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

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

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

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)

Numerical modelling of coupled fluid, heat, and solute transport in deformable fractured rock

International Nuclear Information System (INIS)

Chan, T.; Reid, J.A.K.

1987-01-01

This paper reports on a three-dimensional (3D) finite-element code, MOTIF (model of transport in fractured/porous media), developed to model the coupled processes of groundwater flow, heat transport, brine transport, and one-species radionuclide transport in geological media. Three types of elements are available: a 3D continuum element, a planar fracture element that can be oriented in any arbitrary direction in 3D space or pipe flow in 3D space, and a line element for simulating fracture flow in 2D space or pipe flow in 3D space. As a quality-assurance measure, the MOTIF code was verified by comparison of its results with analytical solutions and other published numerical solutions

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

Semianalytical solutions for contaminant transport under variable velocity field in a coastal aquifer

Science.gov (United States)

Koohbor, Behshad; Fahs, Marwan; Ataie-Ashtiani, Behzad; Simmons, Craig T.; Younes, Anis

2018-05-01

Existing closed-form solutions of contaminant transport problems are limited by the mathematically convenient assumption of uniform flow. These solutions cannot be used to investigate contaminant transport in coastal aquifers where seawater intrusion induces a variable velocity field. An adaptation of the Fourier-Galerkin method is introduced to obtain semi-analytical solutions for contaminant transport in a confined coastal aquifer in which the saltwater wedge is in equilibrium with a freshwater discharge flow. Two scenarios dealing with contaminant leakage from the aquifer top surface and contaminant migration from a source at the landward boundary are considered. Robust implementation of the Fourier-Galerkin method is developed to efficiently solve the coupled flow, salt and contaminant transport equations. Various illustrative examples are generated and the semi-analytical solutions are compared against an in-house numerical code. The Fourier series are used to evaluate relevant metrics characterizing contaminant transport such as the discharge flux to the sea, amount of contaminant persisting in the groundwater and solute flux from the source. These metrics represent quantitative data for numerical code validation and are relevant to understand the effect of seawater intrusion on contaminant transport. It is observed that, for the surface contamination scenario, seawater intrusion limits the spread of the contaminant but intensifies the contaminant discharge to the sea. For the landward contamination scenario, moderate seawater intrusion affects only the spatial distribution of the contaminant plume while extreme seawater intrusion can increase the contaminant discharge to the sea. The developed semi-analytical solution presents an efficient tool for the verification of numerical models. It provides a clear interpretation of the contaminant transport processes in coastal aquifers subject to seawater intrusion. For practical usage in further studies, the full

Modeling variably saturated subsurface solute transport with MODFLOW-UZF and MT3DMS

Science.gov (United States)

Morway, Eric D.; Niswonger, Richard G.; Langevin, Christian D.; Bailey, Ryan T.; Healy, Richard W.

2013-01-01

The MT3DMS groundwater solute transport model was modified to simulate solute transport in the unsaturated zone by incorporating the unsaturated-zone flow (UZF1) package developed for MODFLOW. The modified MT3DMS code uses a volume-averaged approach in which Lagrangian-based UZF1 fluid fluxes and storage changes are mapped onto a fixed grid. Referred to as UZF-MT3DMS, the linked model was tested against published benchmarks solved analytically as well as against other published codes, most frequently the U.S. Geological Survey's Variably-Saturated Two-Dimensional Flow and Transport Model. Results from a suite of test cases demonstrate that the modified code accurately simulates solute advection, dispersion, and reaction in the unsaturated zone. Two- and three-dimensional simulations also were investigated to ensure unsaturated-saturated zone interaction was simulated correctly. Because the UZF1 solution is analytical, large-scale flow and transport investigations can be performed free from the computational and data burdens required by numerical solutions to Richards' equation. Results demonstrate that significant simulation runtime savings can be achieved with UZF-MT3DMS, an important development when hundreds or thousands of model runs are required during parameter estimation and uncertainty analysis. Three-dimensional variably saturated flow and transport simulations revealed UZF-MT3DMS to have runtimes that are less than one tenth of the time required by models that rely on Richards' equation. Given its accuracy and efficiency, and the wide-spread use of both MODFLOW and MT3DMS, the added capability of unsaturated-zone transport in this familiar modeling framework stands to benefit a broad user-ship.

Progress in nuclear well logging modeling using deterministic transport codes

International Nuclear Information System (INIS)

Kodeli, I.; Aldama, D.L.; Maucec, M.; Trkov, A.

2002-01-01

Further studies in continuation of the work presented in 2001 in Portoroz were performed in order to study and improve the performances, precission and domain of application of the deterministic transport codes with respect to the oil well logging analysis. These codes are in particular expected to complement the Monte Carlo solutions, since they can provide a detailed particle flux distribution in the whole geometry in a very reasonable CPU time. Real-time calculation can be envisaged. The performances of deterministic transport methods were compared to those of the Monte Carlo method. IRTMBA generic benchmark was analysed using the codes MCNP-4C and DORT/TORT. Centric as well as excentric casings were considered using 14 MeV point neutron source and NaI scintillation detectors. Neutron and gamma spectra were compared at two detector positions.(author)

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

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

ITS Version 3.0: The Integrated TIGER Series of coupled electron/photon Monte Carlo transport codes

International Nuclear Information System (INIS)

Halbleib, J.A.; Kensek, R.P.; Valdez, G.D.; Mehlhorn, T.A.; Seltzer, S.M.; Berger, M.J.

1993-01-01

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. It combines operational simplicity and physical accuracy in order to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Flexibility of construction permits tailoring of the codes to specific applications and extension of code capabilities to more complex applications through simple update procedures

ITS Version 3.0: The Integrated TIGER Series of coupled electron/photon Monte Carlo transport codes

Energy Technology Data Exchange (ETDEWEB)

Halbleib, J.A.; Kensek, R.P.; Valdez, G.D.; Mehlhorn, T.A. [Sandia National Labs., Albuquerque, NM (United States); Seltzer, S.M.; Berger, M.J. [National Inst. of Standards and Technology, Gaithersburg, MD (United States). Ionizing Radiation Div.

1993-06-01

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. It combines operational simplicity and physical accuracy in order to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Flexibility of construction permits tailoring of the codes to specific applications and extension of code capabilities to more complex applications through simple update procedures.

Geochemical and numerical modelling of interactions between solid solutions and an aqueous solution. Extension of a reactive transport computer code called Archimede and application to reservoirs diagenesis; Modelisation geochimique et numerique des interactions entre des solutions solides et une solution aqueuse: extension du logiciel de reaction-transport archimede et application a la diagenese des reservoirs

Energy Technology Data Exchange (ETDEWEB)

Nourtier-Mazauric, E.

2003-03-15

This thesis presents a thermodynamic and kinetic model of interactions between a fluid and ideal solid solutions represented by several end-members. The reaction between a solid solution and the aqueous solution results from the competition between the stoichiometric dissolution of the initial solid solution and the co-precipitation of the least soluble solid solution in the fluid at considered time. This model was implemented in ARCHIMEDE, a computer code of reactive transport in porous media, then applied to various examples. In the case of binary solid solutions, a graphical method allowed to determine the compositions of the precipitating solid solutions, with the aid of the end-member chemical potentials. The obtained program could be used to notably model the diagenesis of clayey or carbonated oil reservoirs, or the ground pollutant dispersion. (author)

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

Generation of cross-sections and reference solutions using the code Serpent

International Nuclear Information System (INIS)

Gomez T, A. M.; Delfin L, A.; Del Valle G, E.

2012-10-01

Serpent is a code that solves the neutron transport equations using the Monte Carlo method that besides generating reference solutions in stationary state for complex geometry problems, has been specially designed for physical applications of cells, what includes the generation of homogenized cross-sections for several energy groups. In this work a calculation methodology is described using the code Serpent to generate the necessary cross-sections to carry out calculations with the code TNXY, developed in 1993 in the Nuclear Engineering Department of the Instituto Politecnico Nacional (Mexico) by means of an interface programmed in Octave. The computation program TNXY solves the neutron transport equations for several energy groups in stationary state and geometry X Y using the Discreet Ordinates method (S N ). To verify and to validate the methodology the results of TNXY were compared with those calculated by Serpent giving minor differences to 0.55% in the value of the multiplication factor. (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.

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

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

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.

One-dimensional transport code for one-group problems in plane geometry

International Nuclear Information System (INIS)

Bareiss, E.H.; Chamot, C.

1970-09-01

Equations and results are given for various methods of solution of the one-dimensional transport equation for one energy group in plane geometry with inelastic scattering and an isotropic source. After considerable investigation, a matrix method of solution was found to be faster and more stable than iteration procedures. A description of the code is included which allows for up to 24 regions, 250 points, and 16 angles such that the product of the number of angles and the number of points is less than 600

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

The new deterministic 3-D radiation transport code Multitrans: C5G7 MOX fuel assembly benchmark

International Nuclear Information System (INIS)

Kotiluoto, P.

2003-01-01

The novel deterministic three-dimensional radiation transport code MultiTrans is based on combination of the advanced tree multigrid technique and the simplified P3 (SP3) radiation transport approximation. In the tree multigrid technique, an automatic mesh refinement is performed on material surfaces. The tree multigrid is generated directly from stereo-lithography (STL) files exported by computer-aided design (CAD) systems, thus allowing an easy interface for construction and upgrading of the geometry. The deterministic MultiTrans code allows fast solution of complicated three-dimensional transport problems in detail, offering a new tool for nuclear applications in reactor physics. In order to determine the feasibility of a new code, computational benchmarks need to be carried out. In this work, MultiTrans code is tested for a seven-group three-dimensional MOX fuel assembly transport benchmark without spatial homogenization (NEA C5G7 MOX). (author)

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.)

The simulation of solute transport: An approach free of numerical dispersion

International Nuclear Information System (INIS)

Carrera, J.; Melloni, G.

1987-01-01

The applicability of most algorithms for simulation of solute transport is limited either by instability or by numerical dispersion, as seen by a review of existing methods. A new approach is proposed that is free of these two problems. The method is based on the mixed Eulerian-Lagrangian formulation of the mass-transport problem, thus ensuring stability. Advection is simulated by a variation of reverse-particle tracking that avoids the accumulation of interpolation errors, thus preventing numerical dispersion. The algorithm has been implemented in a one-dimensional code. Excellent results are obtained, in comparison with an analytical solution. 36 refs., 14 figs., 1 tab

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)

The EGS4 Code System: Solution of Gamma-ray and Electron Transport Problems

Science.gov (United States)

Nelson, W. R.; Namito, Yoshihito

1990-03-01

In this paper we present an overview of the EGS4 Code System -- a general purpose package for the Monte Carlo simulation of the transport of electrons and photons. During the last 10-15 years EGS has been widely used to design accelerators and detectors for high-energy physics. More recently the code has been found to be of tremendous use in medical radiation physics and dosimetry. The problem-solving capabilities of EGS4 will be demonstrated by means of a variety of practical examples. To facilitate this review, we will take advantage of a new add-on package, called SHOWGRAF, to display particle trajectories in complicated geometries. These are shown as 2-D laser pictures in the written paper and as photographic slides of a 3-D high-resolution color monitor during the oral presentation. 11 refs., 15 figs.

The EGS4 Code System: Solution of gamma-ray and electron transport problems

International Nuclear Information System (INIS)

Nelson, W.R.; Namito, Yoshihito.

1990-01-01

In this paper we present an overview of the EGS4 Code System -- a general purpose package for the Monte Carlo simulation of the transport of electrons and photons. During the last 10-15 years EGS has been widely used to design accelerators and detectors for high-energy physics. More recently the code has been found to be of tremendous use in medical radiation physics and dosimetry. The problem-solving capabilities of EGS4 will be demonstrated by means of a variety of practical examples. To facilitate this review, we will take advantage of a new add-on package, called SHOWGRAF, to display particle trajectories in complicated geometries. These are shown as 2-D laser pictures in the written paper and as photographic slides of a 3-D high-resolution color monitor during the oral presentation. 11 refs., 15 figs

Solute transport model for radioisotopes in layered soil

International Nuclear Information System (INIS)

Essel, P.

2010-01-01

The study considered the transport of a radioactive solute in solution from the surface of the earth down through the soil to the ground water when there is an accidental or intentional spillage of a radioactive material on the surface. The finite difference method was used to model the spatial and temporal profile of moisture content in a soil column using the θ-based Richard's equation leading to solution of the convective-dispersive equation for non-adsorbing solutes numerically. A matlab code has been generated to predict the transport of the radioactive contaminant, spilled on the surface of a vertically heterogeneous soil made up of two layers to determine the residence time of the solute in the unsaturated zone, the time it takes the contaminant to reach the groundwater and the amount of the solute entering the groundwater in various times and the levels of pollution in those times. The model predicted that, then there is a spillage of 7.2g of tritium, on the surface of the ground at the study area, it will take two years for the radionuclide to enter the groundwater and fifteen years to totally leave the unsaturated zone. There is therefore the need to try as much as possible to avoid intentional or accidental spillage of the radionuclide since it has long term effect. (au)

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Mesh generation and energy group condensation studies for the jaguar deterministic transport code

International Nuclear Information System (INIS)

Kennedy, R. A.; Watson, A. M.; Iwueke, C. I.; Edwards, E. J.

2012-01-01

The deterministic transport code Jaguar is introduced, and the modeling process for Jaguar is demonstrated using a two-dimensional assembly model of the Hoogenboom-Martin Performance Benchmark Problem. This single assembly model is being used to test and analyze optimal modeling methodologies and techniques for Jaguar. This paper focuses on spatial mesh generation and energy condensation techniques. In this summary, the models and processes are defined as well as thermal flux solution comparisons with the Monte Carlo code MC21. (authors)

Mesh generation and energy group condensation studies for the jaguar deterministic transport code

Energy Technology Data Exchange (ETDEWEB)

Kennedy, R. A.; Watson, A. M.; Iwueke, C. I.; Edwards, E. J. [Knolls Atomic Power Laboratory, Bechtel Marine Propulsion Corporation, P.O. Box 1072, Schenectady, NY 12301-1072 (United States)

2012-07-01

The deterministic transport code Jaguar is introduced, and the modeling process for Jaguar is demonstrated using a two-dimensional assembly model of the Hoogenboom-Martin Performance Benchmark Problem. This single assembly model is being used to test and analyze optimal modeling methodologies and techniques for Jaguar. This paper focuses on spatial mesh generation and energy condensation techniques. In this summary, the models and processes are defined as well as thermal flux solution comparisons with the Monte Carlo code MC21. (authors)

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.

Chlorovirus-mediated membrane depolarization of Chlorella alters secondary active transport of solutes.

Science.gov (United States)

Agarkova, Irina; Dunigan, David; Gurnon, James; Greiner, Timo; Barres, Julia; Thiel, Gerhard; Van Etten, James L

2008-12-01

Paramecium bursaria chlorella virus 1 (PBCV-1) is the prototype of a family of large, double-stranded DNA, plaque-forming viruses that infect certain eukaryotic chlorella-like green algae from the genus Chlorovirus. PBCV-1 infection results in rapid host membrane depolarization and potassium ion release. One interesting feature of certain chloroviruses is that they code for functional potassium ion-selective channel proteins (Kcv) that are considered responsible for the host membrane depolarization and, as a consequence, the efflux of potassium ions. This report examines the relationship between cellular depolarization and solute uptake. Annotation of the virus host Chlorella strain NC64A genome revealed 482 putative transporter-encoding genes; 224 are secondary active transporters. Solute uptake experiments using seven radioactive compounds revealed that virus infection alters the transport of all the solutes. However, the degree of inhibition varied depending on the solute. Experiments with nystatin, a drug known to depolarize cell membranes, produced changes in solute uptake that are similar but not identical to those that occurred during virus infection. Therefore, these studies indicate that chlorovirus infection causes a rapid and sustained depolarization of the host plasma membrane and that this depolarization leads to the inhibition of secondary active transporters that changes solute uptake.

A numerical solution of the coupled proton-H atom transport equations for the proton aurora

International Nuclear Information System (INIS)

Basu, B.; Jasperse, J.R.; Grossbard, N.J.

1990-01-01

A numerical code has been developed to solve the coupled proton-H atom linear transport equations for the proton aurora. The transport equations have been simplified by using plane-parallel geometry and the forward-scattering approximations only. Otherwise, the equations and their numerical solutions are exact. Results are presented for the particle fluxes and the energy deposition rates, and they are compared with the previous analytical results that were obtained by using additional simplifying approximations. It is found that although the analytical solutions for the particle fluxes differ somewhat from the numerical solutions, the energy deposition rates calculated by the two methods agree to within a few percent. The accurate particle fluxes given by the numerical code are useful for accurate calculation of the characteristic quantities of the proton aurora, such as the ionization rates and the emission rates

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)

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)

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

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).

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

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.

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

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

CFEST Coupled Flow, Energy & Solute Transport Version CFEST005 User’s Guide

Energy Technology Data Exchange (ETDEWEB)

Freedman, Vicky L.; Chen, Yousu; Gilca, Alex; Cole, Charles R.; Gupta, Sumant K.

2006-07-20

The CFEST (Coupled Flow, Energy, and Solute Transport) simulator described in this User’s Guide is a three-dimensional finite-element model used to evaluate groundwater flow and solute mass transport. Confined and unconfined aquifer systems, as well as constant and variable density fluid flows can be represented with CFEST. For unconfined aquifers, the model uses a moving boundary for the water table, deforming the numerical mesh so that the uppermost nodes are always at the water table. For solute transport, changes in concentra¬tion of a single dissolved chemical constituent are computed for advective and hydrodynamic transport, linear sorption represented by a retardation factor, and radioactive decay. Although several thermal parameters described in this User’s Guide are required inputs, thermal transport has not yet been fully implemented in the simulator. Once fully implemented, transport of thermal energy in the groundwater and solid matrix of the aquifer can also be used to model aquifer thermal regimes. The CFEST simulator is written in the FORTRAN 77 language, following American National Standards Institute (ANSI) standards. Execution of the CFEST simulator is controlled through three required text input files. These input file use a structured format of associated groups of input data. Example input data lines are presented for each file type, as well as a description of the structured FORTRAN data format. Detailed descriptions of all input requirements, output options, and program structure and execution are provided in this User’s Guide. Required inputs for auxillary CFEST utilities that aide in post-processing data are also described. Global variables are defined for those with access to the source code. Although CFEST is a proprietary code (CFEST, Inc., Irvine, CA), the Pacific Northwest National Laboratory retains permission to maintain its own source, and to distribute executables to Hanford subcontractors.

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

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

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

Method of solution of the neutron transport equation in multidimensional cartesian geometries using spherical harmonics and spatially orthogonal polynomials

International Nuclear Information System (INIS)

Fenstermacher, T.E.

1981-01-01

The solution of the neutron transport equation has long been a subject of intense interest to nuclear engineers. Present computer codes for the solution of this equation, however, are expensive to run for large, multidimensional problems, and also suffer from computational problems such as the ray effect. A method has been developed which eliminates many of these problems. It consists of transforming the transport equation into a set of linear partial differential equations by the use of spherical harmonics. The problem volume is divided into mesh boxes, and the flux components are approximated within each mesh box by spatially orthogonal quadratic polynomials, which need not be continuous at mesh box interfaces. A variational principle is developed, and used to solve for the unknown coefficients of these polynomials. Both one dimensional and two dimensional computer codes using this method have been written. The codes have each been tested on several test cases, and the solutions checked against solutions obtained by other methods. While the codes have some difficulty in modeling sharp transients, they produce excellent results on problems where the characteristic lengths are many mean free paths. On one test case, the two dimensional code, SHOP/2D, required only one-fourth the computer time required by the finite difference, discrete ordinates code TWOTRAN to produce a solution. In addition, SHOP/2D converged much better than TWOTRAN and produced more physical-appearing results

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.

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)

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

Study of reactive solutes transport and PAH migration in unsaturated soils

International Nuclear Information System (INIS)

Gujisaite, V.; Simonnot, M.O.; Gujisaite, V.; Morel, J.L.; Ouvrard, S.; Simonnot, M.O.; Gaudet, J.P.

2005-01-01

-silty uncontaminated soil or a sand and a model porous medium constituted of sand in which coal tar particles are dispersed. In a second time, column experiments will be carried out with a PAH contaminated soil from a former coking plant and with a multi-polluted industrial soil (PAH, heavy metals) to study PAH migration. For each studied soil, we will also determine the water retention curve in order to find the best operating conditions for our experiments with unsaturated flow. Modelling of solutes transfer in soils is also needed to improve understanding of the fate of contaminants and for risk assessment. However, it is difficult to take into account at the same time flow and interactions in models. Different models and numerical codes have been developed for solute transport. We have chosen to use the CXTFIT code, in order to model our results. This code allows indeed modelling of reactive solute transport in unsaturated porous media as well as under saturated conditions. It is usually used to estimate solute transport parameters using a nonlinear least-squares parameter optimization method. It may be used to solve the inverse problem by fitting a variety of mathematical solutions of theoretical transport models, based upon the one-dimensional convection-dispersion equation (CDE), to experimental results. The program may also be used to solve the direct or forward problem to determine concentrations as a function of time and/or position. This study at a bench scale will enable us at first to develop a methodology under unsaturated conditions and also to better understand the dominating mechanisms which control PAH transfer and availability in natural soils, especially by quantifying the impact of parameters like soil water content, water flow or the presence of plants. This is a first step before the change of scale (lysimeter). Modelling of the observed processes will also enable us to predict long term fate of PAH in soils

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)

CFD code verification and the method of manufactured solutions

International Nuclear Information System (INIS)

Pelletier, D.; Roache, P.J.

2002-01-01

This paper presents the Method of Manufactured Solutions (MMS) for CFD code verification. The MMS provides benchmark solutions for direct evaluation of the solution error. The best benchmarks are exact analytical solutions with sufficiently complex solution structure to ensure that all terms of the differential equations are exercised in the simulation. The MMS provides a straight forward and general procedure for generating such solutions. When used with systematic grid refinement studies, which are remarkably sensitive, the MMS provides strong code verification with a theorem-like quality. The MMS is first presented on simple 1-D examples. Manufactured solutions for more complex problems are then presented with sample results from grid convergence studies. (author)

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)

Verification of the MOTIF code version 3.0

International Nuclear Information System (INIS)

Chan, T.; Guvanasen, V.; Nakka, B.W.; Reid, J.A.K.; Scheier, N.W.; Stanchell, F.W.

1996-12-01

As part of the Canadian Nuclear Fuel Waste Management Program (CNFWMP), AECL has developed a three-dimensional finite-element code, MOTIF (Model Of Transport In Fractured/ porous media), for detailed modelling of groundwater flow, heat transport and solute transport in a fractured rock mass. The code solves the transient and steady-state equations of groundwater flow, solute (including one-species radionuclide) transport, and heat transport in variably saturated fractured/porous media. The initial development was completed in 1985 (Guvanasen 1985) and version 3.0 was completed in 1986. This version is documented in detail in Guvanasen and Chan (in preparation). This report describes a series of fourteen verification cases which has been used to test the numerical solution techniques and coding of MOTIF, as well as demonstrate some of the MOTIF analysis capabilities. For each case the MOTIF solution has been compared with a corresponding analytical or independently developed alternate numerical solution. Several of the verification cases were included in Level 1 of the International Hydrologic Code Intercomparison Project (HYDROCOIN). The MOTIF results for these cases were also described in the HYDROCOIN Secretariat's compilation and comparison of results submitted by the various project teams (Swedish Nuclear Power Inspectorate 1988). It is evident from the graphical comparisons presented that the MOTIF solutions for the fourteen verification cases are generally in excellent agreement with known analytical or numerical solutions obtained from independent sources. This series of verification studies has established the ability of the MOTIF finite-element code to accurately model the groundwater flow and solute and heat transport phenomena for which it is intended. (author). 20 refs., 14 tabs., 32 figs

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

Generic programming for deterministic neutron transport codes

International Nuclear Information System (INIS)

Plagne, L.; Poncot, A.

