WorldWideScience

Sample records for transport code developments

  1. 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)

  2. Coupled geochemical and solute transport code development

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  3. 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

  4. 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.)

  5. 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)

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

    International Nuclear Information System (INIS)

    Herbert, A.W.

    1990-06-01

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

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

    International Nuclear Information System (INIS)

    Mihara, Morihiro; Ooi, Takao

    2004-01-01

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

  8. 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)

  9. 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)

  10. Development of a relativistic Particle In Cell code PARTDYN for linear accelerator beam transport

    Energy Technology Data Exchange (ETDEWEB)

    Phadte, D., E-mail: deepraj@rrcat.gov.in [LPD, Raja Ramanna Centre for Advanced Technology, Indore 452013 (India); Patidar, C.B.; Pal, M.K. [MAASD, Raja Ramanna Centre for Advanced Technology, Indore (India)

    2017-04-11

    A relativistic Particle In Cell (PIC) code PARTDYN is developed for the beam dynamics simulation of z-continuous and bunched beams. The code is implemented in MATLAB using its MEX functionality which allows both ease of development as well higher performance similar to a compiled language like C. The beam dynamics calculations carried out by the code are compared with analytical results and with other well developed codes like PARMELA and BEAMPATH. The effect of finite number of simulation particles on the emittance growth of intense beams has been studied. Corrections to the RF cavity field expressions were incorporated in the code so that the fields could be calculated correctly. The deviations of the beam dynamics results between PARTDYN and BEAMPATH for a cavity driven in zero-mode have been discussed. The beam dynamics studies of the Low Energy Beam Transport (LEBT) using PARTDYN have been presented.

  11. 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

  12. Transport theory and codes

    International Nuclear Information System (INIS)

    Clancy, B.E.

    1986-01-01

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

  13. 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

  14. 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)

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  16. Overview of development and design of MPACT: Michigan parallel characteristics transport code

    Energy Technology Data Exchange (ETDEWEB)

    Kochunas, B.; Collins, B.; Jabaay, D.; Downar, T. J.; Martin, W. R. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, 2200 Bonisteel, Ann Arbor, MI 48109 (United States)

    2013-07-01

    MPACT (Michigan Parallel Characteristics Transport Code) is a new reactor analysis tool. It is being developed by students and research staff at the University of Michigan to be used for an advanced pin-resolved transport capability within VERA (Virtual Environment for Reactor Analysis). VERA is the end-user reactor simulation tool being produced by the Consortium for the Advanced Simulation of Light Water Reactors (CASL). The MPACT development project is itself unique for the way it is changing how students do research to achieve the instructional and research goals of an academic institution, while providing immediate value to industry. The MPACT code makes use of modern lean/agile software processes and extensive testing to maintain a level of productivity and quality required by CASL. MPACT's design relies heavily on object-oriented programming concepts and design patterns and is programmed in Fortran 2003. These designs are explained and illustrated as to how they can be readily extended to incorporate new capabilities and research ideas in support of academic research objectives. The transport methods currently implemented in MPACT include the 2-D and 3-D method of characteristics (MOC) and 2-D and 3-D method of collision direction probabilities (CDP). For the cross section resonance treatment, presently the subgroup method and the new embedded self-shielding method (ESSM) are implemented within MPACT. (authors)

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

    Science.gov (United States)

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

    2017-09-01

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

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

    Directory of Open Access Journals (Sweden)

    Chapoutier Nicolas

    2017-01-01

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

  19. 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)

  20. Present status of transport code development based on Monte Carlo method

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki

    1985-01-01

    The present status of development in Monte Carlo code is briefly reviewed. The main items are the followings; Application fields, Methods used in Monte Carlo code (geometry spectification, nuclear data, estimator and variance reduction technique) and unfinished works, Typical Monte Carlo codes and Merits of continuous energy Monte Carlo code. (author)

  1. 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

  2. Development Of A Parallel Performance Model For The THOR Neutral Particle Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Yessayan, Raffi; Azmy, Yousry; Schunert, Sebastian

    2017-02-01

    The THOR neutral particle transport code enables simulation of complex geometries for various problems from reactor simulations to nuclear non-proliferation. It is undergoing a thorough V&V requiring computational efficiency. This has motivated various improvements including angular parallelization, outer iteration acceleration, and development of peripheral tools. For guiding future improvements to the code’s efficiency, better characterization of its parallel performance is useful. A parallel performance model (PPM) can be used to evaluate the benefits of modifications and to identify performance bottlenecks. Using INL’s Falcon HPC, the PPM development incorporates an evaluation of network communication behavior over heterogeneous links and a functional characterization of the per-cell/angle/group runtime of each major code component. After evaluating several possible sources of variability, this resulted in a communication model and a parallel portion model. The former’s accuracy is bounded by the variability of communication on Falcon while the latter has an error on the order of 1%.

  3. 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

  4. Sub-Transport Layer Coding

    DEFF Research Database (Denmark)

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

    2014-01-01

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

  5. 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)

  6. 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.

  7. Accident consequence assessment code development

    International Nuclear Information System (INIS)

    Homma, T.; Togawa, O.

    1991-01-01

    This paper describes the new computer code system, OSCAAR developed for off-site consequence assessment of a potential nuclear accident. OSCAAR consists of several modules which have modeling capabilities in atmospheric transport, foodchain transport, dosimetry, emergency response and radiological health effects. The major modules of the consequence assessment code are described, highlighting the validation and verification of the models. (author)

  8. Development of one-dimensional computational fluid dynamics code 'GFLOW' for groundwater flow and contaminant transport analysis

    International Nuclear Information System (INIS)

    Rahatgaonkar, P. S.; Datta, D.; Malhotra, P. K.; Ghadge, S. G.

    2012-01-01

    Prediction of groundwater movement and contaminant transport in soil is an important problem in many branches of science and engineering. This includes groundwater hydrology, environmental engineering, soil science, agricultural engineering and also nuclear engineering. Specifically, in nuclear engineering it is applicable in the design of spent fuel storage pools and waste management sites in the nuclear power plants. Ground water modeling involves the simulation of flow and contaminant transport by groundwater flow. In the context of contaminated soil and groundwater system, numerical simulations are typically used to demonstrate compliance with regulatory standard. A one-dimensional Computational Fluid Dynamics code GFLOW had been developed based on the Finite Difference Method for simulating groundwater flow and contaminant transport through saturated and unsaturated soil. The code is validated with the analytical model and the benchmarking cases available in the literature. (authors)

  9. Application of the newly developed 3D transport code r3t to selected field cases

    International Nuclear Information System (INIS)

    Rubel, A.; Fein, E.

    2006-01-01

    To model the transport of contaminants in connexion with long-term safety analyses of repositories for radioactive waste recently the numerical model r 3 t was developed. In the following first applications of r 3 t are presented. The results of modelling of the transport of zinc at the sewage treatment facility in Cape Cod, Massachusetts, and of trace elements in the frame work of the in-situ diffusion experiments DI-A and DR at the underground laboratory Mont Terri are shown. (authors)

  10. 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)

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

    International Nuclear Information System (INIS)

    Santamarina, A.

    1991-01-01

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

  12. 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

  13. Developing a friendly I/O graphical interface for the integral transport CP2D computer code

    International Nuclear Information System (INIS)

    Constantin, M.

    2002-01-01

    The code CP 2 D design and developing involved the newest methods and techniques in the first flight collision probability (FFCP) calculations. These methods are strongly connected with the computer developing both in hardware and software. The code CP 2 D was developed in INR Pitesti, between 1997-2001. It is a transport code in the first flight collision probability formalism, able to treat exactly a lot of complicated geometry (such as CANDU clusters, TRIGA and PWR fuel assemblies). The first version CP 2 D1.0 was released in 1998. The second, CP 2 D2.0, was released in 1999 and uses a multistratified coolant model (MM) for CANDU loss of coolant accident analysis. The third version, CP 2 D3.0 (2000), have incorporated a generalized burning scheme. An user-friendly graphical interface was developed in 2001. It is intended to a rapid introduction of the input data and to extract the interest information from the output files. This information is directly converted into graphics and tables contained into a single MsWord document. The introduced input data are validated by the interface if the numerical, physical and mathematical restrictions are fulfilled. The user can see the representation of the model and can interactively modify the input data until the model is correct. The interface and the code were exhaustively documented and the obtained version was released as CP 2 D4.0. The version allows to a low experienced user to build the input files, to correct the possible errors and to extract the information of interest for the analyzed problem. The paper shows the interface as a useful concept for the upgrade of the classical codes. (author)

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

    International Nuclear Information System (INIS)

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

    2015-09-01

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

  15. 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)

  16. Radiation transport and shielding information, computer codes, and nuclear data for use in CTR neutronics research and development

    International Nuclear Information System (INIS)

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

    1976-01-01

    The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components

  17. 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

  18. Monte Carlo code development in Los Alamos

    International Nuclear Information System (INIS)

    Carter, L.L.; Cashwell, E.D.; Everett, C.J.; Forest, C.A.; Schrandt, R.G.; Taylor, W.M.; Thompson, W.L.; Turner, G.D.

    1974-01-01

    The present status of Monte Carlo code development at Los Alamos Scientific Laboratory is discussed. A brief summary is given of several of the most important neutron, photon, and electron transport codes. 17 references. (U.S.)

  19. Radiation transport and shielding information, computer codes, and nuclear data for use in CTR neutronics research and development

    International Nuclear Information System (INIS)

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

    1976-01-01

    The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components. Since 1964, the Center has been involved in the international exchange of information, encouraged and supported by both government and interagency agreements; and to achieve an equally viable and successful program in fusion research, the reciprocal exchange of CTR data and computing technology is encouraged and welcomed

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

    International Nuclear Information System (INIS)

    White, Morgan C.

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  2. 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

  3. 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

  4. 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

  5. The development of high performance numerical simulation code for transient groundwater flow and reactive solute transport problems based on local discontinuous Galerkin method

    International Nuclear Information System (INIS)

    Suzuki, Shunichi; Motoshima, Takayuki; Naemura, Yumi; Kubo, Shin; Kanie, Shunji

    2009-01-01

    The authors develop a numerical code based on Local Discontinuous Galerkin Method for transient groundwater flow and reactive solute transport problems in order to make it possible to do three dimensional performance assessment on radioactive waste repositories at the earliest stage possible. Local discontinuous Galerkin Method is one of mixed finite element methods which are more accurate ones than standard finite element methods. In this paper, the developed numerical code is applied to several problems which are provided analytical solutions in order to examine its accuracy and flexibility. The results of the simulations show the new code gives highly accurate numeric solutions. (author)

  6. 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

  7. Development of a coupled dynamics code with transport theory capability and application to accelerator driven systems transients

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Ama, T.; Palmiotti, G.; Taiwo, T.A.; Yang, W.S.

    2000-01-01

    The VARIANT-K and DIF3D-K nodal spatial kinetics computer codes have been coupled to the SAS4A and SASSYS-1 liquid metal reactor accident and systems analysis codes. SAS4A and SASSYS-1 have been extended with the addition of heavy liquid metal (Pb and Pb-Bi) thermophysical properties, heat transfer correlations, and fluid dynamics correlations. The coupling methodology and heavy liquid metal modeling additions are described. The new computer code suite has been applied to analysis of neutron source and thermal-hydraulics transients in a model of an accelerator-driven minor actinide burner design proposed in an OECD/NEA/NSC benchmark specification. Modeling assumptions and input data generation procedures are described. Results of transient analyses are reported, with emphasis on comparison of P1 and P3 variational nodal transport theory results with nodal diffusion theory results, and on significance of spatial kinetics effects

  8. Development of a neutron transport code many-group two-dimensional heterogeneous calculations by the method of characteristics

    International Nuclear Information System (INIS)

    Petkov, P.T.

    2000-01-01

    The method of characteristics (MOC) is gaining increased popularity in the reactor physics community all over the world because it gives a new degree of freedom in nuclear reactor analysis. The MARIKO code solves the neutron transport equation by the MOC in two-dimensional real geometry. The domain of solution can be a rectangle or right hexagon with periodic boundary conditions on the outer boundary. Any reasonable symmetry inside the domain can be fully accounted for. The geometry is described in three levels-macro-cells, cells, and regions. The macro-cells and cells can be any polygon. The outer boundary of a region can be any combination of straight line and circular arc segments. Any level of embedded regions is allowed. Procedures for automatic geometry description of hexagonal fuel assemblies and reflector macro-cells have been developed. The initial ray tracing procedure is performed for the full rectangular or hexagonal domain, but only azimuthal angles in the smallest symmetry interval are tracked. (Authors)

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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)

  15. Optix: A Monte Carlo scintillation light transport code

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-02-11

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

  16. 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.)

  17. 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.

  18. 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

  19. 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.

  20. 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

  1. 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

  2. Coupling calculation of CFD-ACE computational fluid dynamics code and DeCART whole-core neutron transport code for development of numerical reactor

    International Nuclear Information System (INIS)

    Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog

    2005-03-01

    Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too

  3. Coupling calculation of CFD-ACE computational fluid dynamics code and DeCART whole-core neutron transport code for development of numerical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog

    2005-03-15

    Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too.

  4. 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

  5. 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

  6. 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

  7. Development status of TUF code

    International Nuclear Information System (INIS)

    Liu, W.S.; Tahir, A.; Zaltsgendler

    1996-01-01

    An overview of the important development of the TUF code in 1995 is presented. The development in the following areas is presented: control of round-off error propagation, gas resolution and release models, and condensation induced water hammer. This development is mainly generated from station requests for operational support and code improvement. (author)

  8. Los Alamos neutral particle transport codes: New and enhanced capabilities

    International Nuclear Information System (INIS)

    Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Clark, B.A.; Koch, K.R.; Marr, D.R.

    1992-01-01

    We present new developments in Los Alamos discrete-ordinates transport codes and introduce THREEDANT, the latest in the series of Los Alamos discrete ordinates transport codes. THREEDANT solves the multigroup, neutral-particle transport equation in X-Y-Z and R-Θ-Z geometries. THREEDANT uses computationally efficient algorithms: Diffusion Synthetic Acceleration (DSA) is used to accelerate the convergence of transport iterations, the DSA solution is accelerated using the multigrid technique. THREEDANT runs on a wide range of computers, from scientific workstations to CRAY supercomputers. The algorithms are highly vectorized on CRAY computers. Recently, the THREEDANT transport algorithm was implemented on the massively parallel CM-2 computer, with performance that is comparable to a single-processor CRAY-YMP We present the results of THREEDANT analysis of test problems

  9. 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

  10. 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)

  11. 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)

  12. Development of code PRETOR for stellarator simulation

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  13. A study on the fusion reactor - Development of MHD stability and transport code for KT-2 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Koo; Shin, Kyo Jin [Pohang University of Science and Tecnology, Pohang (Korea, Republic of)

    1996-08-01

    MHD Stability analyses for KT-2 Tokamak were carried out by using CART (Resistive 3-D) Code. Linear Growth rates and linear perturbed eigen function of both N=0 axisymmetric mode and N=1 kink modes of highly elongated tokamak plasmas, in the presence of a conducting wall at various distances are computed and linear and nonlinear evolution of N=0 axisymmetric modes are simulated. 26 refs., 25 figs. (author)

  14. 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)

  15. 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

  16. Multiple application coded switch development report

    International Nuclear Information System (INIS)

    Bernal, E.L.; Kestly, J.D.

    1979-03-01

    The development of the Multiple Application Coded Switch (MACS) and its related controller are documented; the functional and electrical characteristics are described; the interface requirements defined, and a troubleshooting guide provided. The system was designed for the Safe Secure Trailer System used for secure transportation of nuclear material

  17. 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

  18. Numeric implementation of a nucleation, growth and transport model for helium bubbles in lead-lithium HCLL breeding blanket channels: Theory and code development

    Energy Technology Data Exchange (ETDEWEB)

    Batet, L., E-mail: lluis.batet@upc.edu [Technical University of Catalonia (UPC), Energy and Radiation Studies Research Group (GREENER), Technology for Fusion T4F, Barcelona (Spain); UPC, Department of Physics and Nuclear Engineering (DFEN), ETSEIB, Av. Diagonal 647, 08028 Barcelona (Spain); Fradera, J. [Technical University of Catalonia (UPC), Energy and Radiation Studies Research Group (GREENER), Technology for Fusion T4F, Barcelona (Spain); UPC, Department of Physics and Nuclear Engineering (DFEN), ETSEIB, Av. Diagonal 647, 08028 Barcelona (Spain); Valls, E. Mas de les [Technical University of Catalonia (UPC), Energy and Radiation Studies Research Group (GREENER), Technology for Fusion T4F, Barcelona (Spain); UPC, Department of Heat Engines (DMMT), ETSEIB, Av. Diagonal 647, 08028 Barcelona (Spain); Sedano, L.A. [EURATOM-CIEMAT Association, Fusion Technology Division, Av. Complutense 22, 28040 Madrid (Spain)

    2011-06-15

    Large helium (He) production rates in liquid metal breeding blankets of a DT fusion reactor might have a significant influence in the system design. Low He solubility together with high local concentrations may create the conditions for He cavitation, which would have an impact in the components performance. The paper states that such a possibility is not remote in a helium cooled lithium-lead breeding blanket design. A model based on the Classical Nucleation Theory (CNT) has been developed and implemented in order to have a specific tool able to simulate HCLL systems and identify the key parameters and sensitivities. The nucleation and growth model has been implemented in the open source CFD code OpenFOAM so that transport of dissolved atomic He and nucleated He bubbles can be simulated. At the current level of development it is assumed that void fraction is small enough not to affect either the hydrodynamics or the properties of the liquid metal; thus, bubbles can be represented by means of a passive scalar. He growth and transport has been implemented using the mean radius approach in order to save computational time. Limitations and capabilities of the model are shown by means of zero-dimensional simulation and sensitivity analysis under HCLL breeding unit conditions.

  19. LiveCode mobile development

    CERN Document Server

    Lavieri, Edward D

    2013-01-01

    A practical guide written in a tutorial-style, ""LiveCode Mobile Development Hotshot"" walks you step-by-step through 10 individual projects. Every project is divided into sub tasks to make learning more organized and easy to follow along with explanations, diagrams, screenshots, and downloadable material.This book is great for anyone who wants to develop mobile applications using LiveCode. You should be familiar with LiveCode and have access to a smartphone. You are not expected to know how to create graphics or audio clips.

  20. 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)

  1. 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)

  2. Coarse mesh code development

    Energy Technology Data Exchange (ETDEWEB)

    Lieberoth, J.

    1975-06-15

    The numerical solution of the neutron diffusion equation plays a very important role in the analysis of nuclear reactors. A wide variety of numerical procedures has been proposed, at which most of the frequently used numerical methods are fundamentally based on the finite- difference approximation where the partial derivatives are approximated by the finite difference. For complex geometries, typical of the practical reactor problems, the computational accuracy of the finite-difference method is seriously affected by the size of the mesh width relative to the neutron diffusion length and by the heterogeneity of the medium. Thus, a very large number of mesh points are generally required to obtain a reasonably accurate approximate solution of the multi-dimensional diffusion equation. Since the computation time is approximately proportional to the number of mesh points, a detailed multidimensional analysis, based on the conventional finite-difference method, is still expensive even with modern large-scale computers. Accordingly, there is a strong incentive to develop alternatives that can reduce the number of mesh-points and still retain accuracy. One of the promising alternatives is the finite element method, which consists of the expansion of the neutron flux by piecewise polynomials. One of the advantages of this procedure is its flexibility in selecting the locations of the mesh points and the degree of the expansion polynomial. The small number of mesh points of the coarse grid enables to store the results of several of the least outer iterations and to calculate well extrapolated values of them by comfortable formalisms. This holds especially if only one energy distribution of fission neutrons is assumed for all fission processes in the reactor, because the whole information of an outer iteration is contained in a field of fission rates which has the size of all mesh points of the coarse grid.

  3. 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)

  4. 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.

  5. 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)

  6. 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.

  7. 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)

  8. 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)

  9. 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.

  10. Development of the DTNTES code

    International Nuclear Information System (INIS)

    Ortega Prieto, P.; Morales Dorado, M.D.; Alonso Santos, A.

    1987-01-01

    The DTNTES code has been developed in the Department of Nuclear Technology of the Polytechnical University in Madrid as a part of the Research Program on Quantitative Risk Analysis. DTNTES code calculates several time-dependent probabilistic characteristics of basic events, minimal cut sets and the top event of a fault tree. The code assumes that basic events are statistically independent, and they have failure and repair distributions. It computes the minimal cut upper bound approach for the top event unavailability, and the time-dependent unreliability of the top event by means of different methods, selected by the user. These methods are: expected number of system failures, failure rate, Barlow-Proschan bound, steady-state upper bound, and T* method. (author)

  11. 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.)

  12. 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.

  13. Development of TIME2 code

    International Nuclear Information System (INIS)

    1986-02-01

    The paper reviews the progress on the development of a computer model TIME2, for modelling the long term evolution of shallow burial site environments for low- and intermediate-level radioactive waste disposal. The subject is discussed under the five topic headings: 1) background studies, including geomorphology, climate, human-induced effects, and seismicity, 2) development of the TIME2 code, 3) verification and testing, 4) documentation, and, 5) role of TIME2 in radiological risk assessment. (U.K.)

  14. Development of an integrated fission product release and transport code for spatially resolved full-core calculations of V/HTRs

    International Nuclear Information System (INIS)

    Xhonneux, Andre; Allelein, Hans-Josef

    2014-01-01

    The computer codes FRESCO-I, FRESCO-II, PANAMA and SPATRA developed at Forschungszentrum Jülich in Germany in the early 1980s are essential tools to predict the fission product release from spherical fuel elements and the TRISO fuel performance, respectively, under given normal or accidental conditions. These codes are able to calculate a conservative estimation of the source term, i.e. quantity and duration of radionuclide release. Recently, these codes have been reversed engineered, modernized (FORTRAN 95/2003) and combined to form a consistent code named STACY (Source Term Analysis Code System). STACY will later become a module of the V/HTR Code Package (HCP). In addition, further improvements have been implemented to enable more detailed calculations. For example the distinct temperature profile along the pebble radius is now taken into account and coated particle failure rates can be calculated under normal operating conditions. In addition, the absolute fission product release of an V/HTR pebble bed core can be calculated by using the newly developed burnup code Topological Nuclide Transformation (TNT) replacing the former rudimentary approach. As a new functionality, spatially resolved fission product release calculations for normal operating conditions as well as accident conditions can be performed. In case of a full-core calculation, a large number of individual pebbles which follow a random path through the reactor core can be simulated. The history of the individual pebble is recorded, too. Main input data such as spatially resolved neutron fluxes and fluid dynamics data are provided by the VSOP code. Capabilities of the FRESCO-I and SPATRA code which allow for the simulation of the redistribution of fission products within the primary circuit and the deposition of fission products on graphitic and metallic surfaces are also available in STACY. In this paper, details of the STACY model and first results for its application to the 200 MW(th) HTR

  15. 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)

  16. 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)

  17. 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

  18. HELIAS module development for systems codes

    Energy Technology Data Exchange (ETDEWEB)

    Warmer, F., E-mail: Felix.Warmer@ipp.mpg.de; Beidler, C.D.; Dinklage, A.; Egorov, K.; Feng, Y.; Geiger, J.; Schauer, F.; Turkin, Y.; Wolf, R.; Xanthopoulos, P.

    2015-02-15

    In order to study and design next-step fusion devices such as DEMO, comprehensive systems codes are commonly employed. In this work HELIAS-specific models are proposed which are designed to be compatible with systems codes. The subsequently developed models include: a geometry model based on Fourier coefficients which can represent the complex 3-D plasma shape, a basic island divertor model which assumes diffusive cross-field transport and high radiation at the X-point, and a coil model which combines scaling aspects based on the HELIAS 5-B reactor design in combination with analytic inductance and field calculations. In addition, stellarator-specific plasma transport is discussed. A strategy is proposed which employs a predictive confinement time scaling derived from 1-D neoclassical and 3-D turbulence simulations. This paper reports on the progress of the development of the stellarator-specific models while an implementation and verification study within an existing systems code will be presented in a separate work. This approach is investigated to ultimately allow one to conduct stellarator system studies, develop design points of HELIAS burning plasma devices, and to facilitate a direct comparison between tokamak and stellarator DEMO and power plant designs.

  19. Integrated code development for studying laser driven plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Takabe, Hideaki; Nagatomo, Hideo; Sunahara, Atsusi; Ohnishi, Naofumi; Naruo, Syuji; Mima, Kunioki [Osaka Univ., Suita (Japan). Inst. of Laser Engineering

    1998-03-01

    Present status and plan for developing an integrated implosion code are briefly explained by focusing on motivation, numerical scheme and issues to be developed more. Highly nonlinear stage of Rayleigh-Taylor instability of ablation front by laser irradiation has been simulated so as to be compared with model experiments. Improvement in transport and rezoning/remapping algorithms in ILESTA code is described. (author)

  20. The development of code benchmarks

    International Nuclear Information System (INIS)

    Glass, R.E.

    1986-01-01

    Sandia National Laboratories has undertaken a code benchmarking effort to define a series of cask-like problems having both numerical solutions and experimental data. The development of the benchmarks includes: (1) model problem definition, (2) code intercomparison, and (3) experimental verification. The first two steps are complete and a series of experiments are planned. The experiments will examine the elastic/plastic behavior of cylinders for both the end and side impacts resulting from a nine meter drop. The cylinders will be made from stainless steel and aluminum to give a range of plastic deformations. This paper presents the results of analyses simulating the model's behavior using materials properties for stainless steel and aluminum

  1. 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.)

  2. SWAAM code development, verification and application to steam generator design

    International Nuclear Information System (INIS)

    Shin, Y.W.; Valentin, R.A.

    1990-01-01

    This paper describes the family of SWAAM codes developed by Argonne National Laboratory to analyze the effects of sodium/water reactions on LMR steam generators. The SWAAM codes were developed as design tools for analyzing various phenomena related to steam generator leaks and to predict the resulting thermal and hydraulic effects on the steam generator and the intermediate heat transport system (IHTS). The theoretical foundations and numerical treatments on which the codes are based are discussed, followed by a description of code capabilities and limitations, verification of the codes by comparison with experiment, and applications to steam generator and IHTS design. (author). 25 refs, 14 figs

  3. Transportation in African Development.

    Science.gov (United States)

    Altschul, Robert D.

    1980-01-01

    Examines the structure, role, and needs of Africa's national and intracontinental transportation system. Characteristics of rail, water, road, and air transportation are examined. The conclusion is that high investment in transportation systems is essential to the development process. (Author/KC)

  4. 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

  5. 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.

  6. 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

  7. Discrete fracture network code development

    Energy Technology Data Exchange (ETDEWEB)

    Dershowitz, W.; Doe, T.; Shuttle, D.; Eiben, T.; Fox, A.; Emsley, S.; Ahlstrom, E. [Golder Associates Inc., Redmond, Washington (United States)

    1999-02-01

    This report presents the results of fracture flow model development and application performed by Golder Associates Inc. during the fiscal year 1998. The primary objective of the Golder Associates work scope was to provide theoretical and modelling support to the JNC performance assessment effort in fiscal year 2000. In addition, Golder Associates provided technical support to JNC for the Aespoe project. Major efforts for performance assessment support included extensive flow and transport simulations, analysis of pathway simplification, research on excavation damage zone effects, software verification and cross-verification, and analysis of confidence bounds on Monte Carlo simulations. In addition, a Fickian diffusion algorithm was implemented for Laplace Transform Galerkin solute transport. Support for the Aespoe project included predictive modelling of sorbing tracer transport in the TRUE-1 rock block, analysis of 1 km geochemical transport pathways for Task 5', and data analysis and experimental design for the TRUE Block Scale experiment. Technical information about Golder Associates support to JNC is provided in the appendices to this report. (author)

  8. Development and validation of sodium fire codes

    International Nuclear Information System (INIS)

    Morii, Tadashi; Himeno Yoshiaki; Miyake, Osamu

    1989-01-01

    Development, verification, and validation of the spray fire code, SPRAY-3M, the pool fire codes, SOFIRE-M2 and SPM, the aerosol behavior code, ABC-INTG, and the simultaneous spray and pool fires code, ASSCOPS, are presented. In addition, the state-of-the-art of development of the multi-dimensional natural convection code, SOLFAS, for the analysis of heat-mass transfer during a fire, is presented. (author)

  9. Reactor safety computer code development at INEL

    International Nuclear Information System (INIS)

    Johnsen, G.W.

    1985-01-01

    This report provides a brief overview of the computer code development programs being conducted at EG and G Idaho, Inc. on behalf of US Nuclear Regulatory Commission and the Department of Energy, Idaho Operations Office. Included are descriptions of the codes being developed, their development status as of the date of this report, and resident code development expertise

  10. Development of MCNP interface code in HFETR

    International Nuclear Information System (INIS)

    Qiu Liqing; Fu Rong; Deng Caiyu

    2007-01-01

    In order to describe the HFETR core with MCNP method, the interface code MCNPIP for HFETR and MCNP code is developed. This paper introduces the core DXSY and flowchart of MCNPIP code, and the handling of compositions of fuel elements and requirements on hardware and software. Finally, MCNPIP code is validated against the practical application. (authors)

  11. 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

  12. 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.)

  13. 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.

  14. 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

  15. 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.)

  16. 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)

  17. 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

  18. 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

  19. 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.

  20. 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.

  1. Discrete-ordinates electron transport calculations using standard neutron transport codes

    International Nuclear Information System (INIS)

    Morel, J.E.

    1979-01-01

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

  2. 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.)

  3. 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

  4. 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

  5. PRESTO low-level waste transport and risk assessment code

    International Nuclear Information System (INIS)

    Little, C.A.; Fields, D.E.; McDowell-Boyer, L.M.; Emerson, C.J.

