WorldWideScience

Sample records for code input physics

  1. ETFOD: a point model physics code with arbitrary input

    International Nuclear Information System (INIS)

    Rothe, K.E.; Attenberger, S.E.

    1980-06-01

    ETFOD is a zero-dimensional code which solves a set of physics equations by minimization. The technique used is different than normally used, in that the input is arbitrary. The user is supplied with a set of variables from which he specifies which variables are input (unchanging). The remaining variables become the output. Presently the code is being used for ETF reactor design studies. The code was written in a manner to allow easy modificaton of equations, variables, and physics calculations. The solution technique is presented along with hints for using the code

  2. GASFLOW computer code (physical models and input data)

    International Nuclear Information System (INIS)

    Muehlbauer, Petr

    2007-11-01

    The GASFLOW computer code was developed jointly by the Los Alamos National Laboratory, USA, and Forschungszentrum Karlsruhe, Germany. The code is primarily intended for calculations of the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containments and in other facilities. The physical models and the input data are described, and a commented simple calculation is presented

  3. MARS code manual volume II: input requirements

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Kim, Kyung Doo; Bae, Sung Won; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu

    2010-02-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This input manual provides a complete list of input required to run MARS. The manual is divided largely into two parts, namely, the one-dimensional part and the multi-dimensional part. The inputs for auxiliary parts such as minor edit requests and graph formatting inputs are shared by the two parts and as such mixed input is possible. The overall structure of the input is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

  4. Pre-processing of input files for the AZTRAN code

    International Nuclear Information System (INIS)

    Vargas E, S.; Ibarra, G.

    2017-09-01

    The AZTRAN code began to be developed in the Nuclear Engineering Department of the Escuela Superior de Fisica y Matematicas (ESFM) of the Instituto Politecnico Nacional (IPN) with the purpose of numerically solving various models arising from the physics and engineering of nuclear reactors. The code is still under development and is part of the AZTLAN platform: Development of a Mexican platform for the analysis and design of nuclear reactors. Due to the complexity to generate an input file for the code, a script based on D language is developed, with the purpose of making its elaboration easier, based on a new input file format which includes specific cards, which have been divided into two blocks, mandatory cards and optional cards, including a pre-processing of the input file to identify possible errors within it, as well as an image generator for the specific problem based on the python interpreter. (Author)

  5. Input data required for specific performance assessment codes

    International Nuclear Information System (INIS)

    Seitz, R.R.; Garcia, R.S.; Starmer, R.J.; Dicke, C.A.; Leonard, P.R.; Maheras, S.J.; Rood, A.S.; Smith, R.W.

    1992-02-01

    The Department of Energy's National Low-Level Waste Management Program at the Idaho National Engineering Laboratory generated this report on input data requirements for computer codes to assist States and compacts in their performance assessments. This report gives generators, developers, operators, and users some guidelines on what input data is required to satisfy 22 common performance assessment codes. Each of the codes is summarized and a matrix table is provided to allow comparison of the various input required by the codes. This report does not determine or recommend which codes are preferable

  6. Physical layer network coding

    DEFF Research Database (Denmark)

    Fukui, Hironori; Popovski, Petar; Yomo, Hiroyuki

    2014-01-01

    Physical layer network coding (PLNC) has been proposed to improve throughput of the two-way relay channel, where two nodes communicate with each other, being assisted by a relay node. Most of the works related to PLNC are focused on a simple three-node model and they do not take into account...

  7. Physics of codes

    International Nuclear Information System (INIS)

    Cooper, R.K.; Jones, M.E.

    1989-01-01

    The title given this paper is a bit presumptuous, since one can hardly expect to cover the physics incorporated into all the codes already written and currently being written. The authors focus on those codes which have been found to be particularly useful in the analysis and design of linacs. At that the authors will be a bit parochial and discuss primarily those codes used for the design of radio-frequency (rf) linacs, although the discussions of TRANSPORT and MARYLIE have little to do with the time structures of the beams being analyzed. The plan of this paper is first to describe rather simply the concepts of emittance and brightness, then to describe rather briefly each of the codes TRANSPORT, PARMTEQ, TBCI, MARYLIE, and ISIS, indicating what physics is and is not included in each of them. It is expected that the vast majority of what is covered will apply equally well to protons and electrons (and other particles). This material is intended to be tutorial in nature and can in no way be expected to be exhaustive. 31 references, 4 figures

  8. MIFT: GIFT Combinatorial Geometry Input to VCS Code

    Science.gov (United States)

    1977-03-01

    r-w w-^ H ^ß0318is CQ BRL °RCUMr REPORT NO. 1967 —-S: ... MIFT: GIFT COMBINATORIAL GEOMETRY INPUT TO VCS CODE Albert E...TITLE (and Subtitle) MIFT: GIFT Combinatorial Geometry Input to VCS Code S. TYPE OF REPORT & PERIOD COVERED FINAL 6. PERFORMING ORG. REPORT NUMBER...Vehicle Code System (VCS) called MORSE was modified to accept the GIFT combinatorial geometry package. GIFT , as opposed to the geometry package

  9. Influential input parameters for reflood model of MARS code

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Deog Yeon; Bang, Young Seok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-10-15

    Best Estimate (BE) calculation has been more broadly used in nuclear industries and regulations to reduce the significant conservatism for evaluating Loss of Coolant Accident (LOCA). Reflood model has been identified as one of the problems in BE calculation. The objective of the Post BEMUSE Reflood Model Input Uncertainty Methods (PREMIUM) program of OECD/NEA is to make progress the issue of the quantification of the uncertainty of the physical models in system thermal hydraulic codes, by considering an experimental result especially for reflood. It is important to establish a methodology to identify and select the parameters influential to the response of reflood phenomena following Large Break LOCA. For this aspect, a reference calculation and sensitivity analysis to select the dominant influential parameters for FEBA experiment are performed.

  10. Evaluating nuclear physics inputs in core-collapse supernova models

    Science.gov (United States)

    Lentz, E.; Hix, W. R.; Baird, M. L.; Messer, O. E. B.; Mezzacappa, A.

    Core-collapse supernova models depend on the details of the nuclear and weak interaction physics inputs just as they depend on the details of the macroscopic physics (transport, hydrodynamics, etc.), numerical methods, and progenitors. We present preliminary results from our ongoing comparison studies of nuclear and weak interaction physics inputs to core collapse supernova models using the spherically-symmetric, general relativistic, neutrino radiation hydrodynamics code Agile-Boltztran. We focus on comparisons of the effects of the nuclear EoS and the effects of improving the opacities, particularly neutrino--nucleon interactions.

  11. Physical Layer Network Coding

    DEFF Research Database (Denmark)

    Fukui, Hironori; Yomo, Hironori; Popovski, Petar

    2013-01-01

    of interfering nodes and usage of spatial reservation mechanisms. Specifically, we introduce a reserved area in order to protect the nodes involved in two-way relaying from the interference caused by neighboring nodes. We analytically derive the end-to-end rate achieved by PLNC considering the impact......Physical layer network coding (PLNC) has the potential to improve throughput of multi-hop networks. However, most of the works are focused on the simple, three-node model with two-way relaying, not taking into account the fact that there can be other neighboring nodes that can cause....../receive interference. The way to deal with this problem in distributed wireless networks is usage of MAC-layer mechanisms that make a spatial reservation of the shared wireless medium, similar to the well-known RTS/CTS in IEEE 802.11 wireless networks. In this paper, we investigate two-way relaying in presence...

  12. Statistical screening of input variables in a complex computer code

    International Nuclear Information System (INIS)

    Krieger, T.J.

    1982-01-01

    A method is presented for ''statistical screening'' of input variables in a complex computer code. The object is to determine the ''effective'' or important input variables by estimating the relative magnitudes of their associated sensitivity coefficients. This is accomplished by performing a numerical experiment consisting of a relatively small number of computer runs with the code followed by a statistical analysis of the results. A formula for estimating the sensitivity coefficients is derived. Reference is made to an earlier work in which the method was applied to a complex reactor code with good results

  13. Mars 2.2 code manual: input requirements

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Lee, Won Jae; Jeong, Jae Jun; Lee, Young Jin; Hwang, Moon Kyu; Kim, Kyung Doo; Lee, Seung Wook; Bae, Sung Won

    2003-07-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This input manual provides a complete list of input required to run MARS. The manual is divided largely into two parts, namely, the one-dimensional part and the multi-dimensional part. The inputs for auxiliary parts such as minor edit requests and graph formatting inputs are shared by the two parts and as such mixed input is possible. The overall structure of the input is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS. MARS development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

  14. WWER reactor physics code applications

    International Nuclear Information System (INIS)

    Gado, J.; Kereszturi, A.; Gacs, A.; Telbisz, M.

    1994-01-01

    The coupled steady-state reactor physics and thermohydraulic code system KARATE has been developed and applied for WWER-1000 and WWER-440 operational calculations. The 3 D coupled kinetic code KIKO3D has been developed and validated for WWER-440 accident analysis applications. The coupled kinetic code SMARTA developed by VTT Helsinki has been applied for WWER-440 accident analysis. The paper gives a summary of the experience in code development and application. (authors). 10 refs., 2 tabs., 5 figs

  15. Computer code ANISN multiplying media and shielding calculation 2. Code description (input/output)

    International Nuclear Information System (INIS)

    Maiorino, J.R.

    1991-01-01

    The new code CCC-0514-ANISN/PC is described, as well as a ''GENERAL DESCRIPTION OF ANISN/PC code''. In addition to the ANISN/PC code, the transmittal package includes an interactive input generation programme called APE (ANISN Processor and Evaluator), which facilitates the work of the user in giving input. Also, a 21 group photon cross section master library FLUNGP.LIB in ISOTX format, which can be edited by an executable file LMOD.EXE, is included in the package. The input and output subroutines are reviewed. 6 refs, 1 fig., 1 tab

  16. ACCELERATION PHYSICS CODE WEB REPOSITORY.

    Energy Technology Data Exchange (ETDEWEB)

    WEI, J.

    2006-06-26

    In the framework of the CARE HHH European Network, we have developed a web-based dynamic accelerator-physics code repository. We describe the design, structure and contents of this repository, illustrate its usage, and discuss our future plans, with emphasis on code benchmarking.

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

    International Nuclear Information System (INIS)

    Maiorino, J.R.

    1990-01-01

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

  18. 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)

  19. Automation of RELAP5 input calibration and code validation using genetic algorithm

    International Nuclear Information System (INIS)

    Phung, Viet-Anh; Kööp, Kaspar; Grishchenko, Dmitry; Vorobyev, Yury; Kudinov, Pavel

    2016-01-01

    Highlights: • Automated input calibration and code validation using genetic algorithm is presented. • Predictions generally overlap experiments for individual system response quantities (SRQs). • It was not possible to predict simultaneously experimental maximum flow rate and oscillation period. • Simultaneous consideration of multiple SRQs is important for code validation. - Abstract: Validation of system thermal-hydraulic codes is an important step in application of the codes to reactor safety analysis. The goal of the validation process is to determine how well a code can represent physical reality. This is achieved by comparing predicted and experimental system response quantities (SRQs) taking into account experimental and modelling uncertainties. Parameters which are required for the code input but not measured directly in the experiment can become an important source of uncertainty in the code validation process. Quantification of such parameters is often called input calibration. Calibration and uncertainty quantification may become challenging tasks when the number of calibrated input parameters and SRQs is large and dependencies between them are complex. If only engineering judgment is employed in the process, the outcome can be prone to so called “user effects”. The goal of this work is to develop an automated approach to input calibration and RELAP5 code validation against data on two-phase natural circulation flow instability. Multiple SRQs are used in both calibration and validation. In the input calibration, we used genetic algorithm (GA), a heuristic global optimization method, in order to minimize the discrepancy between experimental and simulation data by identifying optimal combinations of uncertain input parameters in the calibration process. We demonstrate the importance of the proper selection of SRQs and respective normalization and weighting factors in the fitness function. In the code validation, we used maximum flow rate as the

  20. Automation of RELAP5 input calibration and code validation using genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Phung, Viet-Anh, E-mail: vaphung@kth.se [Division of Nuclear Power Safety, Royal Institute of Technology, Roslagstullsbacken 21, 10691 Stockholm (Sweden); Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se [Division of Nuclear Power Safety, Royal Institute of Technology, Roslagstullsbacken 21, 10691 Stockholm (Sweden); Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se [Division of Nuclear Power Safety, Royal Institute of Technology, Roslagstullsbacken 21, 10691 Stockholm (Sweden); Vorobyev, Yury, E-mail: yura3510@gmail.com [National Research Center “Kurchatov Institute”, Kurchatov square 1, Moscow 123182 (Russian Federation); Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se [Division of Nuclear Power Safety, Royal Institute of Technology, Roslagstullsbacken 21, 10691 Stockholm (Sweden)

    2016-04-15

    Highlights: • Automated input calibration and code validation using genetic algorithm is presented. • Predictions generally overlap experiments for individual system response quantities (SRQs). • It was not possible to predict simultaneously experimental maximum flow rate and oscillation period. • Simultaneous consideration of multiple SRQs is important for code validation. - Abstract: Validation of system thermal-hydraulic codes is an important step in application of the codes to reactor safety analysis. The goal of the validation process is to determine how well a code can represent physical reality. This is achieved by comparing predicted and experimental system response quantities (SRQs) taking into account experimental and modelling uncertainties. Parameters which are required for the code input but not measured directly in the experiment can become an important source of uncertainty in the code validation process. Quantification of such parameters is often called input calibration. Calibration and uncertainty quantification may become challenging tasks when the number of calibrated input parameters and SRQs is large and dependencies between them are complex. If only engineering judgment is employed in the process, the outcome can be prone to so called “user effects”. The goal of this work is to develop an automated approach to input calibration and RELAP5 code validation against data on two-phase natural circulation flow instability. Multiple SRQs are used in both calibration and validation. In the input calibration, we used genetic algorithm (GA), a heuristic global optimization method, in order to minimize the discrepancy between experimental and simulation data by identifying optimal combinations of uncertain input parameters in the calibration process. We demonstrate the importance of the proper selection of SRQs and respective normalization and weighting factors in the fitness function. In the code validation, we used maximum flow rate as the

  1. An approach for coupled-code multiphysics core simulations from a common input

    International Nuclear Information System (INIS)

    Schmidt, Rodney; Belcourt, Kenneth; Hooper, Russell; Pawlowski, Roger; Clarno, Kevin; Simunovic, Srdjan; Slattery, Stuart; Turner, John; Palmtag, Scott

    2015-01-01

    Highlights: • We describe an approach for coupled-code multiphysics reactor core simulations. • The approach can enable tight coupling of distinct physics codes with a common input. • Multi-code multiphysics coupling and parallel data transfer issues are explained. • The common input approach and how the information is processed is described. • Capabilities are demonstrated on an eigenvalue and power distribution calculation. - Abstract: This paper describes an approach for coupled-code multiphysics reactor core simulations that is being developed by the Virtual Environment for Reactor Applications (VERA) project in the Consortium for Advanced Simulation of Light-Water Reactors (CASL). In this approach a user creates a single problem description, called the “VERAIn” common input file, to define and setup the desired coupled-code reactor core simulation. A preprocessing step accepts the VERAIn file and generates a set of fully consistent input files for the different physics codes being coupled. The problem is then solved using a single-executable coupled-code simulation tool applicable to the problem, which is built using VERA infrastructure software tools and the set of physics codes required for the problem of interest. The approach is demonstrated by performing an eigenvalue and power distribution calculation of a typical three-dimensional 17 × 17 assembly with thermal–hydraulic and fuel temperature feedback. All neutronics aspects of the problem (cross-section calculation, neutron transport, power release) are solved using the Insilico code suite and are fully coupled to a thermal–hydraulic analysis calculated by the Cobra-TF (CTF) code. The single-executable coupled-code (Insilico-CTF) simulation tool is created using several VERA tools, including LIME (Lightweight Integrating Multiphysics Environment for coupling codes), DTK (Data Transfer Kit), Trilinos, and TriBITS. Parallel calculations are performed on the Titan supercomputer at Oak

  2. User input verification and test driven development in the NJOY21 nuclear data processing code

    Energy Technology Data Exchange (ETDEWEB)

    Trainer, Amelia Jo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Conlin, Jeremy Lloyd [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McCartney, Austin Paul [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-08-21

    Before physically-meaningful data can be used in nuclear simulation codes, the data must be interpreted and manipulated by a nuclear data processing code so as to extract the relevant quantities (e.g. cross sections and angular distributions). Perhaps the most popular and widely-trusted of these processing codes is NJOY, which has been developed and improved over the course of 10 major releases since its creation at Los Alamos National Laboratory in the mid-1970’s. The current phase of NJOY development is the creation of NJOY21, which will be a vast improvement from its predecessor, NJOY2016. Designed to be fast, intuitive, accessible, and capable of handling both established and modern formats of nuclear data, NJOY21 will address many issues that many NJOY users face, while remaining functional for those who prefer the existing format. Although early in its development, NJOY21 is quickly providing input validation to check user input. By providing rapid and helpful responses to users while writing input files, NJOY21 will prove to be more intuitive and easy to use than any of its predecessors. Furthermore, during its development, NJOY21 is subject to regular testing, such that its test coverage must strictly increase with the addition of any production code. This thorough testing will allow developers and NJOY users to establish confidence in NJOY21 as it gains functionality. This document serves as a discussion regarding the current state input checking and testing practices of NJOY21.

  3. Computer codes in particle transport physics

    International Nuclear Information System (INIS)

    Pesic, M.

    2004-01-01

    is given. Importance of validation and verification of data and computer codes is underlined briefly. Examples of applications of the MCNPX, FLUKA and SHIELD codes to simulation of some of processes in nature, from reactor physics, ion medical therapy, cross section calculations, design of accelerator driven sub-critical systems to astrophysics and shielding of spaceships, are shown. More reliable and more frequent cross sections data in intermediate and high- energy range for particles transport and interactions with mater are expected in near future, as a result of new experimental investigations that are under way with the aim to validate theoretical models applied currently in the codes. These new data libraries are expected to be much larger and more comprehensive than existing ones requiring more computer memory and faster CPUs. Updated versions of the codes to be developed in future, beside sequential computation versions, will also include the MPI or PVM options to allow faster ru: ming of the code at acceptable cost for an end-user. A new option to be implemented in the codes is expected too - an end-user written application for particular problem could be added relatively simple to the general source code script. Initial works on full implementation of graphic user interface for preparing input and analysing output of codes and ability to interrupt and/or continue code running should be upgraded to user-friendly level. (author)

  4. Flexible and re-configurable optical three-input XOR logic gate of phase-modulated signals with multicast functionality for potential application in optical physical-layer network coding.

    Science.gov (United States)

    Lu, Guo-Wei; Qin, Jun; Wang, Hongxiang; Ji, XuYuefeng; Sharif, Gazi Mohammad; Yamaguchi, Shigeru

    2016-02-08

    Optical logic gate, especially exclusive-or (XOR) gate, plays important role in accomplishing photonic computing and various network functionalities in future optical networks. On the other hand, optical multicast is another indispensable functionality to efficiently deliver information in optical networks. In this paper, for the first time, we propose and experimentally demonstrate a flexible optical three-input XOR gate scheme for multiple input phase-modulated signals with a 1-to-2 multicast functionality for each XOR operation using four-wave mixing (FWM) effect in single piece of highly-nonlinear fiber (HNLF). Through FWM in HNLF, all of the possible XOR operations among input signals could be simultaneously realized by sharing a single piece of HNLF. By selecting the obtained XOR components using a followed wavelength selective component, the number of XOR gates and the participant light in XOR operations could be flexibly configured. The re-configurability of the proposed XOR gate and the function integration of the optical logic gate and multicast in single device offer the flexibility in network design and improve the network efficiency. We experimentally demonstrate flexible 3-input XOR gate for four 10-Gbaud binary phase-shift keying signals with a multicast scale of 2. Error-free operations for the obtained XOR results are achieved. Potential application of the integrated XOR and multicast function in network coding is also discussed.