2005-01-01

This paper discusses the implementation of neutron transport codes via generic programming techniques. Two different Boltzmann equation approximations have been implemented, namely the Sn and SPn methods. This implementation experiment shows that generic programming allows us to improve maintainability and readability of source codes with no performance penalties compared to classical approaches. In the present implementation, matrices and vectors as well as linear algebra algorithms are treated separately from the rest of source code and gathered in a tool library called 'Generic Linear Algebra Solver System' (GLASS). Such a code architecture, based on a linear algebra library, allows us to separate the three different scientific fields involved in transport codes design: numerical analysis, reactor physics and computer science. Our library handles matrices with optional storage policies and thus applies both to Sn code, where the matrix elements are computed on the fly, and to SPn code where stored matrices are used. Thus, using GLASS allows us to share a large fraction of source code between Sn and SPn implementations. Moreover, the GLASS high level of abstraction allows the writing of numerical algorithms in a form which is very close to their textbook descriptions. Hence the GLASS algorithms collection, disconnected from computer science considerations (e.g. storage policy), is very easy to read, to maintain and to extend. (authors)

BREIT code: Analytical solution of the balance rate equations for charge-state evolutions of heavy-ion beams in matter

Energy Technology Data Exchange (ETDEWEB)

Winckler, N., E-mail: n.winckler@gsi.de [GSI Helmholtzzentrum für Schwerionenforschung GmbH, 64291 Darmstadt (Germany); Rybalchenko, A. [GSI Helmholtzzentrum für Schwerionenforschung GmbH, 64291 Darmstadt (Germany); Shevelko, V.P. [P.N. Lebedev Physical Institute, 119991 Moscow (Russian Federation); Al-Turany, M. [GSI Helmholtzzentrum für Schwerionenforschung GmbH, 64291 Darmstadt (Germany); CERN, European Organization for Nuclear Research, 1211 Geneve 23 (Switzerland); Kollegger, T. [GSI Helmholtzzentrum für Schwerionenforschung GmbH, 64291 Darmstadt (Germany); Stöhlker, Th. [GSI Helmholtzzentrum für Schwerionenforschung GmbH, 64291 Darmstadt (Germany); Helmholtz-Institute Jena, D-07743 Jena (Germany); Institut für Optik und Quantenelektronik, Friedrich-Schiller-Universität, D-07743 Jena (Germany)

2017-02-01

A detailed description of a recently developed BREIT computer code (Balance Rate Equations of Ion Transportation) for calculating charge-state fractions of ion beams passing through matter is presented. The code is based on the analytical solutions of the differential balance equations for the charge-state fractions as a function of the target thickness and can be used for calculating the ion evolutions in gaseous, solid and plasma targets. The BREIT code is available on-line and requires the charge-changing cross sections and initial conditions in the input file. The eigenvalue decomposition method, applied to obtain the analytical solutions of the rate equations, is described in the paper. Calculations of non-equilibrium and equilibrium charge-state fractions, performed by the BREIT code, are compared with experimental data and results of other codes for ion beams in gaseous and solid targets. Ability and limitations of the BREIT code are discussed in detail.

Coupled Fluid, Energy, and Solute Transport (CFEST) model: Formulation and user's manual

International Nuclear Information System (INIS)

Gupta, S.K.; Cole, C.R.; Kincaid, C.T.; Monti, A.M.

1987-10-01

The CFEST (Coupled Fluid, Energy, and Solute Transport) code has been developed to analyze coupled hydrologic, thermal, and solute transport processes. It treats single-pahse Darcy ground-water flow in a horizontal or vertical plane, or in fully three-dimensional space under nonisothermal conditions. The code has the capability to model discontinuous and continuous layering, time-dependent and constant sources/sinks, and transient as well as steady-stae ground-water flow. The code offers a wide choice of boundary conditions such as precsribed heads, nodal injection or withdrawal, constant or spatially varying infiltration rates, and welemental source/sink. Initial conditions for the flow analysis can be prescribed pressure or hydraulic head. The heterogeneity in aquifer permeability and porosity can be described by geologic unit or explicity for given elements. Three-dimensional elelments are generated from user-defined well logs at each surface node. To facilitate interaction between disciplines, support programs are provided to plot the finite element grid, well logs, contour maps of input and output parameters, and vertical cross sections. Ground-water travel paths and times and volumetric rates from a specified point can be determined from support programs. This report includes governing partial differential equations, finite element formulation, a use's manual, verification test examples, sample problems, and source listings. 36 refs., 121 figs., 36 tabs

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

Potential of the MCNP computer code

International Nuclear Information System (INIS)

Kyncl, J.

1995-01-01

The MCNP code is designed for numerical solution of neutron, photon, and electron transport problems by the Monte Carlo method. The code is based on the linear transport theory of behavior of the differential flux of the particles. The code directly uses data from the cross section point data library for input. Experience is outlined, gained in the application of the code to the calculation of the effective parameters of fuel assemblies and of the entire reactor core, to the determination of the effective parameters of the elementary fuel cell, and to the numerical solution of neutron diffusion and/or transport problems of the fuel assembly. The agreement between the calculated and observed data gives evidence that the MCNP code can be used with advantage for calculations involving WWER type fuel assemblies. (J.B.). 4 figs., 6 refs

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

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

CHMTRNS, Non-Equilibrium Chemical Transport Code

International Nuclear Information System (INIS)

Noorishad, J.; Carnahan, C.L.; Benson, L.V.

1998-01-01

1 - Description of program or function: CHMTRNS simulates solute transport for steady one-dimensional fluid flow by convection and diffusion or dispersion in a saturated porous medium based on the assumption of local chemical equilibrium. The chemical interactions included in the model are aqueous-phase complexation, solid-phase ion exchange of bare ions and complexes using the surface complexation model, and precipitation or dissolution of solids. The program can simulate the kinetic dissolution or precipitation for calcite and silica as well as irreversible dissolution of glass. Thermodynamic parameters are temperature dependent and are coupled to a companion heat transport simulator; thus, the effects of transient temperature conditions can be considered. Options for oxidation-reduction (redox) and C-13 fractionation as well as non-isothermal conditions are included. 2 - Method of solution: The governing equations for both reactive chemical and heat transport are discretized in time and space. For heat transport, the Crank-Nicolson approximation is used in conjunction with a LU decomposition and backward substitution solution procedure. To deal with the strong nonlinearity of the chemical transport equations, a generalized Newton-Raphson method is used

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

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

The use of non-dimensional representation of the solute transport equations

International Nuclear Information System (INIS)

Laurens, J.-M.

1988-07-01

This report presents the results obtained in a pilot investigation into the use of non-dimensional representations of the solute transport equations, so as to improve the efficiency of the PRA codes used by the DoE and its contractors. A reduced set of parameters was obtained for a single layer transport model. As expected, the response was shown to be highly sensitive on the new parameters. A faster convergence of the system was observed when the sampling technique used was changed to take into account the properties of the new parameters, such that uniform coverage of the reduced parameter hyperspace was achieved. (author)

A compartmentalized solute transport model for redox zones in contaminated aquifers: 1. Theory and development

Science.gov (United States)

Abrams , Robert H.; Loague, Keith

2000-01-01

This paper, the first of two parts [see Abrams and Loague, this issue], takes the compartmentalized approach for the geochemical evolution of redox zones presented by Abrams et al. [1998] and embeds it within a solute transport framework. In this paper the compartmentalized approach is generalized to facilitate the description of its incorporation into a solute transport simulator. An equivalent formulation is developed which removes any discontinuities that may occur when switching compartments. Rate‐limited redox reactions are modeled with a modified Monod relationship that allows either the organic substrate or the electron acceptor to be the rate‐limiting reactant. Thermodynamic constraints are used to inhibit lower‐energy redox reactions from occurring under infeasible geochemical conditions without imposing equilibrium on the lower‐energy reactions. The procedure used allows any redox reaction to be simulated as being kinetically limited or thermodynamically limited, depending on local geochemical conditions. Empirical reaction inhibition methods are not needed. The sequential iteration approach (SIA), a technique which allows the number of solute transport equations to be reduced, is adopted to solve the coupled geochemical/solute transport problem. When the compartmentalized approach is embedded within the SIA, with the total analytical concentration of each component as the dependent variable in the transport equation, it is possible to reduce the number of transport equations even further than with the unmodified SIA. A one‐dimensional, coupled geochemical/solute transport simulation is presented in which redox zones evolve dynamically in time and space. The compartmentalized solute transport (COMPTRAN) model described in this paper enables the development of redox zones to be simulated under both kinetic and thermodynamic constraints. The modular design of COMPTRAN facilitates the use of many different, preexisting solute transport and

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

submitter BREIT code: Analytical solution of the balance rate equations for charge-state evolutions of heavy-ion beams in matter

CERN Document Server

Winckler, N; Shevelko, V P; Al-Turany, M; Kollegger, T; Stöhlker, Th

2017-01-01

A detailed description of a recently developed BREIT computer code (Balance Rate Equations of Ion Transportation) for calculating charge-state fractions of ion beams passing through matter is presented. The code is based on the analytical solutions of the differential balance equations for the charge-state fractions as a function of the target thickness and can be used for calculating the ion evolutions in gaseous, solid and plasma targets. The BREIT code is available on-line and requires the charge-changing cross sections and initial conditions in the input file. The eigenvalue decomposition method, applied to obtain the analytical solutions of the rate equations, is described in the paper. Calculations of non-equilibrium and equilibrium charge-state fractions, performed by the BREIT code, are compared with experimental data and results of other codes for ion beams in gaseous and solid targets. Ability and limitations of the BREIT code are discussed in detail.

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

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.

PBMC: Pre-conditioned Backward Monte Carlo code for radiative transport in planetary atmospheres

Science.gov (United States)

García Muñoz, A.; Mills, F. P.

2017-08-01

PBMC (Pre-Conditioned Backward Monte Carlo) solves the vector Radiative Transport Equation (vRTE) and can be applied to planetary atmospheres irradiated from above. The code builds the solution by simulating the photon trajectories from the detector towards the radiation source, i.e. in the reverse order of the actual photon displacements. In accounting for the polarization in the sampling of photon propagation directions and pre-conditioning the scattering matrix with information from the scattering matrices of prior (in the BMC integration order) photon collisions, PBMC avoids the unstable and biased solutions of classical BMC algorithms for conservative, optically-thick, strongly-polarizing media such as Rayleigh atmospheres.

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.

NEACRP comparison of codes for the radiation protection assessment of transportation packages. Solutions to problems 1 - 4

International Nuclear Information System (INIS)

Avery, A.F.; Locke, H.F.

1992-03-01

In 1985 the Reactor Physics Committee of the Nuclear Energy Agency initiated an intercomparison of codes for the calculation of the performance of shielding for the transportation of spent reactor fuel. The results of the application of a range of codes to the prediction of the dose-rates in the four theoretical benchmarks set to examine the attenuation of radiation through a variety of cask geometries are presented in this report. The contributions from neutrons, fission product gamma-rays and secondary gamma-rays are tabulated separately, and grouped according to the type of method of calculation employed. A brief discussion is included for each set of results, and overall comparisons of the methods, codes, and nuclear data are made. A number of conclusions are drawn on the advantages and disadvantages of the various methods of calculation, based upon the results of their application to these four benchmark problems

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

timescales. To account for one-dimensional matrix diffusion into homogeneous planar or cylindrical rock layers, analytical relations in the Laplace domain are used. To deal with one-dimensional or two-dimensional matrix diffusion into heterogeneous rock matrices, a finite-element method is embedded. The capability of the code for handling two-dimensional matrix diffusion is - to our knowledge - unique in fracture network modelling. To ensure the reliability of the code, which merges methods from graph theory, Laplace transformation, finite-element methods, analytical and algebraic transformations and a convolution to calculate complex radionuclide transport processes over a large and diverse application range, implementation of the code and careful verification have been alternated for iterative improvement and especially the elimination of bugs. The internal mathematical structure of PICNIC forms the basis of the verification strategy. The code is verified in a series of seven steps with increasing complexity of the rock matrix. Calculations for single nuclides and nuclide decay chains are carefully tested and analysed for radionuclide transport in single legs, in pathways and in networks. Different sources and boundary conditions are considered. Quantitative estimates of the accuracy of the code are derived from comparisons with analytical solutions, cross-comparisons with other codes and different types of self -consistency tests, including extended testing of different refinements of the embedded finite- element method for different rock matrix geometries. The geosphere barrier efficiency is a good single indicator of the code accuracy. Application ranges with reduced accuracy of the code are also considered. For one-dimensional matrix diffusion into homogeneous and heterogeneous rock matrices, cross-comparisons with other codes are performed. For two-dimensional matrix diffusion, however, no code for cross-comparison is available. Consequently, the verification for

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

different timescales. To account for one-dimensional matrix diffusion into homogeneous planar or cylindrical rock layers, analytical relations in the Laplace domain are used. To deal with one-dimensional or two-dimensional matrix diffusion into heterogeneous rock matrices, a finite-element method is embedded. The capability of the code for handling two-dimensional matrix diffusion is - to our knowledge - unique in fracture network modelling. To ensure the reliability of the code, which merges methods from graph theory, Laplace transformation, finite-element methods, analytical and algebraic transformations and a convolution to calculate complex radionuclide transport processes over a large and diverse application range, implementation of the code and careful verification have been alternated for iterative improvement and especially the elimination of bugs. The internal mathematical structure of PICNIC forms the basis of the verification strategy. The code is verified in a series of seven steps with increasing complexity of the rock matrix. Calculations for single nuclides and nuclide decay chains are carefully tested and analysed for radionuclide transport in single legs, in pathways and in networks. Different sources and boundary conditions are considered. Quantitative estimates of the accuracy of the code are derived from comparisons with analytical solutions, cross-comparisons with other codes and different types of self -consistency tests, including extended testing of different refinements of the embedded finite- element method for different rock matrix geometries. The geosphere barrier efficiency is a good single indicator of the code accuracy. Application ranges with reduced accuracy of the code are also considered. For one-dimensional matrix diffusion into homogeneous and heterogeneous rock matrices, cross-comparisons with other codes are performed. For two-dimensional matrix diffusion, however, no code for cross-comparison is available. Consequently, the

A Mathematical Model of Solute Coupled Water Transport in Toad Intestine Incorporating Recirculation of the Actively Transported Solute

DEFF Research Database (Denmark)

Larsen, Erik Hviid; Sørensen, Jakob Balslev; Sørensen, Jens Nørkær

2000-01-01

those of tight junction and interspace basement membrane by convection-diffusion. With solute permeability of paracellular pathway large relative to paracellular water flow, the paracellular flux ratio of the solute (influx/outflux) is small (2-4) in agreement with experiments. The virtual solute......A mathematical model of an absorbing leaky epithelium is developed for analysis of solute coupled water transport. The non-charged driving solute diffuses into cells and is pumped from cells into the lateral intercellular space (lis). All membranes contain water channels with the solute passing...... increases with hydraulic conductance of the pathway carrying water from mucosal solution into lis. Uphill water transport is accomplished, but with high hydraulic conductance of cell membranes strength of transport is obscured by water flow through cells. Anomalous solvent drag occurs when back flux...

Verification of T2VOC using an analytical solution for VOC transport in vadose zone

Energy Technology Data Exchange (ETDEWEB)

Shan, C. [Lawrence Berkeley Laboratory, Berkeley, CA (United States)

1995-03-01

T2VOC represents an adaption of the STMVOC to the TOUGH2 environment. In may contaminated sites, transport of volatile organic chemicals (VOC) is a serious problem which can be simulated by T2VOC. To demonstrate the accuracy and robustness of the code, we chose a practical problem of VOC transport as the test case, conducted T2VOC simulations, and compared the results of T2VOC with those of an analytical solution. The agreements between T2VOC and the analytical solutions are excellent. In addition, the numerical results of T2VOC are less sensitive to grid size and time step to a certain extent.

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)

Reactor Dosimetry Applications Using RAPTOR-M3G:. a New Parallel 3-D Radiation Transport Code

Science.gov (United States)

Longoni, Gianluca; Anderson, Stanwood L.

2009-08-01

The numerical solution of the Linearized Boltzmann Equation (LBE) via the Discrete Ordinates method (SN) requires extensive computational resources for large 3-D neutron and gamma transport applications due to the concurrent discretization of the angular, spatial, and energy domains. This paper will discuss the development RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries), a new 3-D parallel radiation transport code, and its application to the calculation of ex-vessel neutron dosimetry responses in the cavity of a commercial 2-loop Pressurized Water Reactor (PWR). RAPTOR-M3G is based domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architectures. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor, yielding an efficient solution methodology for large 3-D problems. Measured neutron dosimetry responses in the reactor cavity air gap will be compared to the RAPTOR-M3G predictions. This paper is organized as follows: Section 1 discusses the RAPTOR-M3G methodology; Section 2 describes the 2-loop PWR model and the numerical results obtained. Section 3 addresses the parallel performance of the code, and Section 4 concludes this paper with final remarks and future work.

Algebraic solution of the synthesis problem for coded sequences

International Nuclear Information System (INIS)

Leukhin, Anatolii N

2005-01-01

The algebraic solution of a 'complex' problem of synthesis of phase-coded (PC) sequences with the zero level of side lobes of the cyclic autocorrelation function (ACF) is proposed. It is shown that the solution of the synthesis problem is connected with the existence of difference sets for a given code dimension. The problem of estimating the number of possible code combinations for a given code dimension is solved. It is pointed out that the problem of synthesis of PC sequences is related to the fundamental problems of discrete mathematics and, first of all, to a number of combinatorial problems, which can be solved, as the number factorisation problem, by algebraic methods by using the theory of Galois fields and groups. (fourth seminar to the memory of d.n. klyshko)

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

Finite element simulation of moisture movement and solute transport in a large caisson

International Nuclear Information System (INIS)

Huyakorn, P.S.; Jones, B.G.; Parker, J.C.; Wadsworth, T.D.; White, H.O. Jr.

1987-01-01

The results of the solute transport experiments performed on compacted, crushed Bandelier Tuff in caisson B of the experimental cluster described by DePoorter (1981) are simulated. Both one- and three-dimensional simulations of solute transport have been performed using two selected finite element codes. Results of bromide and iodide tracer experiments conducted during near-steady flow conditions have been analyzed for pulse additions made on December 6, 1984, and followed over a period of up to 60 days. In addition, a pulse addition of nonconservative strontium tracer on September 28, 1984, during questionably steady flow conditions has been analyzed over a period of 240 days. One-dimensional finite element flow and transport simulations were carried out assuming the porous medium to be homogeneous and the injection source uniformly distributed. To evaluate effects of the nonuniform source distribution and also to investigate effects of inhomogeneous porous medium properties, three dimensional finite element analyses of transport were carried out. Implications of the three-dimensional effects for the design and analysis of future tracer studies are discussed

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)

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.)

C5 Benchmark Problem with Discrete Ordinate Radiation Transport Code DENOVO

Energy Technology Data Exchange (ETDEWEB)

Yesilyurt, Gokhan [ORNL; Clarno, Kevin T [ORNL; Evans, Thomas M [ORNL; Davidson, Gregory G [ORNL; Fox, Patricia B [ORNL

2011-01-01

The C5 benchmark problem proposed by the Organisation for Economic Co-operation and Development/Nuclear Energy Agency was modeled to examine the capabilities of Denovo, a three-dimensional (3-D) parallel discrete ordinates (S{sub N}) radiation transport code, for problems with no spatial homogenization. Denovo uses state-of-the-art numerical methods to obtain accurate solutions to the Boltzmann transport equation. Problems were run in parallel on Jaguar, a high-performance supercomputer located at Oak Ridge National Laboratory. Both the two-dimensional (2-D) and 3-D configurations were analyzed, and the results were compared with the reference MCNP Monte Carlo calculations. For an additional comparison, SCALE/KENO-V.a Monte Carlo solutions were also included. In addition, a sensitivity analysis was performed for the optimal angular quadrature and mesh resolution for both the 2-D and 3-D infinite lattices of UO{sub 2} fuel pin cells. Denovo was verified with the C5 problem. The effective multiplication factors, pin powers, and assembly powers were found to be in good agreement with the reference MCNP and SCALE/KENO-V.a Monte Carlo calculations.

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)

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

Diffusion Dominant Solute Transport Modelling in Fractured Media Under Deep Geological Environment - 12211

Energy Technology Data Exchange (ETDEWEB)

Kwong, S. [National Nuclear Laboratory (United Kingdom); Jivkov, A.P. [Research Centre for Radwaste and Decommissioning and Modelling and Simulation Centre, University of Manchester (United Kingdom)

2012-07-01

Deep geologic disposal of high activity and long-lived radioactive waste is gaining increasing support in many countries, where suitable low permeability geological formation in combination with engineered barriers are used to provide long term waste contaminant and minimise the impacts to the environment and risk to the biosphere. This modelling study examines the solute transport in fractured media under low flow velocities that are relevant to a deep geological environment. In particular, reactive solute transport through fractured media is studied using a 2-D model, that considers advection and diffusion, to explore the coupled effects of kinetic and equilibrium chemical processes. The effects of water velocity in the fracture, matrix porosity and diffusion on solute transport are investigated and discussed. Some illustrative modelled results are presented to demonstrate the use of the model to examine the effects of media degradation on solute transport, under the influences of hydrogeological (diffusion dominant) and microbially mediated chemical processes. The challenges facing the prediction of long term degradation such as cracks evolution, interaction and coalescence are highlighted. The potential of a novel microstructure informed modelling approach to account for these effects is discussed, particularly with respect to investigating multiple phenomena impact on material performance. The GRM code is used to examine the effects of media degradation for a geological waste disposal package, under the combined hydrogeological (diffusion dominant) and chemical effects in low groundwater flow conditions that are typical of deep geological disposal systems. An illustrative reactive transport modelling application demonstrates the use of the code to examine the interplay of kinetic controlled biogeochemical reactive processes with advective and diffusive transport, under the influence of media degradation. The initial model results are encouraging which show the

Evaluation of Monte Carlo electron-Transport algorithms in the integrated Tiger series codes for stochastic-media simulations

International Nuclear Information System (INIS)

Franke, B.C.; Kensek, R.P.; Prinja, A.K.

2013-01-01

Stochastic-media simulations require numerous boundary crossings. We consider two Monte Carlo electron transport approaches and evaluate accuracy with numerous material boundaries. In the condensed-history method, approximations are made based on infinite-medium solutions for multiple scattering over some track length. Typically, further approximations are employed for material-boundary crossings where infinite-medium solutions become invalid. We have previously explored an alternative 'condensed transport' formulation, a Generalized Boltzmann-Fokker-Planck (GBFP) method, which requires no special boundary treatment but instead uses approximations to the electron-scattering cross sections. Some limited capabilities for analog transport and a GBFP method have been implemented in the Integrated Tiger Series (ITS) codes. Improvements have been made to the condensed history algorithm. The performance of the ITS condensed-history and condensed-transport algorithms are assessed for material-boundary crossings. These assessments are made both by introducing artificial material boundaries and by comparison to analog Monte Carlo simulations. (authors)

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

Solute carrier transporters: Pharmacogenomics research ...