    1981-01-01

    PRESTO (Prediction of Radiation Effects from Shallow Trench Operations) is a computer code developed under US Environmental Protection Agency (EPA) funding to evaluate possible health effects from shallow land burial trenches. The model is intended to be generic and to assess radionuclide transport, ensuing exposure, and health impact to a static local population for a 1000-y period following the end of burial operations. Human exposure scenarios considered by the model include normal releases (including leaching and operational spillage), human intrusion, and site farming or reclamation. Pathways and processes of transit from the trench to an individual or population inlude: groundwater transport, overland flow, erosion, surface water dilution, resuspension, atmospheric transport, deposition, inhalation, and ingestion of contaminated beef, milk, crops, and water. Both population doses and individual doses are calculated as well as doses to the intruder and farmer. Cumulative health effects in terms of deaths from cancer are calculated for the population over the thousand-year period using a life-table approach. Data bases are being developed for three extant shallow land burial sites: Barnwell, South Carolina; Beatty, Nevada; and West Valley, New York

  6. Recent activities in accelerator code development

    International Nuclear Information System (INIS)

    Copper, R.K.; Ryne, R.D.

    1992-01-01

    In this paper we will review recent activities in the area of code development as it affects the accelerator community. We will first discuss the changing computing environment. We will review how the computing environment has changed in the last 10 years, with emphasis on computing power, operating systems, computer languages, graphics standards, and massively parallel processing. Then we will discuss recent code development activities in the areas of electromagnetics codes and beam dynamics codes

  7. 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

  8. Status of SPACE Safety Analysis Code Development

    International Nuclear Information System (INIS)

    Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun

    2009-01-01

    In 2006, the Korean the Korean nuclear industry started developing a thermal-hydraulic analysis code for safety analysis of PWR(Pressurized Water Reactor). The new code is named as SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). The SPACE code can solve two-fluid, three-field governing equations in one dimensional or three dimensional geometry. The SPACE code has many component models required for modeling a PWR, such as reactor coolant pump, safety injection tank, etc. The programming language used in the new code is C++, for new generation of engineers who are more comfortable with C/C++ than old FORTRAN language. This paper describes general characteristics of SPACE code and current status of SPACE code development

  9. Development and application of the waste code

    International Nuclear Information System (INIS)

    Morison, I.W.

    1984-01-01

    This paper discusses the objectives and general approach underlying the Australian Code of Practice on the Management of Radioactive Wastes arising from the Mining and Milling of Radioactive Ores 1982. Background to the development of the Code is provided and the guidelines which supplement the Code are considered

  10. Development of computer code in PNC, 3

    International Nuclear Information System (INIS)

    Ohtaki, Akira; Ohira, Hiroaki

    1990-01-01

    Super-COPD, a code which is integrated by calculation modules, has been developed in order to evaluate kinds of dynamics of LMFBR plant by improving COPD. The code involves all models and its advanced models of COPD in module structures. The code makes it possible to simulate the system dynamics of LMFBR plant of any configurations and components. (author)

  11. 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...

  12. ER@CEBAF: Modeling code developments

    Energy Technology Data Exchange (ETDEWEB)

    Meot, F. [Brookhaven National Lab. (BNL), Upton, NY (United States); Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States); Roblin, Y. [Brookhaven National Lab. (BNL), Upton, NY (United States); Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States)

    2016-04-13

    A proposal for a multiple-pass, high-energy, energy-recovery experiment using CEBAF is under preparation in the frame of a JLab-BNL collaboration. In view of beam dynamics investigations regarding this project, in addition to the existing model in use in Elegant a version of CEBAF is developed in the stepwise ray-tracing code Zgoubi, Beyond the ER experiment, it is also planned to use the latter for the study of polarization transport in the presence of synchrotron radiation, down to Hall D line where a 12 GeV polarized beam can be delivered. This Note briefly reports on the preliminary steps, and preliminary outcomes, based on an Elegant to Zgoubi translation.

  13. Multidimensional electron-photon transport with standard discrete ordinates codes

    International Nuclear Information System (INIS)

    Drumm, C.R.

    1995-01-01

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

  14. Recent developments in discrete ordinates electron transport

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  15. 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

  16. 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.

  17. Final Report for National Transport Code Collaboration PTRANSP

    International Nuclear Information System (INIS)

    Kritz, Arnold H.

    2012-01-01

    PTRANSP, which is the predictive version of the TRANSP code, was developed in a collaborative effort involving the Princeton Plasma Physics Laboratory, General Atomics Corporation, Lawrence Livermore National Laboratory, and Lehigh University. The PTRANSP/TRANSP suite of codes is the premier integrated tokamak modeling software in the United States. A production service for PTRANSP/TRANSP simulations is maintained at the Princeton Plasma Physics Laboratory; the server has a simple command line client interface and is subscribed to by about 100 researchers from tokamak projects in the US, Europe, and Asia. This service produced nearly 13000 PTRANSP/TRANSP simulations in the four year period FY 2005 through FY 2008. Major archives of TRANSP results are maintained at PPPL, MIT, General Atomics, and JET. Recent utilization, counting experimental analysis simulations as well as predictive simulations, more than doubled from slightly over 2000 simulations per year in FY 2005 and FY 2006 to over 4300 simulations per year in FY 2007 and FY 2008. PTRANSP predictive simulations applied to ITER increased eight fold from 30 simulations per year in FY 2005 and FY 2006 to 240 simulations per year in FY 2007 and FY 2008, accounting for more than half of combined PTRANSP/TRANSP service CPU resource utilization in FY 2008. PTRANSP studies focused on ITER played a key role in journal articles. Examples of validation studies carried out for momentum transport in PTRANSP simulations were presented at the 2008 IAEA conference. The increase in number of PTRANSP simulations has continued (more than 7000 TRANSP/PTRANSP simulations in 2010) and results of PTRANSP simulations appear in conference proceedings, for example the 2010 IAEA conference, and in peer reviewed papers. PTRANSP provides a bridge to the Fusion Simulation Program (FSP) and to the future of integrated modeling. Through years of widespread usage, each of the many parts of the PTRANSP suite of codes has been thoroughly

  18. 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)

  19. 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)

  20. Development of in-vessel source term analysis code, tracer

    International Nuclear Information System (INIS)

    Miyagi, K.; Miyahara, S.

    1996-01-01

    Analyses of radionuclide transport in fuel failure accidents (generally referred to source terms) are considered to be important especially in the severe accident evaluation. The TRACER code has been developed to realistically predict the time dependent behavior of FPs and aerosols within the primary cooling system for wide range of fuel failure events. This paper presents the model description, results of validation study, the recent model advancement status of the code, and results of check out calculations under reactor conditions. (author)

  1. 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.

  2. 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

  3. 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)

  4. 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

  5. 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.

  6. 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

  7. 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...

  8. 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)

  9. 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)

  10. 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.)

  11. 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.)

  12. 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

  13. 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

  14. 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

  15. Preliminary investigation study of code of developed country for developing Korean fuel cycle code

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Ko, Won Il; Lee, Ho Hee; Cho, Dong Keun; Park, Chang Je

    2012-01-01

    In order to develop Korean fuel cycle code, the analyses has been performed with the fuel cycle codes which are used in advanced country. Also, recommendations were proposed for future development. The fuel cycle codes are AS FLOOWS: VISTA which has been developed by IAEA, DANESS code which developed by ANL and LISTO, and VISION developed by INL for the Advanced Fuel Cycle Initiative (AFCI) system analysis. The recommended items were proposed for software, program scheme, material flow model, isotope decay model, environmental impact analysis model, and economics analysis model. The described things will be used for development of Korean nuclear fuel cycle code in future

  16. Development of Indian passenger transport

    Energy Technology Data Exchange (ETDEWEB)

    Ramanathan, R. [Indira Ghandi Institute of Development Research, Mumbai (India)

    1998-05-01

    The Indian transport sector has been studied using logistic substitution. The share of rail transport is declining, while road and air transport are increasing. These developments are not desirable from an energy-efficiency perspective. (author)

  17. 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

  18. 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)

  19. Development of a new EMP code at LANL

    Science.gov (United States)

    Colman, J. J.; Roussel-Dupré, R. A.; Symbalisty, E. M.; Triplett, L. A.; Travis, B. J.

    2006-05-01

    A new code for modeling the generation of an electromagnetic pulse (EMP) by a nuclear explosion in the atmosphere is being developed. The source of the EMP is the Compton current produced by the prompt radiation (γ-rays, X-rays, and neutrons) of the detonation. As a first step in building a multi- dimensional EMP code we have written three kinetic codes, Plume, Swarm, and Rad. Plume models the transport of energetic electrons in air. The Plume code solves the relativistic Fokker-Planck equation over a specified energy range that can include ~ 3 keV to 50 MeV and computes the resulting electron distribution function at each cell in a two dimensional spatial grid. The energetic electrons are allowed to transport, scatter, and experience Coulombic drag. Swarm models the transport of lower energy electrons in air, spanning 0.005 eV to 30 keV. The swarm code performs a full 2-D solution to the Boltzmann equation for electrons in the presence of an applied electric field. Over this energy range the relevant processes to be tracked are elastic scattering, three body attachment, two body attachment, rotational excitation, vibrational excitation, electronic excitation, and ionization. All of these occur due to collisions between the electrons and neutral bodies in air. The Rad code solves the full radiation transfer equation in the energy range of 1 keV to 100 MeV. It includes effects of photo-absorption, Compton scattering, and pair-production. All of these codes employ a spherical coordinate system in momentum space and a cylindrical coordinate system in configuration space. The "z" axis of the momentum and configuration spaces is assumed to be parallel and we are currently also assuming complete spatial symmetry around the "z" axis. Benchmarking for each of these codes will be discussed as well as the way forward towards an integrated modern EMP code.

  20. 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

  1. 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.

  2. 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.

  3. 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

  4. CATHARE code development and assessment methodologies

    International Nuclear Information System (INIS)

    Micaelli, J.C.; Barre, F.; Bestion, D.

    1995-01-01

    The CATHARE thermal-hydraulic code has been developed jointly by Commissariat a l'Energie Atomique (CEA), Electricite de France (EdF), and Framatorne for safety analysis. Since the beginning of the project (September 1979), development and assessment activities have followed a methodology supported by two series of experimental tests: separate effects tests and integral effects tests. The purpose of this paper is to describe this methodology, the code assessment status, and the evolution to take into account two new components of this program: the modeling of three-dimensional phenomena and the requirements of code uncertainty evaluation

  5. 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

  6. Computer code development plant for SMART design

    International Nuclear Information System (INIS)

    Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H.

    1999-03-01

    In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the

  7. Computer code development plant for SMART design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Kyoo Hwan; Choi, S.; Cho, B.H.; Kim, K.K.; Lee, J.C.; Kim, J.P.; Kim, J.H.; Chung, M.; Kang, D.J.; Chang, M.H

    1999-03-01

    In accordance with the localization plan for the nuclear reactor design driven since the middle of 1980s, various computer codes have been transferred into the korea nuclear industry through the technical transfer program from the worldwide major pressurized water reactor supplier or through the international code development program. These computer codes have been successfully utilized in reactor and reload core design works. As the results, design- related technologies have been satisfactorily accumulated. However, the activities for the native code development activities to substitute the some important computer codes of which usages are limited by the original technique owners have been carried out rather poorly. Thus, it is most preferentially required to secure the native techniques on the computer code package and analysis methodology in order to establish the capability required for the independent design of our own model of reactor. Moreover, differently from the large capacity loop-type commercial reactors, SMART (SYSTEM-integrated Modular Advanced ReacTor) design adopts a single reactor pressure vessel containing the major primary components and has peculiar design characteristics such as self-controlled gas pressurizer, helical steam generator, passive residual heat removal system, etc. Considering those peculiar design characteristics for SMART, part of design can be performed with the computer codes used for the loop-type commercial reactor design. However, most of those computer codes are not directly applicable to the design of an integral reactor such as SMART. Thus, they should be modified to deal with the peculiar design characteristics of SMART. In addition to the modification efforts, various codes should be developed in several design area. Furthermore, modified or newly developed codes should be verified their reliability through the benchmarking or the test for the object design. Thus, it is necessary to proceed the design according to the

  8. SCDAP/RELAP5/MOD3 code development

    International Nuclear Information System (INIS)

    Allison, C.M.; Siefken, J.L.; Coryell, E.W.

    1992-01-01

    The SCOAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the US Nuclear Regulatory Commission (NRC). Code development activities are currently focused on three main areas - (a) code usability, (b) early phase melt progression model improvements, and (c) advanced reactor thermal-hydraulic model extensions. This paper describes the first two activities. A companion paper describes the advanced reactor model improvements being performed under RELAP5/MOD3 funding

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  10. Development of ADINA-J-integral code

    International Nuclear Information System (INIS)

    Kurihara, Ryoichi

    1988-07-01

    A general purpose finite element program ADINA (Automatic Dynamic Incremental Nonlinear Analysis), which was developed by Bathe et al., was revised to be able to calculate the J- and J-integral. This report introduced the numerical method to add this capability to the code, and the evaluation of the revised ADINA-J code by using a few of examples of the J estimation model, i.e. a compact tension specimen, a center cracked panel subjected to dynamic load, and a thick shell cylinder having inner axial crack subjected to thermal load. The evaluation testified the function of the revised code. (author)

  11. Development of HTGR plant dynamics simulation code

    International Nuclear Information System (INIS)

    Ohashi, Kazutaka; Tazawa, Yujiro; Mitake, Susumu; Suzuki, Katsuo.

    1987-01-01

    Plant dynamics simulation analysis plays an important role in the design work of nuclear power plant especially in the plant safety analysis, control system analysis, and transient condition analysis. The authors have developed the plant dynamics simulation code named VESPER, which is applicable to the design work of High Temperature Engineering Test Reactor, and have been improving the code corresponding to the design changes made in the subsequent design works. This paper describes the outline of VESPER code and shows its sample calculation results selected from the recent design work. (author)

  12. Status of development and verification of the CTFD code FLUBOX

    International Nuclear Information System (INIS)

    Graf, U.; Paradimitriou, P.

    2004-01-01

    The Computational Two-Fluid Dynamics (CTFD) code FLUBOX is developed at GRS for the multidimensional simulation of two-phase flows. FLUBOX will also be used as a multidimensional module for the German system code ATHLET. The Benchmark test cases of the European ASTAR project were used to verify the ability of the code FLUBOX to calculate typical two-phase flow phenomena and conditions: void and pressure wave propagation, phase transitions, countercurrent flows, sharp interface movements, compressible (vapour) and nearly incompressible (water) conditions, thermal and mechanical non-equilibrium, stiff source terms due to mass and heat transfer between the phases. Realistic simulations of two-phase require beside the pure conservation equations additional transport equations for the interfacial area, turbulent energy and dissipation. A transport equation for the interfacial area density covering the whole two-phase flow range is in development. First validation calculations are presented in the paper. Turbulent shear stress for two-phase flows will be modelled by the development of transport equations for the turbulent kinetic energy and the turbulent dissipation rate. The development of the transport equations is mainly based on first principles on bubbles or drops and is largely free from empiricism. (author)

  13. 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

  14. 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.

  15. 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

  16. 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...

  17. 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

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

    Directory of Open Access Journals (Sweden)

    Ilić Radovan D.

    2002-01-01

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

  19. 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.

  20. 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)

  1. 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)

  2. Optics code development at Los Alamos

    International Nuclear Information System (INIS)

    Mottershead, C.T.; Lysenko, W.P.

    1988-01-01

    This paper is an overview of part of the beam optics code development effort in the Accelerator Technology Division at Los Alamos National Laboratory. The aim of this effort is to improve our capability to design advanced beam optics systems. The work reported is being carried out by a collaboration of permanent staff members, visiting consultants, and student research assistants. The main components of the effort are building a new framework of common supporting utilities and software tools to facilitate further development. research and development on basic computational techniques in classical mechanics and electrodynamics, and evaluation and comparison of existing beam optics codes, and support for their continuing development

  3. Optics code development at Los Alamos

    International Nuclear Information System (INIS)

    Mottershead, C.T.; Lysenko, W.P.

    1988-01-01

    This paper is an overview of part of the beam optics code development effort in the Accelerator Technology Division at Los Alamos National Laboratory. The aim of this effort is to improve our capability to design advanced beam optics systems. The work reported is being carried out by a collaboration of permanent staff members, visiting consultants, and student research assistants. The main components of the effort are: building a new framework of common supporting utilities and software tools to facilitate further development; research and development on basic computational techniques in classical mechanics and electrodynamics; and evaluation and comparison of existing beam optics codes, and support for their continuing development. 17 refs

  4. Development of Visual CINDER Code with Visual C⧣.NET

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Oyeon [Institute for Modeling and Simulation Convergence, Daegu (Korea, Republic of)

    2016-10-15

    CINDER code, CINDER' 90 or CINDER2008 that is integrated with the Monte Carlo code, MCNPX, is widely used to calculate the inventory of nuclides in irradiated materials. The MCNPX code provides decay processes to the particle transport scheme that traditionally only covered prompt processes. The integration schemes serve not only the reactor community (MCNPX burnup) but also the accelerator community as well (residual production information). The big benefit for providing these options lies in the easy cross comparison of the transmutation codes since the calculations are based on exactly the same material, neutron flux and isotope production/destruction inputs. However, it is just frustratingly cumbersome to use. In addition, multiple human interventions may increase the possibility of making errors. The number of significant digits in the input data varies in steps, which may cause big errors for highly nonlinear problems. Thus, it is worthwhile to find a new way to wrap all the codes and procedures in one consistent package which can provide ease of use. The visual CINDER code development is underway with visual C .NET framework. It provides a few benefits for the atomic transmutation simulation with CINDER code. A few interesting and useful properties of visual C .NET framework are introduced. We also showed that the wrapper could make the simulation accurate for highly nonlinear transmutation problems and also increase the possibility of direct combination a radiation transport code MCNPX with CINDER code. Direct combination of CINDER with MCNPX in a wrapper will provide more functionalities for the radiation shielding and prevention study.

  5. Development of Visual CINDER Code with Visual C⧣.NET

    International Nuclear Information System (INIS)

    Kim, Oyeon

    2016-01-01

    CINDER code, CINDER' 90 or CINDER2008 that is integrated with the Monte Carlo code, MCNPX, is widely used to calculate the inventory of nuclides in irradiated materials. The MCNPX code provides decay processes to the particle transport scheme that traditionally only covered prompt processes. The integration schemes serve not only the reactor community (MCNPX burnup) but also the accelerator community as well (residual production information). The big benefit for providing these options lies in the easy cross comparison of the transmutation codes since the calculations are based on exactly the same material, neutron flux and isotope production/destruction inputs. However, it is just frustratingly cumbersome to use. In addition, multiple human interventions may increase the possibility of making errors. The number of significant digits in the input data varies in steps, which may cause big errors for highly nonlinear problems. Thus, it is worthwhile to find a new way to wrap all the codes and procedures in one consistent package which can provide ease of use. The visual CINDER code development is underway with visual C .NET framework. It provides a few benefits for the atomic transmutation simulation with CINDER code. A few interesting and useful properties of visual C .NET framework are introduced. We also showed that the wrapper could make the simulation accurate for highly nonlinear transmutation problems and also increase the possibility of direct combination a radiation transport code MCNPX with CINDER code. Direct combination of CINDER with MCNPX in a wrapper will provide more functionalities for the radiation shielding and prevention study

  6. 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)

  7. 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.

  8. 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

  9. 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

  10. 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

  11. SQA of finite element method (FEM) codes used for analyses of pit storage/transport packages

    Energy Technology Data Exchange (ETDEWEB)

    Russel, E. [Lawrence Livermore National Lab., CA (United States)

    1997-11-01

    This report contains viewgraphs on the software quality assurance of finite element method codes used for analyses of pit storage and transport projects. This methodology utilizes the ISO 9000-3: Guideline for application of 9001 to the development, supply, and maintenance of software, for establishing well-defined software engineering processes to consistently maintain high quality management approaches.

  12. 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)

  13. Development of 2-d cfd code

    International Nuclear Information System (INIS)

    Mirza, S.A.

    1999-01-01

    In the present study, a two-dimensional computer code has been developed in FORTRAN using CFD technique, which is basically a numerical scheme. This computer code solves the Navier Stokes equations and continuity equation to find out the velocity and pressure fields within a given domain. This analysis has been done for the developed within a square cavity driven by the upper wall which has become a bench mark for testing and comparing the newly developed numerical schemes. Before to handle this task, different one-dimensional cases have been studied by CFD technique and their FORTRAN programs written. The cases studied are Couette flow, Poiseuille flow with and without using symmetric boundary condition. Finally a comparison between CFD results and analytical results has also been made. For the cavity flow the results from the developed code have been obtained for different Reynolds numbers which are finally presented in the form of velocity vectors. The comparison of the developed code results have been made with the results obtained from the share ware version of a commercially available code for Reynolds number of 10.0. The disagreement in the results quantitatively and qualitatively at some grid points of the calculation domain have been discussed and future recommendations in this regard have also been made. (author)

  14. 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

  15. Development of the point-depletion code DEPTH

    International Nuclear Information System (INIS)

    She, Ding; Wang, Kan; Yu, Ganglin

    2013-01-01

    Highlights: ► The DEPTH code has been developed for the large-scale depletion system. ► DEPTH uses the data library which is convenient to couple with MC codes. ► TTA and matrix exponential methods are implemented and compared. ► DEPTH is able to calculate integral quantities based on the matrix inverse. ► Code-to-code comparisons prove the accuracy and efficiency of DEPTH. -- Abstract: The burnup analysis is an important aspect in reactor physics, which is generally done by coupling of transport calculations and point-depletion calculations. DEPTH is a newly-developed point-depletion code of handling large burnup depletion systems and detailed depletion chains. For better coupling with Monte Carlo transport codes, DEPTH uses data libraries based on the combination of ORIGEN-2 and ORIGEN-S and allows users to assign problem-dependent libraries for each depletion step. DEPTH implements various algorithms of treating the stiff depletion systems, including the Transmutation trajectory analysis (TTA), the Chebyshev Rational Approximation Method (CRAM), the Quadrature-based Rational Approximation Method (QRAM) and the Laguerre Polynomial Approximation Method (LPAM). Three different modes are supported by DEPTH to execute the decay, constant flux and constant power calculations. In addition to obtaining the instantaneous quantities of the radioactivity, decay heats and reaction rates, DEPTH is able to calculate the integral quantities by a time-integrated solver. Through calculations compared with ORIGEN-2, the validity of DEPTH in point-depletion calculations is proved. The accuracy and efficiency of depletion algorithms are also discussed. In addition, an actual pin-cell burnup case is calculated to illustrate the DEPTH code performance in coupling with the RMC Monte Carlo code

  16. Health Code Number (HCN) Development Procedure

    Energy Technology Data Exchange (ETDEWEB)

    Petrocchi, Rocky; Craig, Douglas K.; Bond, Jayne-Anne; Trott, Donna M.; Yu, Xiao-Ying

    2013-09-01

    This report provides the detailed description of health code numbers (HCNs) and the procedure of how each HCN is assigned. It contains many guidelines and rationales of HCNs. HCNs are used in the chemical mixture methodology (CMM), a method recommended by the department of energy (DOE) for assessing health effects as a result of exposures to airborne aerosols in an emergency. The procedure is a useful tool for proficient HCN code developers. Intense training and quality assurance with qualified HCN developers are required before an individual comprehends the procedure to develop HCNs for DOE.

  17. Development of Coolant Radioactivity Interpretation Code

    International Nuclear Information System (INIS)

    Kim, Kiyoung; Jung, Youngsuk; Kim, Kyounghyun; Kim, Jangwook

    2013-01-01

    In Korea, the coolant radioactivity analysis has been performed by using the computer codes of foreign companies such as CADE (Westinghouse), IODYNE and CESIUM (ABB-CE). However, these computer codes are too conservative and have involved considerable errors. Furthermore, since these codes are DOS-based program, their easy operability is not satisfactory. Therefore it is required development of an enhanced analysis algorithm applying an analytical method reflecting the change of operational environments of domestic nuclear power plants and a fuel failure evaluation software considering user' conveniences. We have developed a nuclear fuel failure evaluation code able to estimate the number of failed fuel rods and the burn-up of failed fuels during nuclear power plant operation cycle. A Coolant Radio-activity Interpretation Code (CRIC) for LWR has been developed as the output of the project 'Development of Fuel Reliability Enhanced Technique' organized by Korea Institute of Energy Technology Evaluation and Planning (KETEP). The CRIC is Windows based-software able to evaluate the number of failed fuel rods and the burn-up of failed fuel region by analyzing coolant radioactivity of LWR in operation. The CRIC is based on the model of fission products release commonly known as 'three region model' (pellet region, gap region, and coolant region), and we are verifying the CRIC results based on the cases of domestic fuel failures. CRIC users are able to estimate the number of failed fuel rods, burn-up and regions of failed fuel considered enrichment and power distribution of fuel region by using operational cycle data, coolant activity data, fuel loading pattern, Cs-134/Cs-137 ratio according to burn-up and U-235 enrichment provided in the code. Due to development of the CRIC, it is secured own unique fuel failure evaluation code. And, it is expected to have the following significant meaning. This is that the code reflecting a proprietary technique for quantitatively

  18. 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

  19. 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

  20. Computer codes in nuclear safety, radiation transport and dosimetry

    International Nuclear Information System (INIS)

    Bordy, J.M.; Kodeli, I.; Menard, St.; Bouchet, J.L.; Renard, F.; Martin, E.; Blazy, L.; Voros, S.; Bochud, F.; Laedermann, J.P.; Beaugelin, K.; Makovicka, L.; Quiot, A.; Vermeersch, F.; Roche, H.; Perrin, M.C.; Laye, F.; Bardies, M.; Struelens, L.; Vanhavere, F.; Gschwind, R.; Fernandez, F.; Quesne, B.; Fritsch, P.; Lamart, St.; Crovisier, Ph.; Leservot, A.; Antoni, R.; Huet, Ch.; Thiam, Ch.; Donadille, L.; Monfort, M.; Diop, Ch.; Ricard, M.

    2006-01-01

    The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations

  1. 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.

  2. Transport infrastructure development in China

    Directory of Open Access Journals (Sweden)

    Bouraima Mouhamed Bayane

    2017-05-01

    Full Text Available This paper reviews the historical configuration process of transportation systems in China and examines the relationship between economic development and transport system at three different levels. The current status of transport infrastructure system development in China is summarized at national and regional level. The investment trends for transport infrastructure in China are also depicted. The keys issues relating to government initiatives are presented.

  3. Development of the versatile reactor analysis code system, MARBLE2

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Jin, Tomoyuki; Hazama, Taira; Hirai, Yasushi

    2015-07-01

    The second version of the versatile reactor analysis code system, MARBLE2, has been developed. A lot of new functions have been added in MARBLE2 by using the base technology developed in the first version (MARBLE1). Introducing the remaining functions of the conventional code system (JOINT-FR and SAGEP-FR), MARBLE2 enables one to execute almost all analysis functions of the conventional code system with the unified user interfaces of its subsystem, SCHEME. In particular, the sensitivity analysis functionality is available in MARBLE2. On the other hand, new built-in solvers have been developed, and existing ones have been upgraded. Furthermore, some other analysis codes and libraries developed in JAEA have been consolidated and prepared in SCHEME. In addition, several analysis codes developed in the other institutes have been additionally introduced as plug-in solvers. Consequently, gamma-ray transport calculation and heating evaluation become available. As for another subsystem, ORPHEUS, various functionality updates and speed-up techniques have been applied based on user experience of MARBLE1 to enhance its usability. (author)

  4. 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.)

  5. 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

  6. 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

  7. Developing an Australian code of construction ethics

    Directory of Open Access Journals (Sweden)

    Sean Francis McCarthy

    2012-05-01

    Full Text Available This article looks at the increasing need to consider the role of ethics in construction. The industry, historically, has been challenged by allegations of a serious shortfall in ethical standards. Only limited attempts to date in Australia have been made to address that concern. Any ethical analysis should consider the definition of ethics and its historical development. This paper considers major historical developments in ethical thinking as well as contemporary thinking on ethics for professional sub-sets. A code could be developed specific to construction. Current methods of addressing ethics in construction and in other industries are also reviewed. This paper argues that developing a code of ethics, supported by other measures is the way forward. The author’s aim is to promote further discussion and promote the drafting of a code. This paper includes a summary of other ethical codes that may provide a starting point. The time for reform is upon us, and there is an urgent need for an independent body to take the lead, for fear of floundering and having only found ‘another debating topic’ (Uff 2006.

  8. 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

  9. 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

  10. Development of a coupled code system based on system transient code, RETRAN, and 3-D neutronics code, MASTER

    International Nuclear Information System (INIS)

    Kim, K. D.; Jung, J. J.; Lee, S. W.; Cho, B. O.; Ji, S. K.; Kim, Y. H.; Seong, C. K.

    2002-01-01

    A coupled code system of RETRAN/MASTER has been developed for best-estimate simulations of interactions between reactor core neutron kinetics and plant thermal-hydraulics by incorporation of a 3-D reactor core kinetics analysis code, MASTER into system transient code, RETRAN. The soundness of the consolidated code system is confirmed by simulating the MSLB benchmark problem developed to verify the performance of a coupled kinetics and system transient codes by OECD/NEA

  11. Development of the neutron-transport code TransRay and studies on the two- and three-dimensional calculation of effective group cross sections; Entwicklung des Neutronentransportcodes TransRay und Untersuchungen zur zwei- und dreidimensionalen Berechnung effektiver Gruppenwirkungsquerschnitte

    Energy Technology Data Exchange (ETDEWEB)

    Beckert, C.

