detail. Emphasis is placed on the approaches for dealing with the related problems of computer memory requirements, computational cost, and achievable accuracy. Attention is directed to some areas where problems exist currently and where the need for further work appears to be particularly warranted

Resolution of the neutron transport equation by massively parallel computer in the Cronos code

International Nuclear Information System (INIS)

Zardini, D.M.

1996-01-01

The feasibility of neutron transport problems parallel resolution by CRONOS code's SN module is here studied. In this report we give the first data about the parallel resolution by angular variable decomposition of the transport equation. Problems about parallel resolution by spatial variable decomposition and memory stage limits are also explained here. (author)

Development and preliminary verification of 2-D transport module of radiation shielding code ARES

International Nuclear Information System (INIS)

Zhang Penghe; Chen Yixue; Zhang Bin; Zang Qiyong; Yuan Longjun; Chen Mengteng

2013-01-01

The 2-D transport module of radiation shielding code ARES is two-dimensional neutron and radiation shielding code. The theory model was based on the first-order steady state neutron transport equation, adopting the discrete ordinates method to disperse direction variables. Then a set of differential equations can be obtained and solved with the source iteration method. The 2-D transport module of ARES was capable of calculating k eff and fixed source problem with isotropic or anisotropic scattering in x-y geometry. The theoretical model was briefly introduced and series of benchmark problems were verified in this paper. Compared with the results given by the benchmark, the maximum relative deviation of k eff is 0.09% and the average relative deviation of flux density is about 0.60% in the BWR cells benchmark problem. As for the fixed source problem with isotropic and anisotropic scattering, the results of the 2-D transport module of ARES conform with DORT very well. These numerical results of benchmark problems preliminarily demonstrate that the development process of the 2-D transport module of ARES is right and it is able to provide high precision result. (authors)

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...

Enhanced deterministic phase retrieval using a partially developed speckle field

DEFF Research Database (Denmark)

Almoro, Percival F.; Waller, Laura; Agour, Mostafa

2012-01-01

A technique for enhanced deterministic phase retrieval using a partially developed speckle field (PDSF) and a spatial light modulator (SLM) is demonstrated experimentally. A smooth test wavefront impinges on a phase diffuser, forming a PDSF that is directed to a 4f setup. Two defocused speckle...... intensity measurements are recorded at the output plane corresponding to axially-propagated representations of the PDSF in the input plane. The speckle intensity measurements are then used in a conventional transport of intensity equation (TIE) to reconstruct directly the test wavefront. The PDSF in our...

GANTRAS - a system of codes for the solution of the multigroup transport equation with a rigorous treatment of anisotropic neutron scattering

International Nuclear Information System (INIS)

Schwenk-Ferrero, A.

1986-11-01

GANTRAS is a system of codes for neutron transport calculations in which the anisotropy of elastic and inelastic (including (n,n'x)-reactions) scattering is fully taken into account. This is achieved by employing a rigorous method, so-called I * -method, to represent the scattering term of the transport equation and with the use of double-differential cross-sections for the description of the emission of secondary neutrons. The I * -method was incorporated into the conventional transport code ONETRAN. The ONETRAN subroutines were modified for the new purpose. An implementation of the updated version ANTRA1 was accomplished for plane and spherical geometry. ANTRA1 was included in GANTRAS and linked to another modules which prepare angle-dependent transfer matrices. The GANTRAS code consists of three modules: 1. The CROMIX code which calculates the macroscopic transfer matrices for mixtures on the base of microscopic nuclide-dependent data. 2. The ATP code which generates discretized angular transfer probabilities (i.e. discretizes the I * -function). 3. The ANTRA1 code to perform S N transport calculations in one-dimensional plane and spherical geometries. This structure of GANTRAS allows to accommodate the system to various transport problems. (orig.) [de

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-06-07

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

Open-Source Development of the Petascale Reactive Flow and Transport Code PFLOTRAN

Science.gov (United States)

Hammond, G. E.; Andre, B.; Bisht, G.; Johnson, T.; Karra, S.; Lichtner, P. C.; Mills, R. T.