African Journals Online (AJOL)

Aghogho

2010-12-27

Dec 27, 2010 ... This paper reviews the solute carrier transporters and highlights the fact that there is much to be learnt from .... transporters, drug targets, effect or proteins and meta- ... basolateral or apical plasma membrane of polarized cells,.

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.)

Peritoneal fluid transport in CAPD patients with different transport rates of small solutes.

Science.gov (United States)

Sobiecka, Danuta; Waniewski, Jacek; Weryński, Andrzej; Lindholm, Bengt

2004-01-01

Continuous ambulatory peritoneal dialysis (CAPD) patients with high peritoneal solute transport rate often have inadequate peritoneal fluid transport. It is not known whether this inadequate fluid transport is due solely to a too rapid fall of osmotic pressure, or if the decreased effectiveness of fluid transport is also a contributing factor. To analyze fluid transport parameters and the effectiveness of dialysis fluid osmotic pressure in the induction of fluid flow in CAPD patients with different small solute transport rates. 44 CAPD patients were placed in low (n = 6), low-average (n = 13), high-average (n = 19), and high (n = 6) transport groups according to a modified peritoneal equilibration test (PET). The study involved a 6-hour peritoneal dialysis dwell with 2 L 3.86% glucose dialysis fluid for each patient. Radioisotopically labeled serum albumin was added as a volume marker.The fluid transport parameters (osmotic conductance and fluid absorption rate) were estimated using three mathematical models of fluid transport: (1) Pyle model (model P), which describes ultrafiltration rate as an exponential function of time; (2) model OS, which is based on the linear relationship of ultrafiltration rate and overall osmolality gradient between dialysis fluid and blood; and (3) model G, which is based on the linear relationship between ultrafiltration rate and glucose concentration gradient between dialysis fluid and blood. Diffusive mass transport coefficients (K(BD)) for glucose, urea, creatinine, potassium, and sodium were estimated using the modified Babb-Randerson-Farrell model. The high transport group had significantly lower dialysate volume and glucose and osmolality gradients between dialysate and blood, but significantly higher K(BD) for small solutes compared with the other transport groups. Osmotic conductance, fluid absorption rate, and initial ultrafiltration rate did not differ among the transport groups for model OS and model P. Model G yielded

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

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

A quasilinear model for solute transport under unsaturated flow

International Nuclear Information System (INIS)

Houseworth, J.E.; Leem, J.

2009-01-01

We developed an analytical solution for solute transport under steady-state, two-dimensional, unsaturated flow and transport conditions for the investigation of high-level radioactive waste disposal. The two-dimensional, unsaturated flow problem is treated using the quasilinear flow method for a system with homogeneous material properties. Dispersion is modeled as isotropic and is proportional to the effective hydraulic conductivity. This leads to a quasilinear form for the transport problem in terms of a scalar potential that is analogous to the Kirchhoff potential for quasilinear flow. The solutions for both flow and transport scalar potentials take the form of Fourier series. The particular solution given here is for two sources of flow, with one source containing a dissolved solute. The solution method may easily be extended, however, for any combination of flow and solute sources under steady-state conditions. The analytical results for multidimensional solute transport problems, which previously could only be solved numerically, also offer an additional way to benchmark numerical solutions. An analytical solution for two-dimensional, steady-state solute transport under unsaturated flow conditions is presented. A specific case with two sources is solved but may be generalized to any combination of sources. The analytical results complement numerical solutions, which were previously required to solve this class of problems.

Mathematical modeling of solute transport in the subsurface

International Nuclear Information System (INIS)

Naymik, T.G.

1987-01-01

A review of key works on solute transport models indicates that solute transport processes with the exception of advection are still poorly understood. Solute transport models generally do a good job when they are used to test scientific concepts and hypotheses, investigate natural processes, systematically store and manage data, and simulate mass balance of solutes under certain natural conditions. Solute transport models generally are not good for predicting future conditions with a high degree of certainty, or for determining concentrations precisely. The mathematical treatment of solute transport far surpasses their understanding of the process. Investigations of the extent of groundwater contamination and methods to remedy existing problems show the along-term nature of the hazard. Industrial organic compounds may be immiscible in water, highly volatile, or complexed with inorganic as well as other organic compounds; many remain stable in nature almost indefinitely. In the worst case, future disposal of hazardous waste may be restricted to deep burial, as is proposed for radioactive wastes. For investigations pertinent to transport of radionuclides from a geologic repository, the process cannot be fully understood without adequate thermodynamic and kinetic data bases

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

Modification of the finite element heat and mass transfer code (FEHMN) to model multicomponent reactive transport

International Nuclear Information System (INIS)

Viswanathan, H.S.

1995-01-01

The finite element code FEHMN is a three-dimensional finite element heat and mass transport simulator that can handle complex stratigraphy and nonlinear processes such as vadose zone flow, heat flow and solute transport. Scientists at LANL have been developed hydrologic flow and transport models of the Yucca Mountain site using FEHMN. Previous FEHMN simulations have used an equivalent K d model to model solute transport. In this thesis, FEHMN is modified making it possible to simulate the transport of a species with a rigorous chemical model. Including the rigorous chemical equations into FEHMN simulations should provide for more representative transport models for highly reactive chemical species. A fully kinetic formulation is chosen for the FEHMN reactive transport model. Several methods are available to computationally implement a fully kinetic formulation. Different numerical algorithms are investigated in order to optimize computational efficiency and memory requirements of the reactive transport model. The best algorithm of those investigated is then incorporated into FEHMN. The algorithm chosen requires for the user to place strongly coupled species into groups which are then solved for simultaneously using FEHMN. The complete reactive transport model is verified over a wide variety of problems and is shown to be working properly. The simulations demonstrate that gas flow and carbonate chemistry can significantly affect 14 C transport at Yucca Mountain. The simulations also provide that the new capabilities of FEHMN can be used to refine and buttress already existing Yucca Mountain radionuclide transport studies

The development of a transient neutron flux solution in the PANTHER code

International Nuclear Information System (INIS)

Hutt, P.K.; Knight, M.P.

1990-01-01

In the United Kingdom a new three-dimensional, two-group, homogeneous reactor diffusion code, PANTHER, has been developed for the analysis of pressurized water reactors (PWRs) and advanced gas-cooled reactors (AGRs). The code can perform a comprehensive range of calculations, steady state, depletion, and transient with either a finite difference or analytic nodal flux solution. The nodal solution allows the representation of within-node burnup variation and pin-power reconstruction in either steady-state or transient mode. Specific steady-state and transient thermal feedback modules are included for both PWRs and AGRs. The code is being developed to perform a complete range of reactor calculations from online operational support to fuel management and fault transient analysis. In the area of transient analysis, the code is currently being used for a number of PWR fault transient assessments, including rod ejection and steam-line break. In addition, work is proceeding to incorporate the PANTHER 3D nodal transient solution in the TRAC-P code. This paper outlines the development of the transient flux solutions within PANTHER

Computer code ANISN multiplying media and shielding calculation II. Code description (input/output)

International Nuclear Information System (INIS)

Maiorino, J.R.

1990-01-01

The user manual of the ANISN computer code describing input and output subroutines is presented. ANISN code was developed to solve one-dimensional transport equation for neutron or gamma rays in slab, sphere or cylinder geometry with general anisotropic scattering. The solution technique is the discrete ordinate method. (M.C.K.)

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.

Semianalytical Solutions of Radioactive or Reactive Transport in Variably-Fractured Layered Media: 1. Solutes

International Nuclear Information System (INIS)

George J. Moridis

2001-01-01

In this paper, semianalytical solutions are developed for the problem of transport of radioactive or reactive solute tracers through a layered system of heterogeneous fractured media with misaligned fractures. The tracer transport equations in the non-flowing matrix account for (a) diffusion, (b) surface diffusion, (c) mass transfer between the mobile and immobile water fractions, (d) linear kinetic or equilibrium physical, chemical, or combined solute sorption or colloid filtration, and (e) radioactive decay or first-order chemical reactions. The tracer-transport equations in the fractures account for the same processes, in addition to advection and hydrodynamic dispersion. Any number of radioactive decay daughter products (or products of a linear, first-order reaction chain) can be tracked. The solutions, which are analytical in the Laplace space, are numerically inverted to provide the solution in time and can accommodate any number of fractured and/or porous layers. The solutions are verified using analytical solutions for limiting cases of solute and colloid transport through fractured and porous media. The effect of important parameters on the transport of 3 H, 237 Np and 239 Pu (and its daughters) is investigated in several test problems involving layered geological systems of varying complexity

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

KAMCCO, a reactor physics Monte Carlo neutron transport code

International Nuclear Information System (INIS)

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

1976-06-01

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

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

Optical Code-Division Multiple Access: Challenges and Solutions

Science.gov (United States)

Chen, Lawrence R.

2003-02-01

Optical code-division multiple-access (OCDMA) is a technique well-suited for providing the required photonic connectivity in local access networks. Although the principles of OCDMA have been known for many years, it has never delivered on its potential. In this paper, we will describe the key challenges and impediments that have prevented OCDMA from delivering on its potential, as well as discuss possible solutions. We focus on the limitations of one-dimensional codes and the benefit of exploiting the additional degrees of freedom in using multiple dimensions for defining the codes. We discuss the advantages of using differential detection in order to implement bipolar communications. We then show how two-dimensional wavelength-time codes can be appropriately combined with differential detection in order to achieve high performance OCDMA systems with a large number of users operating with good BER performance for a large aggregate capacity. We also discuss the impact of channel coding techniques, for example forward error correction or turbo coding, on BER performance.

Rn3D: A finite element code for simulating gas flow and radon transport in variably saturated, nonisothermal porous media

International Nuclear Information System (INIS)

Holford, D.J.

1994-01-01

This document is a user's manual for the Rn3D finite element code. Rn3D was developed to simulate gas flow and radon transport in variably saturated, nonisothermal porous media. The Rn3D model is applicable to a wide range of problems involving radon transport in soil because it can simulate either steady-state or transient flow and transport in one-, two- or three-dimensions (including radially symmetric two-dimensional problems). The porous materials may be heterogeneous and anisotropic. This manual describes all pertinent mathematics related to the governing, boundary, and constitutive equations of the model, as well as the development of the finite element equations used in the code. Instructions are given for constructing Rn3D input files and executing the code, as well as a description of all output files generated by the code. Five verification problems are given that test various aspects of code operation, complete with example input files, FORTRAN programs for the respective analytical solutions, and plots of model results. An example simulation is presented to illustrate the type of problem Rn3D is designed to solve. Finally, instructions are given on how to convert Rn3D to simulate systems other than radon, air, and water

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.)

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.)

Murt user's guide: A hybrid Lagrangian-Eulerian finite element model of multiple-pore-region solute transport through subsurface media

International Nuclear Information System (INIS)

Gwo, J.P.; Jardine, P.M.; Yeh, G.T.; Wilson, G.V.

1995-04-01

Matrix diffusion, a diffusive mass transfer process,in the structured soils and geologic units at ORNL, is believe to be an important subsurface mass transfer mechanism; it may affect off-site movement of radioactive wastes and remediation of waste disposal sites by locally exchanging wastes between soil/rock matrix and macropores/fractures. Advective mass transfer also contributes to waste movement but is largely neglected by researchers. This report presents the first documented 2-D multiregion solute transport code (MURT) that incorporates not only diffusive but also advective mass transfer and can be applied to heterogeneous porous media under transient flow conditions. In this report, theoretical background is reviewed and the derivation of multiregion solute transport equations is presented. Similar to MURF (Gwo et al. 1994), a multiregion subsurface flow code, multiplepore domains as suggested by previous investigators (eg, Wilson and Luxmoore 1988) can be implemented in MURT. Transient or steady-state flow fields of the pore domains can be either calculated by MURF or by modelers. The mass transfer process is briefly discussed through a three-pore-region multiregion solute transport mechanism. Mass transfer equations that describe mass flux across pore region interfaces are also presented and parameters needed to calculate mass transfer coefficients detailed. Three applications of MURT (tracer injection problem, sensitivity analysis of advective and diffusive mass transfer, hillslope ponding infiltration and secondary source problem) were simulated and results discussed. Program structure of MURT and functions of MURT subroutiness are discussed so that users can adapt the code; guides for input data preparation are provided in appendices

Composite Transport Model and Water and Solute Transport across Plant Roots: An Update.

Science.gov (United States)

Kim, Yangmin X; Ranathunge, Kosala; Lee, Seulbi; Lee, Yejin; Lee, Deogbae; Sung, Jwakyung

2018-01-01

The present review examines recent experimental findings in root transport phenomena in terms of the composite transport model (CTM). It has been a well-accepted conceptual model to explain the complex water and solute flows across the root that has been related to the composite anatomical structure. There are three parallel pathways involved in the transport of water and solutes in roots - apoplast, symplast, and transcellular paths. The role of aquaporins (AQPs), which facilitate water flows through the transcellular path, and root apoplast is examined in terms of the CTM. The contribution of the plasma membrane bound AQPs for the overall water transport in the whole plant level was varying depending on the plant species, age of roots with varying developmental stages of apoplastic barriers, and driving forces (hydrostatic vs. osmotic). Many studies have demonstrated that the apoplastic barriers, such as Casparian bands in the primary anticlinal walls and suberin lamellae in the secondary cell walls, in the endo- and exodermis are not perfect barriers and unable to completely block the transport of water and some solute transport into the stele. Recent research on water and solute transport of roots with and without exodermis triggered the importance of the extension of conventional CTM adding resistances that arrange in series (epidermis, exodermis, mid-cortex, endodermis, and pericycle). The extension of the model may answer current questions about the applicability of CTM for composite water and solute transport of roots that contain complex anatomical structures with heterogeneous cell layers.

Development of AGNES, a kinetics code for fissile solutions, 1

International Nuclear Information System (INIS)

Nakajima, Ken; Ohnishi, Nobuaki

1986-01-01

A kinetics code for fissile solutions, AGNES (Accidentally Generated Nuclear Excursion Simulation code), has been developed. This code calculates the radiolytic gas void effect as a reactivity feedback. Physical and calculative models of the radiolytic gas void are summarized and the usage of AGNES is described. In addition, some benchmark calculations were performed and results of calculations show good agreement with those of experiments. (author)

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.

Development of a three-dimensional neutron transport code DFEM based on the double finite element method

International Nuclear Information System (INIS)

Fujimura, Toichiro

1996-01-01

A three-dimensional neutron transport code DFEM has been developed by the double finite element method to analyze reactor cores with complex geometry as large fast reactors. Solution algorithm is based on the double finite element method in which the space and angle finite elements are employed. A reactor core system can be divided into some triangular and/or quadrangular prism elements, and the spatial distribution of neutron flux in each element is approximated with linear basis functions. As for the angular variables, various basis functions are applied, and their characteristics were clarified by comparison. In order to enhance the accuracy, a general method is derived to remedy the truncation errors at reflective boundaries, which are inherent in the conventional FEM. An adaptive acceleration method and the source extrapolation method were applied to accelerate the convergence of the iterations. The code structure is outlined and explanations are given on how to prepare input data. A sample input list is shown for reference. The eigenvalue and flux distribution for real scale fast reactors and the NEA benchmark problems were presented and discussed in comparison with the results of other transport codes. (author)

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)

How to Find a Bug in Ten Thousand Lines Transport Solver? Outline of Experiences from AN Advection-Diffusion Code Verification

Science.gov (United States)

Zamani, K.; Bombardelli, F.

2011-12-01

Almost all natural phenomena on Earth are highly nonlinear. Even simplifications to the equations describing nature usually end up being nonlinear partial differential equations. Transport (ADR) equation is a pivotal equation in atmospheric sciences and water quality. This nonlinear equation needs to be solved numerically for practical purposes so academicians and engineers thoroughly rely on the assistance of numerical codes. Thus, numerical codes require verification before they are utilized for multiple applications in science and engineering. Model verification is a mathematical procedure whereby a numerical code is checked to assure the governing equation is properly solved as it is described in the design document. CFD verification is not a straightforward and well-defined course. Only a complete test suite can uncover all the limitations and bugs. Results are needed to be assessed to make a distinction between bug-induced-defect and innate limitation of a numerical scheme. As Roache (2009) said, numerical verification is a state-of-the-art procedure. Sometimes novel tricks work out. This study conveys the synopsis of the experiences we gained during a comprehensive verification process which was done for a transport solver. A test suite was designed including unit tests and algorithmic tests. Tests were layered in complexity in several dimensions from simple to complex. Acceptance criteria defined for the desirable capabilities of the transport code such as order of accuracy, mass conservation, handling stiff source term, spurious oscillation, and initial shape preservation. At the begining, mesh convergence study which is the main craft of the verification is performed. To that end, analytical solution of ADR equation gathered. Also a new solution was derived. In the more general cases, lack of analytical solution could be overcome through Richardson Extrapolation and Manufactured Solution. Then, two bugs which were concealed during the mesh convergence

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

Solutes transport in unsaturated double-porosity medium. Modelling by homogenization and applications

International Nuclear Information System (INIS)

Tran Ngoc, T.D.

2008-07-01

This Ph.D thesis presents the development of the solute transport models in unsaturated double-porosity medium, by using the asymptotic homogenization method. The obtained macroscopic models concern diffusion, diffusion-convection and dispersion-convection, according to the transport regime which is characterized by the non-dimensional numbers. The models consist of two coupled equations that show the local non-equilibrium of concentrations. The double-porosity transport models were numerically implemented using the code COMSOL Multiphysics (finite elements method), and compared with the solution of the same problem at the fine scale. The implementation allows solving the coupled equations in the macro- and micro-porosity domains (two-scale computations). The calculations of the dispersion tensor as a solution of the local boundary value problems, were also conducted. It was shown that the dispersivity depends on the saturation, the physical properties of the macro-porosity domain and the internal structure of the double-porosity medium. Finally, two series of experiments were performed on a physical model of double-porosity that is composed of a periodic assemblage of sintered clay spheres in Hostun sand HN38. The first experiment was a drainage experiment, which was conducted in order to validate the unsaturated flow model. The second series was a dispersion experiment in permanent unsaturated water flow condition (water content measured by gamma ray attenuation technique). A good agreement between the numerical simulations and the experimental observations allows the validation of the developed models. (author)

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)

A new Eulerian-Lagrangian finite element simulator for solute transport in discrete fracture-matrix systems

Energy Technology Data Exchange (ETDEWEB)

Birkholzer, J.; Karasaki, K. [Lawrence Berkeley National Lab., CA (United States). Earth Sciences Div.

1996-07-01

Fracture network simulators have extensively been used in the past for obtaining a better understanding of flow and transport processes in fractured rock. However, most of these models do not account for fluid or solute exchange between the fractures and the porous matrix, although diffusion into the matrix pores can have a major impact on the spreading of contaminants. In the present paper a new finite element code TRIPOLY is introduced which combines a powerful fracture network simulator with an efficient method to account for the diffusive interaction between the fractures and the adjacent matrix blocks. The fracture network simulator used in TRIPOLY features a mixed Lagrangian-Eulerian solution scheme for the transport in fractures, combined with an adaptive gridding technique to account for sharp concentration fronts. The fracture-matrix interaction is calculated with an efficient method which has been successfully used in the past for dual-porosity models. Discrete fractures and matrix blocks are treated as two different systems, and the interaction is modeled by introducing sink/source terms in both systems. It is assumed that diffusive transport in the matrix can be approximated as a one-dimensional process, perpendicular to the adjacent fracture surfaces. A direct solution scheme is employed to solve the coupled fracture and matrix equations. The newly developed combination of the fracture network simulator and the fracture-matrix interaction module allows for detailed studies of spreading processes in fractured porous rock. The authors present a sample application which demonstrate the codes ability of handling large-scale fracture-matrix systems comprising individual fractures and matrix blocks of arbitrary size and shape.

Numerical solution of neutron transport equations in discrete ordinates and slab geometry

International Nuclear Information System (INIS)

Serrano Pedraza, F.

1985-01-01

An unified formalism to solve numerically, between other equation, the neutron transport in discrete ordinates, slab geometry, several energy groups and independents of time, has been developed recently. Such a formalism cover some of the conventional schemes as diamond difference, (WDD) characteristic step (SC) lineal characteristic (LC), quadratic characteristic (QC) and lineal discontinuous. Unified formation gives before hand the convergence order of the previously selected scheme. In fact it allows besides to generate a big amount of numerical schemes, with which is also possible to solve numerical equations as soon as neutron transport. The essential purpose of this work was to solve the neutron transport equations in slab geometry and discrete ordinates considering several energy groups without to take under advisement time dependence based in the above mentioned unified formalism. To reach this purpose it was necesary to design a computer code with the name TNOD1 (Neutron transport in discrete ordinates and 1 dimension) which includes each one of the schemes already pointed out. there exist two numerical schemes, also recently developed, quadratic continuous (QC) and cubic continuous (CN), although covered by unified formalism, it has been possible to include them inside this computer code without make substantial changes in its structure. In chapter I, derivative of neutron transport equation independent of time is taken, for angular flux, including boundary conditions and discontinuity. In chapter II the neutron transport equations are obtained in multigroups, independents of time, for approximation of discrete ordinates. Description of theory related with unified formalism and its relationship with mentioned discretization schemes is presented in chapter III. Chapter IV describes the computer code developed and finally, in chapter V different numerical results obtained with TNOD1 program are shown. In Appendix A theorems and mathematical arguments used

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

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

Composite Transport Model and Water and Solute Transport across Plant Roots: An Update

Directory of Open Access Journals (Sweden)

Yangmin X. Kim

2018-02-01

Full Text Available The present review examines recent experimental findings in root transport phenomena in terms of the composite transport model (CTM. It has been a well-accepted conceptual model to explain the complex water and solute flows across the root that has been related to the composite anatomical structure. There are three parallel pathways involved in the transport of water and solutes in roots – apoplast, symplast, and transcellular paths. The role of aquaporins (AQPs, which facilitate water flows through the transcellular path, and root apoplast is examined in terms of the CTM. The contribution of the plasma membrane bound AQPs for the overall water transport in the whole plant level was varying depending on the plant species, age of roots with varying developmental stages of apoplastic barriers, and driving forces (hydrostatic vs. osmotic. Many studies have demonstrated that the apoplastic barriers, such as Casparian bands in the primary anticlinal walls and suberin lamellae in the secondary cell walls, in the endo- and exodermis are not perfect barriers and unable to completely block the transport of water and some solute transport into the stele. Recent research on water and solute transport of roots with and without exodermis triggered the importance of the extension of conventional CTM adding resistances that arrange in series (epidermis, exodermis, mid-cortex, endodermis, and pericycle. The extension of the model may answer current questions about the applicability of CTM for composite water and solute transport of roots that contain complex anatomical structures with heterogeneous cell layers.

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.

Benchmark testing and independent verification of the VS2DT computer code

International Nuclear Information System (INIS)

McCord, J.T.