    2007-12-19

    Conventionally the data preparation of the neutron cross sections for reactor-core calculations pursues with 2D cell codes. Aim of this thesis was, to develop a 3D cell code, to study with this code 3D effects, and to evaluate the necessarity of a 3D data preparation of the neutron cross sections. For the calculation of the neutron transport the method of the first-collision probabilities, which are calculated with the ray-tracing method, was chosen. The mathematical algorithms were implemented in the 2D/3D cell code TransRay. For the geometry part of the program the geometry module of a Monte Carlo code was used.The ray tracing in 3D was parallelized because of the high computational time. The program TransRay was verified on 2D test problems. For a reference pressured-water reactor following 3D problems were studied: A partly immersed control rod and void (vacuum or steam) around a fuel rod as model of a steam void. All problems were for comparison calculated also with the programs HELIOS(2D) and MCNP(3D). The dependence of the multiplication factor and the averaged two-group cross section on the immersion depth of the control rod respectively of the height of the steam void were studied. The 3D-calculated two-group cross sections were compared with three conventional approximations: Linear interpolation, interpolation with flux weighting, and homogenization, At the 3D problem of the control rod it was shown that the interpolation with flux weighting is a good approximation. Therefore here a 3D data preparation is not necessary. At the test case of the single control rod, which is surrounded by the void, the three approximation for the two-group cross sections were proved as unsufficient. Therefore a 3D data preparation is necessary. The single fuel-rod cell with void can be considered as the limiting case of a reactor, in which a phase interface has been formed. [German] Standardmaessig erfolgt die Datenaufbereitung der Neutronenwirkungsquerschnitte fuer

  12. 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)

  13. Development of chemical equilibrium analysis code 'CHEEQ'

    International Nuclear Information System (INIS)

    Nagai, Shuichiro

    2006-08-01

    'CHEEQ' code which calculates the partial pressure and the mass of the system consisting of ideal gas and pure condensed phase compounds, was developed. Characteristics of 'CHEEQ' code are as follows. All the chemical equilibrium equations were described by the formation reactions from the mono-atomic gases in order to simplify the code structure and input preparation. Chemical equilibrium conditions, Σν i μ i =0 for the gaseous compounds and precipitated condensed phase compounds and Σν i μ i > 0 for the non-precipitated condensed phase compounds, were applied. Where, ν i and μ i are stoichiometric coefficient and chemical potential of component i. Virtual solid model was introduced to perform the calculation of constant partial pressure condition. 'CHEEQ' was consisted of following 3 parts, (1) analysis code, zc132. f. (2) thermodynamic data base, zmdb01 and (3) input data file, zindb. 'CHEEQ' code can calculate the system which consisted of elements (max.20), condensed phase compounds (max.100) and gaseous compounds. (max.200). Thermodynamic data base, zmdb01 contains about 1000 elements and compounds, and 200 of them were Actinide elements and their compounds. This report describes the basic equations, the outline of the solution procedure and instructions to prepare the input data and to evaluate the calculation results. (author)

  14. GRAVE: An Interactive Geometry Construction and Visualization Software System for the TORT Nuclear Radiation Transport Code

    International Nuclear Information System (INIS)

    Blakeman, E.D.

    2000-01-01

    A software system, GRAVE (Geometry Rendering and Visual Editor), has been developed at the Oak Ridge National Laboratory (ORNL) to perform interactive visualization and development of models used as input to the TORT three-dimensional discrete ordinates radiation transport code. Three-dimensional and two-dimensional visualization displays are included. Display capabilities include image rotation, zoom, translation, wire-frame and translucent display, geometry cuts and slices, and display of individual component bodies and material zones. The geometry can be interactively edited and saved in TORT input file format. This system is an advancement over the current, non-interactive, two-dimensional display software. GRAVE is programmed in the Java programming language and can be implemented on a variety of computer platforms. Three- dimensional visualization is enabled through the Visualization Toolkit (VTK), a free-ware C++ software library developed for geometric and data visual display. Future plans include an extension of the system to read inputs using binary zone maps and combinatorial geometry models containing curved surfaces, such as those used for Monte Carlo code inputs. Also GRAVE will be extended to geometry visualization/editing for the DORT two-dimensional transport code and will be integrated into a single GUI-based system for all of the ORNL discrete ordinates transport codes

  15. BBU code development for high-power microwave generators

    International Nuclear Information System (INIS)

    Houck, T.L.; Westenskow, G.A.; Yu, S.S.

    1992-01-01

    We are developing a two-dimensional, time-dependent computer code for the simulation of transverse instabilities in support of relativistic klystron-two beam accelerator research at LLNL. The code addresses transient effects as well as both cumulative and regenerative beam breakup modes. Although designed specifically for the transport of high current (kA) beams through traveling-wave structures, it is applicable to devices consisting of multiple combinations of standing-wave, traveling-wave, and induction accelerator structures. In this paper we compare code simulations to analytical solutions for the case where there is no rf coupling between cavities, to theoretical scaling parameters for coupled cavity structures, and to experimental data involving beam breakup in the two traveling-wave output structure of our microwave generator. (Author) 4 figs., tab., 5 refs

  16. 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)

  17. 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)

  18. Development of Evaluation Code for MUF Uncertainty

    International Nuclear Information System (INIS)

    Won, Byung Hee; Han, Bo Young; Shin, Hee Sung; Ahn, Seong-Kyu; Park, Geun-Il; Park, Se Hwan

    2015-01-01

    Material Unaccounted For (MUF) is the material balance evaluated by measured nuclear material in a Material Balance Area (MBA). Assuming perfect measurements and no diversion from a facility, one can expect a zero MUF. However, non-zero MUF is always occurred because of measurement uncertainty even though the facility is under normal operation condition. Furthermore, there are many measurements using different equipment at various Key Measurement Points (KMPs), and the MUF uncertainty is affected by errors of those measurements. Evaluating MUF uncertainty is essentially required to develop safeguards system including nuclear measurement system in pyroprocessing, which is being developed for reducing radioactive waste from spent fuel in Korea Atomic Energy Research Institute (KAERI). The evaluation code for analyzing MUF uncertainty has been developed and it was verified using sample problem from the IAEA reference. MUF uncertainty can be simply and quickly calculated by using this evaluation code which is made based on graphical user interface for user friendly. It is also expected that the code will make the sensitivity analysis on the MUF uncertainty for the various safeguards systems easy and more systematic. It is suitable for users who want to evaluate the conventional safeguards system as well as to develop a new system for developing facilities

  19. Development of Evaluation Code for MUF Uncertainty

    Energy Technology Data Exchange (ETDEWEB)

    Won, Byung Hee; Han, Bo Young; Shin, Hee Sung; Ahn, Seong-Kyu; Park, Geun-Il; Park, Se Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Material Unaccounted For (MUF) is the material balance evaluated by measured nuclear material in a Material Balance Area (MBA). Assuming perfect measurements and no diversion from a facility, one can expect a zero MUF. However, non-zero MUF is always occurred because of measurement uncertainty even though the facility is under normal operation condition. Furthermore, there are many measurements using different equipment at various Key Measurement Points (KMPs), and the MUF uncertainty is affected by errors of those measurements. Evaluating MUF uncertainty is essentially required to develop safeguards system including nuclear measurement system in pyroprocessing, which is being developed for reducing radioactive waste from spent fuel in Korea Atomic Energy Research Institute (KAERI). The evaluation code for analyzing MUF uncertainty has been developed and it was verified using sample problem from the IAEA reference. MUF uncertainty can be simply and quickly calculated by using this evaluation code which is made based on graphical user interface for user friendly. It is also expected that the code will make the sensitivity analysis on the MUF uncertainty for the various safeguards systems easy and more systematic. It is suitable for users who want to evaluate the conventional safeguards system as well as to develop a new system for developing facilities.

  20. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

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

    International Nuclear Information System (INIS)

    Procassini, Richard; Brantley, Patrick; Dawson, Shawn

    2010-01-01

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

  3. 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

  4. Development of steam explosion simulation code JASMINE

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kiyofumi; Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nagano, Katsuhiro; Araki, Kazuhiro

    1995-11-01

    A steam explosion is considered as a phenomenon which possibly threatens the integrity of the containment vessel of a nuclear power plant in a severe accident condition. A numerical calculation code JASMINE (JAeri Simulator for Multiphase INteraction and Explosion) purposed to simulate the whole process of steam explosions has been developed. The premixing model is based on a multiphase flow simulation code MISTRAL by Fuji Research Institute Co. In JASMINE code, the constitutive equations and the flow regime map are modified for the simulation of premixing related phenomena. The numerical solution method of the original code is succeeded, i.e. the basic equations are discretized semi-implicitly, BCGSTAB method is used for the matrix solver to improve the stability and convergence, also TVD scheme is applied to capture a steep phase distribution accurately. Test calculations have been performed for the conditions correspond to the experiments by Gilbertson et al. and Angelini et al. in which mixing of solid particles and water were observed in iso-thermal condition and with boiling, respectively. (author).

  5. Development of a tritium dispersion code

    International Nuclear Information System (INIS)

    Bell, R.P.; Davis, M.W.; Joseph, S.; Wong, K.Y.

    1985-01-01

    This paper describes the development and verification of a computer code designed to calculate the radiation dose to man following acute or chronic atmospheric releases of tritium gas and oxide from a point source. The Ontario Hydro Tritium Dispersion Code calculates tritium concentrations in air, soil, and vegetation and doses to man resulting from inhalation/immersion and ingestion of food, milk meat and water. The deposition of HT to soil, conversion of HT to HTO by soil enzymes and resuspension of HTO to air have been incorporated into the terrestrial compartment model and are unique features of the code. Sensitivity analysis has identified the HT deposition velocity and the equivalent water depth of the vegetation compartment as two parameters which have a strong influence on dose calculations. Tritium concentrations in vegetation and soil calculated by the code were in reasonable agreement with experimental results. The radiological significance of including the mechanisms of HT to HTO conversion and resuspension of HTO to air is illustrated

  6. Development of steam explosion simulation code JASMINE

    International Nuclear Information System (INIS)

    Moriyama, Kiyofumi; Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Sugimoto, Jun; Nagano, Katsuhiro; Araki, Kazuhiro.

    1995-11-01

    A steam explosion is considered as a phenomenon which possibly threatens the integrity of the containment vessel of a nuclear power plant in a severe accident condition. A numerical calculation code JASMINE (JAeri Simulator for Multiphase INteraction and Explosion) purposed to simulate the whole process of steam explosions has been developed. The premixing model is based on a multiphase flow simulation code MISTRAL by Fuji Research Institute Co. In JASMINE code, the constitutive equations and the flow regime map are modified for the simulation of premixing related phenomena. The numerical solution method of the original code is succeeded, i.e. the basic equations are discretized semi-implicitly, BCGSTAB method is used for the matrix solver to improve the stability and convergence, also TVD scheme is applied to capture a steep phase distribution accurately. Test calculations have been performed for the conditions correspond to the experiments by Gilbertson et al. and Angelini et al. in which mixing of solid particles and water were observed in iso-thermal condition and with boiling, respectively. (author)

  7. Towards Product Lining Model-Driven Development Code Generators

    OpenAIRE

    Roth, Alexander; Rumpe, Bernhard

    2015-01-01

    A code generator systematically transforms compact models to detailed code. Today, code generation is regarded as an integral part of model-driven development (MDD). Despite its relevance, the development of code generators is an inherently complex task and common methodologies and architectures are lacking. Additionally, reuse and extension of existing code generators only exist on individual parts. A systematic development and reuse based on a code generator product line is still in its inf...

  8. SCDAP/RELAP5/MOD3 code development and assessment

    International Nuclear Information System (INIS)

    Allison, C.M.; Heath, C.H.; Siefken, L.J.; Hohorst, J.K.

    1991-01-01

    The SCDAP/RELAP5/MOD3 computer code is designed to describe the overall reactor coolant system (RCS) thermal-hydraulic response, core damage progression, and fission product release and transport during severe accidents. The code is being developed at the Idaho National Engineering Laboratory (INEL) under the primary sponsorship of the Office of Nuclear Regulatory Research of the Nuclear Regulatory Commission (NRC). SCDAP/RELAP5/MOD3, created in January, 1991, is the result of merging RELAP5/MOD3 with SCDAP and TRAP-MELT models from SCDAP/RELAP5/MOD2.5. The RELAP5 models calculate the overall RCS thermal-hydraulics, control system interactions, reactor kinetics, and the transport of noncondensible gases, fission products, and aerosols. The SCDAP models calculate the damage progression in the core structures, the formation, heatup, and melting of debris, and the creep rupture failure of the lower head and other RCS structures. The TRAP-MELT models calculate the deposition of fission products upon aerosols or structural surfaces; the formation, growth, or deposition of aerosols; and the evaporation of species from surfaces. The systematic assessment of modeling uncertainties in SCDAP/RELAP5 code is currently underway. This assessment includes (a) the evaluation of code-to-data comparisons using stand-alone SCDAP and SCDAP/RELAP5/MOD3, (b) the estimation of modeling and experimental uncertainties, and (c) the determination of the influence of those uncertainties on predicted severe accident behavior

  9. Development and validation of a new solver based on the interfacial area transport equation for the numerical simulation of sub-cooled boiling with OpenFOAM CFD code for nuclear safety applications

    Energy Technology Data Exchange (ETDEWEB)

    Alali, Abdullah

    2014-02-21

    The one-group interfacial area transport equation has been coupled to a wall heat flux partitioning model in the framework of two-phase Eulerian approach using the OpenFOAM CFD code for better prediction of subcooled boiling phenomena which is essential for safety analysis of nuclear reactors. The interfacial area transport equation has been modified to include the effect of bubble nucleation at the wall and condensation by subcooled liquid in the bulk that governs the non-uniform bubble size distribution.

  10. Development and validation of a new solver based on the interfacial area transport equation for the numerical simulation of sub-cooled boiling with OpenFOAM CFD code for nuclear safety applications

    International Nuclear Information System (INIS)

    Alali, Abdullah

    2014-01-01

    The one-group interfacial area transport equation has been coupled to a wall heat flux partitioning model in the framework of two-phase Eulerian approach using the OpenFOAM CFD code for better prediction of subcooled boiling phenomena which is essential for safety analysis of nuclear reactors. The interfacial area transport equation has been modified to include the effect of bubble nucleation at the wall and condensation by subcooled liquid in the bulk that governs the non-uniform bubble size distribution.

  11. 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

  12. GEOS Code Development Road Map - May, 2013

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Scott [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Settgast, Randolph [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fu, Pengcheng [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Antoun, Tarabay [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ryerson, F. J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2013-05-03

    GEOS is a massively parallel computational framework designed to enable HPC-based simulations of subsurface reservoir stimulation activities with the goal of optimizing current operations and evaluating innovative stimulation methods. GEOS will enable coupling of different solvers associated with the various physical processes occurring during reservoir stimulation in unique and sophisticated ways, adapted to various geologic settings, materials and stimulation methods. The overall architecture of the framework includes consistent data structures and will allow incorporation of additional physical and materials models as demanded by future applications. Along with predicting the initiation, propagation and reactivation of fractures, GEOS will also generate a seismic source term that can be linked with seismic wave propagation codes to generate synthetic microseismicity at surface and downhole arrays. Similarly, the output from GEOS can be linked with existing fluid/thermal transport codes. GEOS can also be linked with existing, non-intrusive uncertainty quantification schemes to constrain uncertainty in its predictions and sensitivity to the various parameters describing the reservoir and stimulation operations. We anticipate that an implicit-explicit 3D version of GEOS, including a preliminary seismic source model, will be available for parametric testing and validation against experimental and field data by Oct. 1, 2013.

  13. 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

  14. Developments in tokamak transport modeling

    International Nuclear Information System (INIS)

    Houlberg, W.A.; Attenberger; Lao, L.L.

    1981-01-01

    A variety of numerical methods for solving the time-dependent fluid transport equations for tokamak plasmas is presented. Among the problems discussed are techniques for solving the sometimes very stiff parabolic equations for particle and energy flow, treating convection-dominated energy transport that leads to large cell Reynolds numbers, optimizing the flow of a code to reduce the time spent updating the particle and energy source terms, coupling the one-dimensional (1-D) flux-surface-averaged fluid transport equations to solutions of the 2-D Grad-Shafranov equation for the plasma geometry, handling extremely fast transient problems such as internal MHD disruptions and pellet injection, and processing the output to summarize the physics parameters over the potential operating regime for reactors. Emphasis is placed on computational efficiency in both computer time and storage requirements

  15. 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

  16. 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

  17. Using Coding Apps to Support Literacy Instruction and Develop Coding Literacy

    Science.gov (United States)

    Hutchison, Amy; Nadolny, Larysa; Estapa, Anne

    2016-01-01

    In this article the authors present the concept of Coding Literacy and describe the ways in which coding apps can support the development of Coding Literacy and disciplinary and digital literacy skills. Through detailed examples, we describe how coding apps can be integrated into literacy instruction to support learning of the Common Core English…

  18. Development of Nuclear Energy Security Code

    International Nuclear Information System (INIS)

    Shimamura, Takehisa; Suzuki, Atsuyuki; Okubo, Hiroo; Kikuchi, Masahiro.

    1990-01-01

    In establishing of the nuclear fuel cycle in Japan that have a vulnerability in own energy structure, an effectiveness of energy security should be taken into account as well as an economy based on the balance of supply and demand of nuclear fuels. NMCC develops the 'Nuclear Energy Security Code' which was able to evaluate the effectiveness of energy security. Evaluation method adopted in this code is 'Import Premium' which was proposed in 'World Oil', EMF Report 6. The viewpoints of evaluation are as follows: 1. How much uranium fuel quantity can be reduced by using plutonium fuel? 2. How much a sudden rise of fuel cost can be absorbed by establishing the plutonium cycle beforehand the energy crisis? (author)

  19. 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

  20. Development of a subchannel analysis code MATRA (Ver. α)

    International Nuclear Information System (INIS)

    Yoo, Y. J.; Hwang, D. H.

    1998-04-01

    A subchannel analysis code MATRA-α, an interim version of MATRA, has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-IV-I. This MATRA code is a thermal-hydraulic analysis code based on the subchannel approach for calculating the enthalpy and flow distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. MATRA-α has been provided with an improved structure, various functions, and models to give the more convenient user environment and to increase the code accuracy, various functions, and models to give the more convenient user environment and to increase the code accuracy. Among them, the pressure drop model has been improved to be applied to non-square-lattice rod arrays, and the lateral transport models between adjacent subchannels have been improved to increase the accuracy in predicting two-phase flow phenomena. Also included in this report are the detailed instructions for input data preparation and for auxiliary pre-processors to serve as a guide to those who want to use MATRA-α. In addition, we compared the predictions of MATRA-α with the experimental data on the flow and enthalpy distribution in three sample rod-bundle cases to evaluate the performance of MATRA-α. All the results revealed that the prediction of MATRA-α were better than those of COBRA-IV-I. (author). 16 refs., 1 tab., 13 figs

  1. RIVER-RAD: A computer code for simulating the transport of radionuclides in rivers

    International Nuclear Information System (INIS)

    Hetrick, D.M.; McDowell-Boyer, L.M.; Sjoreen, A.L.; Thorne, D.J.; Patterson, M.R.

    1992-11-01

    A screening-level model, RIVER-RAD, has been developed to assess the potential fate of radionuclides released to rivers. The model is simplified in nature and is intended to provide guidance in determining the potential importance of the surface water pathway, relevant transport mechanisms, and key radionuclides in estimating radiological dose to man. The purpose of this report is to provide a description of the model and a user's manual for the FORTRAN computer code

  2. 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)

  3. Development of a domestically-made system code

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    According to lessons learned from the Fukushima-Daiichi NPP accidents, a new safety standard based on state-of-the-art findings has been established by the Japanese Nuclear Regulation Authority (NRA) and will soon come into force in Japan. In order to ensure a precise response to this movement from a technological point of view, it should be required for safety regulation to develop a new system code with much smaller uncertainty and reinforced simulation capability even in application to beyond-DBAs (BDBAs), as well as with the capability of close coupling to a newly developing severe accident code. Accordingly, development of a new domestically-made system code that incorporates 3-dimensional and 3 or more fluid thermal-hydraulics in tandem with a 3-dimensional neutronics has been started in 2012. In 2012, two branches of development activities, the development of 'main body' and advanced features have been started in parallel for development efficiency. The main body has been started from scratch and the following activities have therefore been performed: 1) development and determination of key principles and methodologies to realize a flexible, extensible and robust platform, 2) determination of requirements definition, 3) start of basic program design and coding and 4) start of a development of prototypical GUI-based pre-post processor. As for the advanced features, the following activities have been performed: 1) development of Phenomena Identification and Ranking Tables (PIRTs) and model capability matrix from normal operations to BDBAs in order to address requirements definition for advanced modeling, 2) development of detailed action plan for modification of field equations, numerical schemes and solvers and 3) start of the program development of field equations with an interfacial area concentration transport equation, a robust solver for condensation induced water hammer phenomena and a versatile Newton-Raphson solver. (author)

  4. Development status of the lattice physics code in COSINE project

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Yu, H.; Li, S.; Liu, Z.; Yan, Y. [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software NEKLS, North Third Ring Road, Beijing 100029 (China)

    2013-07-01

    LATC is an essential part of COSINE code package, which stands for Core and System Integrated Engine for design and analysis. LATC performs 2D multi-group assembly transport calculation and generates few group constants and the required cross-section data for CORE, the core simulator code. LATC is designed to have the capability of modeling the API 000 series assemblies. The development is a continuously improved process. Currently, LATC uses well-proven technology to achieve the key functions. In the next stage, more advanced methods and modules will be implemented. At present, WIMS and WIMS improved format library could be read in LATC code. For resonance calculation, equivalent relation with rational approximations is utilized. For transport calculation, two options are available. One choice is collision probability method in cell homogenization while discrete coordinate method in assembly homogenization, the other is method of characteristics in assembly homogenization directly. For depletion calculation, an improved linear rate 'constant power' depletion method has been developed. (authors)

  5. Development status of the lattice physics code in COSINE project

    International Nuclear Information System (INIS)

    Chen, Y.; Yu, H.; Li, S.; Liu, Z.; Yan, Y.

    2013-01-01

    LATC is an essential part of COSINE code package, which stands for Core and System Integrated Engine for design and analysis. LATC performs 2D multi-group assembly transport calculation and generates few group constants and the required cross-section data for CORE, the core simulator code. LATC is designed to have the capability of modeling the API 000 series assemblies. The development is a continuously improved process. Currently, LATC uses well-proven technology to achieve the key functions. In the next stage, more advanced methods and modules will be implemented. At present, WIMS and WIMS improved format library could be read in LATC code. For resonance calculation, equivalent relation with rational approximations is utilized. For transport calculation, two options are available. One choice is collision probability method in cell homogenization while discrete coordinate method in assembly homogenization, the other is method of characteristics in assembly homogenization directly. For depletion calculation, an improved linear rate 'constant power' depletion method has been developed. (authors)

  6. SWAAM-code development and verification and application to steam generator designs

    International Nuclear Information System (INIS)

    Shin, Y.W.; Valentin, R.A.

    1990-01-01

    This paper describes the family of SWAAM codes which were developed by Argonne National Laboratory to analyze the effects of sodium-water reactions on LMR steam generators. The SWAAM codes were developed as design tools for analyzing various phenomena related to steam generator leaks and the resulting thermal and hydraulic effects on the steam generator and the intermediate heat transport system (IHTS). The paper discusses the theoretical foundations and numerical treatments on which the codes are based, followed by a description of code capabilities and limitations, verification of the codes and applications to steam generator and IHTS designs. 25 refs., 14 figs

  7. Development of cask and transportation system

    International Nuclear Information System (INIS)

    Ro, Seong Gy; Kang, Hee Dong; Lee, Heung Young; Seo, Ki Suk; Koo, Jung Hoe; Jung, Sung Hwan; Yoon, Jung Hyun; Lee, Ju Chan; Bang, Kyung Sik; Baek, Chang Yeol

    1992-03-01

    The major goal of this project is to establish the safe transport system and obtain the necessary data for cask development by during research work for the design and safety test of shipping cask. The analysis technique using computer code for design has been studied in the field of structure, thermal and shielding analysis in this study. And also the test and measurement technology was developed for the measuring system of drop and fire test. It is expected that research activity ensured in this job will enable us to ultilize the basic data for the cask development. (Author)

  8. 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

  9. Development of DUST: A computer code that calculates release rates from a LLW disposal unit

    International Nuclear Information System (INIS)

    Sullivan, T.M.

    1992-01-01

    Performance assessment of a Low-Level Waste (LLW) disposal facility begins with an estimation of the rate at which radionuclides migrate out of the facility (i.e., the disposal unit source term). The major physical processes that influence the source term are water flow, container degradation, waste form leaching, and radionuclide transport. A computer code, DUST (Disposal Unit Source Term) has been developed which incorporates these processes in a unified manner. The DUST code improves upon existing codes as it has the capability to model multiple container failure times, multiple waste form release properties, and radionuclide specific transport properties. Verification studies performed on the code are discussed

  10. 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

  11. Development of Probabilistic Internal Dosimetry Computer Code

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Siwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kwon, Tae-Eun [Korea Institute of Radiological and Medical Sciences, Seoul (Korea, Republic of); Lee, Jai-Ki [Korean Association for Radiation Protection, Seoul (Korea, Republic of)

    2017-02-15

    Internal radiation dose assessment involves biokinetic models, the corresponding parameters, measured data, and many assumptions. Every component considered in the internal dose assessment has its own uncertainty, which is propagated in the intake activity and internal dose estimates. For research or scientific purposes, and for retrospective dose reconstruction for accident scenarios occurring in workplaces having a large quantity of unsealed radionuclides, such as nuclear power plants, nuclear fuel cycle facilities, and facilities in which nuclear medicine is practiced, a quantitative uncertainty assessment of the internal dose is often required. However, no calculation tools or computer codes that incorporate all the relevant processes and their corresponding uncertainties, i.e., from the measured data to the committed dose, are available. Thus, the objective of the present study is to develop an integrated probabilistic internal-dose-assessment computer code. First, the uncertainty components in internal dosimetry are identified, and quantitative uncertainty data are collected. Then, an uncertainty database is established for each component. In order to propagate these uncertainties in an internal dose assessment, a probabilistic internal-dose-assessment system that employs the Bayesian and Monte Carlo methods. Based on the developed system, we developed a probabilistic internal-dose-assessment code by using MATLAB so as to estimate the dose distributions from the measured data with uncertainty. Using the developed code, we calculated the internal dose distribution and statistical values (e.g. the 2.5{sup th}, 5{sup th}, median, 95{sup th}, and 97.5{sup th} percentiles) for three sample scenarios. On the basis of the distributions, we performed a sensitivity analysis to determine the influence of each component on the resulting dose in order to identify the major component of the uncertainty in a bioassay. The results of this study can be applied to various

  12. Development of disruption thermal analysis code DREAM

    Energy Technology Data Exchange (ETDEWEB)

    Yamazaki, Seiichiro; Kobayahsi, Takeshi [Kawasaki Heavy Industries Ltd., Kobe (Japan); Seki, Masahiro

    1989-07-01

    When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing componenets such as first wall and divertor/limiter are subjected to a intensse heat load in a short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs. It causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes and radiation heat loss in required in the design of these components. This paper describes the computer code DREAM, developed to perform the disruption thermal analysis, taking phase changes and radiation into account. (author).

  13. Development of Probabilistic Internal Dosimetry Computer Code

    International Nuclear Information System (INIS)

    Noh, Siwan; Kwon, Tae-Eun; Lee, Jai-Ki

    2017-01-01

    Internal radiation dose assessment involves biokinetic models, the corresponding parameters, measured data, and many assumptions. Every component considered in the internal dose assessment has its own uncertainty, which is propagated in the intake activity and internal dose estimates. For research or scientific purposes, and for retrospective dose reconstruction for accident scenarios occurring in workplaces having a large quantity of unsealed radionuclides, such as nuclear power plants, nuclear fuel cycle facilities, and facilities in which nuclear medicine is practiced, a quantitative uncertainty assessment of the internal dose is often required. However, no calculation tools or computer codes that incorporate all the relevant processes and their corresponding uncertainties, i.e., from the measured data to the committed dose, are available. Thus, the objective of the present study is to develop an integrated probabilistic internal-dose-assessment computer code. First, the uncertainty components in internal dosimetry are identified, and quantitative uncertainty data are collected. Then, an uncertainty database is established for each component. In order to propagate these uncertainties in an internal dose assessment, a probabilistic internal-dose-assessment system that employs the Bayesian and Monte Carlo methods. Based on the developed system, we developed a probabilistic internal-dose-assessment code by using MATLAB so as to estimate the dose distributions from the measured data with uncertainty. Using the developed code, we calculated the internal dose distribution and statistical values (e.g. the 2.5 th , 5 th , median, 95 th , and 97.5 th percentiles) for three sample scenarios. On the basis of the distributions, we performed a sensitivity analysis to determine the influence of each component on the resulting dose in order to identify the major component of the uncertainty in a bioassay. The results of this study can be applied to various situations. In cases

  14. Development of disruption thermal analysis code DREAM

    International Nuclear Information System (INIS)

    Yamazaki, Seiichiro; Kobayahsi, Takeshi; Seki, Masahiro.

    1989-01-01

    When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing componenets such as first wall and divertor/limiter are subjected to a intensse heat load in a short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs. It causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes and radiation heat loss in required in the design of these components. This paper describes the computer code DREAM, developed to perform the disruption thermal analysis, taking phase changes and radiation into account. (author)

  15. 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.