  5. HOTSPOT Health Physics codes for the PC

    Energy Technology Data Exchange (ETDEWEB)

    Homann, S.G.

    1994-03-01

    The HOTSPOT Health Physics codes were created to provide Health Physics personnel with a fast, field-portable calculation tool for evaluating accidents involving radioactive materials. HOTSPOT codes are a first-order approximation of the radiation effects associated with the atmospheric release of radioactive materials. HOTSPOT programs are reasonably accurate for a timely initial assessment. More importantly, HOTSPOT codes produce a consistent output for the same input assumptions and minimize the probability of errors associated with reading a graph incorrectly or scaling a universal nomogram during an emergency. The HOTSPOT codes are designed for short-term (less than 24 hours) release durations. Users requiring radiological release consequences for release scenarios over a longer time period, e.g., annual windrose data, are directed to such long-term models as CAPP88-PC (Parks, 1992). Users requiring more sophisticated modeling capabilities, e.g., complex terrain; multi-location real-time wind field data; etc., are directed to such capabilities as the Department of Energy`s ARAC computer codes (Sullivan, 1993). Four general programs -- Plume, Explosion, Fire, and Resuspension -- calculate a downwind assessment following the release of radioactive material resulting from a continuous or puff release, explosive release, fuel fire, or an area contamination event. Other programs deal with the release of plutonium, uranium, and tritium to expedite an initial assessment of accidents involving nuclear weapons. Additional programs estimate the dose commitment from the inhalation of any one of the radionuclides listed in the database of radionuclides; calibrate a radiation survey instrument for ground-survey measurements; and screen plutonium uptake in the lung (see FIDLER Calibration and LUNG Screening sections).

  6. HOTSPOT Health Physics codes for the PC

    International Nuclear Information System (INIS)

    Homann, S.G.

    1994-03-01

    The HOTSPOT Health Physics codes were created to provide Health Physics personnel with a fast, field-portable calculation tool for evaluating accidents involving radioactive materials. HOTSPOT codes are a first-order approximation of the radiation effects associated with the atmospheric release of radioactive materials. HOTSPOT programs are reasonably accurate for a timely initial assessment. More importantly, HOTSPOT codes produce a consistent output for the same input assumptions and minimize the probability of errors associated with reading a graph incorrectly or scaling a universal nomogram during an emergency. The HOTSPOT codes are designed for short-term (less than 24 hours) release durations. Users requiring radiological release consequences for release scenarios over a longer time period, e.g., annual windrose data, are directed to such long-term models as CAPP88-PC (Parks, 1992). Users requiring more sophisticated modeling capabilities, e.g., complex terrain; multi-location real-time wind field data; etc., are directed to such capabilities as the Department of Energy's ARAC computer codes (Sullivan, 1993). Four general programs -- Plume, Explosion, Fire, and Resuspension -- calculate a downwind assessment following the release of radioactive material resulting from a continuous or puff release, explosive release, fuel fire, or an area contamination event. Other programs deal with the release of plutonium, uranium, and tritium to expedite an initial assessment of accidents involving nuclear weapons. Additional programs estimate the dose commitment from the inhalation of any one of the radionuclides listed in the database of radionuclides; calibrate a radiation survey instrument for ground-survey measurements; and screen plutonium uptake in the lung (see FIDLER Calibration and LUNG Screening sections)

  7. Automation of Geometry Input for Building Code Compliance Check

    DEFF Research Database (Denmark)

    Petrova, Ekaterina Aleksandrova; Johansen, Peter Lind; Jensen, Rasmus Lund

    2017-01-01

    Documentation of compliance with the energy performance regulations at the end of the detailed design phase is mandatory for building owners in Denmark. Therefore, besides multidisciplinary input, the building design process requires various iterative analyses, so that the optimal solutions can...... be identified amongst multiple alternatives. However, meeting performance criteria is often associated with manual data inputs and retroactive modifications of the design. Due to poor interoperability between the authoring tools and the compliance check program, the processes are redundant and inefficient...... from building geometry created in Autodesk Revit and its translation to input for compliance check analysis....

  8. Utility subroutine package used by Applied Physics Division export codes

    International Nuclear Information System (INIS)

    Adams, C.H.; Derstine, K.L.; Henryson, H. II; Hosteny, R.P.; Toppel, B.J.

    1983-04-01

    This report describes the current state of the utility subroutine package used with codes being developed by the staff of the Applied Physics Division. The package provides a variety of useful functions for BCD input processing, dynamic core-storage allocation and managemnt, binary I/0 and data manipulation. The routines were written to conform to coding standards which facilitate the exchange of programs between different computers

  9. TASS/SMR Code Topical Report for SMART Plant, Vol II: User's Guide and Input Requirement

    International Nuclear Information System (INIS)

    Kim, See Darl; Kim, Soo Hyoung; Kim, Hyung Rae

    2008-10-01

    The TASS/SMR code has been developed with domestic technologies for the safety analysis of the SMART plant which is an integral type pressurized water reactor. It can be applied to the analysis of design basis accidents including non-LOCA (loss of coolant accident) and LOCA of the SMART plant. The TASS/SMR code can be applied to any plant regardless of the structural characteristics of a reactor since the code solves the same governing equations for both the primary and secondary system. The code has been developed to meet the requirements of the safety analysis code. This report describes the overall structure of the TASS/SMR, input processing, and the processes of a steady state and transient calculations. In addition, basic differential equations, finite difference equations, state relationships, and constitutive models are described in the report. First, the conservation equations, a discretization process for numerical analysis, search method for state relationship are described. Then, a core power model, heat transfer models, physical models for various components, and control and trip models are explained

  10. The Gift Code User Manual. Volume I. Introduction and Input Requirements

    Science.gov (United States)

    1975-07-01

    REPORT & PERIOD COVERED ‘TII~ GIFT CODE USER MANUAL; VOLUME 1. INTRODUCTION AND INPUT REQUIREMENTS FINAL 6. PERFORMING ORG. REPORT NUMBER ?. AuTHOR(#) 8...reverua side if neceaeary and identify by block number] (k St) The GIFT code is a FORTRANcomputerprogram. The basic input to the GIFT ode is data called

  11. Automation of Geometry Input for Building Code Compliance Check

    DEFF Research Database (Denmark)

    Petrova, Ekaterina Aleksandrova; Johansen, Peter Lind; Jensen, Rasmus Lund

    2017-01-01

    Documentation of compliance with the energy performance regulations at the end of the detailed design phase is mandatory for building owners in Denmark. Therefore, besides multidisciplinary input, the building design process requires various iterative analyses, so that the optimal solutions can....... That has left the industry in constant pursuit of possibilities for integration of the tool within the Building Information Modelling environment so that the potential provided by the latter can be harvested and the processed can be optimized. This paper presents a solution for automated data extraction...... from building geometry created in Autodesk Revit and its translation to input for compliance check analysis....

  12. Automation of Geometry Input for Building Code Compliance Check

    DEFF Research Database (Denmark)

    Petrova, Ekaterina Aleksandrova; Johansen, Peter Lind; Jensen, Rasmus Lund

    2017-01-01

    Documentation of compliance with the energy performance regulations at the end of the detailed design phase is mandatory for building owners in Denmark. Therefore, besides multidisciplinary input, the building design process requires various iterative analyses, so that the optimal solutions can b...

  13. Accelerator Physics Code Web Repository

    CERN Document Server

    Zimmermann, Frank; Bellodi, G; Benedetto, E; Dorda, U; Giovannozzi, Massimo; Papaphilippou, Y; Pieloni, T; Ruggiero, F; Rumolo, G; Schmidt, F; Todesco, E; Zotter, Bruno W; Payet, J; Bartolini, R; Farvacque, L; Sen, T; Chin, Y H; Ohmi, K; Oide, K; Furman, M; Qiang, J; Sabbi, G L; Seidl, P A; Vay, J L; Friedman, A; Grote, D P; Cousineau, S M; Danilov, V; Holmes, J A; Shishlo, A; Kim, E S; Cai, Y; Pivi, M; Kaltchev, D I; Abell, D T; Katsouleas, Thomas C; Boine-Frankenheim, O; Franchetti, G; Hofmann, I; Machida, S; Wei, J

    2006-01-01

    In the framework of the CARE HHH European Network, we have developed a web-based dynamic acceleratorphysics code repository. We describe the design, structure and contents of this repository, illustrate its usage, and discuss our future plans, with emphasis on code benchmarking.

  14. Pre-processing of input files for the AZTRAN code; Pre procesamiento de archivos de entrada para el codigo AZTRAN

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ibarra, G., E-mail: samuel.vargas@inin.gob.mx [IPN, Av. Instituto Politecnico Nacional s/n, 07738 Ciudad de Mexico (Mexico)

    2017-09-15

    The AZTRAN code began to be developed in the Nuclear Engineering Department of the Escuela Superior de Fisica y Matematicas (ESFM) of the Instituto Politecnico Nacional (IPN) with the purpose of numerically solving various models arising from the physics and engineering of nuclear reactors. The code is still under development and is part of the AZTLAN platform: Development of a Mexican platform for the analysis and design of nuclear reactors. Due to the complexity to generate an input file for the code, a script based on D language is developed, with the purpose of making its elaboration easier, based on a new input file format which includes specific cards, which have been divided into two blocks, mandatory cards and optional cards, including a pre-processing of the input file to identify possible errors within it, as well as an image generator for the specific problem based on the python interpreter. (Author)

  15. Data exchange between zero dimensional code and physics platform in the CFETR integrated system code

    Energy Technology Data Exchange (ETDEWEB)

    Xu, Guoliang [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Shi, Nan [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Zhou, Yifu; Mao, Shifeng [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Jian, Xiang [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, School of Electrical and Electronics Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China); Chen, Jiale [Institute of Plasma Physics, Chinese Academy of Sciences, No. 350 Shushanhu Road, Hefei (China); Liu, Li; Chan, Vincent [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China); Ye, Minyou, E-mail: yemy@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 China (China)

    2016-11-01

    Highlights: • The workflow of the zero dimensional code and the multi-dimension physics platform of CFETR integrated system codeis introduced. • The iteration process among the codes in the physics platform. • The data transfer between the zero dimensionalcode and the physical platform, including data iteration and validation, and justification for performance parameters.. - Abstract: The China Fusion Engineering Test Reactor (CFETR) integrated system code contains three parts: a zero dimensional code, a physics platform and an engineering platform. We use the zero dimensional code to identify a set of preliminary physics and engineering parameters for CFETR, which is used as input to initiate multi-dimension studies using the physics and engineering platform for design, verification and validation. Effective data exchange between the zero dimensional code and the physical platform is critical for the optimization of CFETR design. For example, in evaluating the impact of impurity radiation on core performance, an open field line code is used to calculate the impurity transport from the first-wall boundary to the pedestal. The impurity particle in the pedestal are used as boundary conditions in a transport code for calculating impurity transport in the core plasma and the impact of core radiation on core performance. Comparison of the results from the multi-dimensional study to those from the zero dimensional code is used to further refine the controlled radiation model. The data transfer between the zero dimensional code and the physical platform, including data iteration and validation, and justification for performance parameters will be presented in this paper.

  16. VOA: a 2-d plasma physics code

    International Nuclear Information System (INIS)

    Eltgroth, P.G.

    1975-12-01

    A 2-dimensional relativistic plasma physics code was written and tested. The non-thermal components of the particle distribution functions are represented by expansion into moments in momentum space. These moments are computed directly from numerical equations. Currently three species are included - electrons, ions and ''beam electrons''. The computer code runs on either the 7600 or STAR machines at LLL. Both the physics and the operation of the code are discussed

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

    International Nuclear Information System (INIS)

    Douglass, M.; Bezak, E.

    2010-01-01

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

  18. Input/output manual of light water reactor fuel analysis code FEMAXI-7 and its related codes

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Saitou, Hiroaki [ITOCHU Techno-Solutions Corporation, Tokyo (Japan)

    2013-10-15

    A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/Output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7. (author)

  19. Input/output manual of light water reactor fuel analysis code FEMAXI-7 and its related codes

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa; Saitou, Hiroaki

    2013-10-01

    A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/Output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7. (author)

  20. Input research and testing of code TOODY. Quarterly report, July--September 1971

    International Nuclear Information System (INIS)

    Haynie, G.A.

    1997-01-01

    The purpose of this report is to simplify and further explain input instructions for Code TOODY and to demonstrate the ability of the code to reproduce cylinder test results. This input is intended to be a supplement to, and not a replacement for, the existing TOODY manual. The TOODY manual should be read and understood before attempting to read this report. Problems arise in the preparation of the input data in four areas: material definition, initial shape definition, the restart feature, and the limiting of output. Aside from these areas, the code is adequately discussed in the manual, 'TOODY, A Computer Program For Calculating Problems Of Motion In Two Dimensions'

  1. Input/output manual of light water reactor fuel performance code FEMAXI-7 and its related codes

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Saitou, Hiroaki [ITOCHU Techno-Solutions Corp., Tokyo (Japan)

    2012-07-15

    A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which has been fully disclosed in the code model description published recently as JAEA-Data/Code 2010-035. The present manual, which is the counterpart of this description, gives detailed explanations of operation method of FEMAXI-7 code and its related codes, methods of Input/Output, methods of source code modification, features of subroutine modules, and internal variables in a specific manner in order to facilitate users to perform a fuel analysis with FEMAXI-7. This report includes some descriptions which are modified from the original contents of JAEA-Data/Code 2010-035. A CD-ROM is attached as an appendix. (author)

  2. Input/output manual of light water reactor fuel performance code FEMAXI-7 and its related codes

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa; Saitou, Hiroaki

    2012-07-01

    A light water reactor fuel analysis code FEMAXI-7 has been developed for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which has been fully disclosed in the code model description published recently as JAEA-Data/Code 2010-035. The present manual, which is the counterpart of this description, gives detailed explanations of operation method of FEMAXI-7 code and its related codes, methods of Input/Output, methods of source code modification, features of subroutine modules, and internal variables in a specific manner in order to facilitate users to perform a fuel analysis with FEMAXI-7. This report includes some descriptions which are modified from the original contents of JAEA-Data/Code 2010-035. A CD-ROM is attached as an appendix. (author)

  3. Development of the GUI environments of MIDAS code for convenient input and output processing

    International Nuclear Information System (INIS)

    Kim, K. L.; Kim, D. H.

    2003-01-01

    MIDAS is being developed at KAERI as an integrated Severe Accident Analysis Code with easy model modification and addition by restructuring the data transfer scheme. In this paper, the input file management system, IEDIT and graphic simulation system, SATS, are presented as MIDAS input and output GUI systems. These two systems would form the basis of the MIDAS GUI system for input and output processing, and they are expected to be useful tools for severe accidents analysis and simulation

  4. DOG -II input generator program for DOT3.5 code

    International Nuclear Information System (INIS)

    Hayashi, Katsumi; Handa, Hiroyuki; Yamada, Koubun; Kamogawa, Susumu; Takatsu, Hideyuki; Koizumi, Kouichi; Seki, Yasushi

    1992-01-01

    DOT3.5 is widely used for radiation transport analysis of fission reactors, fusion experimental facilities and particle accelerators. We developed the input generator program for DOT3.5 code in aim to prepare input data effectively. Formar program DOG was developed and used internally in Hitachi Engineering Company. In this new version DOG-II, limitation for R-Θ geometry was removed. All the input data is created by interactive method in front of color display without using DOT3.5 manual. Also the geometry related input are easily created without calculation of precise curved mesh point. By using DOG-II, reliable input data for DOT3.5 code is obtained easily and quickly

  5. Evaluation of severe accident risks: Quantification of major input parameters: MAACS [MELCOR Accident Consequence Code System] input

    International Nuclear Information System (INIS)

    Sprung, J.L.; Jow, H-N; Rollstin, J.A.; Helton, J.C.

    1990-12-01

    Estimation of offsite accident consequences is the customary final step in a probabilistic assessment of the risks of severe nuclear reactor accidents. Recently, the Nuclear Regulatory Commission reassessed the risks of severe accidents at five US power reactors (NUREG-1150). Offsite accident consequences for NUREG-1150 source terms were estimated using the MELCOR Accident Consequence Code System (MACCS). Before these calculations were performed, most MACCS input parameters were reviewed, and for each parameter reviewed, a best-estimate value was recommended. This report presents the results of these reviews. Specifically, recommended values and the basis for their selection are presented for MACCS atmospheric and biospheric transport, emergency response, food pathway, and economic input parameters. Dose conversion factors and health effect parameters are not reviewed in this report. 134 refs., 15 figs., 110 tabs

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

    International Nuclear Information System (INIS)

    Serfontein, Dawid E.; Mulder, Eben J.; Reitsma, Frederik

    2014-01-01

    A computer code was developed for the semi-automatic translation of input models for the VSOP-A diffusion neutronics simulation code to the format of the newer VSOP 99/05 code. In this paper, this algorithm is presented as a generic method for producing codes for the automatic translation of input models from the format of one code version to another, or even to that of a completely different code. Normally, such translations are done manually. However, input model files, such as for the VSOP codes, often are very large and may consist of many thousands of numeric entries that make no particular sense to the human eye. Therefore the task, of for instance nuclear regulators, to verify the accuracy of such translated files can be very difficult and cumbersome. This may cause translation errors not to be picked up, which may have disastrous consequences later on when a reactor with such a faulty design is built. Therefore a generic algorithm for producing such automatic translation codes may ease the translation and verification process to a great extent. It will also remove human error from the process, which may significantly enhance the accuracy and reliability of the process. The developed algorithm also automatically creates a verification log file which permanently record the names and values of each variable used, as well as the list of meanings of all the possible values. This should greatly facilitate reactor licensing applications

  7. A user input manual for single fuel rod behaviour analysis code FEMAXI-III

    International Nuclear Information System (INIS)

    Saito, Hiroaki; Yanagisawa, Kazuaki; Fujita, Misao.

    1983-03-01

    Principal objectives of Safety related research in connection with lighr water reactor fuel rods under normal operating condition are mainly addressed 1) to assess fuel integrity under steady state condition and 2) to generate initial condition under hypothetical accident. These assessments have to be relied principally upon steady state fuel behaviour computing code that is able to calculate fuel conditions to tbe occurred in a various manner. To achieve these objectives, efforts have been made to develope analytical computer code that calculates in-reactor fuel rod behaviour in best estimate manner. The computer code developed for the prediction of the long-term burnup response of single fuel rod under light water reactor condition is the third in a series of code versions:FEMAMI-III. The code calculates temperature, rod internal gas pressure, fission gas release and pellet-cladding interaction related rod deformation as a function of time-dependent fuel rod power and coolant boundary conditions. This document serves as a user input manual for the code FEMAMI-III which has opened to the public in year of 1982. A general description of the code input and output are included together with typical examples of input data. A detailed description of structures, analytical submodels and solution schemes in the code shall be given in the separate document to be published. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Serfontein, Dawid E., E-mail: Dawid.Serfontein@nwu.ac.za [School of Mechanical and Nuclear Engineering, North West University (PUK-Campus), PRIVATE BAG X6001 (Internal Post Box 360), Potchefstroom 2520 (South Africa); Mulder, Eben J. [School of Mechanical and Nuclear Engineering, North West University (South Africa); Reitsma, Frederik [Calvera Consultants (South Africa)

    2014-05-01

    A computer code was developed for the semi-automatic translation of input models for the VSOP-A diffusion neutronics simulation code to the format of the newer VSOP 99/05 code. In this paper, this algorithm is presented as a generic method for producing codes for the automatic translation of input models from the format of one code version to another, or even to that of a completely different code. Normally, such translations are done manually. However, input model files, such as for the VSOP codes, often are very large and may consist of many thousands of numeric entries that make no particular sense to the human eye. Therefore the task, of for instance nuclear regulators, to verify the accuracy of such translated files can be very difficult and cumbersome. This may cause translation errors not to be picked up, which may have disastrous consequences later on when a reactor with such a faulty design is built. Therefore a generic algorithm for producing such automatic translation codes may ease the translation and verification process to a great extent. It will also remove human error from the process, which may significantly enhance the accuracy and reliability of the process. The developed algorithm also automatically creates a verification log file which permanently record the names and values of each variable used, as well as the list of meanings of all the possible values. This should greatly facilitate reactor licensing applications.