1994-11-01

The finite difference flow and transport simulator VS2DT was benchmark tested against several other codes which solve the same equations (Richards equation for flow and the Advection-Dispersion equation for transport). The benchmark problems investigated transient two-dimensional flow in a heterogeneous soil profile with a localized water source at the ground surface. The VS2DT code performed as well as or better than all other codes when considering mass balance characteristics and computational speed. It was also rated highly relative to the other codes with regard to ease-of-use. Following the benchmark study, the code was verified against two analytical solutions, one for two-dimensional flow and one for two-dimensional transport. These independent verifications show reasonable agreement with the analytical solutions, and complement the one-dimensional verification problems published in the code's original documentation

Exact solution of the neutron transport equation in spherical geometry

Energy Technology Data Exchange (ETDEWEB)

Anli, Fikret; Akkurt, Abdullah; Yildirim, Hueseyin; Ates, Kemal [Kahramanmaras Suetcue Imam Univ. (Turkey). Faculty of Sciences and Letters

2017-03-15

Solution of the neutron transport equation in one dimensional slab geometry construct a basis for the solution of neutron transport equation in a curvilinear geometry. Therefore, in this work, we attempt to derive an exact analytical benchmark solution for both neutron transport equations in slab and spherical medium by using P{sub N} approximation which is widely used in neutron transport theory.

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...

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)

A multi scale approximation solution for the time dependent Boltzmann-transport equation

International Nuclear Information System (INIS)

Merk, B.

2004-03-01

once more compared with the exact analytical solution obtaining good agreement. In the next steps multiple scale expansion solutions are developed for the space-time dependent P 1 and P 3 transport equations for the homogenized cell and 2 delayed neutorn groups. These results are analysed versus the solution for the diffusion equation emphasizing the differences in the space-time structure between the time dependent diffusion- and transport solutions. The effect of the additional derivation terms in the transport equations can be observed during the analytical expansion process and in the graphical analysis of the differences between the solutions. The developed solution is tested for direct calculation of the time behaviour of single nodes in the framework of a nodal code and the results are compared. It is evident that the nature of the inserted perturbation has major impact on the discrepancy of the results compared to the reference nodal method. (orig.) [de

Assessment of assembly homogenized two-steps core dynamic calculations using direct whole core transport solutions

International Nuclear Information System (INIS)

Hursin, Mathieu; Downar, Thomas J.; Yoon, Joo Il; Joo, Han Gyu

2016-01-01

Highlights: • Reactivity initiated accident analysis with direct whole core transient transport code. • Comparison with usual “two steps” procedure. • Effect of effective delayed neutron fraction definition on energy deposition in the fuel. • Effect of homogenized few-group cross sections generation at the assembly level on energy deposition in the fuel. • Effect of effective fuel temperature definition on energy deposition in the fuel. - Abstract: The impact of the approximations in the “two-steps” procedure used in the current generation of nodal simulators for core transient calculations is assessed by using a higher order solution obtained from a direct, whole core, transient transport calculation. A control rod ejection accident in an idealized minicore is analyzed with PARCS, which uses the two-steps procedure and DeCART which provides the higher order solution. DeCART is used as lattice code to provide the homogenized cross sections and kinetics parameters to PARCS. The approximations made by using (1) the homogenized few-group cross sections and kinetic parameters generated at the assembly level, (2) an effective delayed neutrons fraction, (3) an effective fuel temperature and (4) the few-group formulation are investigated in terms of global and local core power behavior. The results presented in the paper show that the current two-steps procedure produces sufficiently accurate transient results with respect to the direct whole core calculation solution, provided that its parameters are carefully generated using the prescriptions described in the present article.

Nonrelativistic grey Sn-transport radiative-shock solutions

International Nuclear Information System (INIS)

Ferguson, J. M.; Morel, J. E.; Lowrie, R. B.

2017-01-01

We present semi-analytic radiative-shock solutions in which grey Sn-transport is used to model the radiation, and we include both constant cross sections and cross sections that depend on temperature and density. These new solutions solve for a variable Eddington factor (VEF) across the shock domain, which allows for interesting physics not seen before in radiative-shock solutions. Comparisons are made with the grey nonequilibrium-diffusion radiative-shock solutions of Lowrie and Edwards [1], which assumed that the Eddington factor is constant across the shock domain. It is our experience that the local Mach number is monotonic when producing nonequilibrium-diffusion solutions, but that this monotonicity may disappear while integrating the precursor region to produce Sn-transport solutions. For temperature- and density-dependent cross sections we show evidence of a spike in the VEF in the far upstream portion of the radiative-shock precursor. We show evidence of an adaptation zone in the precursor region, adjacent to the embedded hydrodynamic shock, as conjectured by Drake [2, 3], and also confirm his expectation that the precursor temperatures adjacent to the Zel’dovich spike take values that are greater than the downstream post-shock equilibrium temperature. We also show evidence that the radiation energy density can be nonmonotonic under the Zel’dovich spike, which is indicative of anti-diffusive radiation flow as predicted by McClarren and Drake [4]. We compare the angle dependence of the radiation flow for the Sn-transport and nonequilibriumdiffusion radiation solutions, and show that there are considerable differences in the radiation flow between these models across the shock structure. Lastly, we analyze the radiation flow to understand the cause of the adaptation zone, as well as the structure of the Sn-transport radiation-intensity solutions across the shock structure.

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)

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

Transport Visualization for Studying Mass Transfer and Solute Transport in Permeable Media

International Nuclear Information System (INIS)

Roy Haggerty

2004-01-01

Understanding and predicting mass transfer coupled with solute transport in permeable media is central to several energy-related programs at the US Department of Energy (e.g., CO 2 sequestration, nuclear waste disposal, hydrocarbon extraction, and groundwater remediation). Mass transfer is the set of processes that control movement of a chemical between mobile (advection-dominated) domains and immobile (diffusion- or sorption-dominated) domains within a permeable medium. Consequences of mass transfer on solute transport are numerous and may include (1) increased sequestration time within geologic formations; (2) reduction in average solute transport velocity by as much as several orders of magnitude; (3) long ''tails'' in concentration histories during removal of a solute from a permeable medium; (4) poor predictions of solute behavior over long time scales; and (5) changes in reaction rates due to mass transfer influences on pore-scale mixing of solutes. Our work produced four principle contributions: (1) the first comprehensive visualization of solute transport and mass transfer in heterogeneous porous media; (2) the beginnings of a theoretical framework that encompasses both macrodispersion and mass transfer within a single set of equations; (3) experimental and analytical tools necessary for understanding mixing and aqueous reaction in heterogeneous, granular porous media; (4) a clear experimental demonstration that reactive transport is often not accurately described by a simple coupling of the convection-dispersion equation with chemical reaction equations. The work shows that solute transport in heterogeneous media can be divided into 3 regimes--macrodispersion, advective mass transfer, and diffusive mass transfer--and that these regimes can be predicted quantitatively in binary media. We successfully predicted mass transfer in each of these regimes and verified the prediction by completing quantitative visualization experiments in each of the regimes, the

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...

CFD analysis for the hydrogen transport in the primary contention of a BWR using the codes OpenFOAM and Gas-Flow

International Nuclear Information System (INIS)

Jimenez P, D. A.

2014-01-01

The accidents in Unit 2 of the Three Mile Island Nuclear Power Plant (NPP) in the United States (March 28 th , 1979), the one in Unit 4 of the NPP Chernobyl in Ukraine (April 26 th , 1986) and the explosions in some units of Fukushima NPP in Japan (March 11 th , 2011) boosted the investigations on severe accidents with core damage and, in particular, the threat to the ultimate barrier by an eventual explosion from uncontrolled Hydrogen combustion within the containment was considered of particular relevance. Research programs for analyzing Hydrogen behavior and control during this kind of accidents were early initiated by research and regulatory bodies. Assessment on Hydrogen behavior once it has been postulated to be released on the containment system can be divided into two phases, in the first one, transport and the concentrations of the gas mixtures and steam in each volume or area comprised between the structures of the containment are calculated, in the second one, the propagation of the detonation of the Hydrogen is calculated if there are the conditions to occur. Currently, there are computer programs that can be used in one, or both stages of computation, and they are based on one of the two solution methods in current use, one of them are integrated codes (e.g. MELCOR), which consists in assuming the containment as a network composed of hydraulic tanks or nodes on which the balance equations of mass and energy have to be solved, the network is connected by ducts or connections where the momentum balance equation arise. This methodology relies on the use of semi-empirical relationships and the criteria used to define a geometric pattern, are subjective. The second method, which is having relevance due to the large computing power of modern computers, is the numerical solution of the three-dimensional Navier-Stokes equations in complex geometries. This method of solution is known as Computational Fluid Dynamics (CFD), and offers the advantage of using a

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

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

Solute transport across the articular surface of injured cartilage.

Science.gov (United States)

Chin, Hooi Chuan; Moeini, Mohammad; Quinn, Thomas M

2013-07-15

Solute transport through extracellular matrix (ECM) is important to physiology and contrast agent-based clinical imaging of articular cartilage. Mechanical injury is likely to have important effects on solute transport since it involves alteration of ECM structure. Therefore it is of interest to characterize effects of mechanical injury on solute transport in cartilage. Using cartilage explants injured by an established mechanical compression protocol, effective partition coefficients and diffusivities of solutes for transport across the articular surface were measured. A range of fluorescent solutes (fluorescein isothiocyanate, 4 and 40kDa dextrans, insulin, and chondroitin sulfate) and an X-ray contrast agent (sodium iodide) were used. Mechanical injury was associated with a significant increase in effective diffusivity versus uninjured explants for all solutes studied. On the other hand, mechanical injury had no effects on effective partition coefficients for most solutes tested, except for 40kDa dextran and chondroitin sulfate where small but significant changes in effective partition coefficient were observed in injured explants. Findings highlight enhanced diffusive transport across the articular surface of injured cartilage, which may have important implications for injury and repair situations. Results also support development of non-equilibrium methods for identification of focal cartilage lesions by contrast agent-based clinical imaging. Copyright © 2013 Elsevier Inc. All rights reserved.

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 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

Development and applications of the channel network model for simulations of flow and solute transport in fractured rock

International Nuclear Information System (INIS)

Gylling, B.

1997-01-01

The Channel Network model and its computer implementation, the code CHAN3D, for simulations of fluid flow and transport of solutes have been developed. The tool may be used for performance and safety assessments of deep lying repositories in fractured rocks for nuclear and other hazardous wastes, e.g. chemical wastes. It may also be used to simulate and interpret field experiments of flow and transport in large or small scale. Fluid flow and solute transport in fractured media are of interest in the performance assessment of a repository for hazardous waste, located at depth in crystalline rock, with potential release of solutes. Fluid flow in fractured rock is found to be very unevenly distributed due to the heterogeneity of the medium. The water will seek the easiest path, channels, under a prevailing pressure gradient. Solutes in the flowing water may be transported through preferential paths and migrate from the water in the fractures into the stagnant water in the rock matrix. There, sorbing solutes may be sorbed on the micro surfaces within the matrix. The diffusion into the matrix and the sorption process may significantly retard the transport of species and increase the time available for radionuclide decay. Channelling and matrix diffusion contribute to the dispersion of solutes in the water. Important for performance assessment is that channeling may cause a portion of the solutes to arrive much faster than the rest of the solutes. Simulations of field experiments at the Aespoe Hard Rock Laboratory using the Channel Network model have been performed. The application of the model to the site and the simulation results of the pumping and tracer tests are presented. The results show that the model is capable of describing the hydraulic gradient and of predicting flow rates and tracer transport obtained in the experiments. The data requirements for the Channel Network model have been investigated to determine which data are the most important for predictions

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.

Solute transport in soil

NARCIS (Netherlands)

Zee, van der S.E.A.T.M.; Leijnse, A.

2013-01-01

Solute transport is of importance in view of the movement of nutrient elements, e.g. towards the plant root system, and because of a broad range of pollutants. Pollution is not necessarily man induced, but may be due to geological or geohydrological causes, e.g. in the cases of pollution with

Effects of turbulent hyporheic mixing on reach-scale solute transport

Science.gov (United States)

Roche, K. R.; Li, A.; Packman, A. I.

2017-12-01

Turbulence rapidly mixes solutes and fine particles into coarse-grained streambeds. Both hyporheic exchange rates and spatial variability of hyporheic mixing are known to be controlled by turbulence, but it is unclear how turbulent mixing influences mass transport at the scale of stream reaches. We used a process-based particle-tracking model to simulate local- and reach-scale solute transport for a coarse-bed stream. Two vertical mixing profiles, one with a smooth transition from in-stream to hyporheic transport conditions and a second with enhanced turbulent transport at the sediment-water interface, were fit to steady-state subsurface concentration profiles observed in laboratory experiments. The mixing profile with enhanced interfacial transport better matched the observed concentration profiles and overall mass retention in the streambed. The best-fit mixing profiles were then used to simulate upscaled solute transport in a stream. Enhanced mixing coupled in-stream and hyporheic solute transport, causing solutes exchanged into the shallow subsurface to have travel times similar to the water column. This extended the exponential region of the in-stream solute breakthrough curve, and delayed the onset of the heavy power-law tailing induced by deeper and slower hyporheic porewater velocities. Slopes of observed power-law tails were greater than those predicted from stochastic transport theory, and also changed in time. In addition, rapid hyporheic transport velocities truncated the hyporheic residence time distribution by causing mass to exit the stream reach via subsurface advection, yielding strong exponential tempering in the in-stream breakthrough curves at the timescale of advective hyporheic transport through the reach. These results show that strong turbulent mixing across the sediment-water interface violates the conventional separation of surface and subsurface flows used in current models for solute transport in rivers. Instead, the full distribution of

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

Modification of the finite element heat and mass transfer code (FEHM) to model multicomponent reactive transport

International Nuclear Information System (INIS)

Viswanathan, H.S.

1996-08-01

The finite element code FEHMN, developed by scientists at Los Alamos National Laboratory (LANL), is a three-dimensional finite element heat and mass transport simulator that can handle complex stratigraphy and nonlinear processes such as vadose zone flow, heat flow and solute transport. Scientists at LANL have been developing hydrologic flow and transport models of the Yucca Mountain site using FEHMN. Previous FEHMN simulations have used an equivalent Kd model to model solute transport. In this thesis, FEHMN is modified making it possible to simulate the transport of a species with a rigorous chemical model. Including the rigorous chemical equations into FEHMN simulations should provide for more representative transport models for highly reactive chemical species. A fully kinetic formulation is chosen for the FEHMN reactive transport model. Several methods are available to computationally implement a fully kinetic formulation. Different numerical algorithms are investigated in order to optimize computational efficiency and memory requirements of the reactive transport model. The best algorithm of those investigated is then incorporated into FEHMN. The algorithm chosen requires for the user to place strongly coupled species into groups which are then solved for simultaneously using FEHMN. The complete reactive transport model is verified over a wide variety of problems and is shown to be working properly. The new chemical capabilities of FEHMN are illustrated by using Los Alamos National Laboratory's site scale model of Yucca Mountain to model two-dimensional, vadose zone 14 C transport. The simulations demonstrate that gas flow and carbonate chemistry can significantly affect 14 C transport at Yucca Mountain. The simulations also prove that the new capabilities of FEHMN can be used to refine and buttress already existing Yucca Mountain radionuclide transport studies

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.

The secret to successful solute-transport modeling

Science.gov (United States)

Konikow, Leonard F.

2011-01-01

Modeling subsurface solute transport is difficult—more so than modeling heads and flows. The classical governing equation does not always adequately represent what we see at the field scale. In such cases, commonly used numerical models are solving the wrong equation. Also, the transport equation is hyperbolic where advection is dominant, and parabolic where hydrodynamic dispersion is dominant. No single numerical method works well for all conditions, and for any given complex field problem, where seepage velocity is highly variable, no one method will be optimal everywhere. Although we normally expect a numerically accurate solution to the governing groundwater-flow equation, errors in concentrations from numerical dispersion and/or oscillations may be large in some cases. The accuracy and efficiency of the numerical solution to the solute-transport equation are more sensitive to the numerical method chosen than for typical groundwater-flow problems. However, numerical errors can be kept within acceptable limits if sufficient computational effort is expended. But impractically long

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)

Fluid transport with time on peritoneal dialysis: the contribution of free water transport and solute coupled water transport

NARCIS (Netherlands)

Coester, Annemieke M.; Smit, Watske; Struijk, Dirk G.; Krediet, Raymond T.

2009-01-01

Ultrafiltration in peritoneal dialysis occurs through endothelial water channels (free water transport) and together with solutes across small pores: solute coupled water transport. A review is given of cross-sectional studies and on the results of longitudinal follow-up

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.)

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)

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.

Recent developments of the MARC/PN transport theory code including a treatment of anisotropic scatter

International Nuclear Information System (INIS)

Fletcher, J.K.

1987-12-01

The computer code MARC/PN provides a solution of the multigroup transport equation by expanding the flux in spherical harmonics. The coefficients of the series so obtained satisfy linked first order differential equations, and on eliminating terms associated with odd parity harmonics a second order system results which can be solved by established finite difference or finite element techniques. This report describes modifications incorporated in MARC/PN to allow for anisotropic scattering, and the modelling of irregular exterior boundaries in the finite element option. The latter development leads to substantial reductions in problem size, particularly for three dimensions. Also, links to an interactive graphics mesh generator (SUPERTAB) have been added. The final section of the report contains results from problems showing the effects of anisotropic scatter and the ability of the code to model irregular geometries. (author)

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

Solute transport in aggregated and layered porous media

International Nuclear Information System (INIS)

Koch, S.

1993-01-01

This work is a contribution to research in soil physics dealing with solute transport in porous media. The influence of structural inhomogeneities on solute transport is investigated. Detailed experiments at the laboratory scale are used to enlighten distinct processes which cannot be studied separately at field scale. Two main aspects are followed up: (i) to show the influence of aggregation of a porous medium on breakthrough time and spreading of an inert tracer and consequences on the estimation of parameter values of models describing solute transport in aggregated systems, (ii) to investigate the influences on the dispersion process when stratification is perpendicular to the direction of flow. Several concepts of modelling solute transport in soil are discussed. Models based on the convection-dispersion equation (CDE) are emphasized because they are used here to model solute transport experiments conducted with aggregated porous media. Stochastic concepts are introduced to show the limitations of the deterministic CDE approaches. Experiments are done in columns containing two kinds of solid phases and were saturated with water. The solid phases are porous and solid glass beads exhibiting a distinctly unimodal or bimodal pore size distribution. Experimental breakthrough curves (BTCs) are modelled with the CDE, a bicontinuum model with a phenomenological mass transfer rate and a bicontinuum spherical diffusion model. Experiments are also done in columns that are unsaturated containing porous materials that are layered. Flow is made at a steady rate. It is shown that layer boundaries have a severe influence on lateral mixing. They may force streamlines to converge or cause a lateral redistribution of solutes. (author) figs., tabs., 122 refs

The CFEST-INV stochastic hydrology code: Mathematical formulation, application, and user's manual

International Nuclear Information System (INIS)

Devary, J.L.

1987-06-01

Performance assessments of a nuclear waste repository must consider the hydrologic, thermal, mechanical, and geochemical environments of a candidate site. Predictions of radionuclide transport requires estimating water movement as a function of pressure, temperature, and solute concentration. CFEST (Coupled Fluid, Energy, and Solute Transport), is a finite-element based groundwater code that can be used to simultaneously solve the partial differential equations for pressure head, solute temperature, and solute concentration. The CFEST code has been designed to support site, repository, and waste package subsystem assessments. CFEST-INV is a stochastic hydrology code that was developed to augment the CFEST code in data processing; model calibration; performance prediction; error propagation; and data collection guidance. The CFEST-INV code utilizes kriging, finite-element modeling, adjoint-sensitivity, statistical-inverse, first-order variance, and Monte-Carlo techniques to develop performance (measure) driven data collection schemes and to determine the waste isolation capabilities (including uncertainties) of candidate repository sites. This report contains the basic physical and numerical principles of the CFEST-INV code, its input parameters, verification exercises, a user's manual, and the code's application history. 18 refs., 16 figs., 6 tabs

A three-dimensional neutron transport benchmark solution

International Nuclear Information System (INIS)

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

1993-01-01

For one-group neutron transport theory in one dimension, several powerful analytical techniques have been developed to solve the neutron transport equation, including Caseology, Wiener-Hopf factorization, and Fourier and Laplace transform methods. In addition, after a Fourier transform in the transverse plane and formulation of a pseudo problem, two-dimensional (2-D) and three-dimensional (3-D) problems can be solved using the techniques specifically developed for the one-dimensional (1-D) case. Numerical evaluation of the resulting expressions requiring an inversion in the transverse plane have been successful for 2-D problems but becomes exceedingly difficult in the 3-D case. In this paper, we show that by using the symmetry along the beam direction, a 2-D problem can be transformed into a 3-D problem in an infinite medium. The numerical solution to the 3-D problem is then demonstrated. Thus, a true 3-D transport benchmark solution can be obtained from a well-established numerical solution to a 2-D problem

Coupling between solute transport and chemical reactions models

International Nuclear Information System (INIS)

Samper, J.; Ajora, C.

1993-01-01

During subsurface transport, reactive solutes are subject to a variety of hydrodynamic and chemical processes. The major hydrodynamic processes include advection and convection, dispersion and diffusion. The key chemical processes are complexation including hydrolysis and acid-base reactions, dissolution-precipitation, reduction-oxidation, adsorption and ion exchange. The combined effects of all these processes on solute transport must satisfy the principle of conservation of mass. The statement of conservation of mass for N mobile species leads to N partial differential equations. Traditional solute transport models often incorporate the effects of hydrodynamic processes rigorously but oversimplify chemical interactions among aqueous species. Sophisticated chemical equilibrium models, on the other hand, incorporate a variety of chemical processes but generally assume no-flow systems. In the past decade, coupled models accounting for complex hydrological and chemical processes, with varying degrees of sophistication, have been developed. The existing models of reactive transport employ two basic sets of equations. The transport of solutes is described by a set of partial differential equations, and the chemical processes, under the assumption of equilibrium, are described by a set of nonlinear algebraic equations. An important consideration in any approach is the choice of primary dependent variables. Most existing models cannot account for the complete set of chemical processes, cannot be easily extended to include mixed chemical equilibria and kinetics, and cannot handle practical two and three dimensional problems. The difficulties arise mainly from improper selection of the primary variables in the transport equations. (Author) 38 refs

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)

A solution for automatic parallelization of sequential assembly code

Directory of Open Access Journals (Sweden)

Kovačević Đorđe

2013-01-01

Full Text Available Since modern multicore processors can execute existing sequential programs only on a single core, there is a strong need for automatic parallelization of program code. Relying on existing algorithms, this paper describes one new software solution tool for parallelization of sequential assembly code. The main goal of this paper is to develop the parallelizator which reads sequential assembler code and at the output provides parallelized code for MIPS processor with multiple cores. The idea is the following: the parser translates assembler input file to program objects suitable for further processing. After that the static single assignment is done. Based on the data flow graph, the parallelization algorithm separates instructions on different cores. Once sequential code is parallelized by the parallelization algorithm, registers are allocated with the algorithm for linear allocation, and the result at the end of the program is distributed assembler code on each of the cores. In the paper we evaluate the speedup of the matrix multiplication example, which was processed by the parallelizator of assembly code. The result is almost linear speedup of code execution, which increases with the number of cores. The speed up on the two cores is 1.99, while on 16 cores the speed up is 13.88.

Pathogen transport in groundwater systems: contrasts with traditional solute transport

Science.gov (United States)

Hunt, Randall J.; Johnson, William P.