  16. 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)

  17. Code development for nuclear reactor simulation

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  18. 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

  19. 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

  20. Development of INCTAC code for analyzing criticality accident phenomena

    International Nuclear Information System (INIS)

    Mitake, Susumu; Hayashi, Yamato; Sakurai, Shungo

    2003-01-01

    Aiming at understanding nuclear transients and thermal- and hydraulic-phenomena of the criticality accident, a code named INCTAC has been newly developed at the Institute of Nuclear Safety. The code is applicable to the analysis of criticality accident transients of aqueous homogenous fuel solution system. Neutronic transient model is composed of equations for the kinetics and for the spatial distributions, which are deduced from the time dependent multi-group transport equations with the quasi steady state assumption. Thermal-hydraulic transient model is composed of a complete set of the mass, momentum and energy equations together with the two-phase flow assumptions. Validation tests of INCTAC were made using the data obtained at TRACY, a transient experiment criticality facility of JAERI. The calculated results with INCTAC showed a very good agreement with the experiment data, except a slight discrepancy of the time when the peak of reactor power was attained. But, the discrepancy was resolved with the use of an adequate model for movement and transfer of the void in the fuel solution mostly generated by radiolysis. With a simulation model for the transport of radioactive materials through ventilation systems to the environment, INCTAC will be used as an overall safety evaluation code of the criticality accident. (author)

  1. Transport calculations with the BALDUR code. Pt. 1

    International Nuclear Information System (INIS)

    Lackner, K.; Wunderlich, R.

    1979-12-01

    1-d transport calculations with the BALDUR-code are described for predicting the performance of ZEPHYR under D-T operation. Results presented in this report refer to the impurity-free case, and ion and electron heat conduction losses described by CHIsub(i) = neoclassical and CHIsub(e) = 6.25 x 10 17 /nsub(e) (cgs-units). A simple refuelling scenario taking account of the density limit for the ohmic heating phase, the contribution of neutral injection to the refuelling rate and the need for an approximately balanced D-T mixture at the instance of ignition is adopted. The heating scenario assumes a neutral injection beam with 160 keV particle energy in the main component, with a duration of 1.1 sec. Major radius compression by a factor of 1.5 starts 1 sec after the onset of neutral injection and lasts 100 msec. For this standard scenario the performance is studied in different density regimes and for different neutral injection powers. Under the above assumption ignition is predicted for total neutral injection powers < approx. 16 MW (9.6 MW in the main energy component) and average total β-values < 2.8%. Results including impurities, alternative scaling laws, and deviations from the standard scenario will be presented in another report. (orig.) 891 GG/orig. 892 HIS

  2. Preparing diagnostic data for the SNAP transport code

    International Nuclear Information System (INIS)

    Murphy, J.A.; Scott, S.D.; Towner, H.H.

    1992-01-01

    This paper describes the program SNAPIN which is used to prepare data for transport analysis with the SNAP code. The data input to SNAP includes diagnostic profiles [n e (R), T e (R), T i (R), v φ (R), Z eff (R), P rad (R)] and measurements such as total plasma current, R major , beam power, gas puff rate, etc. SNAPIN reads in the necessary TFTR data, allows editing of that data, including graphical editing of profile data and the selection of physics models. SNAPIN allows comparison of profile data from all diagnostics that measure a quantity, for example, electron temperature profiles from Thomson scattering and electron cyclotron emission (ECE). A powerful user interface is important to help the user prepare input data sets quickly and consistently, because hundreds of variables must be specified for each analysis. SNAPIN facilitates this by a careful organization of menus, display of all scalar data and switch settings within the menus, the graphical editing and comparison of profiles, and step-by-step checking for consistent physics controls [J. Murphy, S. Scott, and H. Towner, The SNAP User's Guide, Technical Report PPPL-TM-393, Princeton Plasma Physics Laboratory (1992)

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

    International Nuclear Information System (INIS)

    2004-01-01

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

  4. 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

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

    Indian Academy of Sciences (India)

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

  6. Qualifying codes under software quality assurance: Two examples as guidelines for codes that are existing or under development

    Energy Technology Data Exchange (ETDEWEB)

    Mangold, D.

    1993-05-01

    Software quality assurance is an area of concem for DOE, EPA, and other agencies due to the poor quality of software and its documentation they have received in the past. This report briefly summarizes the software development concepts and terminology increasingly employed by these agencies and provides a workable approach to scientific programming under the new requirements. Following this is a practical description of how to qualify a simulation code, based on a software QA plan that has been reviewed and officially accepted by DOE/OCRWM. Two codes have recently been baselined and qualified, so that they can be officially used for QA Level 1 work under the DOE/OCRWM QA requirements. One of them was baselined and qualified within one week. The first of the codes was the multi-phase multi-component flow code TOUGH version 1, an already existing code, and the other was a geochemistry transport code STATEQ that was under development The way to accomplish qualification for both types of codes is summarized in an easy-to-follow step-by step fashion to illustrate how to baseline and qualify such codes through a relatively painless procedure.

  7. Qualifying codes under software quality assurance: Two examples as guidelines for codes that are existing or under development

    International Nuclear Information System (INIS)

    Mangold, D.

    1993-05-01

    Software quality assurance is an area of concern for DOE, EPA, and other agencies due to the poor quality of software and its documentation they have received in the past. This report briefly summarizes the software development concepts and terminology increasingly employed by these agencies and provides a workable approach to scientific programming under the new requirements. Following this is a practical description of how to qualify a simulation code, based on a software QA plan that has been reviewed and officially accepted by DOE/OCRWM. Two codes have recently been baselined and qualified, so that they can be officially used for QA Level 1 work under the DOE/OCRWM QA requirements. One of them was baselined and qualified within one week. The first of the codes was the multi-phase multi-component flow code TOUGH version 1, an already existing code, and the other was a geochemistry transport code STATEQ that was under development The way to accomplish qualification for both types of codes is summarized in an easy-to-follow step-by step fashion to illustrate how to baseline and qualify such codes through a relatively painless procedure

  8. 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)

  9. 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

  10. FDA Developments: Food Code 2013 and Proposed Trans Fat Determination

    NARCIS (Netherlands)

    Grossman, M.R.

    2014-01-01

    268 Reports EFFL 4|2014 USA FDA Developments: Food Code 2013 and Proposed Trans Fat Determination Margaret Rosso Grossman* I. Food Code 2013 and Food Code Reference System Since 1993, the US Food and Drug Administration has published a Food Code, now updated every four years. In November 2013, the

  11. 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.)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  15. 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

  16. TRAC code development status and plans

    International Nuclear Information System (INIS)

    Spore, J.W.; Liles, D.R.; Nelson, R.A.

    1986-01-01

    This report summarizes the characteristics and current status of the TRAC-PF1/MOD1 computer code. Recent error corrections and user-convenience features are described, and several user enhancements are identified. Current plans for the release of the TRAC-PF1/MOD2 computer code and some preliminary MOD2 results are presented. This new version of the TRAC code implements stability-enhancing two-step numerics into the 3-D vessel, using partial vectorization to obtain a code that has run 400% faster than the MOD1 code

  17. Theoretical Atomic Physics code development II: ACE: Another collisional excitation code

    International Nuclear Information System (INIS)

    Clark, R.E.H.; Abdallah, J. Jr.; Csanak, G.; Mann, J.B.; Cowan, R.D.

    1988-12-01

    A new computer code for calculating collisional excitation data (collision strengths or cross sections) using a variety of models is described. The code uses data generated by the Cowan Atomic Structure code or CATS for the atomic structure. Collisional data are placed on a random access file and can be displayed in a variety of formats using the Theoretical Atomic Physics Code or TAPS. All of these codes are part of the Theoretical Atomic Physics code development effort at Los Alamos. 15 refs., 10 figs., 1 tab

  18. The PARTRAC code: Status and recent developments

    Science.gov (United States)

    Friedland, Werner; Kundrat, Pavel

    Biophysical modeling is of particular value for predictions of radiation effects due to manned space missions. PARTRAC is an established tool for Monte Carlo-based simulations of radiation track structures, damage induction in cellular DNA and its repair [1]. Dedicated modules describe interactions of ionizing particles with the traversed medium, the production and reactions of reactive species, and score DNA damage determined by overlapping track structures with multi-scale chromatin models. The DNA repair module describes the repair of DNA double-strand breaks (DSB) via the non-homologous end-joining pathway; the code explicitly simulates the spatial mobility of individual DNA ends in parallel with their processing by major repair enzymes [2]. To simulate the yields and kinetics of radiation-induced chromosome aberrations, the repair module has been extended by tracking the information on the chromosome origin of ligated fragments as well as the presence of centromeres [3]. PARTRAC calculations have been benchmarked against experimental data on various biological endpoints induced by photon and ion irradiation. The calculated DNA fragment distributions after photon and ion irradiation reproduce corresponding experimental data and their dose- and LET-dependence. However, in particular for high-LET radiation many short DNA fragments are predicted below the detection limits of the measurements, so that the experiments significantly underestimate DSB yields by high-LET radiation [4]. The DNA repair module correctly describes the LET-dependent repair kinetics after (60) Co gamma-rays and different N-ion radiation qualities [2]. First calculations on the induction of chromosome aberrations have overestimated the absolute yields of dicentrics, but correctly reproduced their relative dose-dependence and the difference between gamma- and alpha particle irradiation [3]. Recent developments of the PARTRAC code include a model of hetero- vs euchromatin structures to enable

  19. Development of numerical methods for reactive transport

    International Nuclear Information System (INIS)

    Bouillard, N.

    2006-12-01

    When a radioactive waste is stored in deep geological disposals, it is expected that the waste package will be damaged under water action (concrete leaching, iron corrosion). Then, to understand these damaging processes, chemical reactions and solutes transport are modelled. Numerical simulations of reactive transport can be done sequentially by the coupling of several codes. This is the case of the software platform ALLIANCES which is developed jointly with CEA, ANDRA and EDF. Stiff reactions like precipitation-dissolution are crucial for the radioactive waste storage applications, but standard sequential iterative approaches like Picard's fail in solving rapidly reactive transport simulations with such stiff reactions. In the first part of this work, we focus on a simplified precipitation and dissolution process: a system made up with one solid species and two aqueous species moving by diffusion is studied mathematically. It is assumed that a precipitation dissolution reaction occurs in between them, and it is modelled by a discontinuous kinetics law of unknown sign. By using monotonicity properties, the convergence of a finite volume scheme on admissible mesh is proved. Existence of a weak solution is obtained as a by-product of the convergence of the scheme. The second part is dedicated to coupling algorithms which improve Picard's method and can be easily used in an existing coupling code. By extending previous works, we propose a general and adaptable framework to solve nonlinear systems. Indeed by selecting special options, we can either recover well known methods, like nonlinear conjugate gradient methods, or design specific method. This algorithm has two main steps, a preconditioning one and an acceleration one. This algorithm is tested on several examples, some of them being rather academical and others being more realistic. We test it on the 'three species model'' example. Other reactive transport simulations use an external chemical code CHESS. For a

  20. TIERCE: A code system for particles and radiation transport in thick targets

    Energy Technology Data Exchange (ETDEWEB)

    Bersillon, O.; Bauge, E.; Borne, F.; Clergeau, J.F.; Collin, M.; Cotten, D.; Delaroche, J.P.; Duarte, H.; Flament, J.L.; Girod, M.; Gosselin, G.; Granier, T.; Hilaire, S.; Morel, P.; Perrier, R.; Romain, P.; Roux, L. [CEA, Bruyeres-le-Chatel (France). Service de Physique Nucleaire

    1997-09-01

    Over the last few years, a great effort at Bruyeres-le-Chatel has been the development of the TIERCE code system for the transport of particles and radiations in complex geometry. The comparison of calculated results with experimental data, either microscopic (double differential spectra, residual nuclide yield...) or macroscopic (energy deposition, neutron leakage...), shows the need to improve the nuclear reaction models used. We present some new developments concerning data required for the evaporation model in the framework of a microscopic approach. 22 refs., 6 figs.

  1. 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)

  2. 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.)

  3. Contribution to the development of methods for nuclear reactor core calculations with APOLLO3 code: domain decomposition in transport theory with nonlinear diffusion acceleration for 2D and 3D geometries

    International Nuclear Information System (INIS)

    Lenain, Roland

    2015-01-01

    This thesis is devoted to the implementation of a domain decomposition method applied to the neutron transport equation. The objective of this work is to access high-fidelity deterministic solutions to properly handle heterogeneities located in nuclear reactor cores, for problems' size ranging from color-sets of assemblies to large reactor cores configurations in 2D and 3D. The innovative algorithm developed during the thesis intends to optimize the use of parallelism and memory. The approach also aims to minimize the influence of the parallel implementation on the performances. These goals match the needs of APOLLO3 project, developed at CEA and supported by EDF and AREVA, which must be a portable code (no optimization on a specific architecture) in order to achieve best estimate modeling with resources ranging from personal computer to compute cluster available for engineers analyses. The proposed algorithm is a Parallel Multigroup-Block Jacobi one. Each sub-domain is considered as a multi-group fixed-source problem with volume-sources (fission) and surface-sources (interface flux between the sub-domains). The multi-group problem is solved in each sub-domain and a single communication of the interface flux is required at each power iteration. The spectral radius of the resolution algorithm is made similar to the one of a classical resolution algorithm with a nonlinear diffusion acceleration method: the well-known Coarse Mesh Finite Difference. In this way an ideal scalability is achievable when the calculation is parallelized. The memory organization, taking advantage of shared memory parallelism, optimizes the resources by avoiding redundant copies of the data shared between the sub-domains. Distributed memory architectures are made available by a hybrid parallel method that combines both paradigms of shared memory parallelism and distributed memory parallelism. For large problems, these architectures provide a greater number of processors and the amount of

  4. TRANSNET -- access to radioactive and hazardous materials transportation codes and databases

    International Nuclear Information System (INIS)

    Cashwell, J.W.

    1992-01-01

    TRANSNET has been developed and maintained by Sandia National Laboratories under the sponsorship of the United States Department of Energy (DOE) Office of Environmental Restoration and Waste Management to permit outside access to computerized routing, risk and systems analysis models, and associated databases. The goal of the TRANSNET system is to enable transfer of transportation analytical methods and data to qualified users by permitting direct, timely access to the up-to-date versions of the codes and data. The TRANSNET facility comprises a dedicated computer with telephone ports on which these codes and databases are adapted, modified, and maintained. To permit the widest spectrum of outside users, TRANSNET is designed to minimize hardware and documentation requirements. The user is thus required to have an IBM-compatible personal computer, Hayes-compatible modem with communications software, and a telephone. Maintenance and operation of the TRANSNET facility are underwritten by the program sponsor(s) as are updates to the respective models and data, thus the only charges to the user of the system are telephone hookup charges. TRANSNET provides access to the most recent versions of the models and data developed by or for Sandia National Laboratories. Code modifications that have been made since the last published documentation are noted to the user on the introductory screens. User friendly interfaces have been developed for each of the codes and databases on TRANSNET. In addition, users are provided with default input data sets for typical problems which can either be used directly or edited. Direct transfers of analytical or data files between codes are provided to permit the user to perform complex analyses with a minimum of input. Recent developments to the TRANSNET system include use of the system to directly pass data files between both national and international users as well as development and integration of graphical depiction techniques

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

    International Nuclear Information System (INIS)

    Popescu, Lucretiu M.

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  7. 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

  8. Self-shielding phenomenon modelling in multigroup transport code Apollo-2; Modelisation du phenomene d'autoprotection dans le code de transport multigroupe Apollo 2

    Energy Technology Data Exchange (ETDEWEB)

    Coste-Delclaux, M

    2006-03-15

    This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)

  9. Self-shielding phenomenon modelling in multigroup transport code Apollo-2; Modelisation du phenomene d'autoprotection dans le code de transport multigroupe Apollo 2

    Energy Technology Data Exchange (ETDEWEB)

    Coste-Delclaux, M

    2006-03-15

    This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)

  10. Verification and Validation of the Tritium Transport Code TMAP7

    International Nuclear Information System (INIS)

    Longhurst, Glen R.; Ambrosek, James

    2005-01-01

    The TMAP code has been upgraded to version 7, which includes radioactive decay along with many features implemented in prior versions. Pursuant to acceptance and release for distribution, the code was exercised in a variety of problem types to demonstrate that it provides results in agreement with theoretical results for cases where those are available. It has also been used to model certain experimental results. In this paper, the capabilities of the TMAP7 code are demonstrated by presenting some of the results from the verification and validation process

  11. 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

  12. Evaluation of angular integrals in the generation of transfer matrices for multigroup transport codes

    International Nuclear Information System (INIS)

    Garcia, R.D.M.

    1985-01-01

    The generalization of a semi-analytical technique for the evaluation of angular integrals that appear in the generation of elastic and discrete inelastic tranfer matrices for transport codes is carried out. In contrast to the generalized series expansions which are found to be too complex and thus of little practical value, when compared to the Gaussian quadrature technique, the recursion relations developed in this work are superior to the quadrature scheme, for those cases where the round-off error propagation is not significant. (Author) [pt

  13. MORSE-CGT Monte Carlo radiation transport code with the capability of the torus geometric treatment

    International Nuclear Information System (INIS)

    Deng Li

    1990-01-01

    The combinatorial geometry package CGT with the capability of the torus geometric treatment is introduced. It is get by developing the combinatorial geometry package CG. The CGT package can be transplanted to those codes which the CG package is being used and makes them also with the capability. The MORSE-CGT code can be used to solve the neutron, gamma-ray or coupled neutron-gamma-ray transport problems and time dependence for both shielding and criticality problems in torus system or system which is produced by arbitrary finite combining torus with torus or other bodies in CG package and it can also be used to design the blanket and compute shielding for TOKAMAK Fusion-Fission Hybrid Reactor

  14. Abstracts of digital computer code packages. Assembled by the Radiation Shielding Information Center. [Radiation transport codes

    Energy Technology Data Exchange (ETDEWEB)

    McGill, B.; Maskewitz, B.F.; Anthony, C.M.; Comolander, H.E.; Hendrickson, H.R.

    1976-01-01

    The term ''code package'' is used to describe a miscellaneous grouping of materials which, when interpreted in connection with a digital computer, enables the scientist--user to solve technical problems in the area for which the material was designed. In general, a ''code package'' consists of written material--reports, instructions, flow charts, listings of data, and other useful material and IBM card decks (or, more often, a reel of magnetic tape) on which the source decks, sample problem input (including libraries of data) and the BCD/EBCDIC output listing from the sample problem are written. In addition to the main code, and any available auxiliary routines are also included. The abstract format was chosen to give to a potential code user several criteria for deciding whether or not he wishes to request the code package. (RWR)

  15. Development of a code for description of sodium spray and pool fires. Pt. 1

    International Nuclear Information System (INIS)

    Alexas, A.

    1979-09-01

    In the scope of the development of a code to describe both, sodium pool- and spray fires, the well-known codes SOFIRE II and NABRAND have been compared. Regarding the program technique of both codes, the NABRAND-code seems to be the better one, though it includes some conservatisms in the modelling and in the transport coefficients used. For a realistic estimation of the consequences of large sodium fires in an LMFBR, an elimination of these conservatisms is necessary. After that it must be investigated if a combination of the modificated version of the NABRAND-code and of a spray fire-code (for example the code SPRAY) is efficient. (orig.) [de

  16. Development of a definitive internal dosimetry code

    International Nuclear Information System (INIS)

    Miller, G.; Inkret, W.C.; Schillaci, M.E.

    1996-01-01

    Internal dosimetry may be divided into tow main problems: (1) the forward (scientific) problem of determining biokinetics models that describe how radionuclides are taken into the body, distributed in body tissues, and excreted, and (2) the inverse (mathematical) problem: given the measured amounts in excreta and assuming a biokinetic model, to determine the times and amounts of intakes into the body. The inverse problem of internal dosimetry is, in fact, a generic problem studied in other fields (e.g., image reconstruction, spectral deconvulution, and model parameter fitting). We have developed a code for plutonium internal dosimetry using the maximum entropy method, a method for solving underdetermined inverse problems with a positivity constraint. Within the framework of Bayesian statistics, we believe the definitive approach is to examine the Bayesian posterior probability describing the probability of an intake scenario (X i ) read ( ... ) as open-quotes the set of,close quotes where X i denotes the intake amount that occurs on the with day. For plutonium, for a worker with a long employment history, this is a very high dimensional probability space, since there may be on the order of 10,000 days when intakes may have occurred. Within this high dimensional space, we calculate the mean intake scenario as i > where denotes the expectation value over the posterior probability distribution. Similarly, we calculate uncertainties and other relevant quantities, such as X 2 , as expectation values over the posterior distribution. Thanks to a recent breakthrough in describing the mathematical structure of the intake process (a Poisson sum representation of intakes), we have developed the initial version of a Bayesian expectation-value algorithm for internal dosimetry reconstructions

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

    International Nuclear Information System (INIS)

    Barten, Werner; Robinson, Peter C.

    2001-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-02-01

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

  19. Development of Monte Carlo-based pebble bed reactor fuel management code

    International Nuclear Information System (INIS)

    Setiadipura, Topan; Obara, Toru

    2014-01-01

    Highlights: • A new Monte Carlo-based fuel management code for OTTO cycle pebble bed reactor was developed. • The double-heterogeneity was modeled using statistical method in MVP-BURN code. • The code can perform analysis of equilibrium and non-equilibrium phase. • Code-to-code comparisons for Once-Through-Then-Out case were investigated. • Ability of the code to accommodate the void cavity was confirmed. - Abstract: A fuel management code for pebble bed reactors (PBRs) based on the Monte Carlo method has been developed in this study. The code, named Monte Carlo burnup analysis code for PBR (MCPBR), enables a simulation of the Once-Through-Then-Out (OTTO) cycle of a PBR from the running-in phase to the equilibrium condition. In MCPBR, a burnup calculation based on a continuous-energy Monte Carlo code, MVP-BURN, is coupled with an additional utility code to be able to simulate the OTTO cycle of PBR. MCPBR has several advantages in modeling PBRs, namely its Monte Carlo neutron transport modeling, its capability of explicitly modeling the double heterogeneity of the PBR core, and its ability to model different axial fuel speeds in the PBR core. Analysis at the equilibrium condition of the simplified PBR was used as the validation test of MCPBR. The calculation results of the code were compared with the results of diffusion-based fuel management PBR codes, namely the VSOP and PEBBED codes. Using JENDL-4.0 nuclide library, MCPBR gave a 4.15% and 3.32% lower k eff value compared to VSOP and PEBBED, respectively. While using JENDL-3.3, MCPBR gave a 2.22% and 3.11% higher k eff value compared to VSOP and PEBBED, respectively. The ability of MCPBR to analyze neutron transport in the top void of the PBR core and its effects was also confirmed

  20. 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

  1. Development of codes for physical calculations of WWER

    International Nuclear Information System (INIS)

    Novikov, A.N.

    2000-01-01

    A package of codes for physical calculations of WWER reactors, used at the RRC 'Kurchatov Institute' is discussed including the purpose of these codes, approximations used, degree of data verification, possibilities of automation of calculations and presentation of results, trends of further development of the codes. (Authors)

  2. Resuspension of toxic aerosol using MATHEW--ADPIC wind field--transport and diffusion codes

    International Nuclear Information System (INIS)

    Porch, W.M.

    1979-01-01

    Computer codes have been written which estimate toxic aerosol resuspension based on computed deposition from a primary source, wind, and surface characteristics. The primary deposition pattern and the transport, diffusion, and redeposition of the resuspended toxic aerosol are calculated using a mass-consistent wind field model including topography (MATHEW) and a particle-in-cell diffusion and transport model (ADPIC) which were developed at LLL. The source term for resuspended toxic aerosol is determined by multiplying the total aerosol flux as a function of wind speed by the area of highest concentration and the fraction of suspended material estimated to be toxic. Preliminary calculations based on a test problem at the Nevada Test Site determined an hourly averaged maximum resuspension factor of 10 -4 for a 15 m/sec wind which is within an admittedly large range of resuspension factor measurements using experimental data

  3. The bidimensional neutron transport code TWOTRAN-GG. Users manual and input data TWOTRAN-TRACA version

    International Nuclear Information System (INIS)

    Ahnert, C.; Aragones, J. M.

    1981-01-01

    This Is a users manual of the neutron transport code TWOTRAN-TRACA, which is a version of the original TWOTRAN-GG from the Los Alamos Laboratory, with some modifications made at JEN. A detailed input data description is given as well as the new modifications developed at JEN. (Author) 8 refs

  4. The bidimensional neutron transport code Twotran-GG. User's manual and input data. Twotran-Traca version

    International Nuclear Information System (INIS)

    Ahnert, C.; Aragones, J.M.

    1981-01-01

    A user's manual of the neutron transport code Twotran-Traca is presented; it is a version of the original Twotran-GG from the Los Alamos Laboratory, with some modifications made at J.E.N., Spain. A detailed input data description is given as well as the new modifications developped at J.E.N. (author) [es

  5. 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)

  6. 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).

  7. 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)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  9. 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

  10. A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom

    International Nuclear Information System (INIS)

    Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H.

    2014-08-01

    As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)

  11. A GPU-based Monte Carlo dose calculation code for photon transport in a voxel phantom

    Energy Technology Data Exchange (ETDEWEB)

    Bellezzo, M.; Do Nascimento, E.; Yoriyaz, H., E-mail: mbellezzo@gmail.br [Instituto de Pesquisas Energeticas e Nucleares / CNEN, Av. Lineu Prestes 2242, Cidade Universitaria, 05508-000 Sao Paulo (Brazil)

    2014-08-15

    As the most accurate method to estimate absorbed dose in radiotherapy, Monte Carlo method has been widely used in radiotherapy treatment planning. Nevertheless, its efficiency can be improved for clinical routine applications. In this paper, we present the CUBMC code, a GPU-based Mc photon transport algorithm for dose calculation under the Compute Unified Device Architecture platform. The simulation of physical events is based on the algorithm used in Penelope, and the cross section table used is the one generated by the Material routine, als present in Penelope code. Photons are transported in voxel-based geometries with different compositions. To demonstrate the capabilities of the algorithm developed in the present work four 128 x 128 x 128 voxel phantoms have been considered. One of them is composed by a homogeneous water-based media, the second is composed by bone, the third is composed by lung and the fourth is composed by a heterogeneous bone and vacuum geometry. Simulations were done considering a 6 MeV monoenergetic photon point source. There are two distinct approaches that were used for transport simulation. The first of them forces the photon to stop at every voxel frontier, the second one is the Woodcock method, where the photon stop in the frontier will be considered depending on the material changing across the photon travel line. Dose calculations using these methods are compared for validation with Penelope and MCNP5 codes. Speed-up factors are compared using a NVidia GTX 560-Ti GPU card against a 2.27 GHz Intel Xeon CPU processor. (Author)

  12. TRU waste transportation package development

    International Nuclear Information System (INIS)

    Eakes, R.G.; Lamoreaux, G.H.; Romesberg, L.E.; Sutherland, S.H.; Duffey, T.A.

    1980-01-01

    Inventories of the transuranic wastes buried or stored at various US DOE sites are tabulated. The leading conceptual design of Type-B packaging for contact-handled transuranic waste is the Transuranic Package Transporter (TRUPACT), a large metal container comprising inner and outer tubular steel frameworks which are separated by rigid polyurethane foam and sheathed with steel plate. Testing of TRUPACT is reported. The schedule for its development is given. 6 figures

  13. Developing sustainable transportation performance measures for ALDOT.

    Science.gov (United States)

    2013-06-01

    Sustainable transportation is generally used to refer to transportation that contributes to the sustainable development of the community that owns and uses the system. The Transportation Research Board defines sustainability as: Sustainability is ...

  14. 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.

  15. Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code

    International Nuclear Information System (INIS)

    Ramsthaler, J.A.; Lime, J.F.; Sahota, M.S.

    1978-12-01

    A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A

  16. Development of authentication code for multi-access optical code division multiplexing based quantum key distribution

    Science.gov (United States)

    Taiwo, Ambali; Alnassar, Ghusoon; Bakar, M. H. Abu; Khir, M. F. Abdul; Mahdi, Mohd Adzir; Mokhtar, M.

    2018-05-01

    One-weight authentication code for multi-user quantum key distribution (QKD) is proposed. The code is developed for Optical Code Division Multiplexing (OCDMA) based QKD network. A unique address assigned to individual user, coupled with degrading probability of predicting the source of the qubit transmitted in the channel offer excellent secure mechanism against any form of channel attack on OCDMA based QKD network. Flexibility in design as well as ease of modifying the number of users are equally exceptional quality presented by the code in contrast to Optical Orthogonal Code (OOC) earlier implemented for the same purpose. The code was successfully applied to eight simultaneous users at effective key rate of 32 bps over 27 km transmission distance.

  17. Theoretical atomic physics code development III TAPS: A display code for atomic physics data

    International Nuclear Information System (INIS)

    Clark, R.E.H.; Abdallah, J. Jr.; Kramer, S.P.

    1988-12-01

    A large amount of theoretical atomic physics data is becoming available through use of the computer codes CATS and ACE developed at Los Alamos National Laboratory. A new code, TAPS, has been written to access this data, perform averages over terms and configurations, and display information in graphical or text form. 7 refs., 13 figs., 1 tab

  18. A predictive transport modeling code for ICRF-heated tokamaks

    International Nuclear Information System (INIS)

    Phillips, C.K.; Hwang, D.Q.

    1992-02-01

    In this report, a detailed description of the physic included in the WHIST/RAZE package as well as a few illustrative examples of the capabilities of the package will be presented. An in depth analysis of ICRF heating experiments using WHIST/RAZE will be discussed in a forthcoming report. A general overview of philosophy behind the structure of the WHIST/RAZE package, a summary of the features of the WHIST code, and a description of the interface to the RAZE subroutines are presented in section 2 of this report. Details of the physics contained in the RAZE code are examined in section 3. Sample results from the package follow in section 4, with concluding remarks and a discussion of possible improvements to the package discussed in section 5

  19. HYDROCOIN [HYDROlogic COde INtercomparison] Level 1: Benchmarking and verification test results with CFEST [Coupled Fluid, Energy, and Solute Transport] code: Draft report

    International Nuclear Information System (INIS)

    Yabusaki, S.; Cole, C.; Monti, A.M.; Gupta, S.K.