  9. SERPENT Monte Carlo reactor physics code

    International Nuclear Information System (INIS)

    Leppaenen, J.

    2010-01-01

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

  10. INTRA/Mod3.2. Manual and Code Description. Volume I - Physical Modelling

    International Nuclear Information System (INIS)

    Andersson, Jenny; Edlund, O.; Hermann, J.; Johansson, Lise-Lotte

    1999-01-01

    The INTRA Manual consists of two volumes. Volume I of the manual is a thorough description of the code INTRA, the Physical modelling of INTRA and the ruling numerical methods and volume II, the User's Manual is an input description. This document, the Physical modelling of INTRA, contains code characteristics, integration methods and applications

  11. INTRA/Mod3.2. Manual and Code Description. Volume I - Physical Modelling

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Jenny; Edlund, O; Hermann, J; Johansson, Lise-Lotte

    1999-01-01

    The INTRA Manual consists of two volumes. Volume I of the manual is a thorough description of the code INTRA, the Physical modelling of INTRA and the ruling numerical methods and volume II, the User`s Manual is an input description. This document, the Physical modelling of INTRA, contains code characteristics, integration methods and applications

  12. Applications of the ARGUS code in accelerator physics

    International Nuclear Information System (INIS)

    Petillo, J.J.; Mankofsky, A.; Krueger, W.A.; Kostas, C.; Mondelli, A.A.; Drobot, A.T.

    1993-01-01

    ARGUS is a three-dimensional, electromagnetic, particle-in-cell (PIC) simulation code that is being distributed to U.S. accelerator laboratories in collaboration between SAIC and the Los Alamos Accelerator Code Group. It uses a modular architecture that allows multiple physics modules to share common utilities for grid and structure input., memory management, disk I/O, and diagnostics, Physics modules are in place for electrostatic and electromagnetic field solutions., frequency-domain (eigenvalue) solutions, time- dependent PIC, and steady-state PIC simulations. All of the modules are implemented with a domain-decomposition architecture that allows large problems to be broken up into pieces that fit in core and that facilitates the adaptation of ARGUS for parallel processing ARGUS operates on either Cray or workstation platforms, and MOTIF-based user interface is available for X-windows terminals. Applications of ARGUS in accelerator physics and design are described in this paper

  13. Physics options in the plasma code VOA

    International Nuclear Information System (INIS)

    Eltgroth, P.G.

    1976-06-01

    A two dimensional relativistic plasma physics code has been modified to accomodate general electromagnetic boundary conditions and various approximations of basic physics. The code can treat internal conductors and insulators, imposed electromagnetic fields, the effects of external circuitry and non-equilibrium starting conditions. Particle dynamics options include a full microscopic treatment, fully relaxed electrons, a low frequency electron approximation and a combination of approximations for specified zones. Electromagnetic options include the full wave treatment, an electrostatic approximation and two varieties of magnetohydrodynamic approximations in specified zones

  14. Design of a Code-Maker Translator Assistive Input Device with a Contest Fuzzy Recognition Algorithm for the Severely Disabled

    Directory of Open Access Journals (Sweden)

    Chung-Min Wu

    2015-01-01

    Full Text Available This study developed an assistive system for the severe physical disabilities, named “code-maker translator assistive input device” which utilizes a contest fuzzy recognition algorithm and Morse codes encoding to provide the keyboard and mouse functions for users to access a standard personal computer, smartphone, and tablet PC. This assistive input device has seven features that are small size, easy installing, modular design, simple maintenance, functionality, very flexible input interface selection, and scalability of system functions, when this device combined with the computer applications software or APP programs. The users with severe physical disabilities can use this device to operate the various functions of computer, smartphone, and tablet PCs, such as sending e-mail, Internet browsing, playing games, and controlling home appliances. A patient with a brain artery malformation participated in this study. The analysis result showed that the subject could make himself familiar with operating of the long/short tone of Morse code in one month. In the future, we hope this system can help more people in need.

  15. Input-dependent frequency modulation of cortical gamma oscillations shapes spatial synchronization and enables phase coding.

    Science.gov (United States)

    Lowet, Eric; Roberts, Mark; Hadjipapas, Avgis; Peter, Alina; van der Eerden, Jan; De Weerd, Peter

    2015-02-01

    Fine-scale temporal organization of cortical activity in the gamma range (∼25-80Hz) may play a significant role in information processing, for example by neural grouping ('binding') and phase coding. Recent experimental studies have shown that the precise frequency of gamma oscillations varies with input drive (e.g. visual contrast) and that it can differ among nearby cortical locations. This has challenged theories assuming widespread gamma synchronization at a fixed common frequency. In the present study, we investigated which principles govern gamma synchronization in the presence of input-dependent frequency modulations and whether they are detrimental for meaningful input-dependent gamma-mediated temporal organization. To this aim, we constructed a biophysically realistic excitatory-inhibitory network able to express different oscillation frequencies at nearby spatial locations. Similarly to cortical networks, the model was topographically organized with spatially local connectivity and spatially-varying input drive. We analyzed gamma synchronization with respect to phase-locking, phase-relations and frequency differences, and quantified the stimulus-related information represented by gamma phase and frequency. By stepwise simplification of our models, we found that the gamma-mediated temporal organization could be reduced to basic synchronization principles of weakly coupled oscillators, where input drive determines the intrinsic (natural) frequency of oscillators. The gamma phase-locking, the precise phase relation and the emergent (measurable) frequencies were determined by two principal factors: the detuning (intrinsic frequency difference, i.e. local input difference) and the coupling strength. In addition to frequency coding, gamma phase contained complementary stimulus information. Crucially, the phase code reflected input differences, but not the absolute input level. This property of relative input-to-phase conversion, contrasting with latency codes

  16. Modular ORIGEN-S for multi-physics code systems

    International Nuclear Information System (INIS)

    Yesilyurt, Gokhan; Clarno, Kevin T.; Gauld, Ian C.; Galloway, Jack

    2011-01-01

    The ORIGEN-S code in the SCALE 6.0 nuclear analysis code suite is a well-validated tool to calculate the time-dependent concentrations of nuclides due to isotopic depletion, decay, and transmutation for many systems in a wide range of time scales. Application areas include nuclear reactor and spent fuel storage analyses, burnup credit evaluations, decay heat calculations, and environmental assessments. Although simple to use within the SCALE 6.0 code system, especially with the ORIGEN-ARP graphical user interface, it is generally complex to use as a component within an externally developed code suite because of its tight coupling within the infrastructure of the larger SCALE 6.0 system. The ORIGEN2 code, which has been widely integrated within other simulation suites, is no longer maintained by Oak Ridge National Laboratory (ORNL), has obsolete data, and has a relatively small validation database. Therefore, a modular version of the SCALE/ORIGEN-S code was developed to simplify its integration with other software packages to allow multi-physics nuclear code systems to easily incorporate the well-validated isotopic depletion, decay, and transmutation capability to perform realistic nuclear reactor and fuel simulations. SCALE/ORIGEN-S was extensively restructured to develop a modular version that allows direct access to the matrix solvers embedded in the code. Problem initialization and the solver were segregated to provide a simple application program interface and fewer input/output operations for the multi-physics nuclear code systems. Furthermore, new interfaces were implemented to access and modify the ORIGEN-S input variables and nuclear cross-section data through external drivers. Three example drivers were implemented, in the C, C++, and Fortran 90 programming languages, to demonstrate the modular use of the new capability. This modular version of SCALE/ORIGEN-S has been embedded within several multi-physics software development projects at ORNL, including

  17. Modular ORIGEN-S for multi-physics code systems

    Energy Technology Data Exchange (ETDEWEB)

    Yesilyurt, Gokhan; Clarno, Kevin T.; Gauld, Ian C., E-mail: yesilyurtg@ornl.gov, E-mail: clarnokt@ornl.gov, E-mail: gauldi@ornl.gov [Oak Ridge National Laboratory, TN (United States); Galloway, Jack, E-mail: jack@galloways.net [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2011-07-01

    The ORIGEN-S code in the SCALE 6.0 nuclear analysis code suite is a well-validated tool to calculate the time-dependent concentrations of nuclides due to isotopic depletion, decay, and transmutation for many systems in a wide range of time scales. Application areas include nuclear reactor and spent fuel storage analyses, burnup credit evaluations, decay heat calculations, and environmental assessments. Although simple to use within the SCALE 6.0 code system, especially with the ORIGEN-ARP graphical user interface, it is generally complex to use as a component within an externally developed code suite because of its tight coupling within the infrastructure of the larger SCALE 6.0 system. The ORIGEN2 code, which has been widely integrated within other simulation suites, is no longer maintained by Oak Ridge National Laboratory (ORNL), has obsolete data, and has a relatively small validation database. Therefore, a modular version of the SCALE/ORIGEN-S code was developed to simplify its integration with other software packages to allow multi-physics nuclear code systems to easily incorporate the well-validated isotopic depletion, decay, and transmutation capability to perform realistic nuclear reactor and fuel simulations. SCALE/ORIGEN-S was extensively restructured to develop a modular version that allows direct access to the matrix solvers embedded in the code. Problem initialization and the solver were segregated to provide a simple application program interface and fewer input/output operations for the multi-physics nuclear code systems. Furthermore, new interfaces were implemented to access and modify the ORIGEN-S input variables and nuclear cross-section data through external drivers. Three example drivers were implemented, in the C, C++, and Fortran 90 programming languages, to demonstrate the modular use of the new capability. This modular version of SCALE/ORIGEN-S has been embedded within several multi-physics software development projects at ORNL, including

  18. Input to the PRAST computer code used in the SRS probabilistic risk assessment

    International Nuclear Information System (INIS)

    Kearnaghan, D.P.

    1992-01-01

    The PRAST (Production Reactor Algorithm for Source Terms) computer code was developed by Westinghouse Savannah River Company and Science Application International Corporation for the quantification of source terms for the SRS Savannah River Site (SRS) Reactor Probabilistic Risk Assessment. PRAST requires as input a set of release fractions, decontamination factors, transfer fractions and source term characteristics that accurately reflect the conditions that are evaluated by PRAST. This document links the analyses which form the basis for the PRAST input parameters. In addition, it gives the distribution of the input parameters that are uncertain and considered to be important to the evaluation of the source terms to the environment

  19. 78 FR 24725 - National Fire Codes: Request for Public Input for Revision of Codes and Standards

    Science.gov (United States)

    2013-04-26

    ... Production, Storage, and Handling of Liquefied Natural Gas (LNG). NFPA 61--2013 Standard for the 7/6/2015... Nitrate Film. NFPA 51--2013 Standard for the Design 7/6/2015 and Installation of Oxygen-Fuel Gas Systems... Charging Plants. NFPA 52--2013 Vehicular Gaseous Fuel 1/3/2014 Systems Code. NFPA 53--2011 Recommended...

  20. RELAP5/MOD3 code manual: User's guide and input requirements. Volume 2

    International Nuclear Information System (INIS)

    1995-08-01

    The RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during postulated accidents. The code models the coupled behavior of the reactor coolant system and the core for loss-of-coolant accidents, and operational transients, such as anticipated transient without scram, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits simulating a variety of thermal hydraulic systems. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater systems. Volume II contains detailed instructions for code application and input data preparation

  1. Input parameters to codes which analyze LMFBR wire-wrapped bundles

    International Nuclear Information System (INIS)

    Hawley, J.T.; Chan, Y.N.; Todreas, N.E.

    1980-12-01

    This report provides a current summary of recommended values of key input parameters required by ENERGY code analysis of LMFBR wire wrapped bundles. This data is based on the interpretation of experimental results from the MIT and other available laboratory programs

  2. 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.

  3. Spatiotemporal coding of inputs for a system of globally coupled phase oscillators

    Science.gov (United States)

    Wordsworth, John; Ashwin, Peter

    2008-12-01

    We investigate the spatiotemporal coding of low amplitude inputs to a simple system of globally coupled phase oscillators with coupling function g(ϕ)=-sin(ϕ+α)+rsin(2ϕ+β) that has robust heteroclinic cycles (slow switching between cluster states). The inputs correspond to detuning of the oscillators. It was recently noted that globally coupled phase oscillators can encode their frequencies in the form of spatiotemporal codes of a sequence of cluster states [P. Ashwin, G. Orosz, J. Wordsworth, and S. Townley, SIAM J. Appl. Dyn. Syst. 6, 728 (2007)]. Concentrating on the case of N=5 oscillators we show in detail how the spatiotemporal coding can be used to resolve all of the information that relates the individual inputs to each other, providing that a long enough time series is considered. We investigate robustness to the addition of noise and find a remarkable stability, especially of the temporal coding, to the addition of noise even for noise of a comparable magnitude to the inputs.

  4. The Wims-Traca code for the calculation of fuel elements. User's manual and input data

    International Nuclear Information System (INIS)

    Anhert, C.

    1980-01-01

    The set of modifications and new options developped for the Wims-D code is explained. The input data of the new version Wims-Traca are described. The printed output of results is also explained. The contents and the source of the nuclear data in the basic library is exposed. (author)

  5. 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

  6. 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)

  7. Graphical interface for the physics-based generation of inputs to 3D MEEC SGEMP and SREMP simulations

    International Nuclear Information System (INIS)

    Bland, M; Walters, D; Wondra, J

    1999-01-01

    A graphical user interface (GUI) is under development for the MEEC family of SGEMP and SREMP simulation codes [1,2]. These codes are ''workhorse'' legacy codes that have been in use for nearly two decades, with modifications and enhanced physics models added throughout the years. The MEEC codes are currently being evaluated for use by the DOE in the Dual Revalidation Program and experiments at NIF. The new GUI makes the codes more accessible and less prone to input errors by automatically generating the parameters and grids that previously had to be designed ''by hand''. Physics-based algorithms define the simulation volume with expanding meshes. Users are able to specify objects, materials, and emission surfaces through dialogs and input boxes. 3D and orthographic views are available to view objects in the volume. Zone slice views are available for stepping through the overlay of objects on the mesh in planes aligned with the primary axes

  8. Development and validation of gui based input file generation code for relap

    International Nuclear Information System (INIS)

    Anwar, M.M.; Khan, A.A.; Chughati, I.R.; Chaudri, K.S.; Inyat, M.H.; Hayat, T.

    2009-01-01

    Reactor Excursion and Leak Analysis Program (RELAP) is a widely acceptable computer code for thermal hydraulics modeling of Nuclear Power Plants. It calculates thermal- hydraulic transients in water-cooled nuclear reactors by solving approximations to the one-dimensional, two-phase equations of hydraulics in an arbitrarily connected system of nodes. However, the preparation of input file and subsequent analysis of results in this code is a tedious task. The development of a Graphical User Interface (GUI) for preparation of the input file for RELAP-5 is done with the validation of GUI generated Input File. The GUI is developed in Microsoft Visual Studio using Visual C Sharp (C) as programming language. The Nodalization diagram is drawn graphically and the program contains various component forms along with the starting data form, which are launched for properties assignment to generate Input File Cards serving as GUI for the user. The GUI is provided with Open / Save function to store and recall the Nodalization diagram along with Components' properties. The GUI generated Input File is validated for several case studies and individual component cards are compared with the originally required format. The generated Input File of RELAP is found consistent with the requirement of RELAP. The GUI provided a useful platform for simulating complex hydrodynamic problems efficiently with RELAP. (author)

  9. The CAIN computer code for the generation of MABEL input data sets: a user's manual

    International Nuclear Information System (INIS)

    Tilley, D.R.

    1983-03-01

    CAIN is an interactive FORTRAN computer code designed to overcome the substantial effort involved in manually creating the thermal-hydraulics input data required by MABEL-2. CAIN achieves this by processing output from either of the whole-core codes, RELAP or TRAC, interpolating where necessary, and by scanning RELAP/TRAC output in order to generate additional information. This user's manual describes the actions required in order to create RELAP/TRAC data sets from magnetic tape, to create the other input data sets required by CAIN, and to operate the interactive command procedure for the execution of CAIN. In addition, the CAIN code is described in detail. This programme of work is part of the Nuclear Installations Inspectorate (NII)'s contribution to the United Kingdom Atomic Energy Authority's independent safety assessment of pressurized water reactors. (author)

  10. Operational reactor physics analysis codes (ORPAC)

    International Nuclear Information System (INIS)

    Kumar, Jainendra; Singh, K.P.; Singh, Kanchhi

    2007-07-01

    For efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are regularly required. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation of samples requires a prior estimation of the reactivity load due to the sample, the heating rate and the activity developed in it during irradiation. For the safety of personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr.Therefore, a proper shielding and radioactive cooling of the irradiated sample are required to meet the said requirement. Knowledge of xenon load variation with time (Startup-curve) helps in estimating Xenon override time. Monitoring of power in individual fuel channels during reactor operation is essential to know any abnormal power distribution to avoid unsafe situations. Complexities in the estimation of above mentioned reactor parameters and their frequent requirement compel one to use computer codes to avoid possible human errors. For efficient and quick evaluation of parameters related to reactor operations such as xenon load, critical moderator height and nuclear heating and reactivity load of isotope samples/experimental assembly, a computer code ORPAC (Operational Reactor Physics Analysis Codes) has been developed. This code is being used for regular assessment of reactor physics parameters in Dhruva and Cirus. The code ORPAC written in Visual Basic 6.0 environment incorporates several important operational reactor physics aspects on a single platform with graphical user interfaces (GUI) to make it more user-friendly and presentable. (author)

  11. 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

  12. Single-phase multi-dimensional thermohydraulics direct numerical simulation code DINUS-3. Input data description

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-08-01

    This report explains the numerical methods and the set-up method of input data for a single-phase multi-dimensional thermohydraulics direct numerical simulation code DINUS-3 (Direct Numerical Simulation using a 3rd-order upwind scheme). The code was developed to simulate non-stationary temperature fluctuation phenomena related to thermal striping phenomena, developed at Power Reactor and Nuclear Fuel Development Corporation (PNC). The DINUS-3 code was characterized by the use of a third-order upwind scheme for convection terms in instantaneous Navier-Stokes and energy equations, and an adaptive control system based on the Fuzzy theory to control time step sizes. Author expect this report is very useful to utilize the DINUS-3 code for the evaluation of various non-stationary thermohydraulic phenomena in reactor applications. (author)

  13. Development a computer codes to couple PWR-GALE output and PC-CREAM input

    Science.gov (United States)

    Kuntjoro, S.; Budi Setiawan, M.; Nursinta Adi, W.; Deswandri; Sunaryo, G. R.

    2018-02-01

    Radionuclide dispersion analysis is part of an important reactor safety analysis. From the analysis it can be obtained the amount of doses received by radiation workers and communities around nuclear reactor. The radionuclide dispersion analysis under normal operating conditions is carried out using the PC-CREAM code, and it requires input data such as source term and population distribution. Input data is derived from the output of another program that is PWR-GALE and written Population Distribution data in certain format. Compiling inputs for PC-CREAM programs manually requires high accuracy, as it involves large amounts of data in certain formats and often errors in compiling inputs manually. To minimize errors in input generation, than it is make coupling program for PWR-GALE and PC-CREAM programs and a program for writing population distribution according to the PC-CREAM input format. This work was conducted to create the coupling programming between PWR-GALE output and PC-CREAM input and programming to written population data in the required formats. Programming is done by using Python programming language which has advantages of multiplatform, object-oriented and interactive. The result of this work is software for coupling data of source term and written population distribution data. So that input to PC-CREAM program can be done easily and avoid formatting errors. Programming sourceterm coupling program PWR-GALE and PC-CREAM is completed, so that the creation of PC-CREAM inputs in souceterm and distribution data can be done easily and according to the desired format.

  14. The computer code SEURBNUK/EURDYN (Release 1). Input and output specification

    International Nuclear Information System (INIS)

    Broadhouse, B.J.; Yerkess, A.