2017-06-01

Water quality affects many aspects of water availability, from precluding use to societal perceptions of fit-for-purpose. Pathogen source and transport processes are drivers of water quality because they have been responsible for numerous outbreaks resulting in large economic losses due to illness and, in some cases, loss of life. Outbreaks result from very small exposure (e.g., less than 20 viruses) from very strong sources (e.g., trillions of viruses shed by a single infected individual). Thus, unlike solute contaminants, an acute exposure to a very small amount of contaminated water can cause immediate adverse health effects. Similarly, pathogens are larger than solutes. Thus, interactions with surfaces and settling become important even as processes important for solutes such as diffusion become less important. These differences are articulated in "Colloid Filtration Theory", a separate branch of pore-scale transport. Consequently, understanding pathogen processes requires changes in how groundwater systems are typically characterized, where the focus is on the leading edges of plumes and preferential flow paths, even if such features move only a very small fraction of the aquifer flow. Moreover, the relatively short survival times of pathogens in the subsurface require greater attention to very fast (solute transport mechanisms discussed here, a more encompassing view of water quality and source water protection is attained. With this more holistic view and theoretical understanding, better evaluations can be made regarding drinking water vulnerability and the relation between groundwater and human health.

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...

Modeling solute transport in a heterogeneous unsaturated porous medium under dynamic boundary conditions on different spatial scales

Science.gov (United States)

Cremer, Clemens; Neuweiler, Insa; Bechtold, Michel

2013-04-01

Understanding transport of solutes/contaminants through unsaturated soil in the shallow subsurface is vital to assess groundwater quality, nutrient cycling or to plan remediation projects. Alternating precipitation and evaporation conditions causing upward and downward flux with differing flow paths, changes in saturation and related structural heterogeneity make the description of transport in the unsaturated zone near the soil-surface a complex problem. Preferential flow paths strongly depend, among other things, on the saturation of a medium. Recent studies (e.g. Bechtold et al., 2011) showed lateral flow and solute transport during evaporation conditions (upward flux) in vertically layered sand columns. Results revealed that during evaporation water and solute are redistributed laterally from coarse to fine media deeper in the soil, and towards zones of lowest hydraulic head near to the soil surface. These zones at the surface can be coarse or fine grained depending on saturation status and evaporation flux. However, if boundary conditions are reversed and precipitation is applied, the flow field is not reversed in the same manner, resulting in entirely different transport patterns for downward and upward flow. Therefore, considering net-flow rates alone is misleading when describing transport in the shallow unsaturated zone. In this contribution, we analyze transport of a solute in the shallow subsurface to assess effects resulting from the superposition of heterogeneous soil structures and dynamic flow conditions on various spatial scales. Two-dimensional numerical simulations of unsaturated flow and transport in heterogeneous porous media under changing boundary conditions are carried out using a finite-volume code coupled to a particle tracking algorithm to quantify solute transport and leaching rates. In order to validate numerical simulations, results are qualitatively compared to those of a physical experiment (Bechtold et al., 2011). Numerical

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)

Assessment of applications of transport models on regional scale solute transport

Science.gov (United States)

Guo, Z.; Fogg, G. E.; Henri, C.; Pauloo, R.

2017-12-01

Regional scale transport models are needed to support the long-term evaluation of groundwater quality and to develop management strategies aiming to prevent serious groundwater degradation. The purpose of this study is to evaluate the capacity of previously-developed upscaling approaches to accurately describe main solute transport processes including the capture of late-time tails under changing boundary conditions. Advective-dispersive contaminant transport in a 3D heterogeneous domain was simulated and used as a reference solution. Equivalent transport under homogeneous flow conditions were then evaluated applying the Multi-Rate Mass Transfer (MRMT) model. The random walk particle tracking method was used for both heterogeneous and homogeneous-MRMT scenarios under steady state and transient conditions. The results indicate that the MRMT model can capture the tails satisfactorily for plume transported with ambient steady-state flow field. However, when boundary conditions change, the mass transfer model calibrated for transport under steady-state conditions cannot accurately reproduce the tailing effect observed for the heterogeneous scenario. The deteriorating impact of transient boundary conditions on the upscaled model is more significant for regions where flow fields are dramatically affected, highlighting the poor applicability of the MRMT approach for complex field settings. Accurately simulating mass in both mobile and immobile zones is critical to represent the transport process under transient flow conditions and will be the future focus of our study.

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)

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

Fluid flow and convective transport of solutes within the intervertebral disc.

Science.gov (United States)

Ferguson, Stephen J; Ito, Keita; Nolte, Lutz P

2004-02-01

Previous experimental and analytical studies of solute transport in the intervertebral disc have demonstrated that for small molecules diffusive transport alone fulfils the nutritional needs of disc cells. It has been often suggested that fluid flow into and within the disc may enhance the transport of larger molecules. The goal of the study was to predict the influence of load-induced interstitial fluid flow on mass transport in the intervertebral disc. An iterative procedure was used to predict the convective transport of physiologically relevant molecules within the disc. An axisymmetric, poroelastic finite-element structural model of the disc was developed. The diurnal loading was divided into discrete time steps. At each time step, the fluid flow within the disc due to compression or swelling was calculated. A sequentially coupled diffusion/convection model was then employed to calculate solute transport, with a constant concentration of solute being provided at the vascularised endplates and outer annulus. Loading was simulated for a complete diurnal cycle, and the relative convective and diffusive transport was compared for solutes with molecular weights ranging from 400 Da to 40 kDa. Consistent with previous studies, fluid flow did not enhance the transport of low-weight solutes. During swelling, interstitial fluid flow increased the unidirectional penetration of large solutes by approximately 100%. Due to the bi-directional temporal nature of disc loading, however, the net effect of convective transport over a full diurnal cycle was more limited (30% increase). Further study is required to determine the significance of large solutes and the timing of their delivery for disc physiology.

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

Semianalytical solutions of radioactive or reactive tracer transport in layered fractured media

International Nuclear Information System (INIS)

Moridis, G.J.; Bodvarsson, G.S.

2001-01-01

In this paper, semianalytical solutions are developed for the problem of transport of radioactive or reactive tracers (solutes or colloids) through a layered system of heterogeneous fractured media with misaligned fractures. The tracer transport equations in the matrix account for (a) diffusion, (b) surface diffusion (for solutes only), (c) mass transfer between the mobile and immobile water fractions, (d) linear kinetic or equilibrium physical, chemical, or combined solute sorption or colloid filtration, and (e) radioactive decay or first order chemical reactions. Any number of radioactive decay daughter products (or products of a linear, first-order reaction chain) can be tracked. The tracer-transport equations in the fractures account for the same processes, in addition to advection and hydrodynamic dispersion. Additionally, the colloid transport equations account for straining and velocity adjustments related to the colloidal size. The solutions, which are analytical in the Laplace space, are numerically inverted to provide the solution in time and can accommodate any number of fractured and/or porous layers. The solutions are verified using analytical solutions for limiting cases of solute and colloid transport through fractured and porous media. The effect of important parameters on the transport of 3 H, 237 Np and 239 Pu (and its daughters) is investigated in several test problems involving layered geological systems of varying complexity. 239 Pu colloid transport problems in multilayered systems indicate significant colloid accumulations at straining interfaces but much faster transport of the colloid than the corresponding strongly sorbing solute species

Reevaluation of the case, de Hoffman, and Placzek one-group neutron transport benchmark solution in plane geometry

International Nuclear Information System (INIS)

Ganapol, B.D.

1986-01-01

In a course on neutron transport theory and also in the analytical neutron transport theory literature, the pioneering work of Case et al. (CdHP) is often referenced. This work was truly a monumental effort in that it treated the fundamental mathematical properties of the one-group neutron Boltzmann equation in detail as well as the numerical evaluation of most of the resulting solutions. Many mathematically and numerically oriented dissertations were based on this classic monograph. In light of the considerable advances made both in numerical methods and computer technology since 1953, when the historic CdHP monograph first appeared, it seems appropriate to reevaluate the numerical benchmark solutions found therein with present-day computational technology. In most transport theory courses, the subject of proper benchmarking of numerical algorithms and transport codes is seldom addressed at any great length. This may be the reason that the benchmarking procedure is so rarely practiced in the nuclear community and when practiced is improperly applied. In this presentation, the development of a new benchmark for the one-group neutron flux in an infinite medium will be detailed with emphasis placed on the educational aspects of the benchmarking activity

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

IPOLE - semi-analytic scheme for relativistic polarized radiative transport

Science.gov (United States)

Mościbrodzka, M.; Gammie, C. F.

2018-03-01

We describe IPOLE, a new public ray-tracing code for covariant, polarized radiative transport. The code extends the IBOTHROS scheme for covariant, unpolarized transport using two representations of the polarized radiation field: In the coordinate frame, it parallel transports the coherency tensor; in the frame of the plasma it evolves the Stokes parameters under emission, absorption, and Faraday conversion. The transport step is implemented to be as spacetime- and coordinate- independent as possible. The emission, absorption, and Faraday conversion step is implemented using an analytic solution to the polarized transport equation with constant coefficients. As a result, IPOLE is stable, efficient, and produces a physically reasonable solution even for a step with high optical depth and Faraday depth. We show that the code matches analytic results in flat space, and that it produces results that converge to those produced by Dexter's GRTRANS polarized transport code on a complicated model problem. We expect IPOLE will mainly find applications in modelling Event Horizon Telescope sources, but it may also be useful in other relativistic transport problems such as modelling for the IXPE mission.

Sustainable freight transport in South Africa:Domestic intermodal solutions

Directory of Open Access Journals (Sweden)

Jan H. Havenga

2011-11-01

Full Text Available Due to the rapid deregulation of freight transport in South Africa two decades ago, and low historical investment in rail (with resultant poor service delivery, an integrated alternative to road and rail competition was never developed. High national freight logistics costs, significant road infrastructure challenges and environmental impact concerns of a road-dominated freight transport market have, however, fuelled renewed interest in intermodal transport solutions. In this article, a high-level business case for domestic intermodal solutions in South Africa is presented. The results demonstrate that building three intermodal terminals to connect the three major industrial hubs (i.e. Gauteng, Durban and Cape Town through an intermodal solution could reduce transport costs (including externalities for the identified 11.5 million tons of intermodalfriendly freight flows on the Cape and Natal corridors by 42% (including externalities.

The solute carrier 6 family of transporters

DEFF Research Database (Denmark)

Bröer, Stefan; Gether, Ulrik

2012-01-01

of these transporters is associated with a variety of diseases. Pharmacological inhibition of the neurotransmitter transporters in this family is an important strategy in the management of neurological and psychiatric disorders. This review provides an overview of the biochemical and pharmacological properties......The solute carrier 6 (SLC6) family of the human genome comprises transporters for neurotransmitters, amino acids, osmolytes and energy metabolites. Members of this family play critical roles in neurotransmission, cellular and whole body homeostasis. Malfunction or altered expression...... of the SLC6 family transporters....

Solute transport and storage mechanisms in wetlands of the Everglades, south Florida

Science.gov (United States)

Harvey, Judson W.; Saiers, James E.; Newlin, Jessica T.

2005-01-01

Solute transport and storage processes in wetlands play an important role in biogeochemical cycling and in wetland water quality functions. In the wetlands of the Everglades, there are few data or guidelines to characterize transport through the heterogeneous flow environment. Our goal was to conduct a tracer study to help quantify solute exchange between the relatively fast flowing water in the open part of the water column and much more slowly moving water in thick floating vegetation and in the pore water of the underlying peat. We performed a tracer experiment that consisted of a constant‐rate injection of a sodium bromide (NaBr) solution for 22 hours into a 3 m wide, open‐ended flume channel in Everglades National Park. Arrival of the bromide tracer was monitored at an array of surface water and subsurface samplers for 48 hours at a distance of 6.8 m downstream of the injection. A one‐dimensional transport model was used in combination with an optimization code to identify the values of transport parameters that best explained the tracer observations. Parameters included dimensions and mass transfer coefficients describing exchange with both short (hours) and longer (tens of hours) storage zones as well as the average rates of advection and longitudinal dispersion in the open part of the water column (referred to as the “main flow zone”). Comparison with a more detailed set of tracer measurements tested how well the model's storage zones approximated the average characteristics of tracer movement into and out of the layer of thick floating vegetation and the pore water in the underlying peat. The rate at which the relatively fast moving water in the open water column was exchanged with slowly moving water in the layer of floating vegetation and in sediment pore water amounted to 50 and 3% h−1, respectively. Storage processes decreased the depth‐averaged velocity of surface water by 50% relative to the water velocity in the open part of the water

A single continuum approximation of the solute transport in fractured porous media

International Nuclear Information System (INIS)

Jeong, J.T.; Lee, K.J.

1992-01-01

Solute transport in fractured porous media is described by the single continuum model, i.e., equivalent porous medium model. In this model, one-dimensional solute transport in the fracture and two-dimensional solute transport in the porous rock matrix is considered. The network of fractures embedded in the porous rock matrix is idealized as two orthogonally intersecting families of equally spaced, parallel fractures directed at 45 o to the regional groundwater flow direction. Governing equations are solved by the finite element method, and an upstream weighting technique is used in order to prevent the oscillation of the solution in the case of highly advection dominated transport. Breakthrough curves, similar to those of the one-dimensional solute transport problem in ordinary porous media, are obtained as a function of time according to volume or flux averaging of the concentration profile across the width of the flow region. The equivalent parameters, i.e., porosity and overall coefficient of longitudinal dispersivity, are obtained by a trial-and-error method. Analyses for the non-sorbing solute transport case show that within the range of considered parameters, and except for the region very close to the source, application of the single continuum model in the idealized fracture system is sufficient for modeling solute transport in fractured porous media. This numerical scheme is shown to be applicable to a sorbing solute and radionuclide transport. (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.

From analytical solutions of solute transport equations to multidimensional time-domain random walk (TDRW) algorithms

Science.gov (United States)

Bodin, Jacques

2015-03-01

In this study, new multi-dimensional time-domain random walk (TDRW) algorithms are derived from approximate one-dimensional (1-D), two-dimensional (2-D), and three-dimensional (3-D) analytical solutions of the advection-dispersion equation and from exact 1-D, 2-D, and 3-D analytical solutions of the pure-diffusion equation. These algorithms enable the calculation of both the time required for a particle to travel a specified distance in a homogeneous medium and the mass recovery at the observation point, which may be incomplete due to 2-D or 3-D transverse dispersion or diffusion. The method is extended to heterogeneous media, represented as a piecewise collection of homogeneous media. The particle motion is then decomposed along a series of intermediate checkpoints located on the medium interface boundaries. The accuracy of the multi-dimensional TDRW method is verified against (i) exact analytical solutions of solute transport in homogeneous media and (ii) finite-difference simulations in a synthetic 2-D heterogeneous medium of simple geometry. The results demonstrate that the method is ideally suited to purely diffusive transport and to advection-dispersion transport problems dominated by advection. Conversely, the method is not recommended for highly dispersive transport problems because the accuracy of the advection-dispersion TDRW algorithms degrades rapidly for a low Péclet number, consistent with the accuracy limit of the approximate analytical solutions. The proposed approach provides a unified methodology for deriving multi-dimensional time-domain particle equations and may be applicable to other mathematical transport models, provided that appropriate analytical solutions are available.

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)

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

Foreign trip report: International Nuclide Transport Code Intercomparison Study (INTRACOIN) workshop and coordinating committee meeting, Interlaken, Switzerland, February 15-20, 1982

International Nuclear Information System (INIS)

Jansen, G. Jr.; Hewitt, W.M.

1986-01-01

The INTRACOIN (International Nuclide TRAnsport COde INtercomparison study) workshop provided material with which to benchmark codes and demonstrated a wide variety of capabilities for several types of transport problems in radionuclide migrations. Review lectures pointed to the general methods for solving ground-water transport problems and the difficulties to be expected in applying them to performance assessment. Numerical examples of solutions to benchmark problems important in waste repository safety assessment were presented. A framework of solved analytical models was provided by the work of Dr. Thomas H. Pigford of the University of California, Berkeley. Results from the US Department of Energy (DOE)-sponsored participants showed that the analytical codes are working properly and give results comparable to those of others. Further analysis of the results is needed to pinpoint discrepancies and resolve conflicts in boundary assumptions. In the coordinating group meeting, plans were made to complete a Level 1 report, and three problems with field test pumping data were fully defined and agreed upon for Level 2. The Office of Nuclear Waste Isolation (ONWT) agreed to host the Level 2 workshop in November 1982 and to appoint technical advisors to the Level 3 technical subcommittee of the coordinating group. Some information is included in microfiche form only. 8 refs., 11 tabs

The future of public transport in light of solutions for sustainable transport development

Directory of Open Access Journals (Sweden)

Kazimierz LEJDA

2017-06-01

Full Text Available The paper highlights possible directions in the development of sustainable public transport solutions. When appropriately identified and implemented, such solutions can contribute to improved quality of urban life by reducing the adverse effects of transport on human health and the natural environment. The paper also raises questions about implementing pedestrian traffic zones, expanding the level of cycling, and introducing a workable parking policy, congestion charges, electromobility and intelligent systems for road traffic management in conurbations.

The fusion code XGC: Enabling kinetic study of multi-scale edge turbulent transport in ITER

Energy Technology Data Exchange (ETDEWEB)

D' Azevedo, Eduardo [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Abbott, Stephen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Koskela, Tuomas [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Worley, Patrick [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ku, Seung-Hoe [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Ethier, Stephane [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Yoon, Eisung [Rensselaer Polytechnic Inst., Troy, NY (United States); Shephard, Mark [Rensselaer Polytechnic Inst., Troy, NY (United States); Hager, Robert [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Lang, Jianying [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Intel Corporation, Santa Clara, CA (United States); Choi, Jong [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Podhorszki, Norbert [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Klasky, Scott [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Parashar, Manish [Rutgers Univ., Piscataway, NJ (United States); Chang, Choong-Seock [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

2017-01-01

The XGC fusion gyrokinetic code combines state-of-the-art, portable computational and algorithmic technologies to enable complicated multiscale simulations of turbulence and transport dynamics in ITER edge plasma on the largest US open-science computer, the CRAY XK7 Titan, at its maximal heterogeneous capability, which have not been possible before due to a factor of over 10 shortage in the time-to-solution for less than 5 days of wall-clock time for one physics case. Frontier techniques such as nested OpenMP parallelism, adaptive parallel I/O, staging I/O and data reduction using dynamic and asynchronous applications interactions, dynamic repartitioning.

Determination of chemical solute transport parameters effecting radiostrontium interbed sediments

International Nuclear Information System (INIS)

Hemming, C.; Bunde, R.L.; Rosentreter, J.J.

1993-01-01

The extent to which radionuclides migrate in an aquifer system is a function of various physical, chemical, and biological processes. A measure of this migration rate is of primary concern when locating suitable storage sites for such species. Parameters including water-rock interactions, infiltration rates, chemical phase modification, and biochemical reactions all affect solute transport. While these different types of chemical reactions can influence solute transport in subsurface waters, distribution coefficients (Kd) can be send to effectively summarize the net chemical factors which dictate transport efficiency. This coefficient describes the partitioning of the solute between the solution and solid phase. Methodology used in determining and interpreting the distribution coefficient for radiostrontium in well characterized sediments will be presented

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.

Reactive solute transport in an asymmetrical fracture-rock matrix system

Science.gov (United States)

Zhou, Renjie; Zhan, Hongbin

2018-02-01

The understanding of reactive solute transport in a single fracture-rock matrix system is the foundation of studying transport behavior in the complex fractured porous media. When transport properties are asymmetrically distributed in the adjacent rock matrixes, reactive solute transport has to be considered as a coupled three-domain problem, which is more complex than the symmetric case with identical transport properties in the adjacent rock matrixes. This study deals with the transport problem in a single fracture-rock matrix system with asymmetrical distribution of transport properties in the rock matrixes. Mathematical models are developed for such a problem under the first-type and the third-type boundary conditions to analyze the spatio-temporal concentration and mass distribution in the fracture and rock matrix with the help of Laplace transform technique and de Hoog numerical inverse Laplace algorithm. The newly acquired solutions are then tested extensively against previous analytical and numerical solutions and are proven to be robust and accurate. Furthermore, a water flushing phase is imposed on the left boundary of system after a certain time. The diffusive mass exchange along the fracture/rock matrixes interfaces and the relative masses stored in each of three domains (fracture, upper rock matrix, and lower rock matrix) after the water flushing provide great insights of transport with asymmetric distribution of transport properties. This study has the following findings: 1) Asymmetric distribution of transport properties imposes greater controls on solute transport in the rock matrixes. However, transport in the fracture is mildly influenced. 2) The mass stored in the fracture responses quickly to water flushing, while the mass stored in the rock matrix is much less sensitive to the water flushing. 3) The diffusive mass exchange during the water flushing phase has similar patterns under symmetric and asymmetric cases. 4) The characteristic distance

Advances in the solution of three-dimensional nodal neutron transport equation

International Nuclear Information System (INIS)

Pazos, Ruben Panta; Hauser, Eliete Biasotto; Vilhena, Marco Tullio de

2003-01-01

In this paper we study the three-dimensional nodal discrete-ordinates approximations of neutron transport equation in a convex domain with piecewise smooth boundaries. We use the combined collocation method of the angular variables and nodal approach for the spatial variables. By nodal approach we mean the iterated transverse integration of the S N equations. This procedure leads to the set of one-dimensional averages angular fluxes in each spatial variable. The resulting system of equations is solved with the LTS N method, first applying the Laplace transform to the set of the nodal S N equations and then obtaining the solution by symbolic computation. We include the LTS N method by diagonalization to solve the nodal neutron transport equation and then we outline the convergence of these nodal-LTS N approximations with the help of a norm associated to the quadrature formula used to approximate the integral term of the neutron transport equation. We give numerical results obtained with an algebraic computer system (for N up to 8) and with a code for higher values of N. We compare our results for the geometry of a box with a source in a vertex and a leakage zone in the opposite with others techniques used in this problem. (author)

One-dimensional spatially dependent solute transport in semi ...

African Journals Online (AJOL)

Initially porous domain is considered solute free and the input source condition is ... parameters for description of solute transport in porous media. ... flow assuming uniform initial concentration with first and third type boundary conditions. Aral.

Mass transfer processes and field-scale transport of organic solutes

International Nuclear Information System (INIS)

Brusseau, M.L.

1990-01-01

The influence of mass transfer processes, such as sorption/desorption and mass transfer between immiscible liquids and water, on the transport of organic solutes is discussed. Rate-limited sorption of organic solutes caused by a diffusion-constrained mechanism is shown to be significant under laboratory conditions. The significance of the impact of nonequilibrium sorption on field-scale transport is scale dependent. The impact of organic liquids on mass transfer and transport of organic solutes depends upon the nature of the solute and the nature and form of the organic liquid. For example, while retardation of nonionic solutes is decreased in mixed-solvent systems, (i.e. systems comprised of water and a miscible organic liquid or an immiscible liquid present in concentrations below phase separation), the retardation of organic acids may, in some cases, increase with addition of a cosolvent. While the presence of an immiscible liquid existing as a mobile phase will reduce retention of organic solutes, the presence of residual saturation of an immiscible liquid can significantly increase retention. A model is presented that incorporates the effects of retention resulting from residual saturation, as well as nonequilibrium sorption, on the transport of organic solutes. (Author) (70 refs., 3 figs.)