    1987-04-01

    Part of the safety analysis is evaluating groundwater flow through the repository and the host rock to the accessible environment by developing mathematical or analytical models and numerical computer codes describing the flow mechanisms. This need led to the establishment of an international project called HYDROCOIN (HYDROlogic COde INtercomparison) organized by the Swedish Nuclear Power Inspectorate, a forum for discussing techniques and strategies in subsurface hydrologic modeling. The major objective of the present effort, HYDROCOIN Level 1, is determining the numerical accuracy of the computer codes. The definition of each case includes the input parameters, the governing equations, the output specifications, and the format. The Coupled Fluid, Energy, and Solute Transport (CFEST) code was applied to solve cases 1, 2, 4, 5, and 7; the Finite Element Three-Dimensional Groundwater (FE3DGW) Flow Model was used to solve case 6. Case 3 has been ignored because unsaturated flow is not pertinent to SRP. This report presents the Level 1 results furnished by the project teams. The numerical accuracy of the codes is determined by (1) comparing the computational results with analytical solutions for cases that have analytical solutions (namely cases 1 and 4), and (2) intercomparing results from codes for cases which do not have analytical solutions (cases 2, 5, 6, and 7). Cases 1, 2, 6, and 7 relate to flow analyses, whereas cases 4 and 5 require nonlinear solutions. 7 refs., 71 figs., 9 tabs

  20. 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

  1. Environmental, Transient, Three-Dimensional, Hydrothermal, Mass Transport Code - FLESCOT

    Energy Technology Data Exchange (ETDEWEB)

    Onishi, Yasuo [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bao, Jie [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Glass, Kevin A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Eyler, L. L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Okumura, Masahiko [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-03-28

    The purpose of the project was to modify and apply the transient, three-dimensional FLESCOT code to be able to effectively simulate cesium behavior in Fukushima lakes/dam reservoirs, river mouths, and coastal areas. The ultimate objective of the FLESCOT simulation is to predict future changes of cesium accumulation in Fukushima area reservoirs and costal water. These evaluation results will assist ongoing and future environmental remediation activities and policies in a systematic and comprehensive manner.

  2. MIDAS/PK code development using point kinetics model

    International Nuclear Information System (INIS)

    Song, Y. M.; Park, S. H.

    1999-01-01

    In this study, a MIDAS/PK code has been developed for analyzing the ATWS (Anticipated Transients Without Scram) which can be one of severe accident initiating events. The MIDAS is an integrated computer code based on the MELCOR code to develop a severe accident risk reduction strategy by Korea Atomic Energy Research Institute. In the mean time, the Chexal-Layman correlation in the current MELCOR, which was developed under a BWR condition, is appeared to be inappropriate for a PWR. So as to provide ATWS analysis capability to the MIDAS code, a point kinetics module, PKINETIC, has first been developed as a stand-alone code whose reference model was selected from the current accident analysis codes. In the next step, the MIDAS/PK code has been developed via coupling PKINETIC with the MIDAS code by inter-connecting several thermal hydraulic parameters between the two codes. Since the major concern in the ATWS analysis is the primary peak pressure during the early few minutes into the accident, the peak pressure from the PKINETIC module and the MIDAS/PK are compared with the RETRAN calculations showing a good agreement between them. The MIDAS/PK code is considered to be valuable for analyzing the plant response during ATWS deterministically, especially for the early domestic Westinghouse plants which rely on the operator procedure instead of an AMSAC (ATWS Mitigating System Actuation Circuitry) against ATWS. This capability of ATWS analysis is also important from the view point of accident management and mitigation

  3. 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.)

  4. 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

  5. Development of time dependent safety analysis code for plasma anomaly events in fusion reactors

    International Nuclear Information System (INIS)

    Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

    1997-01-01

    A safety analysis code SAFALY has been developed to analyze plasma anomaly events in fusion reactors, e.g., a loss of plasma control. The code is a hybrid code comprising a zero-dimensional plasma dynamics and a one-dimensional thermal analysis of in-vessel components. The code evaluates the time evolution of plasma parameters and temperature distributions of in-vessel components. As the plasma-safety interface model, we proposed a robust plasma physics model taking into account updated data for safety assessment. For example, physics safety guidelines for beta limit, density limit and H-L mode confinement transition threshold power, etc. are provided in the model. The model of the in-vessel components are divided into twenty temperature regions in the poloidal direction taking account of radiative heat transfer between each surface of each region. This code can also describe the coolant behavior under hydraulic accidents with the results by hydraulics code and treat vaporization (sublimation) from plasma facing components (PFCs). Furthermore, the code includes the model of impurity transport form PFCs by using a transport probability and a time delay. Quantitative analysis based on the model is possible for a scenario of plasma passive shutdown. We examined the possibility of the code as a safety analysis code for plasma anomaly events in fusion reactors and had a prospect that it would contribute to the safety analysis of the International Thermonuclear Experimental Reactor (ITER). (author)

  6. The FLUKA code for space applications Recent developments

    CERN Document Server

    Andersen, V; Battistoni, G; Campanella, M; Carboni, M; Cerutti, F; Empl, A; Fassò, A; Ferrari, A; Gadioli, E; Garzelli, M V; Lee, K; Ottolenghi, A; Pelliccioni, M; Pinsky, L S; Ranft, J; Roesler, S; Sala, P R; Wilson, T L

    2004-01-01

    The FLUKA Monte Carlo transport code is widely used for fundamental research, radioprotection and dosimetry, hybrid nuclear energy system and cosmic ray calculations. The validity of its physical models has been benchmarked against a variety of experimental data over a wide range of energies, ranging from accelerator data to cosmic ray showers in the earth atmosphere. The code is presently undergoing several developments in order to better fit the needs of space applications. The generation of particle spectra according to up-to- date cosmic ray data as well as the effect of the solar and geomagnetic modulation have been implemented and already successfully applied to a variety of problems. The implementation of suitable models for heavy ion nuclear interactions has reached an operational stage. At medium/high energy FLUKA is using the DPMJET model. The major task of incorporating heavy ion interactions from a few GeV/n down to the threshold for inelastic collisions is also progressing and promising results h...

  7. 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)

  8. Low-discrepancy point sets in transport codes

    Energy Technology Data Exchange (ETDEWEB)

    Warnock, T.T.

    1985-01-01

    A drawback to Monte Carlo methods of computation is its rate of convergence. There are methods of sampling that have a better error estimate than those using random numbers. This paper gives the result of some preliminary experiments with these sampling methods on two neutron transport problems.

  9. 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

  10. 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.

  11. 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)

  12. Development and Application of a Code for Internal Exposure (CINEX) based on the CINDY code

    International Nuclear Information System (INIS)

    Kravchik, T.; Duchan, N.; Sarah, R.; Gabay, Y.; Kol, R.

    2004-01-01

    Internal exposure to radioactive materials at the NRCN is evaluated using the CINDY (Code for Internal Dosimetry) Package. The code was developed by the Pacific Northwest Laboratory to assist the interpretation of bioassay data, provide bioassay projections and evaluate committed and calendar-year doses from intake or bioassay measurement data. It provides capabilities to calculate organ dose and effective dose equivalents using the International Commission on Radiological Protection (ICRP) 30 approach. The CINDY code operates under DOS operating system and consequently its operation needs a relatively long procedure which also includes a lot of manual typing that can lead to personal human mistakes. A new code has been developed at the NRCN, the CINEX (Code for Internal Exposure), which is an Excel application and leads to a significant reduction in calculation time (in the order of 5-10 times) and in the risk of personal human mistakes. The code uses a database containing tables which were constructed by the CINDY and contain the bioassay values predicted by the ICRP30 model after an intake of an activity unit of each isotope. Using the database, the code than calculates the appropriate intake and consequently the committed effective dose and organ dose. Calculations with the CINEX code were compared to similar calculations with the CINDY code. The discrepancies were less than 5%, which is the rounding error of the CINDY code. Attached is a table which compares parameters calculated with the CINEX and the CINDY codes (for a class Y uranium). The CINEX is now used at the NRCN to calculate occupational intakes and doses to workers with radioactive materials

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  14. Computer-assisted Particle-in-Cell code development

    International Nuclear Information System (INIS)

    Kawata, S.; Boonmee, C.; Teramoto, T.; Drska, L.; Limpouch, J.; Liska, R.; Sinor, M.

    1997-12-01

    This report presents a new approach for an electromagnetic Particle-in-Cell (PIC) code development by a computer: in general PIC codes have a common structure, and consist of a particle pusher, a field solver, charge and current density collections, and a field interpolation. Because of the common feature, the main part of the PIC code can be mechanically developed on a computer. In this report we use the packages FIDE and GENTRAN of the REDUCE computer algebra system for discretizations of field equations and a particle equation, and for an automatic generation of Fortran codes. The approach proposed is successfully applied to the development of 1.5-dimensional PIC code. By using the generated PIC code the Weibel instability in a plasma is simulated. The obtained growth rate agrees well with the theoretical value. (author)

  15. SCATTER: Source and Transport of Emplaced Radionuclides: Code documentation

    International Nuclear Information System (INIS)

    Longsine, D.E.

    1987-03-01

    SCATTER simulated several processes leading to the release of radionuclides to the site subsystem and then simulates transport via the groundwater of the released radionuclides to the biosphere. The processes accounted for to quantify release rates to a ground-water migration path include radioactive decay and production, leaching, solubilities, and the mixing of particles with incoming uncontaminated fluid. Several decay chains of arbitrary length can be considered simultaneously. The release rates then serve as source rates to a numerical technique which solves convective-dispersive transport for each decay chain. The decay chains are allowed to have branches and each member can have a different radioactive factor. Results are cast as radionuclide discharge rates to the accessible environment

  16. Automatic modeling for the Monte Carlo transport code Geant4

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  17. Challenges on innovations of newly-developed safety analysis codes

    International Nuclear Information System (INIS)

    Yang, Yanhua; Zhang, Hao

    2016-01-01

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  18. Challenges on innovations of newly-developed safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yanhua [Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering; Zhang, Hao [State Nuclear Power Software Development Center, Beijing (China). Beijing Future Science and Technology City

    2016-05-15

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  19. Development of covariance capabilities in EMPIRE code

    Energy Technology Data Exchange (ETDEWEB)

    Herman,M.; Pigni, M.T.; Oblozinsky, P.; Mughabghab, S.F.; Mattoon, C.M.; Capote, R.; Cho, Young-Sik; Trkov, A.

    2008-06-24

    The nuclear reaction code EMPIRE has been extended to provide evaluation capabilities for neutron cross section covariances in the thermal, resolved resonance, unresolved resonance and fast neutron regions. The Atlas of Neutron Resonances by Mughabghab is used as a primary source of information on uncertainties at low energies. Care is taken to ensure consistency among the resonance parameter uncertainties and those for thermal cross sections. The resulting resonance parameter covariances are formatted in the ENDF-6 File 32. In the fast neutron range our methodology is based on model calculations with the code EMPIRE combined with experimental data through several available approaches. The model-based covariances can be obtained using deterministic (Kalman) or stochastic (Monte Carlo) propagation of model parameter uncertainties. We show that these two procedures yield comparable results. The Kalman filter and/or the generalized least square fitting procedures are employed to incorporate experimental information. We compare the two approaches analyzing results for the major reaction channels on {sup 89}Y. We also discuss a long-standing issue of unreasonably low uncertainties and link it to the rigidity of the model.

  20. Resonance self-shielding methodology of new neutron transport code STREAM

    International Nuclear Information System (INIS)

    Choi, Sooyoung; Lee, Hyunsuk; Lee, Deokjung; Hong, Ser Gi

    2015-01-01

    This paper reports on the development and verification of three new resonance self-shielding methods. The verifications were performed using the new neutron transport code, STREAM. The new methodologies encompass the extension of energy range for resonance treatment, the development of optimum rational approximation, and the application of resonance treatment to isotopes in the cladding region. (1) The extended resonance energy range treatment has been developed to treat the resonances below 4 eV of three resonance isotopes and shows significant improvements in the accuracy of effective cross sections (XSs) in that energy range. (2) The optimum rational approximation can eliminate the geometric limitations of the conventional approach of equivalence theory and can also improve the accuracy of fuel escape probability. (3) The cladding resonance treatment method makes it possible to treat resonances in cladding material which have not been treated explicitly in the conventional methods. These three new methods have been implemented in the new lattice physics code STREAM and the improvement in the accuracy of effective XSs is demonstrated through detailed verification calculations. (author)

  1. Further development of the computer code ATHLET-CD

    International Nuclear Information System (INIS)

    Weber, Sebastian; Austregesilo, Henrique; Bals, Christine; Band, Sebastian; Hollands, Thorsten; Koellein, Carsten; Lovasz, Liviusz; Pandazis, Peter; Schubert, Johann-Dietrich; Sonnenkalb, Martin

    2016-10-01

    In the framework of the reactor safety research program sponsored by the German Federal Ministry for Economic Affairs and Energy (BMWi), the computer code system ATHLET/ATHLET-CD has been further developed as an analysis tool for the simulation of accidents in nuclear power plants with pressurized and boiling water reactors as well as for the evaluation of accident management procedures. The main objective was to provide a mechanistic analysis tool for best estimate calculations of transients, accidents, and severe accidents with core degradation in light water reactors. With the continued development, the capability of the code system has been largely improved, allowing best estimate calculations of design and beyond design base accidents, and the simulation of advanced core degradation with enhanced model extent in a reasonable calculation time. ATHLET comprises inter alia a 6-equation model, models for the simulation of non-condensable gases and tracking of boron concentration, as well as additional component and process models for the complete system simulation. Among numerous model improvements, the code application has been extended to super critical pressures. The mechanistic description of the dynamic development of flow regimes on the basis of a transport equation for the interface area has been further developed. This ATHLET version is completely integrated in ATHLET-CD. ATHLET-CD further comprises dedicated models for the simulation of fuel and control assembly degradation for both pressurized and boiling water reactors, debris bed with melting in the core region, as well as fission product and aerosol release and transport in the cooling system, inclusive of decay of nuclide inventories and of chemical reactions in the gas phase. The continued development also concerned the modelling of absorber material release, of melting, melt relocation and freezing, and the interaction with the wall of the reactor pressure vessel. The following models were newly

  2. 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)

  3. Subchannel analysis code development for CANDU fuel channel

    International Nuclear Information System (INIS)

    Park, J. H.; Suk, H. C.; Jun, J. S.; Oh, D. J.; Hwang, D. H.; Yoo, Y. J.

    1998-07-01

    Since there are several subchannel codes such as COBRA and TORC codes for a PWR fuel channel but not for a CANDU fuel channel in our country, the subchannel analysis code for a CANDU fuel channel was developed for the prediction of flow conditions on the subchannels, for the accurate assessment of the thermal margin, the effect of appendages, and radial/axial power profile of fuel bundles on flow conditions and CHF and so on. In order to develop the subchannel analysis code for a CANDU fuel channel, subchannel analysis methodology and its applicability/pertinence for a fuel channel were reviewed from the CANDU fuel channel point of view. Several thermalhydraulic and numerical models for the subchannel analysis on a CANDU fuel channel were developed. The experimental data of the CANDU fuel channel were collected, analyzed and used for validation of a subchannel analysis code developed in this work. (author). 11 refs., 3 tabs., 50 figs

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    CERN Multimedia

    CERN. Geneva; Ferrari, Alfredo; Silari, Marco

    2006-01-01

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

  6. Realistic integration of sorption processes in transport codes for long-term safety assessments

    Energy Technology Data Exchange (ETDEWEB)

    Noseck, Ulrich; Fluegge, Judith; Britz, Susan; Schneider, Anke [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany); Brendler, Vinzenz; Stockmann, Madlen; Schikora, Johannes [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Lampe, Michael [Frankfurt Univ. (Germany). Goethe Center for Scientific Computing

    2012-09-15

    One important aspect in long-term safety assessment is related to radionuclide transport in geologic formations. In order to assess its consequences over assessment periods of one million years numerical models describing flow and transport are applied. Sorption on mineral surfaces is the most relevant process retarding radionuclide transport. On the one hand an increased transport time might cause a decrease in radionuclide concentration by radioactive decay. On the other hand it might increase concentrations of dose-relevant daughter nuclides in decay chains. In order to treat the radionuclide sorption processes in natural systems close to reality the so-called smart K{sub d}-concept is implemented into the transport program r{sup 3}t, which is applied to large model areas and very long time scales in long-term safety assessment. In the first stage this approach is developed for a typical sedimentary system covering rock salt and clay formations in Northern Germany. The smart K{sub d}-values are based on mechanistic surface complexation models (SCM), varying in time and space and de-pending on the actual geochemical conditions, which might change in the future e. g. due to the impact of climate changes. The concept developed and introduced here is based on a feasible treatment of the most relevant geochemical parameters in the transport code as well as on a matrix of smart K{sub d}-values calculated in dependence on these parameters. The implementation of the concept comprises the selection of relevant elements and minerals to be considered, an experimental program to fill data gaps of the thermody-namic sorption database, an uncertainty and sensitivity analysis to identify the most important environmental parameters influencing sorption of long-term relevant radionu-clides, the creation of a matrix with K{sub d}-values dependent on the selected environmental parameters, and the development and realisation of the conceptual model for treatment of temporal and

  7. Development of multi-physics code systems based on the reactor dynamics code DYN3D

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Schuetze, Jochen [ANSYS Germany GmbH, Darmstadt (Germany); Frank, Thomas [ANSYS Germany GmbH, Otterfing (Germany); Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)

    2011-07-15

    The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

  8. Development of multi-physics code systems based on the reactor dynamics code DYN3D

    International Nuclear Information System (INIS)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich; Schuetze, Jochen; Frank, Thomas; Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo

    2011-01-01

    The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

  9. Development of thermal hydraulic evaluation code for CANDU reactors

    International Nuclear Information System (INIS)

    Kim, Man Woong; Yu, Seon Oh; Choi, Yong Seog; Shin, Chull; Hwang, Soo Hyun

    2004-02-01

    To enhance the safety of operating CANDU reactors, the establishment of the safety analysis codes system for CANDU reactors is in progress. As for the development of thermal-hydraulic analysis code for CANDU system, the studies for improvement of evaluation model inside RELAP/CANDU code and the development of safety assessment methodology for GSI (Generic Safety Issues) are in progress as a part of establishment of CANDU safety assessment system. To develop the 3-D thermal-hydraulic analysis code for moderator system, the CFD models for analyzing the CANDU-6 moderator circulation are developed. One model uses a structured grid system with the porous media approach for the 380 Calandria tubes in the core region. The other uses a unstructured grid system on the real geometry of 380 Calandria tubes, so that the detailed fluid flow between the Calandria tubes can be observed. As to the development of thermal-hydraulic analysis code for containment, the study on the applicability of CONTAIN 2.0 code to a CANDU containment was conducted and a simulation of the thermal-hydraulic phenomena during the accident was performed. Besides, the model comparison of ESFs (Engineered Safety Features) inside CONTAIN 2.0 code and PRESCON code has also conducted

  10. Development of a nuclear power plant system analysis code

    International Nuclear Information System (INIS)

    Sim, Suk K.; Jeong, J. J.; Ha, K. S.; Moon, S. K.; Park, J. W.; Yang, S. K.; Song, C. H.; Chun, S. Y.; Kim, H. C.; Chung, B. D.; Lee, W. J.; Kwon, T. S.

    1997-07-01

    During the period of this study, TASS 1.0 code has been prepared for the non-LOCA licensing and reload safety analyses of the Westinghouse and the Korean Standard Nuclear Power Plants (KSNPP) type reactors operating in Korea. TASS-NPA also has been developed for a real time simulation of the Kori-3/4 transients using on-line graphical interactions. TASS 2.0 code has been further developed to timely apply the TASS 2.0 code for the design certification of the KNGR. The COBRA/RELAP5 code, a multi-dimensional best estimate system code, has been developed by integrating the realistic three-dimensional reactor vessel model with the RELAP5 /MOD3.2 code, a one-dimensional system code. Also, a 3D turbulent two-phase flow analysis code, FEMOTH-TF, has been developed using finite element technique to analyze local thermal hydraulic phenomena in support of the detailed design analysis for the development of the advanced reactors. (author). 84 refs., 27 tabs., 83 figs

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

    Energy Technology Data Exchange (ETDEWEB)

    Cupini, E. [ENEA, Centro Ricerche Ezio Clementel, Bologna, (Italy). Dipt. Innovazione

    1999-07-01

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

  12. Verification and Validation of The Tritium Transport Code TMAP7

    International Nuclear Information System (INIS)

    Glen R. Longhurst; James Ambrosek

    2004-01-01

    The TMAP Code was written at the Idaho National Engineering and Environmental Laboratory in the late 1980s as a tool for safety analysis of systems involving tritium. Since then it has been upgraded several times and has been used in numerous applications including experiments supporting fusion safety, predictions for advanced systems such as the International Thermonuclear Experimental Reactor (ITER), and estimates involving tritium production technologies. Its most recent upgrade to TMAP7 was accomplished in response to several needs. Prior versions had the capacity to deal with only a single trap for diffusing gaseous species in solid structures. TMAP7 includes up to three separate traps and up to 10 diffusing species. The original code had difficulty dealing with heteronuclear molecule formation such as HD and DT. That has been removed. Under pre-specified boundary enclosure conditions and solution-law dependent diffusion boundary conditions, such as Sieverts' law, TMAP7 automatically generates heteronuclear molecular partial pressures when solubilities and partial pressures of the homonuclear molecular species are provided for law-dependent diffusion boundary conditions. A further sophistication is the addition of non-diffusing surface species. Atoms such as oxygen or nitrogen or formation of hydroxyl radicals on metal surfaces are sometimes important in molecule formation with diffusing hydrogen isotopes but do not, themselves, diffuse appreciably in the material. TMAP7 will accommodate up to 30 such surface species, allowing the user to specify relationships between those surface concentrations and partial pressures of gaseous species above the surfaces or to form them dynamically by combining diffusion species or other surface species. Additionally, TMAP7 allows the user to include a surface binding energy and an adsorption barrier energy and includes asymmetrical diffusion between the surface sites and regular diffusion sites in the bulk. All of the

  13. COSINE software development based on code generation technology

    International Nuclear Information System (INIS)

    Ren Hao; Mo Wentao; Liu Shuo; Zhao Guang

    2013-01-01

    The code generation technology can significantly improve the quality and productivity of software development and reduce software development risk. At present, the code generator is usually based on UML model-driven technology, which can not satisfy the development demand of nuclear power calculation software. The feature of scientific computing program was analyzed and the FORTRAN code generator (FCG) based on C# was developed in this paper. FCG can generate module variable definition FORTRAN code automatically according to input metadata. FCG also can generate memory allocation interface for dynamic variables as well as data access interface. FCG was applied to the core and system integrated engine for design and analysis (COSINE) software development. The result shows that FCG can greatly improve the development efficiency of nuclear power calculation software, and reduce the defect rate of software development. (authors)

  14. Recent developments in the CONTAIN-LMR code

    International Nuclear Information System (INIS)

    Murata, K.K.

    1990-01-01

    Through an international collaborative effort, a special version of the CONTAIN code is being developed for integrated mechanistic analysis of the conditions in liquid metal reactor (LMR) containments during severe accidents. The capabilities of the most recent code version, CONTAIN LMR/1B-Mod.1, are discussed. These include new models for the treatment of two condensables, sodium condensation on aerosols, chemical reactions, hygroscopic aerosols, and concrete outgassing. This code version also incorporates all of the previously released LMR model enhancements. The results of an integral demonstration calculation of a sever core-melt accident scenario are given to illustrate the features of this code version. 11 refs., 7 figs., 1 tab

  15. RADHEAT-V4: a code system to generate multigroup constants and analyze radiation transport for shielding safety evaluation

    International Nuclear Information System (INIS)

    Yamano, Naoki; Minami, Kazuyoshi; Koyama, Kinji; Naito, Yoshitaka.

    1989-03-01

    A modular code system RADHEAT-V4 has been developed for performing precisely neutron and photon transport analyses, and shielding safety evaluations. The system consists of the functional modules for producing coupled multi-group neutron and photon cross section sets, for analyzing the neutron and photon transport, and for calculating the atom displacement and the energy deposition due to radiations in nuclear reactor or shielding material. A precise method named Direct Angular Representation (DAR) has been developed for eliminating an error associated with the method of the finite Legendre expansion in evaluating angular distributions of cross sections and radiation fluxes. The DAR method implemented in the code system has been described in detail. To evaluate the accuracy and applicability of the code system, some test calculations on strong anisotropy problems have been performed. From the results, it has been concluded that RADHEAT-V4 is successfully applicable to evaluating shielding problems accurately for fission and fusion reactors and radiation sources. The method employed in the code system is very effective in eliminating negative values and oscillations of angular fluxes in a medium having an anisotropic source or strong streaming. Definitions of the input data required in various options of the code system and the sample problems are also presented. (author)

  16. Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

    2015-07-01

    Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

  17. Verification of three dimensional triangular prismatic discrete ordinates transport code ENSEMBLE-TRIZ by comparison with Monte Carlo code GMVP

    International Nuclear Information System (INIS)

    Homma, Y.; Moriwaki, H.; Ikeda, K.; Ohdi, S.

    2013-01-01

    This paper deals with the verification of the 3 dimensional triangular prismatic discrete ordinates transport calculation code ENSEMBLE-TRIZ by comparison with the multi-group Monte Carlo calculation code GMVP in a large fast breeder reactor. The reactor is a 750 MWe electric power sodium cooled reactor. Nuclear characteristics are calculated at the beginning of cycle of an initial core and at the beginning and the end of cycle of an equilibrium core. According to the calculations, the differences between the two methodologies are smaller than 0.0002 Δk in the multiplication factor, relatively about 1% in the control rod reactivity, and 1% in the sodium void reactivity. (authors)

  18. Nupack, the new Asme code for radioactive material transportation packaging containments

    International Nuclear Information System (INIS)

    Turula, P.

    1998-01-01

    The American Society of Mechanical Engineers (ASME) has added a new division to the nuclear construction section of its Boiler and Pressure Vessel Code (B and PVC). This Division, commonly referred to as 'Nupack', has been written to provide a consistent set of technical requirements for containment vessels of transportation packagings for high-level radioactive materials. This paper provides an introduction to Nupack, discusses some of its technical provisions, and describes how it can be used the design and construction of packaging components. Nupack's general provisions and design requirements are emphasized, while treatment of materials, fabrication and inspection is left for another paper. Participation in the Nupack development work described in this paper was supported by the U.S. Department of Energy. (authors)

  19. CLUB - a multigroup integral transport theory code for lattice calculations of PHWR cells

    International Nuclear Information System (INIS)

    Krishnani, P.D.

    1992-01-01

    The computer code CLUB has been developed to calculate lattice parameters as a function of burnup for a pressurised heavy water reactor (PHWR) lattice cell containing fuel in the form of cluster. It solves the multigroup integral transport equation by the method based on combination of small scale collision probability (CP) method and large scale interface current technique. The calculations are performed by using WIMS 69 group cross section library or its condensed versions of 27 or 28 group libraries. It can also compute Keff from the given geometrical buckling in the input using multigroup diffusion theory in fundamental mode. The first order differential burnup equations can be solved by either Trapezoidal rule or Runge-Kutta method. (author). 17 refs., 2 figs

  20. Computer codes for the evaluation of thermodynamic and transport properties for equilibrium air to 30000 K

    Science.gov (United States)

    Thompson, Richard A.; Lee, Kam-Pui; Gupta, Roop N.

    1991-01-01

    The computer codes developed here provide self-consistent thermodynamic and transport properties for equilibrium air for temperatures from 500 to 30000 K over a temperature range of 10 (exp -4) to 10 (exp -2) atm. These properties are computed through the use of temperature dependent curve fits for discrete values of pressure. Interpolation is employed for intermediate values of pressure. The curve fits are based on mixture values calculated from an 11-species air model. Individual species properties used in the mixture relations are obtained from a recent study by the present authors. A review and discussion of the sources and accuracy of the curve fitted data used herein are given in NASA RP 1260.

  1. Development and Verification of Behavior of Tritium Analytic Code (BOTANIC)

    International Nuclear Information System (INIS)

    Park, Min Young; Kim, Eung Soo

    2014-01-01

    VHTR, one of the Generation IV reactor concepts, has a relatively high operation temperature and is usually suggested as a heat source for many industrial processes, including hydrogen production process. Thus, it is vital to trace tritium behavior in the VHTR system and the potential permeation rate to the industrial process. In other words, tritium is a crucial issue in terms of safety in the fission reactor system. Therefore, it is necessary to understand the behavior of tritium and the development of the tool to enable this is vital.. In this study, a Behavior of Tritium Analytic Code (BOTANIC) an analytic tool which is capable of analyzing tritium behavior is developed using a chemical process code called gPROMS. BOTANIC was then further verified using the analytic solutions and benchmark codes such as Tritium Permeation Analysis Code (TPAC) and COMSOL. In this study, the Behavior of Tritium Analytic Code, BOTANIC, has been developed using a chemical process code called gPROMS. The code has several distinctive features including non-diluted assumption, flexible applications and adoption of distributed permeation model. Due to these features, BOTANIC has the capability to analyze a wide range of tritium level systems and has a higher accuracy as it has the capacity to solve distributed models. BOTANIC was successfully developed and verified using analytical solution and the benchmark code calculation result. The results showed very good agreement with the analytical solutions and the calculation results of TPAC and COMSOL. Future work will be focused on the total system verification

  2. Developments of fuel performance analysis codes in KEPCO NF

    International Nuclear Information System (INIS)

    Han, H. T.; Choi, J. M.; Jung, C. D.; Yoo, J. S.