    1986-05-01

    SEURBNUK/EURODYN is an extension of SEURBNUK-2, a two dimensional, axisymmetric, Eulerian, finite element containment code in which the finite difference thin shell treatment is replaced by a finite element calculation for both thin and thick structures. These codes are designed to model the hydrodynamic development in time of a hypothetical core disruptive accident (HCDA) in a fast breeder reactor. This manual describes the input data specifications needed for the execution of SEURBNUK/EURDYN calculations, with information on output facilities, and aid to users to avoid some common difficulties. (UK)

  15. New approach to derive linear power/burnup history input for CANDU fuel codes

    International Nuclear Information System (INIS)

    Lac Tang, T.; Richards, M.; Parent, G.

    2003-01-01

    The fuel element linear power / burnup history is a required input for the ELESTRES code in order to simulate CANDU fuel behavior during normal operating conditions and also to provide input for the accident analysis codes ELOCA and SOURCE. The purpose of this paper is to present a new approach to derive 'true', or at least more realistic linear power / burnup histories. Such an approach can be used to recreate any typical bundle power history if only a single pair of instantaneous values of bundle power and burnup, together with the position in the channel, are known. The histories obtained could be useful to perform more realistic simulations for safety analyses for cases where the reference (overpower) history is not appropriate. (author)

  16. Input data requirements for special processors in the computation system containing the VENTURE neutronics code

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

    1979-07-01

    User input data requirements are presented for certain special processors in a nuclear reactor computation system. These processors generally read data in formatted form and generate binary interface data files. Some data processing is done to convert from the user oriented form to the interface file forms. The VENTURE diffusion theory neutronics code and other computation modules in this system use the interface data files which are generated

  17. Input data requirements for special processors in the computation system containing the VENTURE neutronics code

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

    1976-11-01

    This report presents user input data requirements for certain special processors in a nuclear reactor computation system. These processors generally read data in formatted form and generate binary interface data files. Some data processing is done to convert from the user-oriented form to the interface file forms. The VENTURE diffusion theory neutronics code and other computation modules in this system use the interface data files which are generated

  18. Safety analysis code input automation using the Nuclear Plant Data Bank

    International Nuclear Information System (INIS)

    Kopp, H.; Leung, J.; Tajbakhsh, A.; Viles, F.

    1985-01-01

    The Nuclear Plant Data Bank (NPDB) is a computer-based system that organizes a nuclear power plant's technical data, providing mechanisms for data storage, retrieval, and computer-aided engineering analysis. It has the specific objective to describe thermohydraulic systems in order to support: rapid information retrieval and display, and thermohydraulic analysis modeling. The Nuclear Plant Data Bank (NPBD) system fully automates the storage and analysis based on this data. The system combines the benefits of a structured data base system and computer-aided modeling with links to large scale codes for engineering analysis. Emphasis on a friendly and very graphically oriented user interface facilitates both initial use and longer term efficiency. Specific features are: organization and storage of thermohydraulic data items, ease in locating specific data items, graphical and tabular display capabilities, interactive model construction, organization and display of model input parameters, input deck construction for TRAC and RELAP analysis programs, and traceability of plant data, user model assumptions, and codes used in the input deck construction process. The major accomplishments of this past year were the development of a RELAP model generation capability and the development of a CRAY version of the code

  19. TASS/SMR Code Topical Report for SMART Plant, Vol II: User's Guide and Input Requirement

    Energy Technology Data Exchange (ETDEWEB)

    Kim, See Darl; Kim, Soo Hyoung; Kim, Hyung Rae (and others)

    2008-10-15

    The TASS/SMR code has been developed with domestic technologies for the safety analysis of the SMART plant which is an integral type pressurized water reactor. It can be applied to the analysis of design basis accidents including non-LOCA (loss of coolant accident) and LOCA of the SMART plant. The TASS/SMR code can be applied to any plant regardless of the structural characteristics of a reactor since the code solves the same governing equations for both the primary and secondary system. The code has been developed to meet the requirements of the safety analysis code. This report describes the overall structure of the TASS/SMR, input processing, and the processes of a steady state and transient calculations. In addition, basic differential equations, finite difference equations, state relationships, and constitutive models are described in the report. First, the conservation equations, a discretization process for numerical analysis, search method for state relationship are described. Then, a core power model, heat transfer models, physical models for various components, and control and trip models are explained.

  20. Sensitivity study of steam explosion characteristics to uncertain input parameters using TEXAS-V code

    International Nuclear Information System (INIS)

    Grishchenko, Dmitry; Basso, Simone; Kudinov, Pavel; Bechta, Sevostian

    2014-01-01

    Release of core melt from failed reactor vessel into a pool of water is adopted in several existing designs of light water reactors (LWRs) as an element of severe accident mitigation strategy. Corium melt is expected to fragment, solidify and form a debris bed coolable by natural circulation. However, steam explosion can occur upon melt release threatening containment integrity and potentially leading to large early release of radioactive products to the environment. There are many factors and parameters that could be considered for prediction of the fuel-coolant interaction (FCI) energetics, but it is not clear which of them are the most influential and should be addressed in risk analysis. The goal of this work is to assess importance of different uncertain input parameters used in FCI code TEXAS-V for prediction of the steam explosion energetics. Both aleatory uncertainty in characteristics of melt release scenarios and water pool conditions, and epistemic uncertainty in modeling are considered. Ranges of the uncertain parameters are selected based on the available information about prototypic severe accident conditions in a reference design of a Nordic BWR. Sensitivity analysis with Morris method is implemented using coupled TEXAS-V and DAKOTA codes. In total 12 input parameters were studied and 2 melt release scenarios were considered. Each scenario is based on 60,000 of TEXAS-V runs. Sensitivity study identified the most influential input parameters, and those which have no statistically significant effect on the explosion energetics. Details of approach to robust usage of TEXAS-V input, statistical enveloping of TEXAS-V output and interpretation of the results are discussed in the paper. We also provide probability density function (PDF) of steam explosion impulse estimated using TEXAS-V for reference Nordic BWR. It can be used for assessment of the uncertainty ranges of steam explosion loads for given ranges of input parameters. (author)

  1. Three-Dimensional Terahertz Coded-Aperture Imaging Based on Single Input Multiple Output Technology

    Directory of Open Access Journals (Sweden)

    Shuo Chen

    2018-01-01

    Full Text Available As a promising radar imaging technique, terahertz coded-aperture imaging (TCAI can achieve high-resolution, forward-looking, and staring imaging by producing spatiotemporal independent signals with coded apertures. In this paper, we propose a three-dimensional (3D TCAI architecture based on single input multiple output (SIMO technology, which can reduce the coding and sampling times sharply. The coded aperture applied in the proposed TCAI architecture loads either purposive or random phase modulation factor. In the transmitting process, the purposive phase modulation factor drives the terahertz beam to scan the divided 3D imaging cells. In the receiving process, the random phase modulation factor is adopted to modulate the terahertz wave to be spatiotemporally independent for high resolution. Considering human-scale targets, images of each 3D imaging cell are reconstructed one by one to decompose the global computational complexity, and then are synthesized together to obtain the complete high-resolution image. As for each imaging cell, the multi-resolution imaging method helps to reduce the computational burden on a large-scale reference-signal matrix. The experimental results demonstrate that the proposed architecture can achieve high-resolution imaging with much less time for 3D targets and has great potential in applications such as security screening, nondestructive detection, medical diagnosis, etc.

  2. Sparse coding can predict primary visual cortex receptive field changes induced by abnormal visual input.

    Science.gov (United States)

    Hunt, Jonathan J; Dayan, Peter; Goodhill, Geoffrey J

    2013-01-01

    Receptive fields acquired through unsupervised learning of sparse representations of natural scenes have similar properties to primary visual cortex (V1) simple cell receptive fields. However, what drives in vivo development of receptive fields remains controversial. The strongest evidence for the importance of sensory experience in visual development comes from receptive field changes in animals reared with abnormal visual input. However, most sparse coding accounts have considered only normal visual input and the development of monocular receptive fields. Here, we applied three sparse coding models to binocular receptive field development across six abnormal rearing conditions. In every condition, the changes in receptive field properties previously observed experimentally were matched to a similar and highly faithful degree by all the models, suggesting that early sensory development can indeed be understood in terms of an impetus towards sparsity. As previously predicted in the literature, we found that asymmetries in inter-ocular correlation across orientations lead to orientation-specific binocular receptive fields. Finally we used our models to design a novel stimulus that, if present during rearing, is predicted by the sparsity principle to lead robustly to radically abnormal receptive fields.

  3. Basic concept of common reactor physics code systems. Final report of working party on common reactor physics code systems (CCS)

    International Nuclear Information System (INIS)

    2004-03-01

    A working party was organized for two years (2001-2002) on common reactor physics code systems under the Research Committee on Reactor Physics of JAERI. This final report is compilation of activity of the working party on common reactor physics code systems during two years. Objectives of the working party is to clarify basic concept of common reactor physics code systems to improve convenience of reactor physics code systems for reactor physics researchers in Japan on their various field of research and development activities. We have held four meetings during 2 years, investigated status of reactor physics code systems and innovative software technologies, and discussed basic concept of common reactor physics code systems. (author)

  4. Does Kaniso activate CASINO?: input coding schemes and phonology in visual-word recognition.

    Science.gov (United States)

    Acha, Joana; Perea, Manuel

    2010-01-01

    Most recent input coding schemes in visual-word recognition assume that letter position coding is orthographic rather than phonological in nature (e.g., SOLAR, open-bigram, SERIOL, and overlap). This assumption has been drawn - in part - by the fact that the transposed-letter effect (e.g., caniso activates CASINO) seems to be (mostly) insensitive to phonological manipulations (e.g., Perea & Carreiras, 2006, 2008; Perea & Pérez, 2009). However, one could argue that the lack of a phonological effect in prior research was due to the fact that the manipulation always occurred in internal letter positions - note that phonological effects tend to be stronger for the initial syllable (Carreiras, Ferrand, Grainger, & Perea, 2005). To reexamine this issue, we conducted a masked priming lexical decision experiment in which we compared the priming effect for transposed-letter pairs (e.g., caniso-CASINO vs. caviro-CASINO) and for pseudohomophone transposed-letter pairs (kaniso-CASINO vs. kaviro-CASINO). Results showed a transposed-letter priming effect for the correctly spelled pairs, but not for the pseudohomophone pairs. This is consistent with the view that letter position coding is (primarily) orthographic in nature.

  5. The computer code SEURBNUK/EURDYN (release 1). Input and output specifications

    International Nuclear Information System (INIS)

    Smith, B.L.; Broadhouse, B.J.; Yerkess, A.

    1986-05-01

    SEURBNUK-2 is a two-dimensional, axisymmetric, Eulerian, finite difference containment code developed initially by AWRE Aldermaston, AEE Winfrith and JRC-Ispra, and more recently by AEEW, JRC and EIR Wuerenlingen. The numerical procedure adopted in SEURBNUK to solve the hydrodynamic equations is based on the semi-implicit ICE method which itself is an extension of the MAC algorithm. SEURBNUK has a finite difference thin shell treatment for vessels and internal structures of arbitrary shape and includes the effects of the compressibility of the fluid. Fluid flow through porous media and porous structures can also be accommodated. SEURBNUK/EURDYN is an extension of SEURBNUK-2 in which the finite difference thin shell treatment is replaced by a finite element calculation for both thin or thick structures. This has been achieved by coupling the finite element code EURDYN with SEURBNUK-2, allowing the use of conical shell elements and axisymmetric triangular elements. Within the code, the equations of motion for the structures are solved quite separately from those for the fluid, and the timestep for the fluid can be an integer multiple of that for the structures. The interaction of the structures with the fluid is then considered as a modification to the coefficients in the pressure equations, the modifications naturally depending on the behaviour of the structures within the fluid cell. The code is limited to dealing with a single fluid, the coolant, and the bubble and the cover gas are treated as cavities of uniform pressure calculated via appropriate pressure-volume-energy relationships. This manual describes the input data specifications needed for the execution of SEURBNUK/EURDYN calculations. After explaining the output facilities information is included to aid users to avoid some common pit-falls. (author)

  6. 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

  7. 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)

  8. The HELIOS-2 lattice physics code

    International Nuclear Information System (INIS)

    Wemple, C.A.; Gheorghiu, H-N.M.; Stamm'ler, R.J.J.; Villarino, E.A.

    2008-01-01

    Major advances have been made in the HELIOS code, resulting in the impending release of a new version, HELIOS-2. The new code includes a method of characteristics (MOC) transport solver to supplement the existing collision probabilities (CP) solver. A 177-group, ENDF/B-VII nuclear data library has been developed for inclusion with the new code package. Computational tests have been performed to verify the performance of the MOC solver against the CP solver, and validation testing against computational and measured benchmarks is underway. Results to-date of the verification and validation testing are presented, demonstrating the excellent performance of the new transport solver and nuclear data library. (Author)

  9. The r-Java 2.0 code: nuclear physics

    Science.gov (United States)

    Kostka, M.; Koning, N.; Shand, Z.; Ouyed, R.; Jaikumar, P.

    2014-08-01

    Aims: We present r-Java 2.0, a nucleosynthesis code for open use that performs r-process calculations, along with a suite of other analysis tools. Methods: Equipped with a straightforward graphical user interface, r-Java 2.0 is capable of simulating nuclear statistical equilibrium (NSE), calculating r-process abundances for a wide range of input parameters and astrophysical environments, computing the mass fragmentation from neutron-induced fission and studying individual nucleosynthesis processes. Results: In this paper we discuss enhancements to this version of r-Java, especially the ability to solve the full reaction network. The sophisticated fission methodology incorporated in r-Java 2.0 that includes three fission channels (beta-delayed, neutron-induced, and spontaneous fission), along with computation of the mass fragmentation, is compared to the upper limit on mass fission approximation. The effects of including beta-delayed neutron emission on r-process yield is studied. The role of Coulomb interactions in NSE abundances is shown to be significant, supporting previous findings. A comparative analysis was undertaken during the development of r-Java 2.0 whereby we reproduced the results found in the literature from three other r-process codes. This code is capable of simulating the physical environment of the high-entropy wind around a proto-neutron star, the ejecta from a neutron star merger, or the relativistic ejecta from a quark nova. Likewise the users of r-Java 2.0 are given the freedom to define a custom environment. This software provides a platform for comparing proposed r-process sites.

  10. OFFSCALE: A PC input processor for the SCALE code system. The ORIGNATE processor for ORIGEN-S

    International Nuclear Information System (INIS)

    Bowman, S.M.

    1994-11-01

    OFFSCALE is a suite of personal computer input processor programs developed at Oak Ridge National Laboratory to provide an easy-to-use interface for modules in the SCALE-4 code system. ORIGNATE is a program in the OFFSCALE suite that serves as a user-friendly interface for the ORIGEN-S isotopic generation and depletion code. It is designed to assist an ORIGEN-S user in preparing an input file for execution of light-water-reactor (LWR) fuel depletion and decay cases. ORIGNATE generates an input file that may be used to execute ORIGEN-S in SCALE-4. ORIGNATE features a pulldown menu system that accesses sophisticated data entry screens. The program allows the user to quickly set up an ORIGEN-S input file and perform error checking. This capability increases productivity and decreases the chance of user error

  11. Preliminary Coupling of MATRA Code for Multi-physics Analysis

    International Nuclear Information System (INIS)

    Kim, Seongjin; Choi, Jinyoung; Yang, Yongsik; Kwon, Hyouk; Hwang, Daehyun

    2014-01-01

    The boundary conditions such as the inlet temperature, mass flux, averaged heat flux, power distributions of the rods, and core geometry is given by constant values or functions of time. These conditions are separately calculated and provided by other codes, such as a neutronics or a system codes, into the MATRA code. In addition, the coupling of several codes in the different physics field is focused and embodied. In this study, multiphysics coupling methods were developed for a subchannel code (MATRA) with neutronics codes (MASTER, DeCART) and a fuel performance code (FRAPCON-3). Preliminary evaluation results for representative sample cases are presented. The MASTER and DeCART codes provide the power distribution of the rods in the core to the MATRA code. In case of the FRAPCON-3 code, the variation of the rod diameter induced by the thermal expansion is yielded and provided. The MATRA code transfers the thermal-hydraulic conditions that each code needs. Moreover, the coupling method with each code is described

  12. Successful vectorization - reactor physics Monte Carlo code

    International Nuclear Information System (INIS)

    Martin, W.R.

    1989-01-01

    Most particle transport Monte Carlo codes in use today are based on the ''history-based'' algorithm, wherein one particle history at a time is simulated. Unfortunately, the ''history-based'' approach (present in all Monte Carlo codes until recent years) is inherently scalar and cannot be vectorized. In particular, the history-based algorithm cannot take advantage of vector architectures, which characterize the largest and fastest computers at the current time, vector supercomputers such as the Cray X/MP or IBM 3090/600. However, substantial progress has been made in recent years in developing and implementing a vectorized Monte Carlo algorithm. This algorithm follows portions of many particle histories at the same time and forms the basis for all successful vectorized Monte Carlo codes that are in use today. This paper describes the basic vectorized algorithm along with descriptions of several variations that have been developed by different researchers for specific applications. These applications have been mainly in the areas of neutron transport in nuclear reactor and shielding analysis and photon transport in fusion plasmas. The relative merits of the various approach schemes will be discussed and the present status of known vectorization efforts will be summarized along with available timing results, including results from the successful vectorization of 3-D general geometry, continuous energy Monte Carlo. (orig.)

  13. 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

  14. 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

  15. Development of M3C code for Monte Carlo reactor physics criticality calculations

    International Nuclear Information System (INIS)

    Kumar, Anek; Kannan, Umasankari; Krishanani, P.D.

    2015-06-01

    The development of Monte Carlo code (M3C) for reactor design entails use of continuous energy nuclear data and Monte Carlo simulations for each of the neutron interaction processes. BARC has started a concentrated effort for developing a new general geometry continuous energy Monte Carlo code for reactor physics calculation indigenously. The code development required a comprehensive understanding of the basic continuous energy cross section sets. The important features of this code are treatment of heterogeneous lattices by general geometry, use of point cross sections along with unionized energy grid approach, thermal scattering model for low energy treatment, capability of handling the microscopic fuel particles dispersed randomly. The capability of handling the randomly dispersed microscopic fuel particles which is very useful for the modeling of High-Temperature Gas-Cooled reactor fuels which are composed of thousands of microscopic fuel particle (TRISO fuel particle), randomly dispersed in a graphite matrix. The Monte Carlo code for criticality calculation is a pioneering effort and has been used to study several types of lattices including cluster geometries. The code has been verified for its accuracy against more than 60 sample problems covering a wide range from simple (like spherical) to complex geometry (like PHWR lattice). Benchmark results show that the code performs quite well for the criticality calculation of the system. In this report, the current status of the code, features of the code, some of the benchmark results for the testing of the code and input preparation etc. are discussed. (author)

  16. OFFSCALE: A PC input processor for the SCALE code system. The CSASIN processor for the criticality sequences

    International Nuclear Information System (INIS)

    Bowman, S.M.

    1994-11-01

    OFFSCALE is a suite of personal computer input processor programs developed at Oak Ridge National Laboratory to provide an easy-to-use interface for modules in the SCALE-4 code system. CSASIN (formerly known as OFFSCALE) is a program in the OFFSCALE suite that serves as a user-friendly interface for the Criticality Safety Analysis Sequences (CSAS) available in SCALE-4. It is designed to assist a SCALE-4 user in preparing an input file for execution of criticality safety problems. Output from CSASIN generates an input file that may be used to execute the CSAS control module in SCALE-4. CSASIN features a pulldown menu system that accesses sophisticated data entry screens. The program allows the user to quickly set up a CSAS input file and perform data checking. This capability increases productivity and decreases the chance of user error

  17. Double-Layer Low-Density Parity-Check Codes over Multiple-Input Multiple-Output Channels

    Directory of Open Access Journals (Sweden)

    Yun Mao

    2012-01-01

    Full Text Available We introduce a double-layer code based on the combination of a low-density parity-check (LDPC code with the multiple-input multiple-output (MIMO system, where the decoding can be done in both inner-iteration and outer-iteration manners. The present code, called low-density MIMO code (LDMC, has a double-layer structure, that is, one layer defines subcodes that are embedded in each transmission vector and another glues these subcodes together. It supports inner iterations inside the LDPC decoder and outeriterations between detectors and decoders, simultaneously. It can also achieve the desired design rates due to the full rank of the deployed parity-check matrix. Simulations show that the LDMC performs favorably over the MIMO systems.