2013-12-01

Open-source software development has become increasingly popular in recent years. Open-source encourages collaborative and transparent software development and promotes unlimited free redistribution of source code to the public. Open-source development is good for science as it reveals implementation details that are critical to scientific reproducibility, but generally excluded from journal publications. In addition, research funds that would have been spent on licensing fees can be redirected to code development that benefits more scientists. In 2006, the developers of PFLOTRAN open-sourced their code under the U.S. Department of Energy SciDAC-II program. Since that time, the code has gained popularity among code developers and users from around the world seeking to employ PFLOTRAN to simulate thermal, hydraulic, mechanical and biogeochemical processes in the Earth's surface/subsurface environment. PFLOTRAN is a massively-parallel subsurface reactive multiphase flow and transport simulator designed from the ground up to run efficiently on computing platforms ranging from the laptop to leadership-class supercomputers, all from a single code base. The code employs domain decomposition for parallelism and is founded upon the well-established and open-source parallel PETSc and HDF5 frameworks. PFLOTRAN leverages modern Fortran (i.e. Fortran 2003-2008) in its extensible object-oriented design. The use of this progressive, yet domain-friendly programming language has greatly facilitated collaboration in the code's software development. Over the past year, PFLOTRAN's top-level data structures were refactored as Fortran classes (i.e. extendible derived types) to improve the flexibility of the code, ease the addition of new process models, and enable coupling to external simulators. For instance, PFLOTRAN has been coupled to the parallel electrical resistivity tomography code E4D to enable hydrogeophysical inversion while the same code base can be used as a third

Development and application of a deterministic-realistic hybrid methodology for LOCA licensing analysis

International Nuclear Information System (INIS)

Liang, Thomas K.S.; Chou, Ling-Yao; Zhang, Zhongwei; Hsueh, Hsiang-Yu; Lee, Min

2011-01-01

Highlights: → A new LOCA licensing methodology (DRHM, deterministic-realistic hybrid methodology) was developed. → DRHM involves conservative Appendix K physical models and statistical treatment of plant status uncertainties. → DRHM can generate 50-100 K PCT margin as compared to a traditional Appendix K methodology. - Abstract: It is well recognized that a realistic LOCA analysis with uncertainty quantification can generate greater safety margin as compared with classical conservative LOCA analysis using Appendix K evaluation models. The associated margin can be more than 200 K. To quantify uncertainty in BELOCA analysis, generally there are two kinds of uncertainties required to be identified and quantified, which involve model uncertainties and plant status uncertainties. Particularly, it will take huge effort to systematically quantify individual model uncertainty of a best estimate LOCA code, such as RELAP5 and TRAC. Instead of applying a full ranged BELOCA methodology to cover both model and plant status uncertainties, a deterministic-realistic hybrid methodology (DRHM) was developed to support LOCA licensing analysis. Regarding the DRHM methodology, Appendix K deterministic evaluation models are adopted to ensure model conservatism, while CSAU methodology is applied to quantify the effect of plant status uncertainty on PCT calculation. Generally, DRHM methodology can generate about 80-100 K margin on PCT as compared to Appendix K bounding state LOCA analysis.

Assessment of shielding analysis methods, codes, and data for spent fuel transport/storage applications

International Nuclear Information System (INIS)

Parks, C.V.; Broadhead, B.L.; Hermann, O.W.; Tang, J.S.; Cramer, S.N.; Gauthey, J.C.; Kirk, B.L.; Roussin, R.W.