    2012-01-01

    The KEPCO NF has developed fuel performance analysis and design code named as ROPER, and utility codes of XGCOL and XDNB in order to perform fuel rod design evaluation for Korean nuclear power plants. The ROPER code intends to cover full range of fuel performance evaluation. The XGCOL code is for the clad flattening evaluation and the XDNB code is for the extensive DNB propagation evaluation. In addition to these, the KEPCO NF is now in the developing stage for 3-dimensional fuel performance analysis code, named as OPER3D, using 3-dimensional FEM for the nest generation within the joint project CANDU ENERGY in order to analyze PCMI behavior and fuel performance under load following operation. Of these, the ROPER code is now in the stage of licensing activities by Korean regulatory body and the other two are almost in the final developing stage. After finishing the developing, licensing activities are to be performed. These activities are intending to acquire competitiveness, originality, vendor-free ownership of fuel performance codes in the KEPCO NF

  3. Twelve gordian knots when developing an organizational code of ethics

    NARCIS (Netherlands)

    Kaptein, Muel; Wempe, Johan

    1998-01-01

    Following the example of the many organizations in the United States which have a code of ethics, an increasing interest on the part of companies, trade organizations, (semi-)governmental organizations and professions in the Netherlands to develop codes of ethics can be witnessed. We have been able

  4. Graphical user interface development for the MARS code

    International Nuclear Information System (INIS)

    Jeong, J.-J.; Hwang, M.; Lee, Y.J.; Kim, K.D.; Chung, B.D.

    2003-01-01

    KAERI has developed the best-estimate thermal-hydraulic system code MARS using the RELAP5/MOD3 and COBRA-TF codes. To exploit the excellent features of the two codes, we consolidated the two codes. Then, to improve the readability, maintainability, and portability of the consolidated code, all the subroutines were completely restructured by employing a modular data structure. At present, a major part of the MARS code development program is underway to improve the existing capabilities. The code couplings with three-dimensional neutron kinetics, containment analysis, and transient critical heat flux calculations have also been carried out. At the same time, graphical user interface (GUI) tools have been developed for user friendliness. This paper presents the main features of the MARS GUI. The primary objective of the GUI development was to provide a valuable aid for all levels of MARS users in their output interpretation and interactive controls. Especially, an interactive control function was designed to allow operator actions during simulation so that users can utilize the MARS code like conventional nuclear plant analyzers (NPAs). (author)

  5. 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)

  6. 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

  7. 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)

  8. 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

  9. 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

  10. Computer codes in nuclear safety, radiation transport and dosimetry; Les codes de calcul en radioprotection, radiophysique et dosimetrie

    Energy Technology Data Exchange (ETDEWEB)

    Bordy, J M; Kodeli, I; Menard, St; Bouchet, J L; Renard, F; Martin, E; Blazy, L; Voros, S; Bochud, F; Laedermann, J P; Beaugelin, K; Makovicka, L; Quiot, A; Vermeersch, F; Roche, H; Perrin, M C; Laye, F; Bardies, M; Struelens, L; Vanhavere, F; Gschwind, R; Fernandez, F; Quesne, B; Fritsch, P; Lamart, St; Crovisier, Ph; Leservot, A; Antoni, R; Huet, Ch; Thiam, Ch; Donadille, L; Monfort, M; Diop, Ch; Ricard, M

    2006-07-01

    The purpose of this conference was to describe the present state of computer codes dedicated to radiation transport or radiation source assessment or dosimetry. The presentations have been parted into 2 sessions: 1) methodology and 2) uses in industrial or medical or research domains. It appears that 2 different calculation strategies are prevailing, both are based on preliminary Monte-Carlo calculations with data storage. First, quick simulations made from a database of particle histories built though a previous Monte-Carlo simulation and secondly, a neuronal approach involving a learning platform generated through a previous Monte-Carlo simulation. This document gathers the slides of the presentations.

  11. Development of computing code system for level 3 PSA

    International Nuclear Information System (INIS)

    Jeong, Jong Tae; Yu, Dong Han; Kim, Seung Hwan.

    1997-07-01

    Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated through wind tunnel experiment. These results will give a physical insight in the development of a new dispersion model. Because there are some discrepancies between the results from Gaussian plume model and those from field test, the effect of terrain on the atmospheric dispersion was investigated by using CTDMPLUS code. Through this study we find that the model which can treat terrain effect is essential in the atmospheric dispersion of radioactive materials and the CTDMPLUS model can be used as a useful tool. And it is suggested that modification of a model and experimental study should be made through the continuous effort. The health effect assessment near the Yonggwang site by using IPE (Individual plant examination) results and its site data was performed. The health effect assessment is an important part of consequence analysis of a nuclear power plant site. The MACCS was used in the assessment. Based on the calculation of CCDF for each risk measure, it is shown that CCDF has a slow slope and thus wide probability distribution in cases of early fatality, early injury, total early fatality risk, and total weighted early fatality risk. And in cases of cancer fatality and population dose within 48km and 80km, the CCDF curve have a steep slope and thus narrow probability distribution. The establishment of methodologies for necessary models for consequence analysis resulting form a server accident in the nuclear power plant was made and a program for consequence analysis was developed. The models include atmospheric transport and diffusion, calculation of exposure doses for various pathways, and assessment of health effects and associated risks. Finally, the economic impact resulting form an accident in a nuclear power plant was investigated. In this study, estimation models for each cost terms that considered in economic

  12. Development of computing code system for level 3 PSA

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Tae; Yu, Dong Han; Kim, Seung Hwan

    1997-07-01

    Among the various research areas of the level 3 PSA, the effect of terrain on the transport of radioactive material was investigated through wind tunnel experiment. These results will give a physical insight in the development of a new dispersion model. Because there are some discrepancies between the results from Gaussian plume model and those from field test, the effect of terrain on the atmospheric dispersion was investigated by using CTDMPLUS code. Through this study we find that the model which can treat terrain effect is essential in the atmospheric dispersion of radioactive materials and the CTDMPLUS model can be used as a useful tool. And it is suggested that modification of a model and experimental study should be made through the continuous effort. The health effect assessment near the Yonggwang site by using IPE (Individual plant examination) results and its site data was performed. The health effect assessment is an important part of consequence analysis of a nuclear power plant site. The MACCS was used in the assessment. Based on the calculation of CCDF for each risk measure, it is shown that CCDF has a slow slope and thus wide probability distribution in cases of early fatality, early injury, total early fatality risk, and total weighted early fatality risk. And in cases of cancer fatality and population dose within 48km and 80km, the CCDF curve have a steep slope and thus narrow probability distribution. The establishment of methodologies for necessary models for consequence analysis resulting form a server accident in the nuclear power plant was made and a program for consequence analysis was developed. The models include atmospheric transport and diffusion, calculation of exposure doses for various pathways, and assessment of health effects and associated risks. Finally, the economic impact resulting form an accident in a nuclear power plant was investigated. In this study, estimation models for each cost terms that considered in economic

  13. 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)

  14. 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.

  15. 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)

  16. 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

  17. 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

  18. Development of the integrated system reliability analysis code MODULE

    International Nuclear Information System (INIS)

    Han, S.H.; Yoo, K.J.; Kim, T.W.

    1987-01-01

    The major components in a system reliability analysis are the determination of cut sets, importance measure, and uncertainty analysis. Various computer codes have been used for these purposes. For example, SETS and FTAP are used to determine cut sets; Importance for importance calculations; and Sample, CONINT, and MOCUP for uncertainty analysis. There have been problems when the codes run each other and the input and output are not linked, which could result in errors when preparing input for each code. The code MODULE was developed to carry out the above calculations simultaneously without linking input and outputs to other codes. MODULE can also prepare input for SETS for the case of a large fault tree that cannot be handled by MODULE. The flow diagram of the MODULE code is shown. To verify the MODULE code, two examples are selected and the results and computation times are compared with those of SETS, FTAP, CONINT, and MOCUP on both Cyber 170-875 and IBM PC/AT. Two examples are fault trees of the auxiliary feedwater system (AFWS) of Korea Nuclear Units (KNU)-1 and -2, which have 54 gates and 115 events, 39 gates and 92 events, respectively. The MODULE code has the advantage that it can calculate the cut sets, importances, and uncertainties in a single run with little increase in computing time over other codes and that it can be used in personal computers

  19. Development of realistic thermal hydraulic system analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Chung, B. D; Kim, K. D. [and others

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.

  20. Development of realistic thermal hydraulic system analysis code

    International Nuclear Information System (INIS)

    Lee, Won Jae; Chung, B. D; Kim, K. D.

    2002-05-01

    The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others

  1. Foundational development of an advanced nuclear reactor integrated safety code

    International Nuclear Information System (INIS)

    Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

    2010-01-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  2. Foundational development of an advanced nuclear reactor integrated safety code.

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  4. 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)

  5. Development of the code for filter calculation

    International Nuclear Information System (INIS)

    Gritzay, O.O.; Vakulenko, M.M.

    2012-01-01

    This paper describes a calculation method, which commonly used in the Neutron Physics Department to develop a new neutron filter or to improve the existing neutron filter. This calculation is the first step of the traditional filter development procedure. It allows easy selection of the qualitative and quantitative contents of a composite filter in order to receive the filtered neutron beam with given parameters

  6. Development of international transport in transboundary regions

    OpenAIRE

    Прокудін, Георгій; Чупайленко, Олексій

    2015-01-01

    Formation of an international cross-border transport and logistics infrastructure meets international standards, increased productivity, transport and innovation activity of enterprises in the cluster, and provide for accelerated socio - economic development of the regions.

  7. 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...

  8. Development of transportation operations requirements

    International Nuclear Information System (INIS)

    Grady, S.T.; Best, R.E.; Danese, F.L.; Peterson, R.W.; Pope, R.B.

    1990-01-01

    Transport conditions at various utility sties vary dramatically in terms of characteristics at and near the site, requirements, administrative procedures, and other factors. Continuation of design efforts for the OCRWM transportation operations system requires that the operating requirements for the transportation system -- quantity of fuel per unit time per site -- be identified so that the effect the variations have on the system can be accommodated. The approach outlined in this paper provides for an identification of specific sites, evaluation of shipment capabilities at each site, and integration of the sites into multi-site shipping campaigns to scope the logistics management problem for the transportation operations system. 1 fig., 1 tab

  9. Development of the next generation reactor analysis code system, MARBLE

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; Kugo, Teruhiko; Ishikawa, Makoto; Tatsumi, Masahiro; Hirai, Yasushi; Hyoudou, Hideaki; Numata, Kazuyuki; Iwai, Takehiko; Jin, Tomoyuki

    2011-03-01

    A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional systems), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system because MARBLE can utilize sub-codes included in the conventional system without any change. On the other hand, burnup analysis functionality for power reactors is improved compared with the conventional system by introducing models on fuel exchange treatment and control rod operation and so on. In addition, MARBLE has newly developed solvers and some new features of burnup calculation by the Krylov sub-space method and nuclear design accuracy evaluation by the extended bias factor method. In the development of MARBLE, the object oriented technology was adopted from the view-point of improvement of the software quality such as flexibility, expansibility, facilitation of the verification by the modularization and assistance of co-development. And, software structure called the two-layer system consisting of scripting language and system development language was applied. As a result, MARBLE is not an independent analysis code system which simply receives input and returns output, but an assembly of components for building an analysis code system (i.e. framework). Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system. SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS. (author)

  10. Spallation neutron production and the current intra-nuclear cascade and transport codes

    International Nuclear Information System (INIS)

    Filges, D.; Goldenbaum, F.

    2001-01-01

    A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models. (orig.)

  11. Modeling a TRIGA Mark II reactor using the Attila three-dimensional deterministic transport code

    International Nuclear Information System (INIS)

    Keller, S.T.; Palmer, T.S.; Wareing, T.A.

    2005-01-01

    A benchmark model of a TRIGA reactor constructed using materials and dimensions similar to existing TRIGA reactors was analyzed using MCNP and the recently developed deterministic transport code Attila TM . The benchmark reactor requires no MCNP modeling approximations, yet is sufficiently complex to validate the new modeling techniques. Geometric properties of the benchmark reactor are specified for use by Attila TM with CAD software. Materials are treated individually in MCNP. Materials used in Attila TM that are clad are homogenized. Attila TM uses multigroup energy discretization. Two cross section libraries were constructed for comparison. A 16 group library collapsed from the SCALE 4.4.a 238 group library provided better results than a seven group library calculated with WIMS-ANL. Values of the k-effective eigenvalue and scalar flux as a function of location and energy were calculated by the two codes. The calculated values for k-effective and spatially averaged neutron flux were found to be in good agreement. Flux distribution by space and energy also agreed well. Attila TM results could be improved with increased spatial and angular resolution and revised energy group structure. (authors)

  12. New Three-Dimensional Neutron Transport Calculation Capability in STREAM Code

    Energy Technology Data Exchange (ETDEWEB)

    Zheng, Youqi [Xi' an Jiaotong University, Xi' an (China); Choi, Sooyoung; Lee, Deokjung [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    The method of characteristics (MOC) is one of the best choices for its powerful capability in the geometry modeling. To reduce the large computational burden in 3D MOC, the 2D/1D schemes were proposed and have achieved great success in the past 10 years. However, such methods have some instability problems during the iterations when the neutron leakage for axial direction is large. Therefore, full 3D MOC methods were developed. A lot of efforts have been devoted to reduce the computational costs. However, it still requires too much memory storage and computational time for the practical modeling of a commercial size reactor core. Recently, a new approach for the 3D MOC calculation without transverse integration has been implemented in the STREAM code. In this approach, the angular flux is expressed as a basis function expansion form of only axial variable z. A new approach based on the axial expansion and 2D MOC sweeping to solve the 3D neutron transport equation is implemented in the STREAM code. This approach avoids using the transverse integration in the traditional 2D/1D scheme of MOC calculation. By converting the 3D equation into the 2D form of angular flux expansion coefficients, it also avoids the complex 3D ray tracing. Current numerical tests using two benchmarks show good accuracy of the new method.

  13. Comparison of two numerical modelling codes for hydraulic and transport calculations in the near-field

    International Nuclear Information System (INIS)

    Kalin, J.; Petkovsek, B.; Montarnal, Ph.; Genty, A.; Deville, E.; Krivic, J.; Ratej, J.

    2011-01-01

    In the past years the Slovenian Performance Analysis/Safety Assessment team has performed many generic studies for the future Slovenian low and intermediate level waste repository, most recently a Special Safety Analysis for the Krsko site. The modelling approach was to split the problem into three parts: near-field (detailed model of the repository), far-field (i.e., geosphere) and biosphere. In the Special Safety Analysis the code used to perform the near-field calculations was Hydrus2D. Recently the team has begun a cooperation with the French Commisariat al'Energie Atomique/Saclay (CEA/Saclay) and, as a part of this cooperation, began investigations into using the Alliances numerical platform for near-field calculations in order to compare the overall approach and calculated results. The article presents the comparison between these two codes for a silo-type repository that was considered in the Special Safety Analysis. The physical layout and characteristics of the repository are presented and a hydraulic and transport model of the repository is developed and implemented in Alliances. Some analysis of sensitivity to mesh fineness and to simulation timestep has been preformed and is also presented. The compared quantity is the output flux of radionuclides on the boundary of the model. Finally the results from Hydrus2D and Alliances are compared and the differences and similarities are commented.

  14. Comparison of two numerical modelling codes for hydraulic and transport calculations in the near-field

    Energy Technology Data Exchange (ETDEWEB)

    Kalin, J., E-mail: jan.kalin@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Petkovsek, B., E-mail: borut.petkovsek@zag.s [Slovenian National Building and Civil Engineering Institute, Dimiceva 12, SI-1000 Ljubljana (Slovenia); Montarnal, Ph., E-mail: philippe.montarnal@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Genty, A., E-mail: alain.genty@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Deville, E., E-mail: estelle.deville@cea.f [CEA/Saclay, DM2S/SFME/LSET, Gif-sur-Yvette, 91191 cedex (France); Krivic, J., E-mail: jure.krivic@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia); Ratej, J., E-mail: joze.ratej@geo-zs.s [Geological Survey of Slovenia, Dimiceva 14, SI-1000 Ljubljana (Slovenia)

    2011-04-15

    In the past years the Slovenian Performance Analysis/Safety Assessment team has performed many generic studies for the future Slovenian low and intermediate level waste repository, most recently a Special Safety Analysis for the Krsko site. The modelling approach was to split the problem into three parts: near-field (detailed model of the repository), far-field (i.e., geosphere) and biosphere. In the Special Safety Analysis the code used to perform the near-field calculations was Hydrus2D. Recently the team has begun a cooperation with the French Commisariat al'Energie Atomique/Saclay (CEA/Saclay) and, as a part of this cooperation, began investigations into using the Alliances numerical platform for near-field calculations in order to compare the overall approach and calculated results. The article presents the comparison between these two codes for a silo-type repository that was considered in the Special Safety Analysis. The physical layout and characteristics of the repository are presented and a hydraulic and transport model of the repository is developed and implemented in Alliances. Some analysis of sensitivity to mesh fineness and to simulation timestep has been preformed and is also presented. The compared quantity is the output flux of radionuclides on the boundary of the model. Finally the results from Hydrus2D and Alliances are compared and the differences and similarities are commented.

  15. Spallation neutron production and the current intra-nuclear cascade and transport codes

    Science.gov (United States)

    Filges, D.; Goldenbaum, F.; Enke, M.; Galin, J.; Herbach, C.-M.; Hilscher, D.; Jahnke, U.; Letourneau, A.; Lott, B.; Neef, R.-D.; Nünighoff, K.; Paul, N.; Péghaire, A.; Pienkowski, L.; Schaal, H.; Schröder, U.; Sterzenbach, G.; Tietze, A.; Tishchenko, V.; Toke, J.; Wohlmuther, M.

    A recent renascent interest in energetic proton-induced production of neutrons originates largely from the inception of projects for target stations of intense spallation neutron sources, like the planned European Spallation Source (ESS), accelerator-driven nuclear reactors, nuclear waste transmutation, and also from the application for radioactive beams. In the framework of such a neutron production, of major importance is the search for ways for the most efficient conversion of the primary beam energy into neutron production. Although the issue has been quite successfully addressed experimentally by varying the incident proton energy for various target materials and by covering a huge collection of different target geometries --providing an exhaustive matrix of benchmark data-- the ultimate challenge is to increase the predictive power of transport codes currently on the market. To scrutinize these codes, calculations of reaction cross-sections, hadronic interaction lengths, average neutron multiplicities, neutron multiplicity and energy distributions, and the development of hadronic showers are confronted with recent experimental data of the NESSI collaboration. Program packages like HERMES, LCS or MCNPX master the prevision of reaction cross-sections, hadronic interaction lengths, averaged neutron multiplicities and neutron multiplicity distributions in thick and thin targets for a wide spectrum of incident proton energies, geometrical shapes and materials of the target generally within less than 10% deviation, while production cross-section measurements for light charged particles on thin targets point out that appreciable distinctions exist within these models.

  16. Development of Regulatory Audit Core Safety Code : COREDAX

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Chae Yong; Jo, Jong Chull; Roh, Byung Hwan [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Jun; Cho, Nam Zin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2005-07-01

    Korea Institute of Nuclear Safety (KINS) has developed a core neutronics simulator, COREDAX code, for verifying core safety of SMART-P reactor, which is technically supported by Korea Advanced Institute of Science and Technology (KAIST). The COREDAX code would be used for regulatory audit calculations of 3- dimendional core neutronics. The COREDAX code solves the steady-state and timedependent multi-group neutron diffusion equation in hexagonal geometry as well as rectangular geometry by analytic function expansion nodal (AFEN) method. AFEN method was developed at KAIST, and it was internationally verified that its accuracy is excellent. The COREDAX code is originally programmed based on the AFEN method. Accuracy of the code on the AFEN method was excellent for the hexagonal 2-dimensional problems, but there was a need for improvement for hexagonal-z 3-dimensional problems. Hence, several solution routines of the AFEN method are improved, and finally the advanced AFEN method is created. COREDAX code is based on the advanced AFEN method . The initial version of COREDAX code is to complete a basic framework, performing eigenvalue calculations and kinetics calculations with thermal-hydraulic feedbacks, for audit calculations of steady-state core design and reactivity-induced accidents of SMART-P reactor. This study describes the COREDAX code for hexagonal geometry.

  17. Development of probabilistic fracture mechanics code PASCAL and user's manual

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Katsuyuki; Onizawa, Kunio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Li, Yinsheng; Kato, Daisuke [Fuji Research Institute Corporation, Tokyo (Japan)

    2001-03-01

    As a part of the aging and structural integrity research for LWR components, a new PFM (Probabilistic Fracture Mechanics) code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed since FY1996. This code evaluates the failure probability of an aged reactor pressure vessel subjected to transient loading such as PTS (Pressurized Thermal Shock). The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the recent development in the fracture mechanics methodologies and computer performance. The code has some new functions in optimized sampling and cell dividing procedure in stratified Monte Carlo simulation, elastic-plastic fracture criterion of R6 method, extension analysis models in semi-elliptical crack, evaluation of effect of thermal annealing and etc. In addition, an input data generator of temperature and stress distribution time histories was also prepared in the code. Functions and performance of the code have been confirmed based on the verification analyses and some case studies on the influence parameters. The present phase of the development will be completed in FY2000. Thus this report provides the user's manual and theoretical background of the code. (author)

  18. Development of the CRIPTE Code for Electromagnetic Coupling

    National Research Council Canada - National Science Library

    Parmantier, Jean-Philippe

    2005-01-01

    .... This code was originally developed as part of an experiment performed under the joint US-France international data exchange program on the atmospheric electricity/aircraft interactions, DEA-AF-79-7336...

  19. Development of code SFINEL (Spent fuel integrity evaluator)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yong Soo; Min, Chin Young; Ohk, Young Kil; Yang, Yong Sik; Kim, Dong Ju; Kim, Nam Ku [Hanyang University, Seoul (Korea)

    1999-01-01

    SFINEL code, an integrated computer program for predicting the spent fuel rod integrity based on burn-up history and major degradation mechanisms, has been developed through this project. This code can sufficiently simulate the power history of a fuel rod during the reactor operation and estimate the degree of deterioration of spent fuel cladding using the recently-developed models on the degradation mechanisms. SFINEL code has been thoroughly benchmarked against the collected in-pile data and operating experiences: deformation and rupture, and cladding oxidation, rod internal pressure creep, then comprehensive whole degradation process. (author). 75 refs., 51 figs., 5 tabs.

  20. Development and application of methods to characterize code uncertainty

    International Nuclear Information System (INIS)

    Wilson, G.E.; Burtt, J.D.; Case, G.S.; Einerson, J.J.; Hanson, R.G.

    1985-01-01

    The United States Nuclear Regulatory Commission sponsors both international and domestic studies to assess its safety analysis codes. The Commission staff intends to use the results of these studies to quantify the uncertainty of the codes with a statistically based analysis method. Development of the methodology is underway. The Idaho National Engineering Laboratory contributions to the early development effort, and testing of two candidate methods are the subjects of this paper

  1. Theoretical Atomic Physics code development IV: LINES, A code for computing atomic line spectra

    International Nuclear Information System (INIS)

    Abdallah, J. Jr.; Clark, R.E.H.

    1988-12-01

    A new computer program, LINES, has been developed for simulating atomic line emission and absorption spectra using the accurate fine structure energy levels and transition strengths calculated by the (CATS) Cowan Atomic Structure code. Population distributions for the ion stages are obtained in LINES by using the Local Thermodynamic Equilibrium (LTE) model. LINES is also useful for displaying the pertinent atomic data generated by CATS. This report describes the use of LINES. Both CATS and LINES are part of the Theoretical Atomic PhysicS (TAPS) code development effort at Los Alamos. 11 refs., 9 figs., 1 tab

  2. Code-first development with Entity Framework

    CERN Document Server

    Barskiy, Sergey

    2015-01-01

    This book is intended for software developers with some prior experience with the Microsoft .NET framework who want to learn how to use Entity Framework. This book will get you up and running quickly, providing many examples that illustrate all the key concepts of Entity Framework.

  3. Development of CANDU 6 Primary Heat Transport System Modeling Program

    International Nuclear Information System (INIS)

    Seo, Hyung-beom; Kim, Sung-min; Park, Joong-woo; Kim, Kwang-su; Ko, Dae-hack; Han, Bong-seob

    2007-01-01

    NUCIRC is a steady-state thermal-hydraulic code used for design and performance analyses of CANDU Heat Transport System. The code is used to build PHT model in Wolsong NPP and to calculate channel flow distribution. Wolsong NPP has to calculate channel flow distribution and quality of coolant at the ROH header after every outage by OPP (Operating Policy and Principal). PHT modeling work is time consuming which need a lot of operation experience and specialty. It is very difficult to build PHT model as plant operator in two weeks which is obligate for plant operation after every outage. That is why Wolsong NPP develop NUMODEL (NUcirc MODELing) with many-years experience and a know-how of using NUCIRC code. NUMODEL is computer program which is used to create PHT model based on utilizing NUCIRC code

  4. Aeroelastic code development activities in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Wright, A.D. [National Renewable Energy Lab., Golden, Colorado (United States)

    1996-09-01

    Designing wind turbines to be fatigue resistant and to have long lifetimes at minimal cost is a major goal of the federal wind program and the wind industry in the United States. To achieve this goal, we must be able to predict critical loads for a wide variety of different wind turbines operating under extreme conditions. The codes used for wind turbine dynamic analysis must be able to analyze a wide range of different wind turbine configurations as well as rapidly predict the loads due to turbulent wind inflow with a minimal set of degrees of freedom. Code development activities in the US have taken a two-pronged approach in order to satisfy both of these criteria: (1) development of a multi-purpose code which can be used to analyze a wide variety of wind turbine configurations without having to develop new equations of motion with each configuration change, and (2) development of specialized codes with minimal sets of specific degrees of freedom for analysis of two- and three-bladed horizontal axis wind turbines and calculation of machine loads due to turbulent inflow. In the first method we have adapted a commercial multi-body dynamics simulation package for wind turbine analysis. In the second approach we are developing specialized codes with limited degrees of freedom, usually specified in the modal domain. This paper will summarize progress to date in the development, validation, and application of these codes. (au) 13 refs.

  5. Computer code selection criteria for flow and transport code(s) to be used in undisturbed vadose zone calculations for TWRS environmental analyses

    International Nuclear Information System (INIS)

    Mann, F.M.

    1998-01-01

    The Tank Waste Remediation System (TWRS) is responsible for the safe storage, retrieval, and disposal of waste currently being held in 177 underground tanks at the Hanford Site. In order to successfully carry out its mission, TWRS must perform environmental analyses describing the consequences of tank contents leaking from tanks and associated facilities during the storage, retrieval, or closure periods and immobilized low-activity tank waste contaminants leaving disposal facilities. Because of the large size of the facilities and the great depth of the dry zone (known as the vadose zone) underneath the facilities, sophisticated computer codes are needed to model the transport of the tank contents or contaminants. This document presents the code selection criteria for those vadose zone analyses (a subset of the above analyses) where the hydraulic properties of the vadose zone are constant in time the geochemical behavior of the contaminant-soil interaction can be described by simple models, and the geologic or engineered structures are complicated enough to require a two-or three dimensional model. Thus, simple analyses would not need to use the fairly sophisticated codes which would meet the selection criteria in this document. Similarly, those analyses which involve complex chemical modeling (such as those analyses involving large tank leaks or those analyses involving the modeling of contaminant release from glass waste forms) are excluded. The analyses covered here are those where the movement of contaminants can be relatively simply calculated from the moisture flow. These code selection criteria are based on the information from the low-level waste programs of the US Department of Energy (DOE) and of the US Nuclear Regulatory Commission as well as experience gained in the DOE Complex in applying these criteria. Appendix table A-1 provides a comparison between the criteria in these documents and those used here. This document does not define the models (that

  6. 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.

  7. Aurora T: a Monte Carlo code for transportation of neutral atoms in a toroidal plasma

    International Nuclear Information System (INIS)

    Bignami, A.; Chiorrini, R.

    1982-01-01

    This paper contains a short description of Aurora code. This code have been developed at Princeton with Monte Carlo method for calculating neutral gas in cylindrical plasma. In this work subroutines such one can take in account toroidal geometry are developed

  8. The Premar Code for the Monte Carlo Simulation of Radiation Transport In the Atmosphere

    International Nuclear Information System (INIS)

    Cupini, E.; Borgia, M.G.; Premuda, M.

    1997-03-01

    The Montecarlo code PREMAR is described, which allows the user to simulate the radiation transport in the atmosphere, in the ultraviolet-infrared frequency interval. A plan multilayer geometry is at present foreseen by the code, witch albedo possibility at the lower boundary surface. For a given monochromatic point source, the main quantities computed by the code are the absorption spatial distributions of aerosol and molecules, together with the related atmospheric transmittances. Moreover, simulation of of Lidar experiments are foreseen by the code, the source and telescope fields of view being assigned. To build-up the appropriate probability distributions, an input data library is assumed to be read by the code. For this purpose the radiance-transmittance LOWTRAN-7 code has been conveniently adapted as a source of the library so as to exploit the richness of information of the code for a large variety of atmospheric simulations. Results of applications of the PREMAR code are finally presented, with special reference to simulations of Lidar system and radiometer experiments carried out at the Brasimone ENEA Centre by the Environment Department

  9. 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)

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

    International Nuclear Information System (INIS)

    Zazula, J.M.