  18. FEMSYN - a code system to solve multigroup diffusion theory equations using a variety of solution techniques. Part 1 : Description of code system - input and sample problems

    International Nuclear Information System (INIS)

    Jagannathan, V.

    1985-01-01

    A modular computer code system called FEMSYN has been developed to solve the multigroup diffusion theory equations. The various methods that are incorporated in FEMSYN are (i) finite difference method (FDM) (ii) finite element method (FEM) and (iii) single channel flux synthesis method (SCFS). These methods are described in detail in parts II, III and IV of the present report. In this report, a comparison of the accuracy and the speed of different methods of solution for some benchmark problems are reported. The input preparation and listing of sample input and output are included in the Appendices. The code FEMSYN has been used to solve a wide variety of reactor core problems. It can be used for both LWR and PHWR applications. (author)

  19. Development of input data to energy code for analysis of reactor fuel bundles

    International Nuclear Information System (INIS)

    Carre, F.O.; Todreas, N.E.

    1975-05-01

    The ENERGY 1 code is a semi-empirical method for predicting temperature distributions in wire wrapped rod bundles of a LMFBR. A comparison of ENERGY 1 and MISTRAL 2 is presented. The predictions of ENERGY 1 for special sets of data taken under geometric conditions at the limits of the code are analyzed. 14 references

  20. Adaptive Iterative Soft-Input Soft-Output Parallel Decision-Feedback Detectors for Asynchronous Coded DS-CDMA Systems

    Directory of Open Access Journals (Sweden)

    Zhang Wei

    2005-01-01

    Full Text Available The optimum and many suboptimum iterative soft-input soft-output (SISO multiuser detectors require a priori information about the multiuser system, such as the users' transmitted signature waveforms, relative delays, as well as the channel impulse response. In this paper, we employ adaptive algorithms in the SISO multiuser detector in order to avoid the need for this a priori information. First, we derive the optimum SISO parallel decision-feedback detector for asynchronous coded DS-CDMA systems. Then, we propose two adaptive versions of this SISO detector, which are based on the normalized least mean square (NLMS and recursive least squares (RLS algorithms. Our SISO adaptive detectors effectively exploit the a priori information of coded symbols, whose soft inputs are obtained from a bank of single-user decoders. Furthermore, we consider how to select practical finite feedforward and feedback filter lengths to obtain a good tradeoff between the performance and computational complexity of the receiver.

  1. Federal Logistics Information System (FLIS) Procedures Manual. Volume 8. Document Identifier Code Input/Output Formats (Fixed Length)

    Science.gov (United States)

    1994-07-01

    REQUIRED MIX OF SEGMENTS OR INDIVIDUAL DATA ELEMENTS TO BE EXTRACTED. IN SEGMENT R ON AN INTERROGATION TRANSACTION (LTI), DATA RECORD NUMBER (DRN 0950) ONLY...and zation and Marketing input DICs. insert the Continuation Indicator Code (DRN 8555) in position 80 of this record. Maximum of OF The assigned NSN...for Procurement KFR, File Data Minus Security Classified Characteristics Data KFC 8.5-2 DoD 4100.39-M Volume 8 CHAPTER 5 ALPHABETIC INDEX OF DIC

  2. A parametric study of MELCOR Accident Consequence Code System 2 (MACCS2) Input Values for the Predicted Health Effect

    International Nuclear Information System (INIS)

    Kim, So Ra; Min, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk

    2016-01-01

    The MELCOR Accident Consequence Code System 2, MACCS2, has been the most widely used through the world among the off-site consequence analysis codes. MACCS2 code is used to estimate the radionuclide concentrations, radiological doses, health effects, and economic consequences that could result from the hypothetical nuclear accidents. Most of the MACCS model parameter values are defined by the user and those input parameters can make a significant impact on the output. A limited parametric study was performed to identify the relative importance of the values of each input parameters in determining the predicted early and latent health effects in MACCS2. These results would not be applicable to every case of the nuclear accidents, because only the limited calculation was performed with Kori-specific data. The endpoints of the assessment were early- and latent cancer-risk in the exposed population, therefore it might produce the different results with the parametric studies for other endpoints, such as contamination level, absorbed dose, and economic cost. Accident consequence assessment is important for decision making to minimize the health effect from radiation exposure, accordingly the sufficient parametric studies are required for the various endpoints and input parameters in further research

  3. High-fidelity plasma codes for burn physics

    Energy Technology Data Exchange (ETDEWEB)

    Cooley, James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Graziani, Frank [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marinak, Marty [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Murillo, Michael [Michigan State Univ., East Lansing, MI (United States)

    2016-10-19

    Accurate predictions of equation of state (EOS), ionic and electronic transport properties are of critical importance for high-energy-density plasma science. Transport coefficients inform radiation-hydrodynamic codes and impact diagnostic interpretation, which in turn impacts our understanding of the development of instabilities, the overall energy balance of burning plasmas, and the efficacy of self-heating from charged-particle stopping. Important processes include thermal and electrical conduction, electron-ion coupling, inter-diffusion, ion viscosity, and charged particle stopping. However, uncertainties in these coefficients are not well established. Fundamental plasma science codes, also called high-fidelity plasma codes, are a relatively recent computational tool that augments both experimental data and theoretical foundations of transport coefficients. This paper addresses the current status of HFPC codes and their future development, and the potential impact they play in improving the predictive capability of the multi-physics hydrodynamic codes used in HED design.

  4. SCDAP/RELAP5/MOD 3.1 code manual: User's guide and input manual. Volume 3

    International Nuclear Information System (INIS)

    Coryell, E.W.; Johnsen, E.C.; Allison, C.M.

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, core, fission product released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume provides guidelines to code users based upon lessons learned during the developmental assessment process. A description of problem control and the installation process is included. Appendix a contains the description of the input requirements

  5. On the average complexity of sphere decoding in lattice space-time coded multiple-input multiple-output channel

    KAUST Repository

    Abediseid, Walid

    2012-12-21

    The exact average complexity analysis of the basic sphere decoder for general space-time codes applied to multiple-input multiple-output (MIMO) wireless channel is known to be difficult. In this work, we shed the light on the computational complexity of sphere decoding for the quasi- static, lattice space-time (LAST) coded MIMO channel. Specifically, we drive an upper bound of the tail distribution of the decoder\\'s computational complexity. We show that when the computational complexity exceeds a certain limit, this upper bound becomes dominated by the outage probability achieved by LAST coding and sphere decoding schemes. We then calculate the minimum average computational complexity that is required by the decoder to achieve near optimal performance in terms of the system parameters. Our results indicate that there exists a cut-off rate (multiplexing gain) for which the average complexity remains bounded. Copyright © 2012 John Wiley & Sons, Ltd.

  6. The DIT nuclear fuel assembly physics design code

    International Nuclear Information System (INIS)

    Jonsson, A.

    1988-01-01

    The DIT code is the Combustion Engineering, Inc. (C-E) nuclear fuel assembly design code. It belongs to a class of codes, all similar in structure and strategy, that may be characterized by the spectrum and spatial calculations being performed in two dimensions and in a single job step for the entire assembly. The forerunner of this class of codes is the United Kingdom Atomic Energy Authority WIMS code, the first version of which was completed 25 yr ago. The structure and strategy of assembly spectrum codes have remained remarkably similar to the original concept thus proving its usefulness. As other organizations, including C-E, have developed their own versions of the concept, many important variations have been added that significantly influence the accuracy and performance of the resulting computational tool. Those features, which are unique to the DIT code and which might be of interest to the community of fuel assembly physics design code users and developers, are described and discussed

  7. 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

  8. Applying Physical-Layer Network Coding in Wireless Networks

    Directory of Open Access Journals (Sweden)

    Liew SoungChang

    2010-01-01

    Full Text Available A main distinguishing feature of a wireless network compared with a wired network is its broadcast nature, in which the signal transmitted by a node may reach several other nodes, and a node may receive signals from several other nodes, simultaneously. Rather than a blessing, this feature is treated more as an interference-inducing nuisance in most wireless networks today (e.g., IEEE 802.11. This paper shows that the concept of network coding can be applied at the physical layer to turn the broadcast property into a capacity-boosting advantage in wireless ad hoc networks. Specifically, we propose a physical-layer network coding (PNC scheme to coordinate transmissions among nodes. In contrast to "straightforward" network coding which performs coding arithmetic on digital bit streams after they have been received, PNC makes use of the additive nature of simultaneously arriving electromagnetic (EM waves for equivalent coding operation. And in doing so, PNC can potentially achieve 100% and 50% throughput increases compared with traditional transmission and straightforward network coding, respectively, in 1D regular linear networks with multiple random flows. The throughput improvements are even larger in 2D regular networks: 200% and 100%, respectively.

  9. Opacity calculations for extreme physical systems: code RACHEL

    Science.gov (United States)

    Drska, Ladislav; Sinor, Milan

    1996-08-01

    Computer simulations of physical systems under extreme conditions (high density, temperature, etc.) require the availability of extensive sets of atomic data. This paper presents basic information on a self-consistent approach to calculations of radiative opacity, one of the key characteristics of such systems. After a short explanation of general concepts of the atomic physics of extreme systems, the structure of the opacity code RACHEL is discussed and some of its applications are presented.

  10. FLIS Procedures Manual. Document Identifier Code Input/Output Formats (Fixed Length). Volume 8.

    Science.gov (United States)

    1997-04-01

    DATA ELE- MENTS. SEGMENT R MAY BE REPEATED A MAXIMUM OF THREE (3) TIMES IN ORDER TO ACQUIRE THE REQUIRED MIX OF SEGMENTS OR INDIVIDUAL DATA ELEMENTS TO...preceding record. Marketing input DICs. QI Next DRN of appropriate segment will be QF The assigned NSN or PSCN being can- reflected in accordance with Table...Classified KFC Notification of Possible Duplicate (Sub- KRP Characteristics Data mitter) Follow-Up Interrogation LFU Notification of Return, SSR Transaction

  11. The Dit nuclear fuel assembly physics design code

    International Nuclear Information System (INIS)

    Jonsson, A.

    1987-01-01

    DIT is the Combustion Engineering, Inc. (C-E) nuclear fuel assembly design code. It belongs to a class of codes, all similar in structure and strategy, which may be characterized by the spectrum and spatial calculations being performed in 2D and in a single job step for the entire assembly. The forerunner of this class of codes is the U.K.A.E.A. WIMS code, the first version of which was completed 25 years ago. The structure and strategy of assembly spectrum codes have remained remarkably similar to the original concept thus proving its usefulness. As other organizations, including C-E, have developed their own versions of the concept, many important variations have been added which significantly influence the accuracy and performance of the resulting computational tool. This paper describes and discusses those features which are unique to the DIT code and which might be of interest to the community of fuel assembly physics design code users and developers

  12. Physical-layer network coding in coherent optical OFDM systems.

    Science.gov (United States)

    Guan, Xun; Chan, Chun-Kit

    2015-04-20

    We present the first experimental demonstration and characterization of the application of optical physical-layer network coding in coherent optical OFDM systems. It combines two optical OFDM frames to share the same link so as to enhance system throughput, while individual OFDM frames can be recovered with digital signal processing at the destined node.

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

    International Nuclear Information System (INIS)

    Bartlett, D.V.

    1983-06-01

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

  14. C2x: A tool for visualisation and input preparation for CASTEP and other electronic structure codes

    Science.gov (United States)

    Rutter, M. J.

    2018-04-01

    The c2x code fills two distinct roles. Its first role is in acting as a converter between the binary format .check files from the widely-used CASTEP [1] electronic structure code and various visualisation programs. Its second role is to manipulate and analyse the input and output files from a variety of electronic structure codes, including CASTEP, ONETEP and VASP, as well as the widely-used 'Gaussian cube' file format. Analysis includes symmetry analysis, and manipulation arbitrary cell transformations. It continues to be under development, with growing functionality, and is written in a form which would make it easy to extend it to working directly with files from other electronic structure codes. Data which c2x is capable of extracting from CASTEP's binary checkpoint files include charge densities, spin densities, wavefunctions, relaxed atomic positions, forces, the Fermi level, the total energy, and symmetry operations. It can recreate .cell input files from checkpoint files. Volumetric data can be output in formats useable by many common visualisation programs, and c2x will itself calculate integrals, expand data into supercells, and interpolate data via combinations of Fourier and trilinear interpolation. It can extract data along arbitrary lines (such as lines between atoms) as 1D output. C2x is able to convert between several common formats for describing molecules and crystals, including the .cell format of CASTEP. It can construct supercells, reduce cells to their primitive form, and add specified k-point meshes. It uses the spglib library [2] to report symmetry information, which it can add to .cell files. C2x is a command-line utility, so is readily included in scripts. It is available under the GPL and can be obtained from http://www.c2x.org.uk. It is believed to be the only open-source code which can read CASTEP's .check files, so it will have utility in other projects.

  15. Validation of the VTT's reactor physics code system

    International Nuclear Information System (INIS)

    Tanskanen, A.

    1998-01-01

    At VTT Energy several international reactor physics codes and nuclear data libraries are used in a variety of applications. The codes and libraries are under constant development and every now and then new updated versions are released, which are taken in use as soon as they have been validated at VTT Energy. The primary aim of the validation is to ensure that the code works properly, and that it can be used correctly. Moreover, the applicability of the codes and libraries are studied in order to establish their advantages and weak points. The capability of generating program-specific nuclear data for different reactor physics codes starting from the same evaluated data is sometimes of great benefit. VTT Energy has acquired a nuclear data processing system based on the NJOY-94.105 and TRANSX-2.15 processing codes. The validity of the processing system has been demonstrated by generating pointwise (MCNP) and groupwise (ANISN) temperature-dependent cross section sets for the benchmark calculations of the Doppler coefficient of reactivity. At VTT Energy the KENO-VI three-dimensional Monte Carlo code is used in criticality safety analyses. The KENO-VI code and the 44GROUPNDF5 data library have been validated at VTT Energy against the ZR-6 and LR-0 critical experiments. Burnup Credit refers to the reduction in reactivity of burned nuclear fuel due to the change in composition during irradiation. VTT Energy has participated in the calculational VVER-440 burnup credit benchmark in order to validate criticality safety calculation tools. (orig.)

  16. Enhanced Verification Test Suite for Physics Simulation Codes

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-10-10

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

  17. The Vulnerability Assessment Code for Physical Protection System

    International Nuclear Information System (INIS)

    Jang, Sung Soon; Yoo, Ho Sik

    2007-01-01

    To neutralize the increasing terror threats, nuclear facilities have strong physical protection system (PPS). PPS includes detectors, door locks, fences, regular guard patrols, and a hot line to a nearest military force. To design an efficient PPS and to fully operate it, vulnerability assessment process is required. Evaluating PPS of a nuclear facility is complicate process and, hence, several assessment codes have been developed. The estimation of adversary sequence interruption (EASI) code analyzes vulnerability along a single intrusion path. To evaluate many paths to a valuable asset in an actual facility, the systematic analysis of vulnerability to intrusion (SAVI) code was developed. KAERI improved SAVI and made the Korean analysis of vulnerability to intrusion (KAVI) code. Existing codes (SAVI and KAVI) have limitations in representing the distance of a facility because they use the simplified model of a PPS called adversary sequence diagram. In adversary sequence diagram the position of doors, sensors and fences is described just as the locating area. Thus, the distance between elements is inaccurate and we cannot reflect the range effect of sensors. In this abstract, we suggest accurate and intuitive vulnerability assessment based on raster map modeling of PPS. The raster map of PPS accurately represents the relative position of elements and, thus, the range effect of sensor can be easily incorporable. Most importantly, the raster map is easy to understand

  18. Enhanced verification test suite for physics simulation codes

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-09-01

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

  19. Generation of initial geometries for the simulation of the physical system in the DualPHYsics code

    International Nuclear Information System (INIS)

    Segura Q, E.

    2013-01-01

    In the diverse research areas of the Instituto Nacional de Investigaciones Nucleares (ININ) are different activities related to science and technology, one of great interest is the study and treatment of the collection and storage of radioactive waste. Therefore at ININ the draft on the simulation of the pollutants diffusion in the soil through a porous medium (third stage) has this problem inherent aspects, hence a need for such a situation is to generate the initial geometry of the physical system For the realization of the simulation method is implemented smoothed particle hydrodynamics (SPH). This method runs in DualSPHysics code, which has great versatility and ability to simulate phenomena of any physical system where hydrodynamic aspects combine. In order to simulate a physical system DualSPHysics code, you need to preset the initial geometry of the system of interest, then this is included in the input file of the code. The simulation sets the initial geometry through regular geometric bodies positioned at different points in space. This was done through a programming language (Fortran, C + +, Java, etc..). This methodology will provide the basis to simulate more complex geometries future positions and form. (Author)

  20. 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.)

  1. 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.)

  2. Cholinergic Inputs from Basal Forebrain Add an Excitatory Bias to Odor Coding in the Olfactory Bulb

    Science.gov (United States)

    Rothermel, Markus; Carey, Ryan M.; Puche, Adam; Shipley, Michael T.

    2014-01-01

    Cholinergic modulation of central circuits is associated with active sensation, attention, and learning, yet the neural circuits and temporal dynamics underlying cholinergic effects on sensory processing remain unclear. Understanding the effects of cholinergic modulation on particular circuits is complicated by the widespread projections of cholinergic neurons to telencephalic structures that themselves are highly interconnected. Here we examined how cholinergic projections from basal forebrain to the olfactory bulb (OB) modulate output from the first stage of sensory processing in the mouse olfactory system. By optogenetically activating their axons directly in the OB, we found that cholinergic projections from basal forebrain regulate OB output by increasing the spike output of presumptive mitral/tufted cells. Cholinergic stimulation increased mitral/tufted cell spiking in the absence of inhalation-driven sensory input and further increased spiking responses to inhalation of odorless air and to odorants. This modulation was rapid and transient, was dependent on local cholinergic signaling in the OB, and differed from modulation by optogenetic activation of cholinergic neurons in basal forebrain, which led to a mixture of mitral/tufted cell excitation and suppression. Finally, bulbar cholinergic enhancement of mitral/tufted cell odorant responses was robust and occurred independent of the strength or even polarity of the odorant-evoked response, indicating that cholinergic modulation adds an excitatory bias to mitral/tufted cells as opposed to increasing response gain or sharpening response spectra. These results are consistent with a role for the basal forebrain cholinergic system in dynamically regulating the sensitivity to or salience of odors during active sensing of the olfactory environment. PMID:24672011

  3. Health physics source document for codes of practice

    International Nuclear Information System (INIS)

    Pearson, G.W.; Meggitt, G.C.

    1989-05-01

    Personnel preparing codes of practice often require basic Health Physics information or advice relating to radiological protection problems and this document is written primarily to supply such information. Certain technical terms used in the text are explained in the extensive glossary. Due to the pace of change in the field of radiological protection it is difficult to produce an up-to-date document. This document was compiled during 1988 however, and therefore contains the principle changes brought about by the introduction of the Ionising Radiations Regulations (1985). The paper covers the nature of ionising radiation, its biological effects and the principles of control. It is hoped that the document will provide a useful source of information for both codes of practice and wider areas and stimulate readers to study radiological protection issues in greater depth. (author)

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

    International Nuclear Information System (INIS)

    Mohd Faiz Salim; Ridha Roslan; Mohd Rizal Mamat

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-02-12

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

  7. A statistical methodology for quantification of uncertainty in best estimate code physical models

    International Nuclear Information System (INIS)

    Vinai, Paolo; Macian-Juan, Rafael; Chawla, Rakesh

    2007-01-01

    A novel uncertainty assessment methodology, based on a statistical non-parametric approach, is presented in this paper. It achieves quantification of code physical model uncertainty by making use of model performance information obtained from studies of appropriate separate-effect tests. Uncertainties are quantified in the form of estimated probability density functions (pdf's), calculated with a newly developed non-parametric estimator. The new estimator objectively predicts the probability distribution of the model's 'error' (its uncertainty) from databases reflecting the model's accuracy on the basis of available experiments. The methodology is completed by applying a novel multi-dimensional clustering technique based on the comparison of model error samples with the Kruskall-Wallis test. This takes into account the fact that a model's uncertainty depends on system conditions, since a best estimate code can give predictions for which the accuracy is affected by the regions of the physical space in which the experiments occur. The final result is an objective, rigorous and accurate manner of assigning uncertainty to coded models, i.e. the input information needed by code uncertainty propagation methodologies used for assessing the accuracy of best estimate codes in nuclear systems analysis. The new methodology has been applied to the quantification of the uncertainty in the RETRAN-3D void model and then used in the analysis of an independent separate-effect experiment. This has clearly demonstrated the basic feasibility of the approach, as well as its advantages in yielding narrower uncertainty bands in quantifying the code's accuracy for void fraction predictions

  8. Structural change of the physical economy. Decomposition analysis of physical and hybrid-unit input-output tables

    Energy Technology Data Exchange (ETDEWEB)

    Hoekstra, R.