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

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

End-Member Formulation of Solid Solutions and Reactive Transport

Energy Technology Data Exchange (ETDEWEB)

Lichtner, Peter C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-09-01

A model for incorporating solid solutions into reactive transport equations is presented based on an end-member representation. Reactive transport equations are solved directly for the composition and bulk concentration of the solid solution. Reactions of a solid solution with an aqueous solution are formulated in terms of an overall stoichiometric reaction corresponding to a time-varying composition and exchange reactions, equivalent to reaction end-members. Reaction rates are treated kinetically using a transition state rate law for the overall reaction and a pseudo-kinetic rate law for exchange reactions. The composition of the solid solution at the onset of precipitation is assumed to correspond to the least soluble composition, equivalent to the composition at equilibrium. The stoichiometric saturation determines if the solid solution is super-saturated with respect to the aqueous solution. The method is implemented for a simple prototype batch reactor using Mathematica for a binary solid solution. Finally, the sensitivity of the results on the kinetic rate constant for a binary solid solution is investigated for reaction of an initially stoichiometric solid phase with an undersaturated aqueous solution.

Reexamining ultrafiltration and solute transport in groundwater

Science.gov (United States)

Neuzil, C. E.; Person, Mark

2017-06-01

Geologic ultrafiltration—slowing of solutes with respect to flowing groundwater—poses a conundrum: it is consistently observed experimentally in clay-rich lithologies, but has been difficult to identify in subsurface data. Resolving this could be important for clarifying clay and shale transport properties at large scales as well as interpreting solute and isotope patterns for applications ranging from nuclear waste repository siting to understanding fluid transport in tectonically active environments. Simulations of one-dimensional NaCl transport across ultrafiltering clay membrane strata constrained by emerging data on geologic membrane properties showed different ultrafiltration effects than have often been envisioned. In relatively high-permeability advection-dominated regimes, salinity increases occurred mostly within membrane units while their effluent salinity initially fell and then rose to match solute delivery. In relatively low-permeability diffusion-dominated regimes, salinity peaked at the membrane upstream boundary and effluent salinity remained low. In both scenarios, however, only modest salinity changes (up to ˜3 g L-1) occurred because of self-limiting tendencies; membrane efficiency declines as salinity rises, and although sediment compaction increases efficiency, it is also decreases permeability and allows diffusive transport to dominate. It appears difficult for ultrafiltration to generate brines as speculated, but widespread and less extreme ultrafiltration effects in the subsurface could be unrecognized. Conditions needed for ultrafiltration are present in settings that include topographically-driven flow systems, confined aquifer systems subjected to injection or withdrawal, compacting basins, and accretionary complexes.

Neutral particle transport modeling with a reflective source in the plasma edge

International Nuclear Information System (INIS)

Valenti, M.E.

1992-01-01

A reflective source term is incorporated into the Boltzmann neutral particle transport equation to account for boundary reflection. This reflective neutral model is integrated over a uniform axis and subsequently discretized. The discrete two-dimensional equations are solved iteratively with a computer code. The results of the reflective neutral model computer code are benchmarked with the neutral particle transport code ONEDANT. The benchmark process demonstrates the validity of the reflective neutral model. The reflective neutral model is coupled to the Braams plasma particle and energy transport code. The coupled system generates self-consistent plasma edge transport solutions. These solutions, which utilize the transport equation are similar to solutions which utilize simple plasma edge neutral models when high recycle divertors are modeled. In the high recycle mode, the high electron density at the divertor plate reduces the mean free path of plate neutrals. Hence, the similarity in results. It is concluded that simple neutral models are sufficient for the analysis of high recycle power reactor edge plasmas. Low recycle edge plasmas were not examined

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.

dispersion equation parameters of solute transport in agricultural

African Journals Online (AJOL)

Jane

2011-08-31

Aug 31, 2011 ... fields for predicting soil quality property. Key words: ... The classical approach of modeling solute transport in porous media uses the deterministic ... concentration of the solution in the liquid phase, u0 is the mean velocity and ...

Solutions to Improve Person Transport System in the Pitesti City by Analyzing Public Transport vs. Private Transport

Science.gov (United States)

Mihaela, Istrate; Alexandru, Boroiu; Viorel, Nicolae; Ionel, Vieru

2017-10-01

One of the major problems facing the Pitesti city is the road congestion that occurs in the central area of the city during the peak hours. With all the measures taken in recent years - the widening of road arteries, increasing the number of parking spaces, the creation of overground road passages - it is obvious that the problem can only be solved by a new philosophy regarding urban mobility: it is no longer possible to continue through solutions to increase the accessibility of the central area of the city, but it is necessary, on the contrary, to promote a policy of discouraging the penetration of vehicles in the city center, coupled with a policy of improving the connection between urban public transport and county public transport. This new approach is also proposed in the new Urban Mobility Plan of Pitesti city, under development. The most convincing argument for the necessity of this new orientation in the Pitesti city mobility plan is based on the analysis of the current situation of passenger transport on the territory of Pitesti city: the analysis of “public transport versus private transport” reveals a very low occupancy rate for cars and the fact that the road surface required for a passenger (the dynamic area) is much higher in the case of private transport than in the case of public transport. Measurements of passenger flows and vehicle flows on the 6 penetration ways in the city have been made and the calculations clearly demonstrate the benefits of an urban public transport system connected by “transshipment buses” to be made at the edge of the city, to the county public transport system. In terms of inter-county transport, it will continue to be connected to the urban public transport system by existing bus Station, within the city: South Bus Station and North Bus Station. The usefulness of the paper is that it identifies the solutions for sustainable mobility in Pitesti city and proposes concrete solutions for the development of the

Electrolyte solution transport in electropolar nanotubes

International Nuclear Information System (INIS)

Zhao Jianbing; Culligan, Patricia J; Chen Xi; Qiao Yu; Zhou Qulan; Li Yibing; Tak, Moonho; Park, Taehyo

2010-01-01

Electrolyte transport in nanochannels plays an important role in a number of emerging areas. Using non-equilibrium molecular dynamics (NEMD) simulations, the fundamental transport behavior of an electrolyte/water solution in a confined model nanoenvironment is systematically investigated by varying the nanochannel dimension, solid phase, electrolyte phase, ion concentration and transport rate. It is found that the shear resistance encountered by the nanofluid strongly depends on these material/system parameters; furthermore, several effects are coupled. The mechanisms of the nanofluidic transport characteristics are explained by considering the unique molecular/ion structure formed inside the nanochannel. The lower shear resistance observed in some of the systems studies could be beneficial for nanoconductors, while the higher shear resistance (or higher effective viscosity) observed in other systems might enhance the performance of energy dissipation devices.

ipole: Semianalytic scheme for relativistic polarized radiative transport

Science.gov (United States)

Moscibrodzka, Monika; Gammie, Charles F.

2018-04-01

ipole is a ray-tracing code for covariant, polarized radiative transport particularly useful for modeling Event Horizon Telescope sources, though may also be used for other relativistic transport problems. The code extends the ibothros scheme for covariant, unpolarized transport using two representations of the polarized radiation field: in the coordinate frame, it parallel transports the coherency tensor, and in the frame of the plasma, it evolves the Stokes parameters under emission, absorption, and Faraday conversion. The transport step is as spacetime- and coordinate- independent as possible; the emission, absorption, and Faraday conversion step is implemented using an analytic solution to the polarized transport equation with constant coefficients. As a result, ipole is stable, efficient, and produces a physically reasonable solution even for a step with high optical depth and Faraday depth.

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.)

3-D neutron transport benchmarks

International Nuclear Information System (INIS)

Takeda, T.; Ikeda, H.

1991-03-01

A set of 3-D neutron transport benchmark problems proposed by the Osaka University to NEACRP in 1988 has been calculated by many participants and the corresponding results are summarized in this report. The results of K eff , control rod worth and region-averaged fluxes for the four proposed core models, calculated by using various 3-D transport codes are compared and discussed. The calculational methods used were: Monte Carlo, Discrete Ordinates (Sn), Spherical Harmonics (Pn), Nodal Transport and others. The solutions of the four core models are quite useful as benchmarks for checking the validity of 3-D neutron transport codes

Solution weighting for the SAND-II Monte Carlo code

International Nuclear Information System (INIS)

Oster, C.A.; McElroy, W.N.; Simons, R.L.; Lippincott, E.P.; Odette, G.R.

1976-01-01

Modifications to the SAND-II Error Analysis Monte Carlo code to include solution weighting based on input data uncertainties have been made and are discussed together with background information on the SAND-II algorithm. The new procedure permits input data having smaller uncertainties to have a greater influence on the solution spectrum than do the data having larger uncertainties. The results of an indepth study to find a practical procedure and the first results of its application to three important Interlaboratory LMFBR Reaction Rate (ILRR) program benchmark spectra (CFRMF, ΣΣ, and 235 U fission) are discussed

Analytic solution to verify code predictions of two-phase flow in a boiling water reactor core channel

International Nuclear Information System (INIS)

Chen, K.F.; Olson, C.A.

1983-01-01

One reliable method that can be used to verify the solution scheme of a computer code is to compare the code prediction to a simplified problem for which an analytic solution can be derived. An analytic solution for the axial pressure drop as a function of the flow was obtained for the simplified problem of homogeneous equilibrium two-phase flow in a vertical, heated channel with a cosine axial heat flux shape. This analytic solution was then used to verify the predictions of the CONDOR computer code, which is used to evaluate the thermal-hydraulic performance of boiling water reactors. The results show excellent agreement between the analytic solution and CONDOR prediction

Water flow and solute transport through fractured rock

International Nuclear Information System (INIS)

Bolt, J.E.; Bourke, P.J.; Pascoe, D.M.; Watkins, V.M.B.; Kingdon, R.D.

1990-09-01

In densely fractured slate at the Nirex research site in Cornwall, the positions, orientations and hydraulic conductivities of the 380 fractures intersecting a drill hole between 9 and 50 m depth have been individually measured. These data have been used: to determine the dimensions of statistically representative volumes of the network of fractures and to predict, using discrete flow path modelling and the NAPSAC code, the total flows into the fractures when large numbers are simultaneously pressurised along various lengths of the hole. Corresponding measurements, which validated the NAPSAC code to factor of two accuracy for the Cornish site, are reported. Possibilities accounting for this factor are noted for experimental investigation, and continuing, more extensive, inter hole flow and transport measurements are outlined. The application of this experimental and theoretical approach for calculating radionuclide transport in less densely fractured rock suitable for waste disposal is discussed. (Author)

Water flow and solute transport through fractured rock

International Nuclear Information System (INIS)

Bourke, P.J.; Kingdon, R.D.; Bolt, J.E.; Pascoe, D.M.; Watkins, V.M.B.

1991-01-01

In densely fractured slate at the Nirex research site in Cornwall, the positions, orientations and hydraulic conductivities of the 380 fractures intersecting a drill hole between 9 and 50 m depths have been individually measured. These data have been used: - to determine the dimensions of statistically representative volumes of the sheetwork of fractures; - to predict; using discrete flowpath modelling and the NAPSAC code; the total flows into the fractures when large numbers are simultaneously pressurised along various lengths of the hole; Corresponding measurements, which proved the modelling and validated the code to factor of two accuracy, are reported. Possibilities accounting for this factor are noted for experimental investigation, and continuing, more extensive inter-hole flow and transport measurements are outlined. The application of this experimental and theoretical approach for calculating radionuclide transport in less densely fractured rock suitable for waste disposal is discussed. 7 figs., 9 refs

A biomechanical triphasic approach to the transport of nondilute solutions in articular cartilage.

Science.gov (United States)

Abazari, Alireza; Elliott, Janet A W; Law, Garson K; McGann, Locksley E; Jomha, Nadr M

2009-12-16

Biomechanical models for biological tissues such as articular cartilage generally contain an ideal, dilute solution assumption. In this article, a biomechanical triphasic model of cartilage is described that includes nondilute treatment of concentrated solutions such as those applied in vitrification of biological tissues. The chemical potential equations of the triphasic model are modified and the transport equations are adjusted for the volume fraction and frictional coefficients of the solutes that are not negligible in such solutions. Four transport parameters, i.e., water permeability, solute permeability, diffusion coefficient of solute in solvent within the cartilage, and the cartilage stiffness modulus, are defined as four degrees of freedom for the model. Water and solute transport in cartilage were simulated using the model and predictions of average concentration increase and cartilage weight were fit to experimental data to obtain the values of the four transport parameters. As far as we know, this is the first study to formulate the solvent and solute transport equations of nondilute solutions in the cartilage matrix. It is shown that the values obtained for the transport parameters are within the ranges reported in the available literature, which confirms the proposed model approach.

Numerical solution of the transport equation describing the radon transport from subsurface soil to buildings

International Nuclear Information System (INIS)

Savovic, S.; Djordjevich, A.; Ristic, G.

2012-01-01

A theoretical evaluation of the properties and processes affecting the radon transport from subsurface soil into buildings is presented in this work. The solution of the relevant transport equation is obtained using the explicit finite difference method (EFDM). Results are compared with analytical steady-state solution reported in the literature. Good agreement is found. It is shown that EFDM is effective and accurate for solving the equation that describes radon diffusion, advection and decay during its transport from subsurface to buildings, which is especially important when arbitrary initial and boundary conditions are required. (authors)

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)

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.

Peritoneal Fluid Transport rather than Peritoneal Solute Transport Associates with Dialysis Vintage and Age of Peritoneal Dialysis Patients

Directory of Open Access Journals (Sweden)

Jacek Waniewski

2016-01-01

Full Text Available During peritoneal dialysis (PD, the peritoneal membrane undergoes ageing processes that affect its function. Here we analyzed associations of patient age and dialysis vintage with parameters of peritoneal transport of fluid and solutes, directly measured and estimated based on the pore model, for individual patients. Thirty-three patients (15 females; age 60 (21–87 years; median time on PD 19 (3–100 months underwent sequential peritoneal equilibration test. Dialysis vintage and patient age did not correlate. Estimation of parameters of the two-pore model of peritoneal transport was performed. The estimated fluid transport parameters, including hydraulic permeability (LpS, fraction of ultrasmall pores (αu, osmotic conductance for glucose (OCG, and peritoneal absorption, were generally independent of solute transport parameters (diffusive mass transport parameters. Fluid transport parameters correlated whereas transport parameters for small solutes and proteins did not correlate with dialysis vintage and patient age. Although LpS and OCG were lower for older patients and those with long dialysis vintage, αu was higher. Thus, fluid transport parameters—rather than solute transport parameters—are linked to dialysis vintage and patient age and should therefore be included when monitoring processes linked to ageing of the peritoneal membrane.

Peritoneal Fluid Transport rather than Peritoneal Solute Transport Associates with Dialysis Vintage and Age of Peritoneal Dialysis Patients

Science.gov (United States)

Waniewski, Jacek; Antosiewicz, Stefan; Baczynski, Daniel; Poleszczuk, Jan; Pietribiasi, Mauro; Lindholm, Bengt; Wankowicz, Zofia

2016-01-01

During peritoneal dialysis (PD), the peritoneal membrane undergoes ageing processes that affect its function. Here we analyzed associations of patient age and dialysis vintage with parameters of peritoneal transport of fluid and solutes, directly measured and estimated based on the pore model, for individual patients. Thirty-three patients (15 females; age 60 (21–87) years; median time on PD 19 (3–100) months) underwent sequential peritoneal equilibration test. Dialysis vintage and patient age did not correlate. Estimation of parameters of the two-pore model of peritoneal transport was performed. The estimated fluid transport parameters, including hydraulic permeability (LpS), fraction of ultrasmall pores (α u), osmotic conductance for glucose (OCG), and peritoneal absorption, were generally independent of solute transport parameters (diffusive mass transport parameters). Fluid transport parameters correlated whereas transport parameters for small solutes and proteins did not correlate with dialysis vintage and patient age. Although LpS and OCG were lower for older patients and those with long dialysis vintage, αu was higher. Thus, fluid transport parameters—rather than solute transport parameters—are linked to dialysis vintage and patient age and should therefore be included when monitoring processes linked to ageing of the peritoneal membrane. PMID:26989432

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)

REVISED STREAM CODE AND WASP5 BENCHMARK

International Nuclear Information System (INIS)

Chen, K

2005-01-01

STREAM is an emergency response code that predicts downstream pollutant concentrations for releases from the SRS area to the Savannah River. The STREAM code uses an algebraic equation to approximate the solution of the one dimensional advective transport differential equation. This approach generates spurious oscillations in the concentration profile when modeling long duration releases. To improve the capability of the STREAM code to model long-term releases, its calculation module was replaced by the WASP5 code. WASP5 is a US EPA water quality analysis program that simulates one-dimensional pollutant transport through surface water. Test cases were performed to compare the revised version of STREAM with the existing version. For continuous releases, results predicted by the revised STREAM code agree with physical expectations. The WASP5 code was benchmarked with the US EPA 1990 and 1991 dye tracer studies, in which the transport of the dye was measured from its release at the New Savannah Bluff Lock and Dam downstream to Savannah. The peak concentrations predicted by the WASP5 agreed with the measurements within ±20.0%. The transport times of the dye concentration peak predicted by the WASP5 agreed with the measurements within ±3.6%. These benchmarking results demonstrate that STREAM should be capable of accurately modeling releases from SRS outfalls

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)

New methods For Modeling Transport Of Water And Solutes In Soils

DEFF Research Database (Denmark)

Møldrup, Per

Recent models for water and solute transport in unsaturated soils have been mechanistically based but numerically very involved. This dissertation concerns the development of mechanistically-based but numerically simple models for calculating and analyzing transport of water and solutes in soil...

Modeling non-isothermal multiphase multi-species reactive chemical transport in geologic media

Energy Technology Data Exchange (ETDEWEB)

Tianfu Xu; Gerard, F.; Pruess, K.; Brimhall, G.

1997-07-01

The assessment of mineral deposits, the analysis of hydrothermal convection systems, the performance of radioactive, urban and industrial waste disposal, the study of groundwater pollution, and the understanding of natural groundwater quality patterns all require modeling tools that can consider both the transport of dissolved species as well as their interactions with solid (or other) phases in geologic media and engineered barriers. Here, a general multi-species reactive transport formulation has been developed, which is applicable to homogeneous and/or heterogeneous reactions that can proceed either subject to local equilibrium conditions or kinetic rates under non-isothermal multiphase flow conditions. Two numerical solution methods, the direct substitution approach (DSA) and sequential iteration approach (SIA) for solving the coupled complex subsurface thermo-physical-chemical processes, are described. An efficient sequential iteration approach, which solves transport of solutes and chemical reactions sequentially and iteratively, is proposed for the current reactive chemical transport computer code development. The coupled flow (water, vapor, air and heat) and solute transport equations are also solved sequentially. The existing multiphase flow code TOUGH2 and geochemical code EQ3/6 are used to implement this SIA. The flow chart of the coupled code TOUGH2-EQ3/6, required modifications of the existing codes and additional subroutines needed are presented.

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)

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)

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)

Approximate solutions of the two-dimensional integral transport equation by collision probability methods

International Nuclear Information System (INIS)

Sanchez, Richard

1977-01-01

A set of approximate solutions for the isotropic two-dimensional neutron transport problem has been developed using the Interface Current formalism. The method has been applied to regular lattices of rectangular cells containing a fuel pin, cladding and water, or homogenized structural material. The cells are divided into zones which are homogeneous. A zone-wise flux expansion is used to formulate a direct collision probability problem within a cell. The coupling of the cells is made by making extra assumptions on the currents entering and leaving the interfaces. Two codes have been written: the first uses a cylindrical cell model and one or three terms for the flux expansion; the second uses a two-dimensional flux representation and does a truly two-dimensional calculation inside each cell. In both codes one or three terms can be used to make a space-independent expansion of the angular fluxes entering and leaving each side of the cell. The accuracies and computing times achieved with the different approximations are illustrated by numerical studies on two benchmark pr

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

Transport-constrained extensions of collision and track length estimators for solutions of radiative transport problems

International Nuclear Information System (INIS)

Kong, Rong; Spanier, Jerome

2013-01-01

In this paper we develop novel extensions of collision and track length estimators for the complete space-angle solutions of radiative transport problems. We derive the relevant equations, prove that our new estimators are unbiased, and compare their performance with that of more conventional estimators. Such comparisons based on numerical solutions of simple one dimensional slab problems indicate the the potential superiority of the new estimators for a wide variety of more general transport problems

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

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

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)

Mathematical modeling of fluid and solute transport in peritoneal dialysis

OpenAIRE

Waniewski, Jacek

2001-01-01

Optimization of peritoneal dialysis schedule and dialysis fluid composition needs, among others, methods for quantitative assessment of fluid and solute transport. Furthermore, an integrative quantitative description of physiological processes within the tissue, which contribute to the net transfer of fluid and solutes, is necessary for interpretation of the data and for predictions of the outcome of possible intervention into the peritoneal transport system. The current pro...

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)

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.

Verification of a dust transport model against theoretical solutions in multidimensional advection diffusion problems

Energy Technology Data Exchange (ETDEWEB)

Xu, Z., E-mail: zhanjie.xu@kit.ed [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Travis, J.R. [Ingenieurbuero DuBois-Pitzer-Travis, 63071 Offenbach (Germany); Breitung, W.; Jordan, T. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany)

2010-12-15

Potentially explosive dust aerosol mobilization in the vacuum vessel is an important safety issue of the ITER facility, especially in scenarios of loss of vacuum accidents. Therefore dust mobilization modeling is ongoing in Research Center Karlsuhe. At first the aerosol particle model in the GASFLOW computer code is introduced briefly. To verify the particle model, a series of particle diffusion problems are simulated in one-, two- and three-dimensions. In each problem a particle source is initially exposed to an advective gas flow. Then a dust cloud is formed in the down stream. To obtain the theoretical solution about the particle concentration in the dust cloud, the governing diffusion partial differential equations with an additional advection term are solved by using Green's function method. Different spatial and temporal characters about the particle sources are also considered, e.g., instantaneous or continuous sources, line, or volume sources and so forth. The GASFLOW simulation results about the particle concentrations and the corresponding Green's function solutions are compared case by case. Very good agreements are found between the theoretical solutions and the GASGLOW simulations, when the drag force between the micron-sized particles and the conveying gas flow meets the Stokes' law about resistance. This situation is corresponding to a very small Reynolds number based on the particle diameter, with a negligible inertia effect of the particles. This verification work shows that the particle model of the GASFLOW code can reproduce numerically particle transport and diffusion in a good way.

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

Multicomponent mass transport model: a model for simulating migration of radionuclides in ground water

International Nuclear Information System (INIS)

Washburn, J.F.; Kaszeta, F.E.; Simmons, C.S.; Cole, C.R.

1980-07-01

This report presents the results of the development of a one-dimensional radionuclide transport code, MMT2D (Multicomponent Mass Transport), for the AEGIS Program. Multicomponent Mass Transport is a numerical solution technique that uses the discrete-parcel-random-wald (DPRW) method to directly simulate the migration of radionuclides. MMT1D accounts for: convection;dispersion; sorption-desorption; first-order radioactive decay; and n-membered radioactive decay chains. Comparisons between MMT1D and an analytical solution for a similar problem show that: MMT1D agrees very closely with the analytical solution; MMT1D has no cumulative numerical dispersion like that associated with solution techniques such as finite differences and finite elements; for current AEGIS applications, relatively few parcels are required to produce adequate results; and the power of MMT1D is the flexibility of the code in being able to handle complex problems for which analytical solution cannot be obtained. Multicomponent Mass Transport (MMT1D) codes were developed at Pacific Northwest Laboratory to predict the movement of radiocontaminants in the saturated and unsaturated sediments of the Hanford Site. All MMT models require ground-water flow patterns that have been previously generated by a hydrologic model. This report documents the computer code and operating procedures of a third generation of the MMT series: the MMT differs from previous versions by simulating the mass transport processes in systems with radionuclide decay chains. Although MMT is a one-dimensional code, the user is referred to the documentation of the theoretical and numerical procedures of the three-dimensional MMT-DPRW code for discussion of expediency, verification, and error-sensitivity analysis

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)

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

Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

Energy Technology Data Exchange (ETDEWEB)

Trambauer, K. [GRS, Garching (Germany)

1997-07-01

The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.