    1983-01-01

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

  11. 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.

  12. 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.

  13. Developing a Pupil Transportation Manual.

    Science.gov (United States)

    Douglas, Dave

    1987-01-01

    District-level pupil transportation manuals that contain clear, concise information about objectives, policies, and regulations are a must. These manuals should also specify procedures concerning evaluation processes, personnel recruitment and selection, and the driver training program. (MLH)

  14. Structural reliability methods: Code development status

    Science.gov (United States)

    Millwater, Harry R.; Thacker, Ben H.; Wu, Y.-T.; Cruse, T. A.

    1991-05-01

    The Probabilistic Structures Analysis Method (PSAM) program integrates state of the art probabilistic algorithms with structural analysis methods in order to quantify the behavior of Space Shuttle Main Engine structures subject to uncertain loadings, boundary conditions, material parameters, and geometric conditions. An advanced, efficient probabilistic structural analysis software program, NESSUS (Numerical Evaluation of Stochastic Structures Under Stress) was developed as a deliverable. NESSUS contains a number of integrated software components to perform probabilistic analysis of complex structures. A nonlinear finite element module NESSUS/FEM is used to model the structure and obtain structural sensitivities. Some of the capabilities of NESSUS/FEM are shown. A Fast Probability Integration module NESSUS/FPI estimates the probability given the structural sensitivities. A driver module, PFEM, couples the FEM and FPI. NESSUS, version 5.0, addresses component reliability, resistance, and risk.

  15. Development of computer code in PNC, 8

    International Nuclear Information System (INIS)

    Ohhira, Mitsuru

    1990-01-01

    Private buildings applied base isolation system, are on the practical stage now. So, under Construction and Maintenance Management Office, we are doing an application study of base isolation system to nuclear fuel facilities. On the process of this study, we have developed Dynamic Analysis Program-Base Isolation System (DAP-BS) which is able to run a 32-bit personal computer. Using this program, we can analyze a 3-dimensional structure, and evaluate the various properties of base isolation parts that are divided into maximum 16 blocks. And from the results of some simulation analyses, we thought that DAP-BS had good reliability and marketability. So, we put DAP-BS on the market. (author)

  16. Development and validation of a nodal code for core calculation

    International Nuclear Information System (INIS)

    Nowakowski, Pedro Mariano

    2004-01-01

    The code RHENO solves the multigroup three-dimensional diffusion equation using a nodal method of polynomial expansion.A comparative study has been made between this code and present internationals nodal diffusion codes, resulting that the RHENO is up to date.The RHENO has been integrated to a calculation line and has been extend to make burnup calculations.Two methods for pin power reconstruction were developed: modulation and imbedded. The modulation method has been implemented in a program, while the implementation of the imbedded method will be concluded shortly.The validation carried out (that includes experimental data of a MPR) show very good results and calculation efficiency

  17. Development and validation of a fuel performance analysis code

    International Nuclear Information System (INIS)

    Majalee, Aaditya V.; Chaturvedi, S.

    2015-01-01

    CAD has been developing a computer code 'FRAVIZ' for calculation of steady-state thermomechanical behaviour of nuclear reactor fuel rods. It contains four major modules viz., Thermal module, Fission Gas Release module, Material Properties module and Mechanical module. All these four modules are coupled to each other and feedback from each module is fed back to others to get a self-consistent evolution in time. The computer code has been checked against two FUMEX benchmarks. Modelling fuel performance in Advance Heavy Water Reactor would require additional inputs related to the fuel and some modification in the code.(author)

  18. Development and validation of sodium fire analysis code ASSCOPS

    International Nuclear Information System (INIS)

    Ohno, Shuji

    2001-01-01

    A version 2.1 of the ASSCOPS sodium fire analysis code was developed to evaluate the thermal consequences of a sodium leak and consequent fire in LMFBRs. This report describes the computational models and the validation studies using the code. The ASSCOPS calculates sodium droplet and pool fire, and consequential heat/mass transfer behavior. Analyses of sodium pool or spray fire experiments confirmed that this code and parameters used in the validation studies gave valid results on the thermal consequences of sodium leaks and fires. (author)

  19. Development of safety analysis codes for light water reactor

    International Nuclear Information System (INIS)

    Akimoto, Masayuki

    1985-01-01

    An overview is presented of currently used major codes for the prediction of thermohydraulic transients in nuclear power plants. The overview centers on the two-phase fluid dynamics of the coolant system and the assessment of the codes. Some of two-phase phenomena such as phase separation are not still predicted with engineering accuracy. MINCS-PIPE are briefly introduced. The MINCS-PIPE code is to assess constitutive relations and to aid development of various experimental correlations for 1V1T model to 2V2T model. (author)

  20. Strategy development management of Multimodal Transport Network

    Directory of Open Access Journals (Sweden)

    Nesterova Natalia S.

    2016-01-01

    Full Text Available The article gives a brief overview of works on the development of transport infrastructure for multimodal transportation and integration of Russian transport system into the international transport corridors. The technology for control of the strategy, that changes shape and capacity of Multi-modal Transport Network (MTN, is considered as part of the methodology for designing and development of MTN. This technology allows to carry out strategic and operational management of the strategy implementation based on the use of the balanced scorecard.

  1. 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

  2. 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

  3. Development of LWR fuel performance code FEMAXI-6

    International Nuclear Information System (INIS)

    Suzuki, Motoe

    2006-01-01

    LWR fuel performance code: FEMAXI-6 (Finite Element Method in AXIs-symmetric system) is a representative fuel analysis code in Japan. Development history, background, design idea, features of model, and future are stated. Characteristic performance of LWR fuel and analysis code, what is model, development history of FEMAXI, use of FEMAXI code, fuel model, and a special feature of FEMAXI model is described. As examples of analysis, PCMI (Pellet-Clad Mechanical Interaction), fission gas release, gap bonding, and fission gas bubble swelling are reported. Thermal analysis and dynamic analysis system of FEMAXI-6, function block at one time step of FEMAXI-6, analytical example of PCMI in the output increase test by FEMAXI-III, analysis of fission gas release in Halden reactor by FEMAXI-V, comparison of the center temperature of fuel in Halden reactor, and analysis of change of diameter of fuel rod in high burn up BWR fuel are shown. (S.Y.)

  4. SCDAP/RELAP5 code development and assessment

    International Nuclear Information System (INIS)

    Allison, C.M.; Hohorst, J.K.

    1996-01-01

    The SCDAP/RELAP5 computer code is designed to describe the overall reactor coolant system thermal-hydraulic response, core damage progression, and fission product release during severe accidents. The code is being developed at the Idaho National Engineering Laboratory under the primary sponsorship of the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission. The current version of the code is SCDAP/RELAP5/MOD3.1e. Although MOD3.1e contains a number of significant improvements since the initial version of MOD3.1 was released, new models to treat the behavior of the fuel and cladding during reflood have had the most dramatic impact on the code's calculations. This paper provides a brief description of the new reflood models, presents highlights of the assessment of the current version of MOD3.1, and discusses future SCDAP/RELAP5/MOD3.2 model development activities

  5. LASER-R a computer code for reactor cell and burnup calculations in neutron transport theory

    International Nuclear Information System (INIS)

    Cristian, I.; Cirstoiu, B.; Dumitrache, I.; Cepraga, D.

    1976-04-01

    The LASER-R code is an IBM 370/135 version of the Westinghouse code, LASER, based on the THERMOS and MUFT codes developped by Poncelet. It can be used to perform thermal reactor cell calculations and burnup calculations. The cell exhibits 3-4 concentric areas: fuel, cladding, moderator and scattering ring. Besides directions for use, a short description of the physical model, numerical methods and output is presented

  6. 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

  7. 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.

  8. 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

  9. 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...

  10. Computer codes developed in FRG to analyse hypothetical meltdown accidents

    International Nuclear Information System (INIS)

    Hassmann, K.; Hosemann, J.P.; Koerber, H.; Reineke, H.

    1978-01-01

    It is the purpose of this paper to give the status of all significant computer codes developed in the core melt-down project which is incorporated in the light water reactor safety research program of the Federal Ministry of Research and Technology. For standard pressurized water reactors, results of some computer codes will be presented, describing the course and the duration of the hypothetical core meltdown accident. (author)

  11. Efficacy of behavioural interventions for transport behaviour change: systematic review, meta-analysis and intervention coding.

    Science.gov (United States)

    Arnott, Bronia; Rehackova, Lucia; Errington, Linda; Sniehotta, Falko F; Roberts, Jennifer; Araujo-Soares, Vera

    2014-11-28

    Reducing reliance on motorised transport and increasing use of more physically active modes of travel may offer an opportunity to address physical inactivity. This review evaluates the evidence for the effects of behavioural interventions to reduce car use for journeys made by adults and codes intervention development and content. The review follows the procedure stated in the registration protocol published in the PROSPERO database (registration number CRD42011001797). Controlled studies evaluating behavioural interventions to reduce car use compared with no interventions or alternative interventions on outcome measures of transport behaviours taken in adult participants are included in this review. Searches were conducted on all records in Applied Social Sciences Index and Abstracts (ASSIA), Ovid Embase, Ovid Medline, Ovid PsycInfo, Scopus, Sociological Abstracts, Transportation Research Information Service (TRIS), Transportation Research International Documentation (TRID), and Web of Science databases. Peer reviewed publications in English language meeting the inclusion criteria are eligible. Methodological quality is assessed using the Cochrane Risk of Bias Tool. Interventions are categorised in terms of behavioural frameworks, theories and techniques. 15 full text articles are included, representing 13 unique studies, with 4895 participants and 27 intervention arms. Risk of bias across the review is appraised as considerable due to the unclear methodological quality of individual studies. Heterogeneity of included studies is considerable. Meta-analyses reveal no significant effect on reduction of frequency of car use or on increasing the proportion of journeys by alternative, more active modes of transport. There is insufficient data relating to alternative outcomes such as distance and duration which may have important health implications. Interventions were top-down but could not be described as theory-based. Intervention efficacy was associated with the use

  12. Code development for eigenvalue total sensitivity analysis and total uncertainty analysis

    International Nuclear Information System (INIS)

    Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Zu, Tiejun; Shen, Wei

    2015-01-01

    Highlights: • We develop a new code for total sensitivity and uncertainty analysis. • The implicit effects of cross sections can be considered. • The results of our code agree well with TSUNAMI-1D. • Detailed analysis for origins of implicit effects is performed. - Abstract: The uncertainties of multigroup cross sections notably impact eigenvalue of neutron-transport equation. We report on a total sensitivity analysis and total uncertainty analysis code named UNICORN that has been developed by applying the direct numerical perturbation method and statistical sampling method. In order to consider the contributions of various basic cross sections and the implicit effects which are indirect results of multigroup cross sections through resonance self-shielding calculation, an improved multigroup cross-section perturbation model is developed. The DRAGON 4.0 code, with application of WIMSD-4 format library, is used by UNICORN to carry out the resonance self-shielding and neutron-transport calculations. In addition, the bootstrap technique has been applied to the statistical sampling method in UNICORN to obtain much steadier and more reliable uncertainty results. The UNICORN code has been verified against TSUNAMI-1D by analyzing the case of TMI-1 pin-cell. The numerical results show that the total uncertainty of eigenvalue caused by cross sections can reach up to be about 0.72%. Therefore the contributions of the basic cross sections and their implicit effects are not negligible

  13. Radiation transport simulation in gamma irradiator systems using E G S 4 Monte Carlo code and dose mapping calculations based on point kernel technique

    International Nuclear Information System (INIS)

    Raisali, G.R.

    1992-01-01

    A series of computer codes based on point kernel technique and also Monte Carlo method have been developed. These codes perform radiation transport calculations for irradiator systems having cartesian, cylindrical and mixed geometries. The monte Carlo calculations, the computer code 'EGS4' has been applied to a radiation processing type problem. This code has been acompanied by a specific user code. The set of codes developed include: GCELLS, DOSMAPM, DOSMAPC2 which simulate the radiation transport in gamma irradiator systems having cylinderical, cartesian, and mixed geometries, respectively. The program 'DOSMAP3' based on point kernel technique, has been also developed for dose rate mapping calculations in carrier type gamma irradiators. Another computer program 'CYLDETM' as a user code for EGS4 has been also developed to simulate dose variations near the interface of heterogeneous media in gamma irradiator systems. In addition a system of computer codes 'PRODMIX' has been developed which calculates the absorbed dose in the products with different densities. validation studies of the calculated results versus experimental dosimetry has been performed and good agreement has been obtained

  14. The Development of the World Anti-Doping Code.

    Science.gov (United States)

    Young, Richard

    2017-01-01

    This chapter addresses both the development and substance of the World Anti-Doping Code, which came into effect in 2003, as well as the subsequent Code amendments, which came into effect in 2009 and 2015. Through an extensive process of stakeholder input and collaboration, the World Anti-Doping Code has transformed the hodgepodge of inconsistent and competing pre-2003 anti-doping rules into a harmonized and effective approach to anti-doping. The Code, as amended, is now widely recognized worldwide as the gold standard in anti-doping. The World Anti-Doping Code originally went into effect on January 1, 2004. The first amendments to the Code went into effect on January 1, 2009, and the second amendments on January 1, 2015. The Code and the related international standards are the product of a long and collaborative process designed to make the fight against doping more effective through the adoption and implementation of worldwide harmonized rules and best practices. © 2017 S. Karger AG, Basel.

  15. Cooperation of experts' opinion, experiment and computer code development

    International Nuclear Information System (INIS)

    Wolfert, K.; Hicken, E.

    The connection between code development, code assessment and confidence in the analysis of transients will be discussed. In this manner, the major sources of errors in the codes and errors in applications of the codes will be shown. Standard problem results emphasize that, in order to have confidence in licensing statements, the codes must be physically realistic and the code user must be qualified and experienced. We will discuss why there is disagreement between the licensing authority and vendor concerning assessment of the fullfillment of safety goal requirements. The answer to the question lies in the different confidence levels of the assessment of transient analysis. It is expected that a decrease in the disagreement will result from an increased confidence level. Strong efforts will be made to increase this confidence level through improvements in the codes, experiments and related organizational strcutures. Because of the low probability for loss-of-coolant-accidents in the nuclear industry, assessment must rely on analytical techniques and experimental investigations. (orig./HP) [de

  16. Software requirements specification document for the AREST code development

    International Nuclear Information System (INIS)

    Engel, D.W.; McGrail, B.P.; Whitney, P.D.; Gray, W.J.; Williford, R.E.; White, M.D.; Eslinger, P.W.; Altenhofen, M.K.

    1993-11-01

    The Analysis of the Repository Source Term (AREST) computer code was selected in 1992 by the U.S. Department of Energy. The AREST code will be used to analyze the performance of an underground high level nuclear waste repository. The AREST code is being modified by the Pacific Northwest Laboratory (PNL) in order to evaluate the engineered barrier and waste package designs, model regulatory compliance, analyze sensitivities, and support total systems performance assessment modeling. The current version of the AREST code was developed to be a very useful tool for analyzing model uncertainties and sensitivities to input parameters. The code has also been used successfully in supplying source-terms that were used in a total systems performance assessment. The current version, however, has been found to be inadequate for the comparison and selection of a design for the waste package. This is due to the assumptions and simplifications made in the selection of the process and system models. Thus, the new version of the AREST code will be designed to focus on the details of the individual processes and implementation of more realistic models. This document describes the requirements of the new models that will be implemented. Included in this document is a section describing the near-field environmental conditions for this waste package modeling, description of the new process models that will be implemented, and a description of the computer requirements for the new version of the AREST code

  17. MILSTAMP TACs: Military Standard Transportation and Movement Procedures Transportation Account Codes. Volume 2

    Science.gov (United States)

    1987-02-15

    82302 F 13211 PT VERDE WPB 82311 F 13212 PT SWIFT WPB 82312 E. 13214 PT THATCHER WPB 82314 E 13218 PT HERRON WPB 82318 C 13232 PT ROBERTS WPB 82332 E...Identifies DOT, FAA Logistica Center, OkIanhea City, as an organization to be billed. 4th Position Code A Ia assigned by DOT, rAA. Identifies appropriation

  18. Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Wagner, John C.; Peplow, Douglas E.; Mosher, Scott W.; Evans, Thomas M.

    2010-01-01

    This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(10 2-4 ), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.

  19. Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes, and Applications at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Wagner, John C.; Peplow, Douglas E.; Mosher, Scott W.; Evans, Thomas M.

    2010-01-01

    This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(102-4), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications.

  20. Review of hybrid (deterministic/Monte Carlo) radiation transport methods, codes, and applications at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Wagner, J.C.; Peplow, D.E.; Mosher, S.W.; Evans, T.M.

    2010-01-01

    This paper provides a review of the hybrid (Monte Carlo/deterministic) radiation transport methods and codes used at the Oak Ridge National Laboratory and examples of their application for increasing the efficiency of real-world, fixed-source Monte Carlo analyses. The two principal hybrid methods are (1) Consistent Adjoint Driven Importance Sampling (CADIS) for optimization of a localized detector (tally) region (e.g., flux, dose, or reaction rate at a particular location) and (2) Forward Weighted CADIS (FW-CADIS) for optimizing distributions (e.g., mesh tallies over all or part of the problem space) or multiple localized detector regions (e.g., simultaneous optimization of two or more localized tally regions). The two methods have been implemented and automated in both the MAVRIC sequence of SCALE 6 and ADVANTG, a code that works with the MCNP code. As implemented, the methods utilize the results of approximate, fast-running 3-D discrete ordinates transport calculations (with the Denovo code) to generate consistent space- and energy-dependent source and transport (weight windows) biasing parameters. These methods and codes have been applied to many relevant and challenging problems, including calculations of PWR ex-core thermal detector response, dose rates throughout an entire PWR facility, site boundary dose from arrays of commercial spent fuel storage casks, radiation fields for criticality accident alarm system placement, and detector response for special nuclear material detection scenarios and nuclear well-logging tools. Substantial computational speed-ups, generally O(10 2-4 ), have been realized for all applications to date. This paper provides a brief review of the methods, their implementation, results of their application, and current development activities, as well as a considerable list of references for readers seeking more information about the methods and/or their applications. (author)

  1. Multi-dimensional Code Development for Safety Analysis of LMR

    International Nuclear Information System (INIS)

    Ha, K. S.; Jeong, H. Y.; Kwon, Y. M.; Lee, Y. B.

    2006-08-01

    A liquid metal reactor loaded a metallic fuel has the inherent safety mechanism due to the several negative reactivity feedback. Although this feature demonstrated through experiments in the EBR-II, any of the computer programs until now did not exactly analyze it because of the complexity of the reactivity feedback mechanism. A multi-dimensional detail program was developed through the International Nuclear Energy Research Initiative(INERI) from 2003 to 2005. This report includes the numerical coupling the multi-dimensional program and SSC-K code which is used to the safety analysis of liquid metal reactors in KAERI. The coupled code has been proved by comparing the analysis results using the code with the results using SAS-SASSYS code of ANL for the UTOP, ULOF, and ULOHS applied to the safety analysis for KALIMER-150

  2. Development of 1D Liner Compression Code for IDL

    Science.gov (United States)

    Shimazu, Akihisa; Slough, John; Pancotti, Anthony

    2015-11-01

    A 1D liner compression code is developed to model liner implosion dynamics in the Inductively Driven Liner Experiment (IDL) where FRC plasmoid is compressed via inductively-driven metal liners. The driver circuit, magnetic field, joule heating, and liner dynamics calculations are performed at each time step in sequence to couple these effects in the code. To obtain more realistic magnetic field results for a given drive coil geometry, 2D and 3D effects are incorporated into the 1D field calculation through use of correction factor table lookup approach. Commercial low-frequency electromagnetic fields solver, ANSYS Maxwell 3D, is used to solve the magnetic field profile for static liner condition at various liner radius in order to derive correction factors for the 1D field calculation in the code. The liner dynamics results from the code is verified to be in good agreement with the results from commercial explicit dynamics solver, ANSYS Explicit Dynamics, and previous liner experiment. The developed code is used to optimize the capacitor bank and driver coil design for better energy transfer and coupling. FRC gain calculations are also performed using the liner compression data from the code for the conceptual design of the reactor sized system for fusion energy gains.

  3. Development of REFLA/TRAC code for engineering work station

    International Nuclear Information System (INIS)

    Ohnuki, Akira; Akimoto, Hajime; Murao, Yoshio

    1994-03-01

    The REFLA/TRAC code is a best-estimate code which is expected to check reactor safety analysis codes for light water reactors (LWRs) and to perform accident analyses for LWRs and also for an advanced LWR. Therefore, a high predictive capability is required and the assessment of each physical model becomes important because the models govern the predictive capability. In the case of the assessment of three-dimensional models in REFLA/TRAC code, a conventional large computer is being used and it is difficult to perform the assessment efficiently because the turnaround time for the calculation and the analysis is long. Then, a REFLA/TRAC code which can run on an engineering work station (EWS) was developed. Calculational speed of the current EWS is the same order as that of large computers and the EWS has an excellent function for multidimensional graphical drawings. Besides, the plotting processors for X-Y drawing and for two-dimensional graphical drawing were developed in order to perform efficient analyses for three-dimensional calculations. In future, we can expect that the assessment of three-dimensional models becomes more efficient by introducing an EWS with higher calculational speed and with improved graphical drawings. In this report, each outline for the following three programs is described: (1) EWS version of REFLA/TRAC code, (2) Plot processor for X-Y drawing and (3) Plot processor for two-dimensional graphical drawing. (author)

  4. 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)

  5. 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.

  6. Transportation planning for sustainable development

    NARCIS (Netherlands)

    Zuidgeest, M.H.P.; van Maarseveen, M.F.A.M.

    2000-01-01

    Over many decades transportation and traffic have grown at a steady pace, and this trend is likely to continue because of a combination of various demand and supply factors [Grübler, 1993], especially in urban areas in both the industrialised and industrialising world. At the same time the negative

  7. Development of interfacial area transport equation

    International Nuclear Information System (INIS)

    Kim, Seung Jin; Ishii, Mamoru; Kelly, Joseph

    2005-01-01

    The interfacial area transport equation dynamically models the changes in interfacial structures along the flow field by mechanistically modeling the creation and destruction of dispersed phase. Hence, when employed in the numerical thermal-hydraulic system analysis codes, it eliminates artificial bifurcations stemming from the use of the static flow regime transition criteria. Accounting for the substantial differences in the transport mechanism for various sizes of bubbles, the transport equation is formulated for two characteristic groups of bubbles. The group 1 equation describes the transport of small-dispersed bubbles, whereas the group 2 equation describes the transport of large cap, slug or churn-turbulent bubbles. To evaluate the feasibility and reliability of interfacial area transport equation available at present, it is benchmarked by an extensive database established in various two-phase flow configurations spanning from bubbly to churn-turbulent flow regimes. The geometrical effect in interfacial area transport is examined by the data acquired in vertical air-water two-phase flow through round pipes of various sizes and a confined flow duct, and by those acquired in vertical co-current downward air-water two-phase flow through round pipes of two different sizes

  8. Parallelization of MCNP Monte Carlo neutron and photon transport code in parallel virtual machine and message passing interface

    International Nuclear Information System (INIS)

    Deng Li; Xie Zhongsheng

    1999-01-01

    The coupled neutron and photon transport Monte Carlo code MCNP (version 3B) has been parallelized in parallel virtual machine (PVM) and message passing interface (MPI) by modifying a previous serial code. The new code has been verified by solving sample problems. The speedup increases linearly with the number of processors and the average efficiency is up to 99% for 12-processor. (author)

  9. Theory and code development for evaluation of tritium retention and exhaust in fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohya, Kaoru; Inai, Kensuke [Univ. of Tokushima, Institute of Technology and Science, Tokushima, Tokushima (Japan); Shimizu, Katsuhiro; Takizuka, Tomonori; Kawashima, Hisato; Hoshino, Kazuo [Japan Atomic Energy Agency, Fusion Research and Development Directorate, Naka, Ibaraki (Japan); Hatayama, Akiyoshi; Toma, Mitsunori [Keio Univ., Faculty of Science and Technology, Yokohama, Kanagawa (Japan); Tomita, Yukihiro; Kawamura, Gakushi; Ashikawa, Naoko; Nakamura, Hiroaki; Ito, Atsushi; Kato, Daiji [National Inst. for Fusion Science, Toki, Gifu (Japan); Tanaka, Yasunori [Kanazawa Univ., College of Science and Engineering, Kanazawa, Ishikawa (Japan); Ono, Tadayoshi; Muramoto, Tetsuya [Okayama Univ. of Science, Faculty of Informatics, Okayama, Okayama (Japan); Kenmotsu, Takahiro [Doshisha Univ., Faculty of Life and Medical Science, Kiyotanabe, Kyoto (Japan)

    2009-10-15

    As a part of the grant-in-aid for scientific research on priority areas entitled 'frontiers of tritium researches toward fusion reactors', coordinated three research programs on the theory and code development for evaluation of tritium retention and exhaust in fusion reactor have been conducted by the A02 team. They include: (1) Tritium transport in fusion plasmas and the adsorption and desorption property of tritium in plasma-facing components. (2) Behavior of dusts in fusion plasmas and their adsorption property of tritium. (3) Development of computer codes to simulate tritium retention in and release from plasma-facing materials. In order to study these issues, considerable effort has been paid to the development of computer codes and the database system. (J.P.N.)

  10. Theory and code development for evaluation of tritium retention and exhaust in fusion reactor

    International Nuclear Information System (INIS)

    Ohya, Kaoru; Inai, Kensuke; Shimizu, Katsuhiro; Takizuka, Tomonori; Kawashima, Hisato; Hoshino, Kazuo; Hatayama, Akiyoshi; Toma, Mitsunori; Tomita, Yukihiro; Kawamura, Gakushi; Ashikawa, Naoko; Nakamura, Hiroaki; Ito, Atsushi; Kato, Daiji; Tanaka, Yasunori; Ono, Tadayoshi; Muramoto, Tetsuya; Kenmotsu, Takahiro

    2009-01-01

    As a part of the grant-in-aid for scientific research on priority areas entitled 'frontiers of tritium researches toward fusion reactors', coordinated three research programs on the theory and code development for evaluation of tritium retention and exhaust in fusion reactor have been conducted by the A02 team. They include: (1) Tritium transport in fusion plasmas and the adsorption and desorption property of tritium in plasma-facing components. (2) Behavior of dusts in fusion plasmas and their adsorption property of tritium. (3) Development of computer codes to simulate tritium retention in and release from plasma-facing materials. In order to study these issues, considerable effort has been paid to the development of computer codes and the database system. (J.P.N.)

  11. Development of a national code of practice for structural masonry ...

    African Journals Online (AJOL)

    The problems and constraints faced by most developing countries, particularly Ghana, in developing codes of practice for structural masonry are highlighted. The steps that must be undertaken through the coordinated efforts of the National Standards Boards, Research Institutions, Universities and Professional Bodies in the ...

  12. Development of FBR integrity system code. Basic concept

    International Nuclear Information System (INIS)

    Asayama, Tai

    2001-05-01

    For fast breeder reactors to be commercialized, they must be more reliable, safer, and at the same, economically competitive with future light water reactors. Innovation of elevated temperature structural design standard is necessary to achieve this goal. The most powerful way is to enlarge the scope of structural integrity code to cover items other than design evaluation that has been addressed in existing codes. Items that must be newly covered are prerequisites of design, fabrication, examination, operation and maintenance, etc. This allows designers to choose the most economical combination of design variations to achieve specific reliability that is needed for a particular component. Designing components by this concept, a cost-minimum design of a whole plant can be realized. By determining the reliability that must be achieved for a component by risk technologies, further economical improvement can be expected by avoiding excessive quality. Recognizing the necessity for the codes based on the new concept, the development of 'FBR integrity system code' began in 2000. Research and development will last 10 years. For this development, the basic logistics and system as well as technologies that materialize the concept are necessary. Original logistics and system must be developed, because no existing researches are available in and out of Japan. This reports presents the results of the work done in the first year regarding the basic idea, methodology, and structure of the code. (author)

  13. New scope covered by PHITS. Particle and heavy ion transport code system

    International Nuclear Information System (INIS)

    Nakamura, Takashi; Niita, Koji; Iwase, Hiroshi; Sato, Tatsuhiko

    2006-01-01

    PHITS is a general high energy transport calculation code from hadron to heavy ions, which embedded in NMTC-JAM with JQMD code. Outline of PHITS and many application examples are stated. PHITS has been used by the shielding calculations of J-PARC, GSI, RIA and Big-RIPS and the good results were reported. The evaluation of exposure dose of astronauts, airmen, proton and heavy ion therapy, and estimation of error frequency of semiconductor software are explained as the application examples. Relation between the event generator and Monte Carlo method and the future are described. (S.Y.)

  14. Accuracy estimation for intermediate and low energy neutron transport calculation with Monte Carlo code MCNP

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi

    1987-02-01

    Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  16. 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

  17. Development of Advanced Suite of Deterministic Codes for VHTR Physics Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog; Cho, J. Y.; Lee, K. H. (and others)

    2007-07-15

    Advanced Suites of deterministic codes for VHTR physics analysis has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. These code suites include the conventional 2-step procedure in which a few group constants are generated by a transport lattice calculation, and the reactor physics analysis is performed by a 3-dimensional diffusion calculation, and a whole core transport code that can model local heterogeneities directly at the core level. Particular modeling issues in physics analysis of the gas-cooled VHTRs were resolved, which include a double heterogeneity of the coated fuel particles, a neutron streaming in the coolant channels, a strong core-reflector interaction, and large spectrum shifts due to changes of the surrounding environment, temperature and burnup. And the geometry handling capability of the DeCART code were extended to deal with the hexagonal fuel elements of the VHTR core. The developed code suites were validated and verified by comparing the computational results with those of the Monte Carlo calculations for the benchmark problems.