    2003-10-01

    Economic processes generate a variety of material flows, which cause resource problems through the depletion of natural resources and environmental issues due to the emission of pollutants. This thesis presents an analytical method to study the relationship between the monetary economy and the 'physical economy'. In particular, this method can assess the impact of structural change in the economy on physical throughput. The starting point for the approach is the development of an elaborate version of the physical input-output table (PIOT), which acts as an economic-environmental accounting framework for the physical economy. In the empirical application, hybrid-unit input-output (I/O) tables, which combine physical and monetary information, are constructed for iron and steel, and plastic products for the Netherlands for the years 1990 and 1997. The impact of structural change on material flows is analyzed using Structural Decomposition Analysis (SDA), whic specifies effects such as sectoral shifts, technological change, and alterations in consumer spending and international trade patterns. The study thoroughly reviews the application of SDA to environmental issues, compares the method with other decomposition methods, and develops new mathematical specifications. An SDA is performed using the hybrid-unit input-output tables for the Netherlands. The results are subsequently used in novel forecasting and backcasting scenario analyses for the period 1997-2030. The results show that dematerialization of iron and steel, and plastics, has generally not occurred in the recent past (1990-1997), and will not occur, under a wide variety of scenario assumptions, in the future (1997-2030)

  9. Structural change of the physical economy. Decomposition analysis of physical and hybrid-unit input-output tables

    International Nuclear Information System (INIS)

    Hoekstra, R.

    2003-01-01

    Economic processes generate a variety of material flows, which cause resource problems through the depletion of natural resources and environmental issues due to the emission of pollutants. This thesis presents an analytical method to study the relationship between the monetary economy and the 'physical economy'. In particular, this method can assess the impact of structural change in the economy on physical throughput. The starting point for the approach is the development of an elaborate version of the physical input-output table (PIOT), which acts as an economic-environmental accounting framework for the physical economy. In the empirical application, hybrid-unit input-output (I/O) tables, which combine physical and monetary information, are constructed for iron and steel, and plastic products for the Netherlands for the years 1990 and 1997. The impact of structural change on material flows is analyzed using Structural Decomposition Analysis (SDA), whic specifies effects such as sectoral shifts, technological change, and alterations in consumer spending and international trade patterns. The study thoroughly reviews the application of SDA to environmental issues, compares the method with other decomposition methods, and develops new mathematical specifications. An SDA is performed using the hybrid-unit input-output tables for the Netherlands. The results are subsequently used in novel forecasting and backcasting scenario analyses for the period 1997-2030. The results show that dematerialization of iron and steel, and plastics, has generally not occurred in the recent past (1990-1997), and will not occur, under a wide variety of scenario assumptions, in the future (1997-2030)

  10. A primer on physical-layer network coding

    CERN Document Server

    Liew, Soung Chang; Zhang, Shengli

    2015-01-01

    The concept of physical-layer network coding (PNC) was proposed in 2006 for application in wireless networks. Since then it has developed into a subfield of communications and networking with a wide following. This book is a primer on PNC. It is the outcome of a set of lecture notes for a course for beginning graduate students at The Chinese University of Hong Kong. The target audience is expected to have some prior background knowledge in communication theory and wireless communications, but not working knowledge at the research level. Indeed, a goal of this book/course is to allow the reader

  11. Channel estimation for physical layer network coding systems

    CERN Document Server

    Gao, Feifei; Wang, Gongpu

    2014-01-01

    This SpringerBrief presents channel estimation strategies for the physical later network coding (PLNC) systems. Along with a review of PLNC architectures, this brief examines new challenges brought by the special structure of bi-directional two-hop transmissions that are different from the traditional point-to-point systems and unidirectional relay systems. The authors discuss the channel estimation strategies over typical fading scenarios, including frequency flat fading, frequency selective fading and time selective fading, as well as future research directions. Chapters explore the performa

  12. The cosmic code quantum physics as the language of nature

    CERN Document Server

    Pagels, Heinz R

    2012-01-01

    ""The Cosmic Code can be read by anyone. I heartily recommend it!"" - The New York Times Book Review""A reliable guide for the nonmathematical reader across the highest ridges of physical theory. Pagels is unfailingly lighthearted and confident."" - Scientific American""A sound, clear, vital work that deserves the attention of anyone who takes an interest in the relationship between material reality and the human mind."" - Science 82This is one of the most important books on quantum mechanics ever written for general readers. Heinz Pagels, an eminent physicist and science writer, discusses and

  13. The Physical Models and Statistical Procedures Used in the RACER Monte Carlo Code

    International Nuclear Information System (INIS)

    Sutton, T.M.; Brown, F.B.; Bischoff, F.G.; MacMillan, D.B.; Ellis, C.L.; Ward, J.T.; Ballinger, C.T.; Kelly, D.J.; Schindler, L.

    1999-01-01

    This report describes the MCV (Monte Carlo - Vectorized)Monte Carlo neutron transport code [Brown, 1982, 1983; Brown and Mendelson, 1984a]. MCV is a module in the RACER system of codes that is used for Monte Carlo reactor physics analysis. The MCV module contains all of the neutron transport and statistical analysis functions of the system, while other modules perform various input-related functions such as geometry description, material assignment, output edit specification, etc. MCV is very closely related to the 05R neutron Monte Carlo code [Irving et al., 1965] developed at Oak Ridge National Laboratory. 05R evolved into the 05RR module of the STEMB system, which was the forerunner of the RACER system. Much of the overall logic and physics treatment of 05RR has been retained and, indeed, the original verification of MCV was achieved through comparison with STEMB results. MCV has been designed to be very computationally efficient [Brown, 1981, Brown and Martin, 1984b; Brown, 1986]. It was originally programmed to make use of vector-computing architectures such as those of the CDC Cyber- 205 and Cray X-MP. MCV was the first full-scale production Monte Carlo code to effectively utilize vector-processing capabilities. Subsequently, MCV was modified to utilize both distributed-memory [Sutton and Brown, 1994] and shared memory parallelism. The code has been compiled and run on platforms ranging from 32-bit UNIX workstations to clusters of 64-bit vector-parallel supercomputers. The computational efficiency of the code allows the analyst to perform calculations using many more neutron histories than is practical with most other Monte Carlo codes, thereby yielding results with smaller statistical uncertainties. MCV also utilizes variance reduction techniques such as survival biasing, splitting, and rouletting to permit additional reduction in uncertainties. While a general-purpose neutron Monte Carlo code, MCV is optimized for reactor physics calculations. It has the

  14. 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

  15. Nucleosynthesis in neutrino-driven winds: Influence of the nuclear physics input

    International Nuclear Information System (INIS)

    Arcones, Almudena; Martinez-Pinedo, Gabriel

    2010-01-01

    We have performed hydrodynamical simulations of the long-time evolution of proto-neutron stars to study the nucleosynthesis using the resulting wind trajectories. Although the conditions found in the present wind models are not favourable for the production of heavy elements, a small enhancement of the entropy results in the production of r-process elements with A ∼ 195. This allows us to explore the sensitivity of their production to the hydrodynamical evolution (wind termination shock) and nuclear physics input used.

  16. TART input manual

    International Nuclear Information System (INIS)

    Kimlinger, J.R.; Plechaty, E.F.

    1982-01-01

    The TART code is a Monte Carlo neutron/photon transport code that is only on the CRAY computer. All the input cards for the TART code are listed, and definitions for all input parameters are given. The execution and limitations of the code are described, and input for two sample problems are given

  17. “PROCESS”: A systems code for fusion power plants—Part 1: Physics

    Energy Technology Data Exchange (ETDEWEB)

    Kovari, M., E-mail: michael.kovari@ccfe.ac.uk; Kemp, R.; Lux, H.; Knight, P.; Morris, J.; Ward, D.J.

    2014-12-15

    Highlights: • PROCESS is a fusion reactor systems code. • It optimises a figure of merit subject to constraints chosen by the user. • CCFE are working to make the assumptions and equations explicit and public. • The PROCESS homepage is (www.ccfe.ac.uk/powerplants.aspx). - Abstract: PROCESS is a reactor systems code – it assesses the engineering and economic viability of a hypothetical fusion power station using simple models of all parts of a reactor system, from the basic plasma physics to the generation of electricity. It has been used for many years, but details of its operation have not been previously published. This paper describes some of its capabilities. PROCESS is usually used in optimisation mode, in which it finds a set of parameters that maximise (or minimise) a figure of merit chosen by the user, while being consistent with the inputs and the specified constraints. Because the user can apply all the physically relevant constraints, while allowing a large number of parameters to vary, it is in principle only necessary to run the code once to produce a self-consistent, physically plausible reactor model. The scope of PROCESS is very wide and goes well beyond reactor physics, including conversion of heat to electricity, buildings, and costs, but this paper describes only the plasma physics and magnetic field calculations. The capabilities of PROCESS in plasma physics are limited, as its main aim is to combine engineering, physics and economics. A model is described which shows the main plasma features of an inductive ITER scenario. Significant differences between the PROCESS results and the published scenario include the bootstrap current and loop voltage. The PROCESS models for these are being revised. Two new models for DEMO have been obtained. The first, DEMO A, is intended to be “conservative” in that it might be possible to build it using the technology of the near future. For example, since current drive technologies are not yet

  18. Role of physical bolus properties as sensory inputs in the trigger of swallowing.

    Science.gov (United States)

    Peyron, Marie-Agnès; Gierczynski, Isabelle; Hartmann, Christoph; Loret, Chrystel; Dardevet, Dominique; Martin, Nathalie; Woda, Alain

    2011-01-01

    Swallowing is triggered when a food bolus being prepared by mastication has reached a defined state. However, although this view is consensual and well supported, the physical properties of the swallowable bolus have been under-researched. We tested the hypothesis that measuring bolus physical changes during the masticatory sequence to deglutition would reveal the bolus properties potentially involved in swallowing initiation. Twenty normo-dentate young adults were instructed to chew portions of cereal and spit out the boluses at different times in the masticatory sequence. The mechanical properties of the collected boluses were measured by a texture profile analysis test currently used in food science. The median particle size of the boluses was evaluated by sieving. In a simultaneous sensory study, twenty-five other subjects expressed their perception of bolus texture dominating at any mastication time. Several physical changes appeared in the food bolus as it was formed during mastication: (1) in rheological terms, bolus hardness rapidly decreased as the masticatory sequence progressed, (2) by contrast, adhesiveness, springiness and cohesiveness regularly increased until the time of swallowing, (3) median particle size, indicating the bolus particle size distribution, decreased mostly during the first third of the masticatory sequence, (4) except for hardness, the rheological changes still appeared in the boluses collected just before swallowing, and (5) physical changes occurred, with sensory stickiness being described by the subjects as a dominant perception of the bolus at the end of mastication. Although these physical and sensory changes progressed in the course of mastication, those observed just before swallowing seem to be involved in swallowing initiation. They can be considered as strong candidates for sensory inputs from the bolus that are probably crucially involved in the triggering of swallowing, since they appeared in boluses prepared in various

  19. Role of physical bolus properties as sensory inputs in the trigger of swallowing.

    Directory of Open Access Journals (Sweden)

    Marie-Agnès Peyron

    Full Text Available BACKGROUND: Swallowing is triggered when a food bolus being prepared by mastication has reached a defined state. However, although this view is consensual and well supported, the physical properties of the swallowable bolus have been under-researched. We tested the hypothesis that measuring bolus physical changes during the masticatory sequence to deglutition would reveal the bolus properties potentially involved in swallowing initiation. METHODS: Twenty normo-dentate young adults were instructed to chew portions of cereal and spit out the boluses at different times in the masticatory sequence. The mechanical properties of the collected boluses were measured by a texture profile analysis test currently used in food science. The median particle size of the boluses was evaluated by sieving. In a simultaneous sensory study, twenty-five other subjects expressed their perception of bolus texture dominating at any mastication time. FINDINGS: Several physical changes appeared in the food bolus as it was formed during mastication: (1 in rheological terms, bolus hardness rapidly decreased as the masticatory sequence progressed, (2 by contrast, adhesiveness, springiness and cohesiveness regularly increased until the time of swallowing, (3 median particle size, indicating the bolus particle size distribution, decreased mostly during the first third of the masticatory sequence, (4 except for hardness, the rheological changes still appeared in the boluses collected just before swallowing, and (5 physical changes occurred, with sensory stickiness being described by the subjects as a dominant perception of the bolus at the end of mastication. CONCLUSIONS: Although these physical and sensory changes progressed in the course of mastication, those observed just before swallowing seem to be involved in swallowing initiation. They can be considered as strong candidates for sensory inputs from the bolus that are probably crucially involved in the triggering of

  20. Physical model of the nuclear fuel cycle simulation code SITON

    International Nuclear Information System (INIS)

    Brolly, Á.; Halász, M.; Szieberth, M.; Nagy, L.; Fehér, S.

    2017-01-01

    Finding answers to main challenges of nuclear energy, like resource utilisation or waste minimisation, calls for transient fuel cycle modelling. This motivation led to the development of SITON v2.0 a dynamic, discrete facilities/discrete materials and also discrete events fuel cycle simulation code. The physical model of the code includes the most important fuel cycle facilities. Facilities can be connected flexibly; their number is not limited. Material transfer between facilities is tracked by taking into account 52 nuclides. Composition of discharged fuel is determined using burnup tables except for the 2400 MW thermal power design of the Gas-Cooled Fast Reactor (GFR2400). For the GFR2400 the FITXS method is used, which fits one-group microscopic cross-sections as polynomial functions of the fuel composition. This method is accurate and fast enough to be used in fuel cycle simulations. Operation of the fuel cycle, i.e. material requests and transfers, is described by discrete events. In advance of the simulation reactors and plants formulate their requests as events; triggered requests are tracked. After that, the events are simulated, i.e. the requests are fulfilled and composition of the material flow between facilities is calculated. To demonstrate capabilities of SITON v2.0, a hypothetical transient fuel cycle is presented in which a 4-unit VVER-440 reactor park was replaced by one GFR2400 that recycled its own spent fuel. It is found that the GFR2400 can be started if the cooling time of its spent fuel is 2 years. However, if the cooling time is 5 years it needs an additional plutonium feed, which can be covered from the spent fuel of a Generation III light water reactor.

  1. The Los Alamos suite of relativistic atomic physics codes

    International Nuclear Information System (INIS)

    Fontes, C J; Zhang, H L; Jr, J Abdallah; Clark, R E H; Kilcrease, D P; Colgan, J; Cunningham, R T; Hakel, P; Magee, N H; Sherrill, M E

    2015-01-01

    The Los Alamos suite of relativistic atomic physics codes is a robust, mature platform that has been used to model highly charged ions in a variety of ways. The suite includes capabilities for calculating data related to fundamental atomic structure, as well as the processes of photoexcitation, electron-impact excitation and ionization, photoionization and autoionization within a consistent framework. These data can be of a basic nature, such as cross sections and collision strengths, which are useful in making predictions that can be compared with experiments to test fundamental theories of highly charged ions, such as quantum electrodynamics. The suite can also be used to generate detailed models of energy levels and rate coefficients, and to apply them in the collisional-radiative modeling of plasmas over a wide range of conditions. Such modeling is useful, for example, in the interpretation of spectra generated by a variety of plasmas. In this work, we provide a brief overview of the capabilities within the Los Alamos relativistic suite along with some examples of its application to the modeling of highly charged ions. (paper)

  2. Resonance interference method in lattice physics code stream

    International Nuclear Information System (INIS)

    Choi, Sooyoung; Khassenov, Azamat; Lee, Deokjung

    2015-01-01

    Newly developed resonance interference model is implemented in the lattice physics code STREAM, and the model shows a significant improvement in computing accurate eigenvalues. Equivalence theory is widely used in production calculations to generate the effective multigroup (MG) cross-sections (XS) for commercial reactors. Although a lot of methods have been developed to enhance the accuracy in computing effective XSs, the current resonance treatment methods still do not have a clear resonance interference model. The conventional resonance interference model simply adds the absorption XSs of resonance isotopes to the background XS. However, the conventional models show non-negligible errors in computing effective XSs and eigenvalues. In this paper, a resonance interference factor (RIF) library method is proposed. This method interpolates the RIFs in a pre-generated RIF library and corrects the effective XS, rather than solving the time consuming slowing down calculation. The RIF library method is verified for homogeneous and heterogeneous problems. The verification results using the proposed method show significant improvements of accuracy in treating the interference effect. (author)

  3. 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

  4. Development of Integrated Code for Risk Assessment (INCORIA) for Physical Protection System

    International Nuclear Information System (INIS)

    Jang, Sung Soon; Seo, Hyung Min; Yoo, Ho Sik

    2010-01-01

    A physical protection system (PPS) integrates people, procedures and equipment for the protection of assets or facilities against theft, sabotage or other malevolent human attacks. Among critical facilities, nuclear facilities and nuclear weapon sites require the highest level of PPS. After the September 11, 2001 terrorist attacks, international communities, including the IAEA, have made substantial efforts to protect nuclear material and nuclear facilities. The international flow on nuclear security is using the concept or risk assessment. The concept of risk assessment is firstly devised by nuclear safety people. They considered nuclear safety including its possible risk, which is the frequency of failure and possible consequence. Nuclear security people also considers security risk, which is the frequency of threat action, vulnerability, and consequences. The concept means that we should protect more when the credible threat exists and the possible radiological consequence is high. Even if there are several risk assessment methods of nuclear security, the application needs the help of tools because of a lot of calculation. It's also hard to find tools for whole phase of risk assessment. Several codes exist for the part of risk assessment. SAVI are used for vulnerability of PPS. Vital area identification code is used for consequence analysis. We are developing Integrated Code for Risk Assessment (INCORIA) to apply risk assessment methods for nuclear facilities. INCORIA evaluates PP-KINAC measures and generation tools for threat scenario. PP-KINAC is risk assessment measures for physical protection system developed by Hosik Yoo and is easy to apply. A threat scenario tool is used to generate threat scenario, which is used as one of input value to PP-KINAC measures

  5. GALEV evolutionary synthesis models – I. Code, input physics and web

    NARCIS (Netherlands)

    Kotulla, R.; Fritze, U.; Weilbacher, P.; Anders, P.

    2009-01-01

    GALEV (GALaxy EVolution) evolutionary synthesis models describe the evolution of stellar populations in general, of star clusters as well as of galaxies, both in terms of resolved stellar populations and of integrated light properties over cosmological time-scales of ≥13 Gyr from the onset of star

  6. "SMART": A Compact and Handy FORTRAN Code for the Physics of Stellar Atmospheres

    Science.gov (United States)

    Sapar, A.; Poolamäe, R.