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)

Positive solution of a time and energy dependent neutron transport problem

International Nuclear Information System (INIS)

Pao, C.V.

1975-01-01

A constructive method is given for the determination of a solution and an existence--uniqueness theorem for some nonlinear time and energy dependent neutron transport problems, including the linear transport system. The geometry of the medium under consideration is allowed to be either bounded or unbounded which includes the geometry of a finite or infinite cylinder, a half-space and the whole space R/subm/ (m=1,2,center-dotcenter-dotcenter-dot). Our approach to the problem is by successive approximation which leads to various recursion formulas for the approximations in terms of explicit integrations. It is shown under some Lipschitz conditions on the nonlinear functions, which describe the process of neutrons absorption, fission, and scattering, that the sequence of approximations converges to a unique positive solution. Since these conditions are satisfied by the linear transport equation, all the results for the nonlinear system are valid for the linear transport problem. In the general nonlinear problem, the existence of both local and global solutions are discussed, and an iterative process for the construction of the solution is given

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)

Transport of organic solutes through amorphous teflon AF films.

Science.gov (United States)

Zhao, Hong; Zhang, Jie; Wu, Nianqiang; Zhang, Xu; Crowley, Katie; Weber, Stephen G

2005-11-02

Fluorous media have great potential for selective extraction (e.g., as applied to organic synthesis). Fluorous polymer films would have significant advantages in fluorous separations. Stable films of Teflon AF 2400 were cast from solution. Films appear defect-free (SEM; AFM). Rigid aromatic solutes are transported (from chloroform solution to chloroform receiving phase) in a size-dependent manner (log permeability is proportional to -0.0067 times critical volume). Benzene's permeability is about 2 orders of magnitude higher than in comparable gas-phase experiments. The films show selectivity for fluorinated solutes in comparison to the hydrogen-containing control. Transport rates are dependent on the solvent making up the source and receiving phases. The effect of solvent is, interestingly, not due to changes in partition ratio, but rather it is due to changes in the solute diffusion coefficient in the film. Solvents plasticize the films. A less volatile compound, -COOH-terminated poly(hexafluoropropylene oxide) (4), plasticizes the films (T(g) = -40 degrees C). Permeabilities are decreased in comparison to 4-free films apparently because of decreased diffusivity of solutes. The slope of dependence of log permeability on critical volume is not changed, however.

Analytical solution to the hybrid diffusion-transport equation

International Nuclear Information System (INIS)

Nanneh, M.M.; Williams, M.M.R.

1986-01-01

A special integral equation was derived in previous work using a hybrid diffusion-transport theory method for calculating the flux distribution in slab lattices. In this paper an analytical solution of this equation has been carried out on a finite reactor lattice. The analytical results of disadvantage factors are shown to be accurate in comparison with the numerical results and accurate transport theory calculations. (author)

Geological entropy and solute transport in heterogeneous porous media

Science.gov (United States)

Bianchi, Marco; Pedretti, Daniele

2017-06-01

We propose a novel approach to link solute transport behavior to the physical heterogeneity of the aquifer, which we fully characterize with two measurable parameters: the variance of the log K values (σY2), and a new indicator (HR) that integrates multiple properties of the K field into a global measure of spatial disorder or geological entropy. From the results of a detailed numerical experiment considering solute transport in K fields representing realistic distributions of hydrofacies in alluvial aquifers, we identify empirical relationship between the two parameters and the first three central moments of the distributions of arrival times of solute particles at a selected control plane. The analysis of experimental data indicates that the mean and the variance of the solutes arrival times tend to increase with spatial disorder (i.e., HR increasing), while highly skewed distributions are observed in more orderly structures (i.e., HR decreasing) or at higher σY2. We found that simple closed-form empirical expressions of the bivariate dependency of skewness on HR and σY2 can be used to predict the emergence of non-Fickian transport in K fields considering a range of structures and heterogeneity levels, some of which based on documented real aquifers. The accuracy of these predictions and in general the results from this study indicate that a description of the global variability and structure of the K field in terms of variance and geological entropy offers a valid and broadly applicable approach for the interpretation and prediction of transport in heterogeneous porous media.

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...

Transport synthetic acceleration scheme for multi-dimensional neutron transport problems

Energy Technology Data Exchange (ETDEWEB)

Modak, R S; Kumar, Vinod; Menon, S V.G. [Theoretical Physics Div., Bhabha Atomic Research Centre, Mumbai (India); Gupta, Anurag [Reactor Physics Design Div., Bhabha Atomic Research Centre, Mumbai (India)

2005-09-15

The numerical solution of linear multi-energy-group neutron transport equation is required in several analyses in nuclear reactor physics and allied areas. Computer codes based on the discrete ordinates (Sn) method are commonly used for this purpose. These codes solve external source problem and K-eigenvalue problem. The overall solution technique involves solution of source problem in each energy group as intermediate procedures. Such a single-group source problem is solved by the so-called Source Iteration (SI) method. As is well-known, the SI-method converges very slowly for optically thick and highly scattering regions, leading to large CPU times. Over last three decades, many schemes have been tried to accelerate the SI; the most prominent being the Diffusion Synthetic Acceleration (DSA) scheme. The DSA scheme, however, often fails and is also rather difficult to implement. In view of this, in 1997, Ramone and others have developed a new acceleration scheme called Transport Synthetic Acceleration (TSA) which is much more robust and easy to implement. This scheme has been recently incorporated in 2-D and 3-D in-house codes at BARC. This report presents studies on the utility of TSA scheme for fairly general test problems involving many energy groups and anisotropic scattering. The scheme is found to be useful for problems in Cartesian as well as Cylindrical geometry. (author)

Transport synthetic acceleration scheme for multi-dimensional neutron transport problems

International Nuclear Information System (INIS)

Modak, R.S.; Vinod Kumar; Menon, S.V.G.; Gupta, Anurag

2005-09-01

The numerical solution of linear multi-energy-group neutron transport equation is required in several analyses in nuclear reactor physics and allied areas. Computer codes based on the discrete ordinates (Sn) method are commonly used for this purpose. These codes solve external source problem and K-eigenvalue problem. The overall solution technique involves solution of source problem in each energy group as intermediate procedures. Such a single-group source problem is solved by the so-called Source Iteration (SI) method. As is well-known, the SI-method converges very slowly for optically thick and highly scattering regions, leading to large CPU times. Over last three decades, many schemes have been tried to accelerate the SI; the most prominent being the Diffusion Synthetic Acceleration (DSA) scheme. The DSA scheme, however, often fails and is also rather difficult to implement. In view of this, in 1997, Ramone and others have developed a new acceleration scheme called Transport Synthetic Acceleration (TSA) which is much more robust and easy to implement. This scheme has been recently incorporated in 2-D and 3-D in-house codes at BARC. This report presents studies on the utility of TSA scheme for fairly general test problems involving many energy groups and anisotropic scattering. The scheme is found to be useful for problems in Cartesian as well as Cylindrical geometry. (author)

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)

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.

Development of numerical solution techniques in the KIKO3D code

International Nuclear Information System (INIS)

Panka, Istvan; Kereszturi, Andras; Hegedus, Csaba

2005-01-01

The paper describes the numerical methods applied in KIKO3D three-dimensional reactor dynamics code and present a new, more effective method (Bi-CGSTAB) for accelerating the large sparse matrix equation solution. The convergence characteristics were investigated in a given macro time step of a Control Rod Ejection transient. The results obtained by the old GMRES and new Bi-CGSTAB methods are compared. It is concluded that the real relative errors of the solutions obtained by GMRES or Bi - CGSTAB algorithms are in fact closer together than the estimated relative errors. The KIKO3D-Bi-CGSTAB method converges safely and it is 7-12 % faster than the old KIKO3D-GMRES solution (Authors)

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

A parallel multi-domain solution methodology applied to nonlinear thermal transport problems in nuclear fuel pins

Energy Technology Data Exchange (ETDEWEB)

Philip, Bobby, E-mail: philipb@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Berrill, Mark A.; Allu, Srikanth; Hamilton, Steven P.; Sampath, Rahul S.; Clarno, Kevin T. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Dilts, Gary A. [Los Alamos National Laboratory, PO Box 1663, Los Alamos, NM 87545 (United States)

2015-04-01

This paper describes an efficient and nonlinearly consistent parallel solution methodology for solving coupled nonlinear thermal transport problems that occur in nuclear reactor applications over hundreds of individual 3D physical subdomains. Efficiency is obtained by leveraging knowledge of the physical domains, the physics on individual domains, and the couplings between them for preconditioning within a Jacobian Free Newton Krylov method. Details of the computational infrastructure that enabled this work, namely the open source Advanced Multi-Physics (AMP) package developed by the authors is described. Details of verification and validation experiments, and parallel performance analysis in weak and strong scaling studies demonstrating the achieved efficiency of the algorithm are presented. Furthermore, numerical experiments demonstrate that the preconditioner developed is independent of the number of fuel subdomains in a fuel rod, which is particularly important when simulating different types of fuel rods. Finally, we demonstrate the power of the coupling methodology by considering problems with couplings between surface and volume physics and coupling of nonlinear thermal transport in fuel rods to an external radiation transport code.

Assessing numerical methods used in nuclear aerosol transport models

International Nuclear Information System (INIS)

McDonald, B.H.

1987-01-01

Several computer codes are in use for predicting the behaviour of nuclear aerosols released into containment during postulated accidents in water-cooled reactors. Each of these codes uses numerical methods to discretize and integrate the equations that govern the aerosol transport process. Computers perform only algebraic operations and generate only numbers. It is in the numerical methods that sense can be made of these numbers and where they can be related to the actual solution of the equations. In this report, the numerical methods most commonly used in the aerosol transport codes are examined as special cases of a general solution procedure, the Method of Weighted Residuals. It would appear that the numerical methods used in the codes are all capable of producing reasonable answers to the mathematical problem when used with skill and care. 27 refs

A Finite-Difference Solution of Solute Transport through a Membrane Bioreactor

Directory of Open Access Journals (Sweden)

B. Godongwana

2015-01-01

Full Text Available The current paper presents a theoretical analysis of the transport of solutes through a fixed-film membrane bioreactor (MBR, immobilised with an active biocatalyst. The dimensionless convection-diffusion equation with variable coefficients was solved analytically and numerically for concentration profiles of the solutes through the MBR. The analytical solution makes use of regular perturbation and accounts for radial convective flow as well as axial diffusion of the substrate species. The Michaelis-Menten (or Monod rate equation was assumed for the sink term, and the perturbation was extended up to second-order. In the analytical solution only the first-order limit of the Michaelis-Menten equation was considered; hence the linearized equation was solved. In the numerical solution, however, this restriction was lifted. The solution of the nonlinear, elliptic, partial differential equation was based on an implicit finite-difference method (FDM. An upwind scheme was employed for numerical stability. The resulting algebraic equations were solved simultaneously using the multivariate Newton-Raphson iteration method. The solution allows for the evaluation of the effect on the concentration profiles of (i the radial and axial convective velocity, (ii the convective mass transfer rates, (iii the reaction rates, (iv the fraction retentate, and (v the aspect ratio.

Ground-water solute transport modeling using a three-dimensional scaled model

International Nuclear Information System (INIS)

Crider, S.S.

1987-01-01

Scaled models are used extensively in current hydraulic research on sediment transport and solute dispersion in free surface flows (rivers, estuaries), but are neglected in current ground-water model research. Thus, an investigation was conducted to test the efficacy of a three-dimensional scaled model of solute transport in ground water. No previous results from such a model have been reported. Experiments performed on uniform scaled models indicated that some historical problems (e.g., construction and scaling difficulties; disproportionate capillary rise in model) were partly overcome by using simple model materials (sand, cement and water), by restricting model application to selective classes of problems, and by physically controlling the effect of the model capillary zone. Results from these tests were compared with mathematical models. Model scaling laws were derived for ground-water solute transport and used to build a three-dimensional scaled model of a ground-water tritium plume in a prototype aquifer on the Savannah River Plant near Aiken, South Carolina. Model results compared favorably with field data and with a numerical model. Scaled models are recommended as a useful additional tool for prediction of ground-water solute transport

Predictability of solute transport in diffusion-controlled hydrogeologic regimes

International Nuclear Information System (INIS)

Gillham, R.W.; Cherry, J.A.

1983-01-01

Hydrogeologic regimes that are favourable for the subsurface management of low-level radioactive wastes must have transport properties that will limit the migration velocity of contaminants to some acceptably low value. Of equal importance, for the purpose of impact assessment and licensing, is the need to be able to predict, with a reasonable degree of certainty and over long time periods, what the migration velocity of the various contaminants of interest will be. This paper presents arguments to show that in addition to having favourable velocity characteristics, transport in saturated, diffusion-controlled hydrogeologic regimes is considerably more predictable than in the most common alternatives. The classical transport models for unsaturated, saturated-advection-controlled and saturated-diffusion-controlled environments are compared, with particular consideration being given to the difficulties associated with the characterization of the respective transport parameters. Results are presented which show that the diffusion of non-reactive solutes and solutes that react according to a constant partitioning ratio (K/sub d/) are highly predictable under laboratory conditions and that the diffusion coefficients for the reactive solutes can be determined with a reasonable degree of accuracy from independent measurements of bulk density, porosity, distribution coefficient and tortuosity. Field evidence is presented which shows that the distribution of environmental isotopes and chloride in thick clayey deposits is consistent with a diffusion-type transport process in these media. These results are particularly important in that they not only demonstrate the occurrence of diffusion-controlled hydrogeologic regimes, but they also demonstrate the predictability of the migration characteristics over very long time periods

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.

Transport of Liquid Phase Organic Solutes in Liquid Crystalline Membranes

OpenAIRE

Han, Sangil

2010-01-01

Porous cellulose nitrate membranes were impregnated with 8CB and PCH5 LCs (liquid crystals) and separations of solutes dissolved in aqueous phases were performed while monitoring solute concentration via UV-VIS spectrometry. The diffusing organic solutes, which consist of one aromatic ring and various functional groups, were selected to exclude molecular size effects on the diffusion and sorption. We studied the effects on solute transport of solute intra-molecular hydrogen bonding and so...

System for sampling active solutions in transport container; Systeme de prelevements de solutions actives sur les recipients de transport

Energy Technology Data Exchange (ETDEWEB)

Fradin, J.

1958-12-03

This report presents a system aimed at sampling active solution from a specific transport container (SCRGR model) while transferring this solution with a maximum safety. The sampling principle is described (a flexible tube connected to the receiving container, with a needle at the other end which goes through a rubber membrane and enters a plunger tube). Its benefits are outlined (operator protection, reduction of contamination risk; only the rubber membrane is removed and replaced). Some manufacturing details are described concerning the membrane and the cover.

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

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

International Nuclear Information System (INIS)

Royston, Katherine K.; Haghighat, Alireza

2011-01-01

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

An Optimal Linear Coding for Index Coding Problem

OpenAIRE

Pezeshkpour, Pouya

2015-01-01

An optimal linear coding solution for index coding problem is established. Instead of network coding approach by focus on graph theoric and algebraic methods a linear coding program for solving both unicast and groupcast index coding problem is presented. The coding is proved to be the optimal solution from the linear perspective and can be easily utilize for any number of messages. The importance of this work is lying mostly on the usage of the presented coding in the groupcast index coding ...

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

Solute transport with periodic input point source in one-dimensional ...

African Journals Online (AJOL)

JOY

groundwater flow velocity is considered proportional to multiple of temporal function and ζ th ... One-dimensional solute transport through porous media with or without .... solute free. ... the periodic concentration at source of the boundary i.e.,. 0.

Development of solute transport models in YMPYRÄ framework to simulate solute migration in military shooting and training areas

Science.gov (United States)

Warsta, L.; Karvonen, T.

2017-12-01

There are currently 25 shooting and training areas in Finland managed by The Finnish Defence Forces (FDF), where military activities can cause contamination of open waters and groundwater reservoirs. In the YMPYRÄ project, a computer software framework is being developed that combines existing open environmental data and proprietary information collected by FDF with computational models to investigate current and prevent future environmental problems. A data centric philosophy is followed in the development of the system, i.e. the models are updated and extended to handle available data from different areas. The results generated by the models are summarized as easily understandable flow and risk maps that can be opened in GIS programs and used in environmental assessments by experts. Substances investigated with the system include explosives and metals such as lead, and both surface and groundwater dominated areas can be simulated. The YMPYRÄ framework is composed of a three dimensional soil and groundwater flow model, several solute transport models and an uncertainty assessment system. Solute transport models in the framework include particle based, stream tube and finite volume based approaches. The models can be used to simulate solute dissolution from source area, transport in the unsaturated layers to groundwater and finally migration in groundwater to water extraction wells and springs. The models can be used to simulate advection, dispersion, equilibrium adsorption on soil particles, solubility and dissolution from solute phase and dendritic solute decay chains. Correct numerical solutions were confirmed by comparing results to analytical 1D and 2D solutions and by comparing the numerical solutions to each other. The particle based and stream tube type solute transport models were useful as they could complement the traditional finite volume based approach which in certain circumstances produced numerical dispersion due to piecewise solution of the

Stochastic dynamics modeling solute transport in porous media modeling solute transport in porous media

CERN Document Server

Kulasiri, Don

2002-01-01

Most of the natural and biological phenomena such as solute transport in porous media exhibit variability which can not be modeled by using deterministic approaches. There is evidence in natural phenomena to suggest that some of the observations can not be explained by using the models which give deterministic solutions. Stochastic processes have a rich repository of objects which can be used to express the randomness inherent in the system and the evolution of the system over time. The attractiveness of the stochastic differential equations (SDE) and stochastic partial differential equations (SPDE) come from the fact that we can integrate the variability of the system along with the scientific knowledge pertaining to the system. One of the aims of this book is to explaim some useufl concepts in stochastic dynamics so that the scientists and engineers with a background in undergraduate differential calculus could appreciate the applicability and appropriateness of these developments in mathematics. The ideas ...

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

Effects of Temperature on Solute Transport Parameters in Differently-Textured Soils at Saturated Condition

Science.gov (United States)

Hamamoto, S.; Arihara, M.; Kawamoto, K.; Nishimura, T.; Komatsu, T.; Moldrup, P.

2014-12-01

Subsurface warming driven by global warming, urban heat islands, and increasing use of shallow geothermal heating and cooling systems such as the ground source heat pump, potentially causes changes in subsurface mass transport. Therefore, understanding temperature dependency of the solute transport characteristics is essential to accurately assess environmental risks due to increased subsurface temperature. In this study, one-dimensional solute transport experiments were conducted in soil columns under temperature control to investigate effects of temperature on solute transport parameters, such as solute dispersion and diffusion coefficients, hydraulic conductivity, and retardation factor. Toyoura sand, Kaolin clay, and intact loamy soils were used in the experiments. Intact loamy soils were taken during a deep well boring at the Arakawa Lowland in Saitama Prefecture, Japan. In the transport experiments, the core sample with 5-cm diameter and 4-cm height was first isotropically consolidated, whereafter 0.01M KCl solution was injected to the sample from the bottom. The concentrations of K+ and Cl- in the effluents were analyzed by an ion chromatograph to obtain solute breakthrough curves. The solute transport parameters were calculated from the breakthrough curves. The experiments were conducted under different temperature conditions (15, 25, and 40 oC). The retardation factor for the intact loamy soils decreased with increasing temperature, while water permeability increased due to reduced viscosity of water at higher temperature. Opposite, the effect of temperature on solute dispersivity for the intact loamy soils was insignificant. The effects of soil texture on the temperature dependency of the solute transport characteristics will be further investigated from comparison of results from differently-textured samples.

Sub-step methodology for coupled Monte Carlo depletion and thermal hydraulic codes

International Nuclear Information System (INIS)

Kotlyar, D.; Shwageraus, E.

2016-01-01

Highlights: • Discretization of time in coupled MC codes determines the results’ accuracy. • The error is due to lack of information regarding the time-dependent reaction rates. • The proposed sub-step method considerably reduces the time discretization error. • No additional MC transport solutions are required within the time step. • The reaction rates are varied as functions of nuclide densities and TH conditions. - Abstract: The governing procedure in coupled Monte Carlo (MC) codes relies on discretization of the simulation time into time steps. Typically, the MC transport solution at discrete points will generate reaction rates, which in most codes are assumed to be constant within the time step. This assumption can trigger numerical instabilities or result in a loss of accuracy, which, in turn, would require reducing the time steps size. This paper focuses on reducing the time discretization error without requiring additional MC transport solutions and hence with no major computational overhead. The sub-step method presented here accounts for the reaction rate variation due to the variation in nuclide densities and thermal hydraulic (TH) conditions. This is achieved by performing additional depletion and TH calculations within the analyzed time step. The method was implemented in BGCore code and subsequently used to analyze a series of test cases. The results indicate that computational speedup of up to a factor of 10 may be achieved over the existing coupling schemes.

New diffusion-like solutions of one-speed transport equations in spherical geometry

International Nuclear Information System (INIS)

Sahni, D.C.

1988-01-01

Stationary, one-speed, spherically symmetric transport equations are considered in a conservative medium. Closed-form expressions are obtained for the angular flux ψ(r, μ) that yield a total flux varying as 1/r by using Sonine transforms. Properties of this solution are studied and it is shown that the solution can not be identified as a diffusion mode solution of the transport equation. Limitations of the Sonine transform technique are noted. (author)

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)

CFD analysis for the hydrogen transport in the primary contention of a BWR using the codes OpenFOAM and Gas-Flow; Analisis CFD para el transporte de hidrogeno en la contencion primaria de un reactor BWR usando los codigos OpenFOAM y GasFlow

Energy Technology Data Exchange (ETDEWEB)

Jimenez P, D. A.

2014-07-01

The accidents in Unit 2 of the Three Mile Island Nuclear Power Plant (NPP) in the United States (March 28{sup th}, 1979), the one in Unit 4 of the NPP Chernobyl in Ukraine (April 26{sup th}, 1986) and the explosions in some units of Fukushima NPP in Japan (March 11{sup th}, 2011) boosted the investigations on severe accidents with core damage and, in particular, the threat to the ultimate barrier by an eventual explosion from uncontrolled Hydrogen combustion within the containment was considered of particular relevance. Research programs for analyzing Hydrogen behavior and control during this kind of accidents were early initiated by research and regulatory bodies. Assessment on Hydrogen behavior once it has been postulated to be released on the containment system can be divided into two phases, in the first one, transport and the concentrations of the gas mixtures and steam in each volume or area comprised between the structures of the containment are calculated, in the second one, the propagation of the detonation of the Hydrogen is calculated if there are the conditions to occur. Currently, there are computer programs that can be used in one, or both stages of computation, and they are based on one of the two solution methods in current use, one of them are integrated codes (e.g. MELCOR), which consists in assuming the containment as a network composed of hydraulic tanks or nodes on which the balance equations of mass and energy have to be solved, the network is connected by ducts or connections where the momentum balance equation arise. This methodology relies on the use of semi-empirical relationships and the criteria used to define a geometric pattern, are subjective. The second method, which is having relevance due to the large computing power of modern computers, is the numerical solution of the three-dimensional Navier-Stokes equations in complex geometries. This method of solution is known as Computational Fluid Dynamics (CFD), and offers the advantage of

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)

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)

Fluid flow and convective transport of solutes within the intervertebral disc

NARCIS (Netherlands)

Ferguson, S.J.; Ito, K.; Nolte, L.P.