  18. Development of throughflow calculation code for axial flow compressors

    International Nuclear Information System (INIS)

    Kim, Ji Hwan; Kim, Hyeun Min; No, Hee Cheon

    2005-01-01

    The power conversion systems of the current HTGRs are based on closed Brayton cycle and major concern is thermodynamic performance of the axial flow helium gas turbines. Particularly, the helium compressor has some unique design challenges compared to the air-breathing compressor such as high hub-to-tip ratios throughout the machine and a large number of stages due to the physical property of the helium and thermodynamic cycle. Therefore, it is necessary to develop a design and analysis code for helium compressor that can estimate the design point and off-design performance accurately. KAIST nuclear system laboratory has developed a compressor design and analysis code by means of throughflow calculation and several loss models. This paper presents the outline of the development of a throughflow calculation code and its verification results

  19. Development of a Fully-Automated Monte Carlo Burnup Code Monteburns

    International Nuclear Information System (INIS)

    Poston, D.I.; Trellue, H.R.

    1999-01-01

    Several computer codes have been developed to perform nuclear burnup calculations over the past few decades. In addition, because of advances in computer technology, it recently has become more desirable to use Monte Carlo techniques for such problems. Monte Carlo techniques generally offer two distinct advantages over discrete ordinate methods: (1) the use of continuous energy cross sections and (2) the ability to model detailed, complex, three-dimensional (3-D) geometries. These advantages allow more accurate burnup results to be obtained, provided that the user possesses the required computing power (which is required for discrete ordinate methods as well). Several linkage codes have been written that combine a Monte Carlo N-particle transport code (such as MCNP TM ) with a radioactive decay and burnup code. This paper describes one such code that was written at Los Alamos National Laboratory: monteburns. Monteburns links MCNP with the isotope generation and depletion code ORIGEN2. The basis for the development of monteburns was the need for a fully automated code that could perform accurate burnup (and other) calculations for any 3-D system (accelerator-driven or a full reactor core). Before the initial development of monteburns, a list of desired attributes was made and is given below. o The code should be fully automated (that is, after the input is set up, no further user interaction is required). . The code should allow for the irradiation of several materials concurrently (each material is evaluated collectively in MCNP and burned separately in 0RIGEN2). o The code should allow the transfer of materials (shuffling) between regions in MCNP. . The code should allow any materials to be added or removed before, during, or after each step in an automated fashion. . The code should not require the user to provide input for 0RIGEN2 and should have minimal MCNP input file requirements (other than a working MCNP deck). . The code should be relatively easy to use

  20. Development and assessment of the COBRA/RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jae Jun; Ha, Kwi Seok; Sim, Seok Ku

    1997-04-01

    The COBRA/RELAP5 code, a merged version of the COBRA-TF and RELAP5/MOD3.2 codes, has been developed to combine the realistic three-dimensional reactor vessel model of COBRA-TF with RELAP5/MOD3, thus to produce an advanced system analysis code with a multidimensional thermal-hydraulic module. This report provides the integration scheme of the two codes and the results of developmental assessments. These includes single channel tests, manometric flow oscillation problem, THTF Test 105, and LOFT L2-3 large-break loss-of-coolant experiment. From the single channel tests the integration scheme and its implementation were proven to be valid. Other simulation results showed good agreement with the experiments. The computational speed was also satisfactory. So it is confirmed that COBRA/RELAP5 can be a promising tool for analysis of complicated, multidimensional two-phase flow transients. The area of further improvements in the code integration are also identified. This report also serves as a user`s manual for the COBRA/RELAP5 code. (author). 6 tabs., 20 figs., 20 refs.

  1. Managing Transportation Infrastructure for Sustainable Development

    NARCIS (Netherlands)

    Akinyemi, Edward O.; Zuidgeest, M.H.P.

    Major requirements for operationalization of the concept of sustainable development in urban transportation infrastructure operations management are presented. In addition, it is shown that the current approach to management is incompatible with the requirements for sustainable urban development.

  2. On transform coding tools under development for VP10

    Science.gov (United States)

    Parker, Sarah; Chen, Yue; Han, Jingning; Liu, Zoe; Mukherjee, Debargha; Su, Hui; Wang, Yongzhe; Bankoski, Jim; Li, Shunyao

    2016-09-01

    Google started the WebM Project in 2010 to develop open source, royaltyfree video codecs designed specifically for media on the Web. The second generation codec released by the WebM project, VP9, is currently served by YouTube, and enjoys billions of views per day. Realizing the need for even greater compression efficiency to cope with the growing demand for video on the web, the WebM team embarked on an ambitious project to develop a next edition codec, VP10, that achieves at least a generational improvement in coding efficiency over VP9. Starting from VP9, a set of new experimental coding tools have already been added to VP10 to achieve decent coding gains. Subsequently, Google joined a consortium of major tech companies called the Alliance for Open Media to jointly develop a new codec AV1. As a result, the VP10 effort is largely expected to merge with AV1. In this paper, we focus primarily on new tools in VP10 that improve coding of the prediction residue using transform coding techniques. Specifically, we describe tools that increase the flexibility of available transforms, allowing the codec to handle a more diverse range or residue structures. Results are presented on a standard test set.

  3. The VENUS-7 benchmarks. Results from state-of-the-art transport codes and nuclear data

    International Nuclear Information System (INIS)

    Zwermann, Winfried; Pautz, Andreas; Timm, Wolf

    2010-01-01

    For the validation of both nuclear data and computational methods, comparisons with experimental data are necessary. Most advantageous are assemblies where not only the multiplication factors or critical parameters were measured, but also additional quantities like reactivity differences or pin-wise fission rate distributions have been assessed. Currently there is a comprehensive activity to evaluate such measure-ments and incorporate them in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. A large number of such experiments was performed at the VENUS zero power reactor at SCK/CEN in Belgium in the sixties and seventies. The VENUS-7 series was specified as an international benchmark within the OECD/NEA Working Party on Scientific Issues of Reactor Systems (WPRS), and results obtained with various codes and nuclear data evaluations were summarized. In the present paper, results of high-accuracy transport codes with full spatial resolution with up-to-date nuclear data libraries from the JEFF and ENDF/B evaluations are presented. The comparisons of the results, both code-to-code and with the measured data are augmented by uncertainty and sensitivity analyses with respect to nuclear data uncertainties. For the multiplication factors, these are performed with the TSUNAMI-3D code from the SCALE system. In addition, uncertainties in the reactivity differences are analyzed with the TSAR code which is available from the current SCALE-6 version. (orig.)

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

    International Nuclear Information System (INIS)

    Shangguan Danhua; Li Gang; Deng Li; Zhang Baoyin

    2013-01-01

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

  5. Energy Conservation Tests of a Coupled Kinetic-kinetic Plasma-neutral Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Stotler, D. P.; Chang, C. S.; Ku, S. H.; Lang, J.; Park, G.

    2012-08-29

    A Monte Carlo neutral transport routine, based on DEGAS2, has been coupled to the guiding center ion-electron-neutral neoclassical PIC code XGC0 to provide a realistic treatment of neutral atoms and molecules in the tokamak edge plasma. The DEGAS2 routine allows detailed atomic physics and plasma-material interaction processes to be incorporated into these simulations. The spatial pro le of the neutral particle source used in the DEGAS2 routine is determined from the uxes of XGC0 ions to the material surfaces. The kinetic-kinetic plasma-neutral transport capability is demonstrated with example pedestal fueling simulations.

  6. 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

  7. Application of the ASME code in designing containment vessels for packages used to transport radioactive materials

    International Nuclear Information System (INIS)

    Raske, D.T.; Wang, Z.

    1992-01-01

    The primary concern governing the design of shipping packages containing radioactive materials is public safety during transport. When these shipments are within the regulatory jurisdiction of the US Department of Energy, the recommended design criterion for the primary containment vessel is either Section III or Section VIII, Division 1, of the ASME Boiler and Pressure Vessel Code, depending on the activity of the contents. The objective of this paper is to discuss the design of a prototypic containment vessel representative of a packaging for the transport of high-level radioactive material

  8. ICARE/CATHARE and ASTEC code development trends

    International Nuclear Information System (INIS)

    Chatelard, P.; Dorsselaere, J.-P. van

    2000-01-01

    Regarding the computer code development for simulation of LWR severe accidents, IPSN developed a two-tier approach based on detailed codes such as ICARE/CATHARE and simplified models to be assembled in the ASTEC integral code. The ICARE/CATHARE code results from the coupling between the ICARE2 code modelling the core degradation phenomena and the thermalhydraulics code CATHARE2. It allows to calculate PWR and VVER severe accident sequences in the whole RCS. The modelling of the early degradation phase can be considered as rather complete in the ICARE/CATHARE V1 mod1 version (to be released by mid-2000) whereas some models are still missing for the late phase. The main future developments (ICARE/CATHARE V2) will concern the multi-dimensional thermalhydraulics, the quenching of partially damaged cores (mechanical and chemical effects), the debris bed two-phase thermalhydraulics (including reflooding) and the corium behaviour in the lower head. The main other physical improvements should concern the behaviour of boron carbide control rods, the processes governing the core loss of geometry (transition phase) and the oxidation of relocated melts. The ASTEC (Accident Source Term Evaluation Code) integral code, commonly developed by IPSN and GRS, aims to predict an entire LWR (PWR, VVER and BWR) severe accident sequence from the initiating event through to FP release out of the containment, for source term, PSA level 2, or accident management studies. The version ASTEC VO.3 to be released by mid-2000 can be considered now as robust and fast-running enough (between 2 and 12 hours for a one day accident) and allows to perform, with a containment multi-compartment configuration, any scenario accident study accounting for the main safety systems and operator procedures (spray, recombiner, etc.). The next version ASTEC V1, to be released beginning of 2002, will include the frontend simulation and improve modelling of in-vessel core degradation. A large validation activity will

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

    International Nuclear Information System (INIS)

    Calloo, A.A.

    2012-01-01

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

  10. Calibration of neutron yield activation measurements at JET using MCNP and furnace neutron transport codes

    International Nuclear Information System (INIS)

    Pillon, M.; Martone, M.; Verschuur, K.A.; Jarvis, O.N.; Kaellne, J.

    1989-01-01

    Neutron transport calculations have been performed using fluence ray tracing (FURNACE code) and Monte Carlo particle trajectory sampling methods (MCNP code) in order to determine the neutron fluence and energy distributions at different locations in the JET tokamak. These calculations were used to calibrate the activation measurements used in the determination of the absolute fusion neutron yields from the JET plasma. We present here the neutron activation response coefficients calculated for three different materials. Comparison of the MCNP and FURNACE results helps identify the sources of error in these neutron transport calculations. The accuracy of these calculations was tested by comparing the total 2.5 MeV neutron yields derived from the activation measurements with those obtained with calibrated fission chambers; agreement at the ±15% level was demonstrate. (orig.)

  11. Current status of the reactor physics code WIMS and recent developments

    International Nuclear Information System (INIS)

    Lindley, B.A.; Hosking, J.G.; Smith, P.J.; Powney, D.J.; Tollit, B.S.; Newton, T.D.; Perry, R.; Ware, T.C.; Smith, P.N.

    2017-01-01

    Highlights: • The current status of the WIMS reactor physics code is presented. • Applications range from 2D lattice calculations up to 3D whole core geometries. • Gamma transport and thermal-hydraulic feedback models added. • Calculations methodologies described for several Gen II, III and IV reactor types. - Abstract: The WIMS modular reactor physics code has been under continuous development for over fifty years. This paper discusses the current status of WIMS and recent developments, in particular developments to the resonance shielding methodology and 3D transport solvers. Traditionally, WIMS is used to perform 2D lattice calculations, typically to generate homogenized reactor physics parameters for a whole core code such as PANTHER. However, with increasing computational resources there has been a growing trend for performing transport calculations on larger problems, up to and including 3D full core models. To this end, a number of the WIMS modules have been parallelised to allow efficient performance for whole core calculations, and WIMS includes a 3D method of characteristics solver with reflective and once-through tracking methods, which can be used to analyse problems of varying size and complexity. A time-dependent flux solver has been incorporated and thermal-hydraulic modelling capability is also being added to allow steady-state and transient coupled calculations to be performed. WIMS has been validated against a range of experimental data and other codes, in particular for water and graphite moderated thermal reactors. Future developments will include improved parallelization, enhancing the thermal-hydraulic feedback models and validating the WIMS/PANTHER code system for BWRs and fast reactors.

  12. Recent developments of JAEA's Monte Carlo Code MVP for reactor physics applications

    International Nuclear Information System (INIS)

    Nagaya, Y.; Okumura, K.; Mori, T.

    2013-01-01

    MVP is a general-purpose continuous-energy Monte Carlo code for neutron and photon transport calculations that has been developed since the late 1980's at Japan Atomic Energy Agency (JAEA, formerly JAERI). The MVP code is designed for nuclear reactor applications such as reactor core design/analysis, criticality safety and reactor shielding. This paper describes the MVP code and present its latest developments. Among the new capabilities of MVP we find: -) the perturbation method has been implemented for the change in k(eff); -) the eigenvalue calculations can be performed with an explicit treatment of delayed neutrons in which their fission spectra are taken into account; -) the capability of tallying the scattering matrix (group-to-group scattering cross sections); -) the implementation of an exact model for resonance elastic scattering; and -) a Monte Carlo perturbation technique is used to calculate reactor kinetics parameters

  13. 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

  14. Recent developments in seismic analysis in the code Aster

    International Nuclear Information System (INIS)

    Guihot, P.; Devesa, G.; Dumond, A.; Panet, M.; Waeckel, F.

    1996-01-01

    Progress in the field of seismic qualification and design methods made these last few years allows physical phenomena actually in play to be better considered, while cutting down the conservatism associated with some simplified design methods. So following the change in methods and developing the most advantageous ones among them contributes to the process of the seismic margins assessment and the preparation of new design tools for future series. In this paper, the main developments and improvements in methods which have been made these last two years in the Code Aster, in order to improve seismic calculation methods and seismic margin assessment are presented. The first development relates to making the MISS3D soil structure interaction code available, thanks to an interface made with the Code Aster. The second relates to the possibility of making modal basis time calculations on multi-supported structures by considering local non linearities like impact, friction or squeeze fluid forces. Recent developments in random dynamics and postprocessing devoted to earthquake designs are then mentioned. Three applications of these developments are then ut forward. The first application relates to a test case for soil structure interaction design using MISS3D-Aster coupling. The second is a test case for a multi-supported structure. The last application, more for manufacturing, refers to seismic qualification of Main Live Steam stop valves. First results of the independent validation of the Code Aster seismic design functionalities, which provide and improve the quality of software, are also recalled. (authors)

  15. 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

  16. 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.

  17. The three-dimensional, discrete ordinates neutral particle transport code TORT: An overview

    International Nuclear Information System (INIS)

    Azmy, Y.Y.

    1996-01-01

    The centerpiece of the Discrete Ordinates Oak Ridge System (DOORS), the three-dimensional neutral particle transport code TORT is reviewed. Its most prominent features pertaining to large applications, such as adjustable problem parameters, memory management, and coarse mesh methods, are described. Advanced, state-of-the-art capabilities including acceleration and multiprocessing are summarized here. Future enhancement of existing graphics and visualization tools is briefly presented

  18. Methods for the development of large computer codes under LTSS

    International Nuclear Information System (INIS)

    Sicilian, J.M.

    1977-06-01

    TRAC is a large computer code being developed by Group Q-6 for the analysis of the transient thermal hydraulic behavior of light-water nuclear reactors. A system designed to assist the development of TRAC is described. The system consists of a central HYDRA dataset, R6LIB, containing files used in the development of TRAC, and a file maintenance program, HORSE, which facilitates the use of this dataset

  19. 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)

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  1. 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

  2. Fuel cell development for transportation: Catalyst development

    Energy Technology Data Exchange (ETDEWEB)

    Doddapaneni, N. [Sandia National Lab., Albuquerque, NM (United States)

    1996-04-01

    Fuel cells are being considered as alternate power sources for transportation and stationary applications. With proton exchange membrane (PEM) fuel cells the fuel crossover to cathodes causes severe thermal management and cell voltage drop due to oxidation of fuel at the platinized cathodes. The main goal of this project was to design, synthesize, and evaluate stable and inexpensive transition metal macrocyclic catalysts for the reduction of oxygen and be electrochemically inert towards anode fuels such as hydrogen and methanol.

  3. Development Of A Navier-Stokes Computer Code

    Science.gov (United States)

    Yoon, Seokkwan; Kwak, Dochan

    1993-01-01

    Report discusses aspects of development of CENS3D computer code, solving three-dimensional Navier-Stokes equations of compressible, viscous, unsteady flow. Implements implicit finite-difference or finite-volume numerical-integration scheme, called "lower-upper symmetric-Gauss-Seidel" (LU-SGS), offering potential for very low computer time per iteration and for fast convergence.

  4. 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

  5. Application of software engineering to development of reactor safety codes

    International Nuclear Information System (INIS)

    Wilburn, N.P.; Niccoli, L.G.

    1981-01-01

    Software Engineering, which is a systematic methodology by which a large scale software development project is partitioned into manageable pieces, has been applied to the development of LMFBR safety codes. The techniques have been applied extensively in the business and aerospace communities and have provided an answer to the drastically increasing cost of developing and maintaining software. The five phases of software engineering (Survey, Analysis, Design, Implementation, and Testing) were applied in turn to development of these codes, along with Walkthroughs (peer review) at each stage. The application of these techniques has resulted in SUPERIOR SOFTWARE which is well documented, thoroughly tested, easy to modify, easier to use and maintain. The development projects have resulted in lower overall cost. (orig.) [de

  6. Development of burnup methods and capabilities in Monte Carlo code RMC

    International Nuclear Information System (INIS)

    She, Ding; Liu, Yuxuan; Wang, Kan; Yu, Ganglin; Forget, Benoit; Romano, Paul K.; Smith, Kord

    2013-01-01

    Highlights: ► The RMC code has been developed aiming at large-scale burnup calculations. ► Matrix exponential methods are employed to solve the depletion equations. ► The Energy-Bin method reduces the time expense of treating ACE libraries. ► The Cell-Mapping method is efficient to handle massive amounts of tally cells. ► Parallelized depletion is necessary for massive amounts of burnup regions. -- Abstract: The Monte Carlo burnup calculation has always been a challenging problem because of its large time consumption when applied to full-scale assembly or core calculations, and thus its application in routine analysis is limited. Most existing MC burnup codes are usually external wrappers between a MC code, e.g. MCNP, and a depletion code, e.g. ORIGEN. The code RMC is a newly developed MC code with an embedded depletion module aimed at performing burnup calculations of large-scale problems with high efficiency. Several measures have been taken to strengthen the burnup capabilities of RMC. Firstly, an accurate and efficient depletion module called DEPTH has been developed and built in, which employs the rational approximation and polynomial approximation methods. Secondly, the Energy-Bin method and the Cell-Mapping method are implemented to speed up the transport calculations with large numbers of nuclides and tally cells. Thirdly, the batch tally method and the parallelized depletion module have been utilized to better handle cases with massive amounts of burnup regions in parallel calculations. Burnup cases including a PWR pin and a 5 × 5 assembly group are calculated, thereby demonstrating the burnup capabilities of the RMC code. In addition, the computational time and memory requirements of RMC are compared with other MC burnup codes.

  7. Development of a code of practice for deep geothermal wells

    International Nuclear Information System (INIS)

    Leaver, J.D.; Bolton, R.S.; Dench, N.D.; Fooks, L.

    1990-01-01

    Recent and on-going changes to the structure of the New Zealand geothermal industry has shifted responsibility for the development of geothermal resources from central government to private enterprise. The need for a code of practice for deep geothermal wells was identified by the Geothermal Inspectorate of the Ministry of Commerce to maintain adequate standards of health and safety and to assist with industry deregulation. This paper reports that the Code contains details of methods, procedures, formulae and design data necessary to attain those standards, and includes information which drilling engineers having experience only in the oil industry could not be expected to be familiar with

  8. Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC): gap analysis for high fidelity and performance assessment code development

    International Nuclear Information System (INIS)

    Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

    2011-01-01

    This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are

  9. Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC) : gap analysis for high fidelity and performance assessment code development.

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joon H.; Siegel, Malcolm Dean; Arguello, Jose Guadalupe, Jr.; Webb, Stephen Walter; Dewers, Thomas A.; Mariner, Paul E.; Edwards, Harold Carter; Fuller, Timothy J.; Freeze, Geoffrey A.; Jove-Colon, Carlos F.; Wang, Yifeng

    2011-03-01

    This report describes a gap analysis performed in the process of developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with rigorous verification, validation, and software quality requirements. The gap analyses documented in this report were are performed during an initial gap analysis to identify candidate codes and tools to support the development and integration of the Waste IPSC, and during follow-on activities that delved into more detailed assessments of the various codes that were acquired, studied, and tested. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. The gap analysis indicates that significant capabilities may already exist in the existing THC codes although there is no single code able to fully account for all physical and chemical processes involved in a waste disposal system. Large gaps exist in modeling chemical processes and their couplings with other processes. The coupling of chemical processes with flow transport and mechanical deformation remains challenging. The data for extreme environments (e.g., for elevated temperature and high ionic strength media) that are

  10. Development of a parallelization strategy for the VARIANT code

    International Nuclear Information System (INIS)

    Hanebutte, U.R.; Khalil, H.S.; Palmiotti, G.; Tatsumi, M.

    1996-01-01

    The VARIANT code solves the multigroup steady-state neutron diffusion and transport equation in three-dimensional Cartesian and hexagonal geometries using the variational nodal method. VARIANT consists of four major parts that must be executed sequentially: input handling, calculation of response matrices, solution algorithm (i.e. inner-outer iteration), and output of results. The objective of the parallelization effort was to reduce the overall computing time by distributing the work of the two computationally intensive (sequential) tasks, the coupling coefficient calculation and the iterative solver, equally among a group of processors. This report describes the code's calculations and gives performance results on one of the benchmark problems used to test the code. The performance analysis in the IBM SPx system shows good efficiency for well-load-balanced programs. Even for relatively small problem sizes, respectable efficiencies are seen for the SPx. An extension to achieve a higher degree of parallelism will be addressed in future work. 7 refs., 1 tab

  11. Theoretical atomic physics code development I: CATS: Cowan Atomic Structure Code

    International Nuclear Information System (INIS)

    Abdallah, J. Jr.; Clark, R.E.H.; Cowan, R.D.

    1988-12-01

    An adaptation of R.D. Cowan's Atomic Structure program, CATS, has been developed as part of the Theoretical Atomic Physics (TAPS) code development effort at Los Alamos. CATS has been designed to be easy to run and to produce data files that can interface with other programs easily. The CATS produced data files currently include wave functions, energy levels, oscillator strengths, plane-wave-Born electron-ion collision strengths, photoionization cross sections, and a variety of other quantities. This paper describes the use of CATS. 10 refs

  12. THE ANALYSIS OF WAYS TO IMPROVE ECONOMIC DEVELOPMENT AND TRADE ORGANIZATION OF TRANSPORTATION ON AIR TRANSPORT

    Directory of Open Access Journals (Sweden)

    I. A. Ivanov

    2015-01-01

    Full Text Available In transport system of Russia air transport is one of main types of passenger and cargo transport. Demand for air transportation constantly increases that allows to consider reasonably improvement of transportations on air transport as the priority direction of development in civil aviation. The article considers issues of development of the branch air transport.

  13. Development of dynamic simulation code for fuel cycle fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, Isao; Seki, Yasushi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan); Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

    1999-02-01

    A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

  14. A development of containment performance analysis methodology using GOTHIC code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B. C.; Yoon, J. I. [Future and Challenge Company, Seoul (Korea, Republic of); Byun, C. S.; Lee, J. Y. [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Lee, J. Y. [Seoul National University, Seoul (Korea, Republic of)

    2003-10-01

    In a circumstance that well-established containment pressure/temperature analysis code, CONTEMPT-LT treats the reactor containment as a single volume, this study introduces, as an alternative, the GOTHIC code for an usage on multi-compartmental containment performance analysis. With a developed GOTHIC methodology, its applicability is verified for containment performance analysis for Korean Nuclear Unit 1. The GOTHIC model for this plant is simply composed of 3 compartments including the reactor containment and RWST. In addition, the containment spray system and containment recirculation system are simulated. As a result of GOTHIC calculation, under the same assumptions and conditions as those in CONTEMPT-LT, the GOTHIC prediction shows a very good result; pressure and temperature transients including their peaks are nearly the same. It can be concluded that the GOTHIC could provide reasonable containment pressure and temperature responses if considering the inherent conservatism in CONTEMPT-LT code.

  15. Development and improvement of safety analysis code for geological disposal

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    In order to confirm the long-term safety concerning geological disposal, probabilistic safety assessment code and other analysis codes, which can evaluate possibility of each event and influence on engineered barrier and natural barrier by the event, were introduced. We confirmed basic functions of those codes and studied the relation between those functions and FEP/PID which should be taken into consideration in safety assessment. We are planning to develop 'Nuclide Migration Assessment System' for the purpose of realizing improvement in efficiency of assessment work, human error prevention for analysis, and quality assurance of the analysis environment and analysis work for safety assessment by using it. As the first step, we defined the system requirements and decided the system composition and functions which should be mounted in them based on those requirements. (author)

  16. A development of containment performance analysis methodology using GOTHIC code

    International Nuclear Information System (INIS)

    Lee, B. C.; Yoon, J. I.; Byun, C. S.; Lee, J. Y.; Lee, J. Y.

    2003-01-01

    In a circumstance that well-established containment pressure/temperature analysis code, CONTEMPT-LT treats the reactor containment as a single volume, this study introduces, as an alternative, the GOTHIC code for an usage on multi-compartmental containment performance analysis. With a developed GOTHIC methodology, its applicability is verified for containment performance analysis for Korean Nuclear Unit 1. The GOTHIC model for this plant is simply composed of 3 compartments including the reactor containment and RWST. In addition, the containment spray system and containment recirculation system are simulated. As a result of GOTHIC calculation, under the same assumptions and conditions as those in CONTEMPT-LT, the GOTHIC prediction shows a very good result; pressure and temperature transients including their peaks are nearly the same. It can be concluded that the GOTHIC could provide reasonable containment pressure and temperature responses if considering the inherent conservatism in CONTEMPT-LT code

  17. Development of simplified decommissioning cost estimation code for nuclear facilities

    International Nuclear Information System (INIS)

    Tachibana, Mitsuo; Shiraishi, Kunio; Ishigami, Tsutomu

    2010-01-01

    The simplified decommissioning cost estimation code for nuclear facilities (DECOST code) was developed in consideration of features and structures of nuclear facilities and similarity of dismantling methods. The DECOST code could calculate 8 evaluation items of decommissioning cost. Actual dismantling in the Japan Atomic Energy Agency (JAEA) was evaluated; unit conversion factors used to calculate the manpower of dismantling activities were evaluated. Consequently, unit conversion factors of general components could be classified into three kinds. Weights of components and structures of the facility were necessary for calculation of manpower. Methods for evaluating weights of components and structures of the facility were studied. Consequently, the weight of components in the facility was proportional to the weight of structures of the facility. The weight of structures of the facility was proportional to the total area of floors in the facility. Decommissioning costs of 7 nuclear facilities in the JAEA were calculated by using the DECOST code. To verify the calculated results, the calculated manpower was compared with the manpower gained from actual dismantling. Consequently, the calculated manpower and actual manpower were almost equal. The outline of the DECOST code, evaluation results of unit conversion factors, the evaluation method of the weights of components and structures of the facility are described in this report. (author)

  18. Health effects estimation code development for accident consequence analysis

    International Nuclear Information System (INIS)

    Togawa, O.; Homma, T.

    1992-01-01

    As part of a computer code system for nuclear reactor accident consequence analysis, two computer codes have been developed for estimating health effects expected to occur following an accident. Health effects models used in the codes are based on the models of NUREG/CR-4214 and are revised for the Japanese population on the basis of the data from the reassessment of the radiation dosimetry and information derived from epidemiological studies on atomic bomb survivors of Hiroshima and Nagasaki. The health effects models include early and continuing effects, late somatic effects and genetic effects. The values of some model parameters are revised for early mortality. The models are modified for predicting late somatic effects such as leukemia and various kinds of cancers. The models for genetic effects are the same as those of NUREG. In order to test the performance of one of these codes, it is applied to the U.S. and Japanese populations. This paper provides descriptions of health effects models used in the two codes and gives comparisons of the mortality risks from each type of cancer for the two populations. (author)

  19. Development of parallel Fokker-Planck code ALLAp

    International Nuclear Information System (INIS)

    Batishcheva, A.A.; Sigmar, D.J.; Koniges, A.E.

    1996-01-01

    We report on our ongoing development of the 3D Fokker-Planck code ALLA for a highly collisional scrape-off-layer (SOL) plasma. A SOL with strong gradients of density and temperature in the spatial dimension is modeled. Our method is based on a 3-D adaptive grid (in space, magnitude of the velocity, and cosine of the pitch angle) and a second order conservative scheme. Note that the grid size is typically 100 x 257 x 65 nodes. It was shown in our previous work that only these capabilities make it possible to benchmark a 3D code against a spatially-dependent self-similar solution of a kinetic equation with the Landau collision term. In the present work we show results of a more precise benchmarking against the exact solutions of the kinetic equation using a new parallel code ALLAp with an improved method of parallelization and a modified boundary condition at the plasma edge. We also report first results from the code parallelization using Message Passing Interface for a Massively Parallel CRI T3D platform. We evaluate the ALLAp code performance versus the number of T3D processors used and compare its efficiency against a Work/Data Sharing parallelization scheme and a workstation version

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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