    2003-01-01

    A new computer code SMART (Spectra from Model Atmospheres by Radiative Transfer) for computing the stellar spectra, forming in plane-parallel atmospheres, has been compiled by us and A. Aret. To guarantee wide compatibility of the code with shell environment, we chose FORTRAN-77 as programming language and tried to confine ourselves to common part of its numerous versions both in WINDOWS and LINUX. SMART can be used for studies of several processes in stellar atmospheres. The current version of the programme is undergoing rapid changes due to our goal to elaborate a simple, handy and compact code. Instead of linearisation (being a mathematical method of recurrent approximations) we propose to use the physical evolutionary changes or in other words relaxation of quantum state populations rates from LTE to NLTE has been studied using small number of NLTE states. This computational scheme is essentially simpler and more compact than the linearisation. This relaxation scheme enables using instead of the Λ-iteration procedure a physically changing emissivity (or the source function) which incorporates in itself changing Menzel coefficients for NLTE quantum state populations. However, the light scattering on free electrons is in the terms of Feynman graphs a real second-order quantum process and cannot be reduced to consequent processes of absorption and emission as in the case of radiative transfer in spectral lines. With duly chosen input parameters the code SMART enables computing radiative acceleration to the matter of stellar atmosphere in turbulence clumps. This also enables to connect the model atmosphere in more detail with the problem of the stellar wind triggering. Another problem, which has been incorporated into the computer code SMART, is diffusion of chemical elements and their isotopes in the atmospheres of chemically peculiar (CP) stars due to usual radiative acceleration and the essential additional acceleration generated by the light-induced drift. As

  7. TVF-NMCRC-A powerful program for writing and executing simulation inputs for the FLUKA Monte Carlo Code system

    International Nuclear Information System (INIS)

    Mark, S.; Khomchenko, S.; Shifrin, M.; Haviv, Y.; Schwartz, J.R.; Orion, I.

    2007-01-01

    We at the Negev Monte Carlo Research Center (NMCRC) have developed a powerful new interface for writing and executing FLUKA input files-TVF-NMCRC. With the TVF tool a FLUKA user has the ability to easily write an input file without requiring any previous experience. The TVF-NMCRC tool is a LINUX program that has been verified for the most common LINUX-based operating systems, and is suitable for the latest version of FLUKA (FLUKA 2006.3)

  8. Derivation of the physical equations solved in the inertial confinement stability code DOC. Informal report

    International Nuclear Information System (INIS)

    Scannapieco, A.J.; Cranfill, C.W.

    1978-11-01

    There now exists an inertial confinement stability code called DOC, which runs as a postprocessor. DOC (a code that has evolved from a previous code, PANSY) is a spherical harmonic linear stability code that integrates, in time, a set of Lagrangian perturbation equations. Effects due to real equations of state, asymmetric energy deposition, thermal conduction, shock propagation, and a time-dependent zeroth-order state are handled in the code. We present here a detailed derivation of the physical equations that are solved in the code

  9. Derivation of the physical equations solved in the inertial confinement stability code DOC. Informal report

    Energy Technology Data Exchange (ETDEWEB)

    Scannapieco, A.J.; Cranfill, C.W.

    1978-11-01

    There now exists an inertial confinement stability code called DOC, which runs as a postprocessor. DOC (a code that has evolved from a previous code, PANSY) is a spherical harmonic linear stability code that integrates, in time, a set of Lagrangian perturbation equations. Effects due to real equations of state, asymmetric energy deposition, thermal conduction, shock propagation, and a time-dependent zeroth-order state are handled in the code. We present here a detailed derivation of the physical equations that are solved in the code.

  10. Development of Teaching Materials for a Physical Chemistry Experiment Using the QR Code

    OpenAIRE

    吉村, 忠与志

    2008-01-01

    The development of teaching materials with the QR code was attempted in an educational environment using a mobile telephone. The QR code is not sufficiently utilized in education, and the current study is one of the first in the field. The QR code is encrypted. However, the QR code can be deciphered by mobile telephones, thus enabling the expression of text in a small space.Contents of "Physical Chemistry Experiment" which are available on the Internet are briefly summarized and simplified. T...

  11. Calculation codes in radiation protection, radiation physics and dosimetry

    International Nuclear Information System (INIS)

    2003-01-01

    These scientific days had for objective to draw up the situation of calculation codes of radiation transport, of sources estimation, of radiation doses managements and to draw the future perspectives. (N.C.)

  12. Recent progress of an integrated implosion code and modeling of element physics

    International Nuclear Information System (INIS)

    Nagatomo, H.; Takabe, H.; Mima, K.; Ohnishi, N.; Sunahara, A.; Takeda, T.; Nishihara, K.; Nishiguchu, A.; Sawada, K.

    2001-01-01

    Physics of the inertial fusion is based on a variety of elements such as compressible hydrodynamics, radiation transport, non-ideal equation of state, non-LTE atomic process, and relativistic laser plasma interaction. In addition, implosion process is not in stationary state and fluid dynamics, energy transport and instabilities should be solved simultaneously. In order to study such complex physics, an integrated implosion code including all physics important in the implosion process should be developed. The details of physics elements should be studied and the resultant numerical modeling should be installed in the integrated code so that the implosion can be simulated with available computer within realistic CPU time. Therefore, this task can be basically separated into two parts. One is to integrate all physics elements into a code, which is strongly related to the development of hydrodynamic equation solver. We have developed 2-D integrated implosion code which solves mass, momentum, electron energy, ion energy, equation of states, laser ray-trace, laser absorption radiation, surface tracing and so on. The reasonable results in simulating Rayleigh-Taylor instability and cylindrical implosion are obtained using this code. The other is code development on each element physics and verification of these codes. We had progress in developing a nonlocal electron transport code and 2 and 3 dimension radiation hydrodynamic code. (author)

  13. Recent Improvement of Measurement Instrumentation to Supervise Nuclear Operations and to Contribute Input Data to 3D Simulation Code - 13289

    Energy Technology Data Exchange (ETDEWEB)

    Mahe, Charly; Chabal, Caroline [CEA, Nuclear Energy Division, Fuel Technology Development Unit, Simulation and Dismantling Technique Laboratory, Marcoule Center, BP 17171, 30207 Bagnols / Ceze (France)

    2013-07-01

    used to quantify the activity of hot spots and can also then be entered in 3D models of nuclear plants to simulate intervention scenarios. Recent developments and results will be presented regarding this. Finally, thanks to a large amount of feedback, the interest of using complementary measurements will be discussed. In fact, the recent use of 3D simulation codes requires very accurate knowledge of nuclear plant radiological data. The use of coupled devices such as imaging devices, (gamma and alpha cameras), gamma spectrometry, dose rate mapping, collimated / un-collimated measurements and many other physical values gives an approach to the radiological knowledge of a process or plant with the lowest possible uncertainty. In line with this, the paper will conclude with the future developments and trials that could be assessed in that field of application. (authors)

  14. Asteroseismic modelling of solar-type stars: internal systematics from input physics and surface correction methods

    Science.gov (United States)

    Nsamba, B.; Campante, T. L.; Monteiro, M. J. P. F. G.; Cunha, M. S.; Rendle, B. M.; Reese, D. R.; Verma, K.

    2018-04-01

    Asteroseismic forward modelling techniques are being used to determine fundamental properties (e.g. mass, radius, and age) of solar-type stars. The need to take into account all possible sources of error is of paramount importance towards a robust determination of stellar properties. We present a study of 34 solar-type stars for which high signal-to-noise asteroseismic data is available from multi-year Kepler photometry. We explore the internal systematics on the stellar properties, that is, associated with the uncertainty in the input physics used to construct the stellar models. In particular, we explore the systematics arising from: (i) the inclusion of the diffusion of helium and heavy elements; and (ii) the uncertainty in solar metallicity mixture. We also assess the systematics arising from (iii) different surface correction methods used in optimisation/fitting procedures. The systematics arising from comparing results of models with and without diffusion are found to be 0.5%, 0.8%, 2.1%, and 16% in mean density, radius, mass, and age, respectively. The internal systematics in age are significantly larger than the statistical uncertainties. We find the internal systematics resulting from the uncertainty in solar metallicity mixture to be 0.7% in mean density, 0.5% in radius, 1.4% in mass, and 6.7% in age. The surface correction method by Sonoi et al. and Ball & Gizon's two-term correction produce the lowest internal systematics among the different correction methods, namely, ˜1%, ˜1%, ˜2%, and ˜8% in mean density, radius, mass, and age, respectively. Stellar masses obtained using the surface correction methods by Kjeldsen et al. and Ball & Gizon's one-term correction are systematically higher than those obtained using frequency ratios.

  15. On the average complexity of sphere decoding in lattice space-time coded multiple-input multiple-output channel

    KAUST Repository

    Abediseid, Walid

    2012-01-01

    complexity of sphere decoding for the quasi- static, lattice space-time (LAST) coded MIMO channel. Specifically, we drive an upper bound of the tail distribution of the decoder's computational complexity. We show that when the computational complexity exceeds

  16. A theory manual for multi-physics code coupling in LIME.

    Energy Technology Data Exchange (ETDEWEB)

    Belcourt, Noel; Bartlett, Roscoe Ainsworth; Pawlowski, Roger Patrick; Schmidt, Rodney Cannon; Hooper, Russell Warren

    2011-03-01

    The Lightweight Integrating Multi-physics Environment (LIME) is a software package for creating multi-physics simulation codes. Its primary application space is when computer codes are currently available to solve different parts of a multi-physics problem and now need to be coupled with other such codes. In this report we define a common domain language for discussing multi-physics coupling and describe the basic theory associated with multiphysics coupling algorithms that are to be supported in LIME. We provide an assessment of coupling techniques for both steady-state and time dependent coupled systems. Example couplings are also demonstrated.

  17. FERRET data analysis code

    International Nuclear Information System (INIS)

    Schmittroth, F.

    1979-09-01

    A documentation of the FERRET data analysis code is given. The code provides a way to combine related measurements and calculations in a consistent evaluation. Basically a very general least-squares code, it is oriented towards problems frequently encountered in nuclear data and reactor physics. A strong emphasis is on the proper treatment of uncertainties and correlations and in providing quantitative uncertainty estimates. Documentation includes a review of the method, structure of the code, input formats, and examples

  18. Physical-mathematical model for cybernetic description of the human organs with trace element concentrations as input variables

    International Nuclear Information System (INIS)

    Mihai, Maria; Popescu, I.V.

    2003-01-01

    In this paper we report a physical-mathematical model for studying the organs and humans fluids by cybernetic principle. The input variables represent the trace elements which are determined by atomic and nuclear methods of elemental analysis. We have determined the health limits between which the organs might function. (authors)

  19. SPOTS4. Group data library and computer code, preparing ENDF/B-4 data for input to LEOPARD

    International Nuclear Information System (INIS)

    Kim, J.D.; Lee, J.T.

    1981-09-01

    The magnetic tape SPOTS4 contains in file 1 a data library to be used as input to the SPOTS4 program which is contained in file 2. The data library is based on ENDF/B-4 and consists of two parts in TEMPEST format (246 groups) and MUFT format (54 groups) respectively. From this library the SPOTS4 program produces a 172 + 54 group library for LEOPARD input. A copy of the magnetic tape is available from the IAEA Nuclear Data Section. (author)

  20. HTR core physics and transient analyses by the Panthermix code system

    Energy Technology Data Exchange (ETDEWEB)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

    2005-07-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes.

  1. HTR core physics and transient analyses by the Panthermix code system

    International Nuclear Information System (INIS)

    Haas, J.B.M. de; Kuijper, J.C.; Oppe, J.

    2005-01-01

    At NRG Petten, core physics analyses on High Temperature gas-cooled Reactors (HTRs) are mainly performed by means of the PANTHERMIX code system. Since some years NRG is developing the HTR reactor physics code system WIMS/PANTHERMIX, based on the lattice code WIMS (Serco Assurance, UK), the 3-dimensional steady-state and transient core physics code PANTHER (British Energy, UK) and the 2-dimensional R-Z HTR thermal hydraulics code THERMIX-DIREKT (Research Centre FZJ Juelich, Germany). By means of the WIMS code nuclear data are being generated to suit the PANTHER code's neutronics. At NRG the PANTHER code has been interfaced with THERMIX-DIREKT to form PANTHERMIX, to enable core-follow/fuel management and transient analyses in a consistent manner on pebble bed type HTR systems. Also provisions have been made to simulate the flow of pebbles through the core of a pebble bed HTR, according to a given (R-Z) flow pattern. As examples of the versatility of the PANTHERMIX code system, calculations are presented on the PBMR, the South African pebble bed reactor design, to show the transient capabilities, and on a plutonium burning MEDUL-reactor, to demonstrate the core-follow/fuel management capabilities. For the investigated cases a good agreement is observed with the results of other HTR core physics codes

  2. Input data preparation and simulation of the second standard problem of IAEA using the Trac/PF1 code

    International Nuclear Information System (INIS)

    Madeira, A.A.; Pontedeiro, A.C.; Silva Galetti, M.R. da; Borges, R.C.

    1989-10-01

    The second Standard Problem sponsored by IAEA consists in the simulation of a small LOCA located in the downcomer of a PMK-NVH integral test facility, which models WWER/440 type reactor. This report presents input data preparation and comparison between TRAC-PF1 results and experimental measurements. (author) [pt

  3. Visual Input Enhancement via Essay Coding Results in Deaf Learners' Long-Term Retention of Improved English Grammatical Knowledge

    Science.gov (United States)

    Berent, Gerald P.; Kelly, Ronald R.; Schmitz, Kathryn L.; Kenney, Patricia

    2009-01-01

    This study explored the efficacy of visual input enhancement, specifically "essay enhancement", for facilitating deaf college students' improvement in English grammatical knowledge. Results documented students' significant improvement immediately after a 10-week instructional intervention, a replication of recent research. Additionally, the…

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

    CERN Multimedia

    2005-01-01

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

  5. 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)

  6. 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)

  7. Check and visualization of input geometry data using the geometrical module of the Monte Carlo code MCU: WWER-440 pressure vessel dosimetry benchmarks

    International Nuclear Information System (INIS)

    Gurevich, M.; Zaritsky, S.; Osmera, B.; Mikus, J.

    1997-01-01

    The Monte Carlo method gives the opportunity to conduct the calculations of neutron and photon flux without any simplifications of the 3-D geometry of the nuclear power and experimental devices. So, each graduated Monte Carlo code includes the combinatorial geometry module and tools for the geometry description giving a possibility to describe very complex systems with a number of hierarchy levels of the geometrical objects. Such codes as usual have special modules for the visual checking of geometry input information. These geometry opportunities could be used for all cases when the accurate 3-D description of the complex geometry becomes a necessity. The description (specification) of benchmark experiments is one of the such cases. Such accurate and uniform description detects all mistakes and ambiguities in the starting information of various kinds (drawings, reports etc.). Usually the quality of different parts of the starting information (generally produced by different persons during the different stages of the device elaboration and operation) is different. After using the above mentioned modules and tools, the resultant geometry description can be used as a standard for this device. One can automatically produce any type of the device figure. The detail geometry description can be used as input for different calculation models carrying out (not only for Monte Carlo). The application of that method to the description of the WWER-440 mock-ups is represented in the report. The mock-ups were created on the reactor LR-O (NRI) and the reactor vessel dosimetry benchmarks were developed on the basis of these mock-up experiments. The NCG-8 module of the Russian Monte Carlo code MCU was used. It is the combinatorial multilingual universal geometrical module. The MCU code was certified by Russian Nuclear Regulatory Body. Almost all figures for mentioned benchmarks specifications were made by the MCU visualization code. The problem of the automatic generation of the

  8. 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...

  9. SEAPATH: A microcomputer code for evaluating physical security effectiveness using adversary sequence diagrams

    International Nuclear Information System (INIS)

    Darby, J.L.

    1986-01-01

    The Adversary Sequence Diagram (ASD) concept was developed by Sandia National Laboratories (SNL) to examine physical security system effectiveness. Sandia also developed a mainframe computer code, PANL, to analyze the ASD. The authors have developed a microcomputer code, SEAPATH, which also analyzes ASD's. The Authors are supporting SNL in software development of the SAVI code; SAVI utilizes the SEAPATH algorithm to identify and quantify paths

  10. Standard interface files and procedures for reactor physics codes. Version IV

    International Nuclear Information System (INIS)

    O'Dell, R.D.

    1977-09-01

    Standards, procedures, and recommendations of the Committee on Computer Code Coordination for promoting the exchange of reactor physics codes are updated to Version IV status. Standards and procedures covering general programming, program structure, standard interface files, and file management and handling subroutines are included

  11. CORESAFE: A Formal Approach against Code Replacement Attacks on Cyber Physical Systems

    Science.gov (United States)

    2018-04-19

    AFRL-AFOSR-JP-TR-2018-0035 CORESAFE:A Formal Approach against Code Replacement Attacks on Cyber Physical Systems Sandeep Shukla INDIAN INSTITUTE OF...Formal Approach against Code Replacement Attacks on Cyber Physical Systems 5a.  CONTRACT NUMBER 5b.  GRANT NUMBER FA2386-16-1-4099 5c.  PROGRAM ELEMENT...SUPPLEMENTARY NOTES 14.  ABSTRACT Industrial Control Systems (ICS) used in manufacturing, power generators and other critical infrastructure monitoring and

  12. The MCUCN simulation code for ultracold neutron physics

    Science.gov (United States)

    Zsigmond, G.

    2018-02-01

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

  13. Calculation codes in radioprotection, radio-physics and dosimetry

    International Nuclear Information System (INIS)

    Jan, S.; Laedermann, J.P.; Bochud, F.; Ferragut, A.; Bordy, J.M.; Parisi, L.L.; Abou-Khalil, R.; Longeot, M.; Kitsos, S.; Groetz, J.E.; Villagrasa, C.; Daures, J.; Martin, E.; Henriet, J.; Tsilanizara, A.; Farah, J.; Uyttenhove, W.; Perrot, Y.; De Carlan, L.; Vivier, A.; Kodeli, I.; Sayah, R.; Hadid, L.; Courageot, E.; Fritsch, P.; Davesne, E.; Michel, X.

    2010-01-01

    This document gathers the slides of the available presentations given during these conference days. Twenty seven presentations are assembled in the document and deal with: 1 - GATE: calculation code for medical imaging, radiotherapy and dosimetry (S. Jan); 2 - estimation of conversion factors for the measurement of the ambient dose equivalent rate by in-situ spectroscopy (J.P. Laedermann); 3 - geometry specific calibration factors for nuclear medicine activity meters (F. Bochud); 4 - Monte Carlo simulation of a rare gases measurement system - calculation and validation, ASGA/VGM system (A. Ferragut); 5 - design of a realistic radiation field for the calibration of the dosemeters used in interventional radiology/cardiology (medical personnel dosimetry) (J.M. Bordy); 6 - determination of the position and height of the KALINA facility chimney at CEA Cadarache (L.L. Parisi); 7 - MERCURAD TM - 3D simulation software for dose rates calculation (R. Abou-Khalil); 8 - PANTHERE - 3D software for gamma dose rates simulation of complex nuclear facilities (M. Longeot); 9 - radioprotection, from the design to the exploitation of radioactive materials transportation containers (S. Kitsos); 10 - post-simulation processing of MCNPX responses in neutron spectroscopy (J.E. Groetz); 11 - last developments of the Geant4 Monte Carlo code for trace amounts simulation in liquid water at the molecular scale (C. Villagrasa); 12 - Calculation of H p (3)/K air conversion coefficients using PENELOPE Monte-Carlo code and comparison with MCNP calculation results (J. Daures); 13 - artificial neural networks, a new alternative to Monte Carlo calculations for radiotherapy (E. Martin); 14 - use of case-based reasoning for the reconstruction and handling of voxelized fantoms (J. Henriet); 15 - resolution of the radioactive decay inverse problem for dose calculation in radioprotection (A. Tsilanizara); 16 - use of NURBS-type fantoms for the study of the morphological factors influencing the pulmonary

  14. Running codes through the web

    International Nuclear Information System (INIS)

    Clark, R.E.H.

    2001-01-01

    Dr. Clark presented a report and demonstration of running atomic physics codes through the WWW. The atomic physics data is generated from Los Alamos National Laboratory (LANL) codes that calculate electron impact excitation, ionization, photoionization, and autoionization, and inversed processes through detailed balance. Samples of Web interfaces, input and output are given in the report

  15. Status of computer codes available in AEOI for reactor physics analysis

    International Nuclear Information System (INIS)

    Karbassiafshar, M.