1988-07-01

This report provides a preliminary assessment of the computational tools and existing methods used to obtain radiation dose rates from shielded spent nuclear fuel and high-level radioactive waste (HLW). Particular emphasis is placed on analysis tools and techniques applicable to facilities/equipment designed for the transport or storage of spent nuclear fuel or HLW. Applications to cask transport, storage, and facility handling are considered. The report reviews the analytic techniques for generating appropriate radiation sources, evaluating the radiation transport through the shield, and calculating the dose at a desired point or surface exterior to the shield. Discrete ordinates, Monte Carlo, and point kernel methods for evaluating radiation transport are reviewed, along with existing codes and data that utilize these methods. A literature survey was employed to select a cadre of codes and data libraries to be reviewed. The selection process was based on specific criteria presented in the report. Separate summaries were written for several codes (or family of codes) that provided information on the method of solution, limitations and advantages, availability, data access, ease of use, and known accuracy. For each data library, the summary covers the source of the data, applicability of these data, and known verification efforts. Finally, the report discusses the overall status of spent fuel shielding analysis techniques and attempts to illustrate areas where inaccuracy and/or uncertainty exist. The report notes the advantages and limitations of several analysis procedures and illustrates the importance of using adequate cross-section data sets. Additional work is recommended to enable final selection/validation of analysis tools that will best meet the US Department of Energy's requirements for use in developing a viable HLW management system. 188 refs., 16 figs., 27 tabs

Proving Non-Deterministic Computations in Agda

Directory of Open Access Journals (Sweden)

Sergio Antoy

2017-01-01

Full Text Available We investigate proving properties of Curry programs using Agda. First, we address the functional correctness of Curry functions that, apart from some syntactic and semantic differences, are in the intersection of the two languages. Second, we use Agda to model non-deterministic functions with two distinct and competitive approaches incorporating the non-determinism. The first approach eliminates non-determinism by considering the set of all non-deterministic values produced by an application. The second approach encodes every non-deterministic choice that the application could perform. We consider our initial experiment a success. Although proving properties of programs is a notoriously difficult task, the functional logic paradigm does not seem to add any significant layer of difficulty or complexity to the task.

Guidelines for selecting codes for ground-water transport modeling of low-level waste burial sites. Executive summary

International Nuclear Information System (INIS)

Simmons, C.S.; Cole, C.R.

1985-05-01

This document was written to provide guidance to managers and site operators on how ground-water transport codes should be selected for assessing burial site performance. There is a need for a formal approach to selecting appropriate codes from the multitude of potentially useful ground-water transport codes that are currently available. Code selection is a problem that requires more than merely considering mathematical equation-solving methods. These guidelines are very general and flexible and are also meant for developing systems simulation models to be used to assess the environmental safety of low-level waste burial facilities. Code selection is only a single aspect of the overall objective of developing a systems simulation model for a burial site. The guidance given here is mainly directed toward applications-oriented users, but managers and site operators need to be familiar with this information to direct the development of scientifically credible and defensible transport assessment models. Some specific advice for managers and site operators on how to direct a modeling exercise is based on the following five steps: identify specific questions and study objectives; establish costs and schedules for achieving answers; enlist the aid of professional model applications group; decide on approach with applications group and guide code selection; and facilitate the availability of site-specific data. These five steps for managers/site operators are discussed in detail following an explanation of the nine systems model development steps, which are presented first to clarify what code selection entails

ITS version 5.0 :the integrated TIGER series of coupled electron/Photon monte carlo transport codes with CAD geometry.

Energy Technology Data Exchange (ETDEWEB)

Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William

2005-09-01

ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-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 5.0, the latest version of ITS, contains (1) improvements to the ITS 3.0 continuous-energy codes, (2) multigroup codes with adjoint transport capabilities, (3) parallel implementations of all ITS codes, (4) a general purpose geometry engine for linking with CAD or other geometry formats, and (5) the Cholla facet geometry library. Moreover, the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability.

SCORCH - a zero dimensional plasma evolution and transport code for use in small and large tokamak systems

International Nuclear Information System (INIS)

Clancy, B.E.; Cook, J.L.

1984-12-01

The zero-dimensional code SCORCH determines number density and temperature evolution in plasmas using concepts derived from the Hinton and Hazeltine transport theory. The code uses the previously reported ADL-1 data library

Use of the APOLLO2 transport code for PWR assembly studies

International Nuclear Information System (INIS)

Belhaffaf, D.; Coste, M.; Lenain, R.; Mathonniere, G.; Sanchez, R.; Stankovski, Z.; Zmijarevic, I.