2004-01-01

Previous experimental and analytical studies of solute transport in the intervertebral disc have demonstrated that for small molecules diffusive transport alone fulfils the nutritional needs of disc cells. It has been often suggested that fluid flow into and within the disc may enhance the transport

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)

Stable solutions of nonlocal electron heat transport equations

International Nuclear Information System (INIS)

Prasad, M.K.; Kershaw, D.S.

1991-01-01

Electron heat transport equations with a nonlocal heat flux are in general ill-posed and intrinsically unstable, as proved by the present authors [Phys. Fluids B 1, 2430 (1989)]. A straightforward numerical solution of these equations will therefore lead to absurd results. It is shown here that by imposing a minimal set of constraints on the problem it is possible to arrive at a globally stable, consistent, and energy conserving numerical solution

Modeling variably saturated multispecies reactive groundwater solute transport with MODFLOW-UZF and RT3D

Science.gov (United States)

Bailey, Ryan T.; Morway, Eric D.; Niswonger, Richard G.; Gates, Timothy K.

2013-01-01

A numerical model was developed that is capable of simulating multispecies reactive solute transport in variably saturated porous media. This model consists of a modified version of the reactive transport model RT3D (Reactive Transport in 3 Dimensions) that is linked to the Unsaturated-Zone Flow (UZF1) package and MODFLOW. Referred to as UZF-RT3D, the model is tested against published analytical benchmarks as well as other published contaminant transport models, including HYDRUS-1D, VS2DT, and SUTRA, and the coupled flow and transport modeling system of CATHY and TRAN3D. Comparisons in one-dimensional, two-dimensional, and three-dimensional variably saturated systems are explored. While several test cases are included to verify the correct implementation of variably saturated transport in UZF-RT3D, other cases are included to demonstrate the usefulness of the code in terms of model run-time and handling the reaction kinetics of multiple interacting species in variably saturated subsurface systems. As UZF1 relies on a kinematic-wave approximation for unsaturated flow that neglects the diffusive terms in Richards equation, UZF-RT3D can be used for large-scale aquifer systems for which the UZF1 formulation is reasonable, that is, capillary-pressure gradients can be neglected and soil parameters can be treated as homogeneous. Decreased model run-time and the ability to include site-specific chemical species and chemical reactions make UZF-RT3D an attractive model for efficient simulation of multispecies reactive transport in variably saturated large-scale subsurface systems.

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)

Recent developments in KTF. Code optimization and improved numerics

International Nuclear Information System (INIS)

Jimenez, Javier; Avramova, Maria; Sanchez, Victor Hugo; Ivanov, Kostadin

2012-01-01

The rapid increase of computer power in the last decade facilitated the development of high fidelity simulations in nuclear engineering allowing a more realistic and accurate optimization as well as safety assessment of reactor cores and power plants compared to the legacy codes. Thermal hydraulic subchannel codes together with time dependent neutron transport codes are the options of choice for an accurate prediction of local safety parameters. Moreover, fast running codes with the best physical models are needed for high fidelity coupled thermal hydraulic / neutron kinetic solutions. Hence at KIT, different subchannel codes such as SUBCHANFLOW and KTF are being improved, validated and coupled with different neutron kinetics solutions. KTF is a subchannel code developed for best-estimate analysis of both Pressurized Water Reactor (PWR) and BWR. It is based on the Pennsylvania State University (PSU) version of COBRA-TF (Coolant Boling in Rod Arrays Two Fluids) named CTF. In this paper, the investigations devoted to the enhancement of the code numeric and informatics structure are presented and discussed. By some examples the gain on code speed-up will be demonstrated and finally an outlook of further activities concentrated on the code improvements will be given. (orig.)

Recent developments in KTF. Code optimization and improved numerics

Energy Technology Data Exchange (ETDEWEB)

Jimenez, Javier; Avramova, Maria; Sanchez, Victor Hugo; Ivanov, Kostadin [Karlsruhe Institute of Technology (KIT) (Germany). Inst. for Neutron Physics and Reactor Technology (INR)

2012-11-01

The rapid increase of computer power in the last decade facilitated the development of high fidelity simulations in nuclear engineering allowing a more realistic and accurate optimization as well as safety assessment of reactor cores and power plants compared to the legacy codes. Thermal hydraulic subchannel codes together with time dependent neutron transport codes are the options of choice for an accurate prediction of local safety parameters. Moreover, fast running codes with the best physical models are needed for high fidelity coupled thermal hydraulic / neutron kinetic solutions. Hence at KIT, different subchannel codes such as SUBCHANFLOW and KTF are being improved, validated and coupled with different neutron kinetics solutions. KTF is a subchannel code developed for best-estimate analysis of both Pressurized Water Reactor (PWR) and BWR. It is based on the Pennsylvania State University (PSU) version of COBRA-TF (Coolant Boling in Rod Arrays Two Fluids) named CTF. In this paper, the investigations devoted to the enhancement of the code numeric and informatics structure are presented and discussed. By some examples the gain on code speed-up will be demonstrated and finally an outlook of further activities concentrated on the code improvements will be given. (orig.)

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...

Development of a coarse mesh code for the solution of two group static diffusion problems

International Nuclear Information System (INIS)

Barros, R.C. de.

1985-01-01

This new coarse mesh code designed for the solution of 2 and 3 dimensional static diffusion problems, is based on an alternating direction method which consists in the solution of one dimensional problem along each coordinate direction with leakage terms for the remaining directions estimated from previous interactions. Four versions of this code have been developed: AD21 - 2D - 1/4, AD21 - 2D - 4/4, AD21 - 3D - 1/4 and AD21 - 3D - 4/4; these versions have been designed for 2 and 3 dimensional problems with or without 1/4 symmetry. (Author) [pt

Recent improvements and new features in the Westinghouse lattice physics codes

International Nuclear Information System (INIS)

Huria, H.C.; Buechel, R.J.

1995-01-01

Westinghouse has been using the ANC three-dimensional, two-energy-group nodal model for nuclear analysis and fuel management calculations for standard pressurized water reactor (PWR) reload design analysis since 1988. The cross sections are obtained from PHOENIX-P, a modified version of the PHOENIX lattice physics code for all square-assembly PWR cores. The PHOENIX-H code was developed for modeling both the VVER-1000 and VVER-440 fuel lattice configurations. The PHOENIX-H code has evolved from PHOENIX-P, the primary difference being in the neutronic solution modules. The PHOENIX-P code determines the assembly flux distribution using integral transport theory-based pin-cell nodal coupling followed by two-dimensional discrete ordinates solution in x-y geometry. The PHOENIX-H code uses the two-dimensional heterogeneous response method. The other infrastructure is identical in both the codes, and they share the same 42-group cross-section library

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.

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

Stochastic analysis of transport of conservative solutes in caisson experiments

International Nuclear Information System (INIS)

Dagan, G.

1995-01-01

The Los Alamos National Laboratory has conducted in the past a series of experiments of transport of conservative and reactive solutes. The experimental setup and the experimental results are presented in a series of reports. The main aim of the experiments was to validate models of transport of solutes in unsaturated flow at the caisson intermediate scale, which is much larger than the one pertaining to laboratory columns. First attempts to analyze the experimental results were by one-dimensional convective-dispersion models. These models could not explain the observed solute breakthrough curves and particularly the large solute dispersion in the caisson effluent Since there were some question marks about the uniformity of water distribution at the caisson top, the transport experiments were repeated under conditions of saturated flow. In these experiments constant heads were applied at the top and the bottom of the caisson and the number of concentration monitoring stations was quadrupled. The analysis of the measurements by the same one-dimensional model indicated clearly that the fitted dispersivity is much larger than the pore-sole dispersivity and that it grows with the distance in an approximately linear fashion. This led to the conclusion, raised before, that transport in the caisson is dominated by heterogeneity effects, i.e. by spatial variability of the material Such effects cannot be captured by traditional one-dimensional models. In order to account for the effect of heterogeneity, the saturated flow experiments have been analyzed by using stochastic transport modeling. The apparent linear growth of dispersivity with distance suggested that the system behaves like a stratified one. Consequently, the model of Dagan and Bresier has been adopted in order to interpret concentration measurements. In this simple model the caisson is viewed as a bundle of columns of different permeabilities, which are characterized by a p.d.f. (probability denasity function)

Code-Based Cryptography: New Security Solutions Against a Quantum Adversary

OpenAIRE

Sendrier , Nicolas; Tillich , Jean-Pierre

2016-01-01

International audience; Cryptography is one of the key tools for providing security in our quickly evolving technological society. An adversary with the ability to use a quantum computer would defeat most of the cryptographic solutions that are deployed today to secure our communications. We do not know when quantum computing will become available, but nevertheless, the cryptographic research community must get ready for it now. Code-based cryptography is among the few cryptographic technique...

Tests of the TRAC code against known analytical solutions for stratified flow

International Nuclear Information System (INIS)

Black, P.S.; Leslie, D.C.; Hewitt, G.F.

1987-01-01

The area averaged equations for gas-liquid flow are briefly summarized and related, for the specific case of stratified flow, to the shallow water equations commonly used in hydraulics. These equations are then compared to the equations used in TRAC-PF/MOD1 and are shown to differ in their treatment of the gravity head terms. A modification of the TRAC code is therefore necessary to bring it into line with established shallow water theory. The corrected form of the code was compared with a number of specific cases, each of which throws further light on the code behavior. The following areas are discussed in the paper: (1) the dam break problem; (2) Kelvin-Helmholtz instability; (3) counter-current flow; and (4) slug flow. It is concluded that detailed comparisons of the code with known analytic solutions and with a number of the more complex phenomenological experiments can give useful insights into its behavior

Colloid transport in porous media: impact of hyper-saline solutions.

Science.gov (United States)

Magal, Einat; Weisbrod, Noam; Yechieli, Yoseph; Walker, Sharon L; Yakirevich, Alexander

2011-05-01

The transport of colloids suspended in natural saline solutions with a wide range of ionic strengths, up to that of Dead Sea brines (10(0.9) M) was explored. Migration of microspheres through saturated sand columns of different sizes was studied in laboratory experiments and simulated with mathematical models. Colloid transport was found to be related to the solution salinity as expected. The relative concentration of colloids at the columns outlet decreased (after 2-3 pore volumes) as the solution ionic strength increased until a critical value was reached (ionic strength > 10(-1.8) M) and then remained constant above this level of salinity. The colloids were found to be mobile even in the extremely saline brines of the Dead Sea. At such high ionic strength no energetic barrier to colloid attachment was presumed to exist and colloid deposition was expected to be a favorable process. However, even at these salinity levels, colloid attachment was not complete and the transport of ∼ 30% of the colloids through the 30-cm long columns was detected. To further explore the deposition of colloids on sand surfaces in Dead Sea brines, transport was studied using 7-cm long columns through which hundreds of pore volumes were introduced. The resulting breakthrough curves exhibited a bimodal shape whereby the relative concentration (C/C(0)) of colloids at the outlet rose to a value of 0.8, and it remained relatively constant (for the ∼ 18 pore volumes during which the colloid suspension was flushed through the column) and then the relative concentration increased to a value of one. The bimodal nature of the breakthrough suggests different rates of colloid attachment. Colloid transport processes were successfully modeled using the limited entrapment model, which assumes that the colloid attachment rate is dependent on the concentration of the attached colloids. Application of this model provided confirmation of the colloid aggregation and their accelerated attachment during

Used Fuel Logistics: Decades of Experience with transportation and Interim storage solutions

Energy Technology Data Exchange (ETDEWEB)

Orban, G.; Shelton, C.

2015-07-01

Used fuel inventories are growing worldwide. While some countries have opted for a closed cycle with recycling, numerous countries must expand their interim storage solutions as implementation of permanent repositories is taking more time than foreseen. In both cases transportation capabilities will have to be developed. AREVA TN has an unparalleled expertise with transportation of used fuel. For more than 50 years AREVA TN has safely shipped more than 7,000 used fuel transport casks. The transportation model that was initially developed in the 1970s has been adapted and enhanced over the years to meet more restrictive regulatory requirements and evolving customer needs, and to address public concerns. The numerous “lessons learned” have offered data and guidance that have allowed for also efficient and consistent improvement over the decades. AREVA TN has also an extensive experience with interim dry storage solutions in many countries on-site but also is working with partners to developed consolidated interim storage facility. Both expertise with storage and transportation contribute to safe, secure and smooth continuity of the operations. This paper will describe decades of experience with a very successful transportation program as well as interim storage solutions. (Author)

Optimal solution of full fuzzy transportation problems using total integral ranking

Science.gov (United States)

Sam’an, M.; Farikhin; Hariyanto, S.; Surarso, B.

2018-03-01

Full fuzzy transportation problem (FFTP) is a transportation problem where transport costs, demand, supply and decision variables are expressed in form of fuzzy numbers. To solve fuzzy transportation problem, fuzzy number parameter must be converted to a crisp number called defuzzyfication method. In this new total integral ranking method with fuzzy numbers from conversion of trapezoidal fuzzy numbers to hexagonal fuzzy numbers obtained result of consistency defuzzyfication on symmetrical fuzzy hexagonal and non symmetrical type 2 numbers with fuzzy triangular numbers. To calculate of optimum solution FTP used fuzzy transportation algorithm with least cost method. From this optimum solution, it is found that use of fuzzy number form total integral ranking with index of optimism gives different optimum value. In addition, total integral ranking value using hexagonal fuzzy numbers has an optimal value better than the total integral ranking value using trapezoidal fuzzy numbers.

Intercomparison of the finite difference and nodal discrete ordinates and surface flux transport methods for a LWR pool-reactor benchmark problem in X-Y geometry

International Nuclear Information System (INIS)

O'Dell, R.D.; Stepanek, J.; Wagner, M.R.

1983-01-01

The aim of the present work is to compare and discuss the three of the most advanced two dimensional transport methods, the finite difference and nodal discrete ordinates and surface flux method, incorporated into the transport codes TWODANT, TWOTRAN-NODAL, MULTIMEDIUM and SURCU. For intercomparison the eigenvalue and the neutron flux distribution are calculated using these codes in the LWR pool reactor benchmark problem. Additionally the results are compared with some results obtained by French collision probability transport codes MARSYAS and TRIDENT. Because the transport solution of this benchmark problem is close to its diffusion solution some results obtained by the finite element diffusion code FINELM and the finite difference diffusion code DIFF-2D are included

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

The next generation in optical transport semiconductors: IC solutions at the system level

Science.gov (United States)

Gomatam, Badri N.

2005-02-01

In this tutorial overview, we survey some of the challenging problems facing Optical Transport and their solutions using new semiconductor-based technologies. Advances in 0.13um CMOS, SiGe/HBT and InP/HBT IC process technologies and mixed-signal design strategies are the fundamental breakthroughs that have made these solutions possible. In combination with innovative packaging and transponder/transceiver architectures IC approaches have clearly demonstrated enhanced optical link budgets with simultaneously lower (perhaps the lowest to date) cost and manufacturability tradeoffs. This paper will describe: *Electronic Dispersion Compensation broadly viewed as the overcoming of dispersion based limits to OC-192 links and extending link budgets, *Error Control/Coding also known as Forward Error Correction (FEC), *Adaptive Receivers for signal quality monitoring for real-time estimation of Q/OSNR, eye-pattern, signal BER and related temporal statistics (such as jitter). We will discuss the theoretical underpinnings of these receiver and transmitter architectures, provide examples of system performance and conclude with general market trends. These Physical layer IC solutions represent a fundamental new toolbox of options for equipment designers in addressing systems level problems. With unmatched cost and yield/performance tradeoffs, it is expected that IC approaches will provide significant flexibility in turn, for carriers and service providers who must ultimately manage the network and assure acceptable quality of service under stringent cost constraints.

FACT (Version 2.0) - Subsurface Flow and Contaminant Transport Documentation and User's Guide

Energy Technology Data Exchange (ETDEWEB)

Aleman, S.E.

2000-05-05

This report documents a finite element code designed to model subsurface flow and contaminant transport, named FACT. FACT is a transient three-dimensional, finite element code designed to simulate isothermal groundwater flow, moisture movement, and solute transport in variably saturated and fully saturated subsurface porous media.

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

Green transportation logistics: the quest for win-win solutions

DEFF Research Database (Denmark)

measures and speed and route optimization; Sulphur emissions; Lifecycle emissions; Green rail transportation; Green air transportation; Green inland navigation and possible areas for further research. Throughout, the book pursues the goal of “win-win” solutions and analyzes the phenomenon of “push......This book examines the state of the art in green transportation logistics from the perspective of balancing environmental performance in the transportation supply chain while also satisfying traditional economic performance criteria. Part of the book is drawn from the recently completed European...... Union project Super Green, a three-year project intended to promote the development of European freight corridors in an environmentally friendly manner. Additional chapters cover both the methodological base and the application context of green transportation logistics. Individual chapters look...

FEHM, Finite Element Heat and Mass Transfer Code

International Nuclear Information System (INIS)

Zyvoloski, G.A.

2002-01-01

1 - Description of program or function: FEHM is a numerical simulation code for subsurface transport processes. It models 3-D, time-dependent, multiphase, multicomponent, non-isothermal, reactive flow through porous and fractured media. It can accurately represent complex 3-D geologic media and structures and their effects on subsurface flow and transport. Its capabilities include flow of gas, water, and heat; flow of air, water, and heat; multiple chemically reactive and sorbing tracers; finite element/finite volume formulation; coupled stress module; saturated and unsaturated media; and double porosity and double porosity/double permeability capabilities. 2 - Methods: FEHM uses a preconditioned conjugate gradient solution of coupled linear equations and a fully implicit, fully coupled Newton Raphson solution of nonlinear equations. It has the capability of simulating transport using either a advection/diffusion solution or a particle tracking method. 3 - Restriction on the complexity of the problem: Disk space and machine memory are the only limitations

Synergism of the method of characteristic, R-functions and diffusion solution for accurate representation of 3D neutron interactions in research reactors using the AGENT code system

International Nuclear Information System (INIS)

Hursin, Mathieu; Xiao Shanjie; Jevremovic, Tatjana

2006-01-01

This paper summarizes the theoretical and numerical aspects of the AGENT code methodology accurately applied for detailed three-dimensional (3D) multigroup steady-state modeling of neutron interactions in complex heterogeneous reactor domains. For the first time we show the fine-mesh neutron scalar flux distribution in Purdue research reactor (that was built over forty years ago). The AGENT methodology is based on the unique combination of the three theories: the method of characteristics (MOC) used to simulate the neutron transport in two-dimensional (2D) whole core heterogeneous calculation, the theory of R-functions used as a mathematical tool to describe the true geometry and fuse with the MOC equations, and one-dimensional (1D) higher-order diffusion correction of 2D transport model to account for full 3D heterogeneous whole core representation. The synergism between the radial 2D transport and the 1D axial transport (to take into account the axial neutron interactions and leakage), called the 2D/1D method (used in DeCART and CHAPLET codes), provides a 3D computational solution. The unique synergism between the AGENT geometrical algorithm capable of modeling any current or future reactor core geometry and 3D neutron transport methodology is described in details. The 3D AGENT accuracy and its efficiency are demonstrated showing the eigenvalues, point-wise flux and reaction rate distributions in representative reactor geometries. The AGENT code, comprising this synergism, represents a building block of the computational system, called the virtual reactor. Its main purpose is to perform 'virtual' experiments and demonstrations of various mainly university research reactor experiments

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)

The effect of low-GDP solution on ultrafiltration and solute transport in continuous ambulatory peritoneal dialysis patients.

Science.gov (United States)

Cho, Kyu-Hyang; Do, Jun-Young; Park, Jong-Won; Yoon, Kyung-Woo; Kim, Yong-Lim

2013-01-01

Several studies have reported benefits for human peritoneal mesothelial cell function of a neutral-pH dialysate low in glucose degradation products (GDPs). However, the effects of low-GDP solution on ultrafiltration (UF), transport of solutes, and control of body water remain elusive. We therefore investigated the effect of low-GDP solution on UF, solute transport, and control of body water. Among 79 new continuous ambulatory peritoneal dialysis (CAPD) patients, 60 completed a 12-month protocol (28 in a lactate-based high-GDP solution group, 32 in a lactate-based low-GDP solution group). Clinical indices--including 24-hour UF volume (UFV), 24-hour urine volume (UV), residual renal function, and dialysis adequacy--were measured at months 1, 6, and 12. At months 1, 6, and 12, UFV, glucose absorption, 4-hour dialysate-to-plasma (D/P) creatinine, and 1-hour D/P Na(+) were assessed during a modified 4.25% peritoneal equilibration test (PET). Body composition by bioelectric impedance analysis was measured at months 1 and 12 in 26 CAPD patients. Daily UFV was lower in the low-GDP group. Despite similar solute transport and aquaporin function, the low-GDP group also showed lower UFV and higher glucose absorption during the PET. Factors associated with UFV during the PET were lactate-based high-GDP solution and 1-hour D/P Na(+). No differences in volume status and obesity at month 12 were observed, and improvements in hypervolemia were equal in both groups. Compared with the high-GDP group, the low-GDP group had a lower UFV during a PET and a lower daily UFV during the first year after peritoneal dialysis initiation. Although the low-GDP group had a lower daily UFV, no difficulties in controlling edema were encountered.

Different nonideality relationships, different databases and their effects on modeling precipitation from concentrated solutions using numerical speciation codes

Energy Technology Data Exchange (ETDEWEB)

Brown, L.F.; Ebinger, M.H.

1996-08-01

Four simple precipitation problems are solved to examine the use of numerical equilibrium codes. The study emphasizes concentrated solutions, assumes both ideal and nonideal solutions, and employs different databases and different activity-coefficient relationships. The study uses the EQ3/6 numerical speciation codes. The results show satisfactory material balances and agreement between solubility products calculated from free-energy relationships and those calculated from concentrations and activity coefficients. Precipitates show slightly higher solubilities when the solutions are regarded as nonideal than when considered ideal, agreeing with theory. When a substance may precipitate from a solution dilute in the precipitating substance, a code may or may not predict precipitation, depending on the database or activity-coefficient relationship used. In a problem involving a two-component precipitation, there are only small differences in the precipitate mass and composition between the ideal and nonideal solution calculations. Analysis of this result indicates that this may be a frequent occurrence. An analytical approach is derived for judging whether this phenomenon will occur in any real or postulated precipitation situation. The discussion looks at applications of this approach. In the solutes remaining after the precipitations, there seems to be little consistency in the calculated concentrations and activity coefficients. They do not appear to depend in any coherent manner on the database or activity-coefficient relationship used. These results reinforce warnings in the literature about perfunctory or mechanical use of numerical speciation codes.

Product Lifecycle Management and the Quest for Sustainable Space Transportation Solutions

Science.gov (United States)

Caruso, Pamela W.

2009-01-01

This viewgraph presentation reviews NASA Marshall's effort to sustain space transportation solutions through product lines that include: 1) Propulsion and Transportation Systems; 2) Life Support Systems; and 3) and Earth and Space Science Spacecraft Systems, and Operations.

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