1992-01-01

This paper presents some validation and application aspects of the APOLLO2, user oriented, modular code for multigroup transport assembly calculation which is developed at the French Commissariat a l'Energies Atomique. The main points approached in this paper are: the two dimensional collision probability convergence, critical leakage calculation schemes, self-shielding spatial discretization, and the equivalence procedure

MAX: an expert system for running the modular transport code APOLLO II

International Nuclear Information System (INIS)

Loussouarn, O.; Ferraris, C.; Boivineau, A.

1990-01-01

MAX is an expert system built to help users of the APOLLO II code to prepare the input data deck to run a job. APOLLO II is a modular transport-theory code for calculating the neutron flux in various geometries. The associated GIBIANE command language allows the user to specify the physical structure and the computational method to be used in the calculation. The purpose of MAX is to bring into play expertise in both neutronic and computing aspects of the code, as well as various computational schemes, in order to generate automatically a batch data set corresponding to the APOLLO II calculation desired by the user. MAX is implemented on the SUN 3/60 workstation with the S1 tool and graphic interface external functions

Radiation transport phenomena and modeling. Part A: Codes; Part B: Applications with examples

International Nuclear Information System (INIS)

Lorence, L.J. Jr.; Beutler, D.E.

1997-09-01

This report contains the notes from the second session of the 1997 IEEE Nuclear and Space Radiation Effects Conference Short Course on Applying Computer Simulation Tools to Radiation Effects Problems. Part A discusses the physical phenomena modeled in radiation transport codes and various types of algorithmic implementations. Part B gives examples of how these codes can be used to design experiments whose results can be easily analyzed and describes how to calculate quantities of interest for electronic devices

A computer code PACTOLE to predict activation and transport of corrosion products in a PWR

International Nuclear Information System (INIS)

Beslu, P.; Frejaville, G.; Lalet, A.

1978-01-01

Theoretical studies on activation and transport of corrosion products in a PWR primary circuit have been concentrated, at CEA on the development of a computer code : PACTOLE. This code takes into account the major phenomena which govern corrosion products transport: 1. Ion solubility is obtained by usual thermodynamics laws in function of water chemistry: pH at operating temperature is calculated by the code. 2. Release rates of base metals, dissolution rates of deposits, precipitation rates of soluble products are derived from solubility variations. 3. Deposition of solid particles is treated by a model taking into account particle size, brownian and turbulent diffusion and inertial effect. Erosion of deposits is accounted for by a semi-empirical model. After a review of calculational models, an application of PACTOLE is presented in view of analyzing the distribution of in core. (author)

A one-and-a-quarter-dimensional transport code for field-reversed configuration studies: A user's guide for CFRX

International Nuclear Information System (INIS)

Hsiao, Ming-Yuan; Werley, K.A.; Ling, Kuok-Mee.

1988-05-01

A one-and-a-quarter-dimensional transport code, which includes radial as well as some two-dimensional effects for field-reversed configurations, is described. The set of transport equations is transformed to a set of new independent and dependent variables and is solved as a coupled initial-boundary value problem. The code simulation includes both the closed and open field regions. The axial effects incorporated include global axial force balance, axial losses in the open field region, and flux surface averaging over the closed field region. Input, output, and structure of the code are described in detail. A typical example of the code results is also given. 20 refs., 21 figs., 7 tabs

Theory and application of deterministic multidimensional pointwise energy lattice physics methods

International Nuclear Information System (INIS)

Zerkle, M.L.

1999-01-01

The theory and application of deterministic, multidimensional, pointwise energy lattice physics methods are discussed. These methods may be used to solve the neutron transport equation in multidimensional geometries using near-continuous energy detail to calculate equivalent few-group diffusion theory constants that rigorously account for spatial and spectral self-shielding effects. A dual energy resolution slowing down algorithm is described which reduces the computer memory and disk storage requirements for the slowing down calculation. Results are presented for a 2D BWR pin cell depletion benchmark problem

Deterministic Echo State Networks Based Stock Price Forecasting