    1986-01-01

    Many of the nuclear computer codes available in Atomic Energy Organization of Iran AEOI can be used for physics analysis of an operating reactor or design purposes. Grasp of the various methods involved and practical experience with these codes would be the starting point for interesting design studies or analysis of operating conditions of presently existing and future reactors. A review of the objectives and flowchart of commonly practiced procedures in reactor physics analysis of LWRs and related computer codes was made, extrapolating to the nationally and internationally available resources. Finally, effective utilization of the existing facilities is discussed and called upon

  16. Optimization and parallelization of the thermal–hydraulic subchannel code CTF for high-fidelity multi-physics applications

    International Nuclear Information System (INIS)

    Salko, Robert K.; Schmidt, Rodney C.; Avramova, Maria N.

    2015-01-01

    Highlights: • COBRA-TF was adopted by the Consortium for Advanced Simulation of LWRs. • We have improved code performance to support running large-scale LWR simulations. • Code optimization has led to reductions in execution time and memory usage. • An MPI parallelization has reduced full-core simulation time from days to minutes. - Abstract: This paper describes major improvements to the computational infrastructure of the CTF subchannel code so that full-core, pincell-resolved (i.e., one computational subchannel per real bundle flow channel) simulations can now be performed in much shorter run-times, either in stand-alone mode or as part of coupled-code multi-physics calculations. These improvements support the goals of the Department Of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL) Energy Innovation Hub to develop high fidelity multi-physics simulation tools for nuclear energy design and analysis. A set of serial code optimizations—including fixing computational inefficiencies, optimizing the numerical approach, and making smarter data storage choices—are first described and shown to reduce both execution time and memory usage by about a factor of ten. Next, a “single program multiple data” parallelization strategy targeting distributed memory “multiple instruction multiple data” platforms utilizing domain decomposition is presented. In this approach, data communication between processors is accomplished by inserting standard Message-Passing Interface (MPI) calls at strategic points in the code. The domain decomposition approach implemented assigns one MPI process to each fuel assembly, with each domain being represented by its own CTF input file. The creation of CTF input files, both for serial and parallel runs, is also fully automated through use of a pressurized water reactor (PWR) pre-processor utility that uses a greatly simplified set of user input compared with the traditional CTF input. To run CTF in

  17. Physical model and calculation code for fuel coolant interactions

    International Nuclear Information System (INIS)

    Goldammer, H.; Kottowski, H.

    1976-01-01

    A physical model is proposed to describe fuel coolant interactions in shock-tube geometry. According to the experimental results, an interaction model which divides each cycle into three phases is proposed. The first phase is the fuel-coolant-contact, the second one is the ejection and recently of the coolant, and the third phase is the impact and fragmentation. Physical background of these phases are illustrated in the first part of this paper. Mathematical expressions of the model are exposed in the second part. A principal feature of the computational method is the consistent application of the fourier-equation throughout the whole interaction process. The results of some calculations, performed for different conditions are compiled in attached figures. (Aoki, K.)

  18. Third order TRANSPORT with MAD [Methodical Accelerator Design] input

    International Nuclear Information System (INIS)

    Carey, D.C.

    1988-01-01

    This paper describes computer-aided design codes for particle accelerators. Among the topics discussed are: input beam description; parameters and algebraic expressions; the physical elements; beam lines; operations; and third-order transfer matrix

  19. FADDEEV: A fortran code for the calculation of the frequency response matrix of multiple-input, multiple-output dynamic systems

    International Nuclear Information System (INIS)

    Owens, D.H.

    1972-06-01

    The KDF9/EGDON programme FADDEEV has been written to investigate a technique for the calculation of the matrix of frequency responses G(jw) describing the response of the output vector y from the multivariable differential/algebraic system S to the drive of the system input vector u. S: Ex = Ax + Bu, y = Cx, G(jw) = C(jw E - A ) -1 B. The programme uses an algorithm due to Faddeev and has been written with emphasis upon: (a) simplicity of programme structure and computational technique which should enable a user to find his way through the programme fairly easily, and hence facilitate its manipulation as a subroutine in a larger code; (b) rapid computational ability, particularly in systems with fairly large number of inputs and outputs and requiring the evaluation of the frequency responses at a large number of frequencies. Transport or time delays must be converted by the user to Pade or Bode approximations prior to input. Conditions under which the algorithm fails to give accurate results are identified, and methods for increasing the accuracy of the calculations are discussed. The conditions for accurate results using FADDEEV indicate that its application is specialized. (author)

  20. The physical closure laws in the CATHARE code

    International Nuclear Information System (INIS)

    Bestion, D.

    1990-01-01

    CATHARE is a 2-fluid thermal-hydraulic code capable of simulating thermal and mechanical phenomena occurring in the primary and secondary circuits of PWRs for a wide variety of accidental situations. The description of the flow is essentially 1-dimensional. Closure laws concerning mass, momentum and energy exchanges between phases and between each phase and the walls are required. A set of specifically designed separate effect experiments were performed and analysed. Having regard for some development principles, correlations are established on the basis of experimental data. The mechanical transfer laws are derived first from experiments where thermal non equilibrium is negligible. Using them as a basis for further interpretation of experimental data, interfacial heat transfer laws are then developed. Wall heat transfer correlations then have to be fixed. All these steps are presented with emphasis being placed on the most recent developments. These last investigations concern the direct contact condensation, stratification model, wall friction, droplet break up and the scale effect, geometrical effect and pressure effect on interfacial friction. (orig.)

  1. Quo vadis code optimization in high energy physics

    International Nuclear Information System (INIS)

    Jarp, S.

    1994-01-01

    Although performance tuning and optimization can be considered less critical than in the past, there are still many High Energy Physics (HEP) applications and application domains that can profit from such an undertaking. In CERN's CORE (Centrally Operated RISC Environment) where all major RISC vendors are present, this implies an understanding of the various computer architectures, instruction sets and performance analysis tools from each of these vendors. This paper discusses some initial observations after having evaluated the situation and makes some recommendations for further progress

  2. ALE3D: An Arbitrary Lagrangian-Eulerian Multi-Physics Code

    Energy Technology Data Exchange (ETDEWEB)

    Noble, Charles R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anderson, Andrew T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Barton, Nathan R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bramwell, Jamie A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Capps, Arlie [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chang, Michael H. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chou, Jin J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dawson, David M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Diana, Emily R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dunn, Timothy A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Faux, Douglas R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fisher, Aaron C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Greene, Patrick T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Heinz, Ines [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kanarska, Yuliya [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Khairallah, Saad A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Liu, Benjamin T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Margraf, Jon D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Nichols, Albert L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Nourgaliev, Robert N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Puso, Michael A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reus, James F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Robinson, Peter B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Shestakov, Alek I. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Solberg, Jerome M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Taller, Daniel [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Tsuji, Paul H. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); White, Christopher A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); White, Jeremy L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-05-23

    ALE3D is a multi-physics numerical simulation software tool utilizing arbitrary-Lagrangian- Eulerian (ALE) techniques. The code is written to address both two-dimensional (2D plane and axisymmetric) and three-dimensional (3D) physics and engineering problems using a hybrid finite element and finite volume formulation to model fluid and elastic-plastic response of materials on an unstructured grid. As shown in Figure 1, ALE3D is a single code that integrates many physical phenomena.

  3. Research on V and V strategy of reactor physics code of COSINE

    International Nuclear Information System (INIS)

    Liu Zhanquan; Chen Yixue; Yang Chao; Dang Halei

    2013-01-01

    Verification and validation (V and V) is very important for the software quality assurance. Reasonable and efficient V and V strategy can achieve twice the result with half the effort. Core and system integrated engine for design and analysis (COSINE) software package contains three reactor physics codes, the lattice code (LATC), the core simulator (CORE) and the kinetics code (KIND), which is called the reactor physics subsystem. The V and V strategy for the physics subsystem was researched based on the foundation of scientific software's V and V method. The module based verification method and the function based validation method were proposed, composing the physical subsystem V and V strategy of COSINE software package. (authors)

  4. A computer code for the prediction of mill gases and hot air distribution between burners sections as input parameters for 3D CFD furnace calculation

    International Nuclear Information System (INIS)

    Tucakovic, Dragan; Zivanovic, Titoslav; Beloshevic, Srdjan

    2006-01-01

    Current computer technology development enables application of powerful software packages that can provide a reliable insight into real operating conditions of a steam boiler in the Thermal Power Plant. Namely, an application of CFD code to the 3D analysis of combustion and heat transfer in a furnace provides temperature, velocity and concentration fields in both cross sectional and longitudinal planes of the observed furnace. In order to obtain reliable analytical results, which corresponds to real furnace conditions, it is necessary to accurately predict a distribution of mill gases and hot air between burners' sections, because these parameters are input values for the furnace 3D calculation. Regarding these tasks, the computer code for the prediction of mill gases and hot air distribution has been developed at the Department for steam boilers of the Faculty of Mechanical Engineering in Belgrade. The code is based on simultaneous calculations of material and heat balances for fan mill and air tracts. The aim of this paper is to present a methodology of performed calculations and results obtained for the steam boiler furnace of 350 MWe Thermal Power Plant equipped with eight fan mills. Key words: mill gases, hot air, aerodynamic calculation, air tract, mill tract.

  5. PREMIUM - Benchmark on the quantification of the uncertainty of the physical models in the system thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Skorek, Tomasz; Crecy, Agnes de

    2013-01-01

    PREMIUM (Post BEMUSE Reflood Models Input Uncertainty Methods) is an activity launched with the aim to push forward the methods of quantification of physical models uncertainties in thermal-hydraulic codes. It is endorsed by OECD/NEA/CSNI/WGAMA. The benchmark PREMIUM is addressed to all who applies uncertainty evaluation methods based on input uncertainties quantification and propagation. The benchmark is based on a selected case of uncertainty analysis application to the simulation of quench front propagation in an experimental test facility. Application to an experiment enables evaluation and confirmation of the quantified probability distribution functions on the basis of experimental data. The scope of the benchmark comprises a review of the existing methods, selection of potentially important uncertain input parameters, preliminary quantification of the ranges and distributions of the identified parameters, evaluation of the probability density function using experimental results of tests performed on FEBA test facility and confirmation/validation of the performed quantification on the basis of blind calculation of Reflood 2-D PERICLES experiment. (authors)

  6. Upgrade and benchmarking of the NIFS physics-engineering-cost code

    International Nuclear Information System (INIS)

    Dolan, T.J.; Yamazaki, K.

    2004-07-01

    The NIFS Physics-Engineering-Cost (PEC) code for helical and tokamak fusion reactors is upgraded by adding data from three blanket-shield designs, a new cost section based on the ARIES cost schedule, more recent unit costs, and improved algorithms for various computations. The PEC code is also benchmarked by modeling the ARIES-AT (advanced technology) tokamak and the ARIES-SPPS (stellarator power plant system). The PEC code succeeds in predicting many of the pertinent plasma parameters and reactor component masses within about 10%. There are cost differences greater than 10% for some fusion power core components, which may be attributed to differences of unit costs used by the codes. The COEs estimated by the PEC code differ from the COEs of the ARIES-AT and ARIES-SPPS studies by 5%. (author)

  7. Simplified models for new physics in vector boson scattering. Input for Snowmass 2013

    International Nuclear Information System (INIS)

    Reuter, Juergen; Kilian, Wolfgang; Sekulla, Marco

    2013-07-01

    In this contribution to the Snowmass process 2013 we give a brief review of how new physics could enter in the electroweak (EW) sector of the Standard Model (SM). This new physics, if it is directly accessible at low energies, can be parameterized by explicit resonances having certain quantum numbers. The extreme case is the decoupling limit where those resonances are very heavy and leave only traces in the form of deviations in the SM couplings. Translations are given into higher-dimensional operators leading to such deviations. As long as such resonances are introduced without a UV-complete theory behind it, these models suffer from unitarity violation of perturbative scattering amplitudes. We show explicitly how theoretically sane descriptions could be achieved by using a unitarization prescription that allows a correct description of such a resonance without specifying a UV-complete model.

  8. Simplified models for new physics in vector boson scattering. Input for Snowmass 2013

    Energy Technology Data Exchange (ETDEWEB)

    Reuter, Juergen [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany); Kilian, Wolfgang; Sekulla, Marco [Siegen Univ. (Germany). Theoretische Physik I

    2013-07-15

    In this contribution to the Snowmass process 2013 we give a brief review of how new physics could enter in the electroweak (EW) sector of the Standard Model (SM). This new physics, if it is directly accessible at low energies, can be parameterized by explicit resonances having certain quantum numbers. The extreme case is the decoupling limit where those resonances are very heavy and leave only traces in the form of deviations in the SM couplings. Translations are given into higher-dimensional operators leading to such deviations. As long as such resonances are introduced without a UV-complete theory behind it, these models suffer from unitarity violation of perturbative scattering amplitudes. We show explicitly how theoretically sane descriptions could be achieved by using a unitarization prescription that allows a correct description of such a resonance without specifying a UV-complete model.

  9. Porting plasma physics simulation codes to modern computing architectures using the libmrc framework

    Science.gov (United States)

    Germaschewski, Kai; Abbott, Stephen

    2015-11-01

    Available computing power has continued to grow exponentially even after single-core performance satured in the last decade. The increase has since been driven by more parallelism, both using more cores and having more parallelism in each core, e.g. in GPUs and Intel Xeon Phi. Adapting existing plasma physics codes is challenging, in particular as there is no single programming model that covers current and future architectures. We will introduce the open-source libmrc framework that has been used to modularize and port three plasma physics codes: The extended MHD code MRCv3 with implicit time integration and curvilinear grids; the OpenGGCM global magnetosphere model; and the particle-in-cell code PSC. libmrc consolidates basic functionality needed for simulations based on structured grids (I/O, load balancing, time integrators), and also introduces a parallel object model that makes it possible to maintain multiple implementations of computational kernels, on e.g. conventional processors and GPUs. It handles data layout conversions and enables us to port performance-critical parts of a code to a new architecture step-by-step, while the rest of the code can remain unchanged. We will show examples of the performance gains and some physics applications.

  10. Study of no-man's land physics in the total-f gyrokinetic code XGC1

    Science.gov (United States)

    Ku, Seung Hoe; Chang, C. S.; Lang, J.

    2014-10-01

    While the ``transport shortfall'' in the ``no-man's land'' has been observed often in delta-f codes, it has not yet been observed in the global total-f gyrokinetic particle code XGC1. Since understanding the interaction between the edge and core transport appears to be a critical element in the prediction for ITER performance, understanding the no-man's land issue is an important physics research topic. Simulation results using the Holland case will be presented and the physics causing the shortfall phenomenon will be discussed. Nonlinear nonlocal interaction of turbulence, secondary flows, and transport appears to be the key.

  11. Sizing and scaling requirements of a large-scale physical model for code validation

    International Nuclear Information System (INIS)

    Khaleel, R.; Legore, T.

    1990-01-01

    Model validation is an important consideration in application of a code for performance assessment and therefore in assessing the long-term behavior of the engineered and natural barriers of a geologic repository. Scaling considerations relevant to porous media flow are reviewed. An analysis approach is presented for determining the sizing requirements of a large-scale, hydrology physical model. The physical model will be used to validate performance assessment codes that evaluate the long-term behavior of the repository isolation system. Numerical simulation results for sizing requirements are presented for a porous medium model in which the media properties are spatially uncorrelated

  12. Statistical physics inspired energy-efficient coded-modulation for optical communications.

    Science.gov (United States)

    Djordjevic, Ivan B; Xu, Lei; Wang, Ting

    2012-04-15

    Because Shannon's entropy can be obtained by Stirling's approximation of thermodynamics entropy, the statistical physics energy minimization methods are directly applicable to the signal constellation design. We demonstrate that statistical physics inspired energy-efficient (EE) signal constellation designs, in combination with large-girth low-density parity-check (LDPC) codes, significantly outperform conventional LDPC-coded polarization-division multiplexed quadrature amplitude modulation schemes. We also describe an EE signal constellation design algorithm. Finally, we propose the discrete-time implementation of D-dimensional transceiver and corresponding EE polarization-division multiplexed system. © 2012 Optical Society of America

  13. Reactor physics computer code development for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs

    International Nuclear Information System (INIS)

    Rastogi, B.P.

    1989-01-01

    This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)

  14. Physical activity and influenza-coded outpatient visits, a population-based cohort study.

    Directory of Open Access Journals (Sweden)

    Eric Siu

    Full Text Available Although the benefits of physical activity in preventing chronic medical conditions are well established, its impacts on infectious diseases, and seasonal influenza in particular, are less clearly defined. We examined the association between physical activity and influenza-coded outpatient visits, as a proxy for influenza infection.We conducted a cohort study of Ontario respondents to Statistics Canada's population health surveys over 12 influenza seasons. We assessed physical activity levels through survey responses, and influenza-coded physician office and emergency department visits through physician billing claims. We used logistic regression to estimate the risk of influenza-coded outpatient visits during influenza seasons. The cohort comprised 114,364 survey respondents who contributed 357,466 person-influenza seasons of observation. Compared to inactive individuals, moderately active (OR 0.83; 95% CI 0.74-0.94 and active (OR 0.87; 95% CI 0.77-0.98 individuals were less likely to experience an influenza-coded visit. Stratifying by age, the protective effect of physical activity remained significant for individuals <65 years (active OR 0.86; 95% CI 0.75-0.98, moderately active: OR 0.85; 95% CI 0.74-0.97 but not for individuals ≥ 65 years. The main limitations of this study were the use of influenza-coded outpatient visits rather than laboratory-confirmed influenza as the outcome measure, the reliance on self-report for assessing physical activity and various covariates, and the observational study design.Moderate to high amounts of physical activity may be associated with reduced risk of influenza for individuals <65 years. Future research should use laboratory-confirmed influenza outcomes to confirm the association between physical activity and influenza.

  15. Carbon Inputs From Riparian Vegetation Limit Oxidation of Physically Bound Organic Carbon Via Biochemical and Thermodynamic Processes

    Science.gov (United States)

    Graham, Emily B.; Tfaily, Malak M.; Crump, Alex R.; Goldman, Amy E.; Bramer, Lisa M.; Arntzen, Evan; Romero, Elvira; Resch, C. Tom; Kennedy, David W.; Stegen, James C.

    2017-12-01

    In light of increasing terrestrial carbon (C) transport across aquatic boundaries, the mechanisms governing organic carbon (OC) oxidation along terrestrial-aquatic interfaces are crucial to future climate predictions. Here we investigate the biochemistry, metabolic pathways, and thermodynamics corresponding to OC oxidation in the Columbia River corridor using ultrahigh-resolution C characterization. We leverage natural vegetative differences to encompass variation in terrestrial C inputs. Our results suggest that decreases in terrestrial C deposition associated with diminished riparian vegetation induce oxidation of physically bound OC. We also find that contrasting metabolic pathways oxidize OC in the presence and absence of vegetation and—in direct conflict with the "priming" concept—that inputs of water-soluble and thermodynamically favorable terrestrial OC protect bound-OC from oxidation. In both environments, the most thermodynamically favorable compounds appear to be preferentially oxidized regardless of which OC pool microbiomes metabolize. In turn, we suggest that the extent of riparian vegetation causes sediment microbiomes to locally adapt to oxidize a particular pool of OC but that common thermodynamic principles govern the oxidation of each pool (i.e., water-soluble or physically bound). Finally, we propose a mechanistic conceptualization of OC oxidation along terrestrial-aquatic interfaces that can be used to model heterogeneous patterns of OC loss under changing land cover distributions.

  16. Validation of physics and thermalhydraulic computer codes for advanced Candu reactor applications

    International Nuclear Information System (INIS)

    Wren, D.J.; Popov, N.; Snell, V.G.

    2004-01-01

    Atomic Energy of Canada Ltd. (AECL) is developing an Advanced Candu Reactor (ACR) that is an evolutionary advancement of the currently operating Candu 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular Candu concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in Candu 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the Candu 6, most of the key phenomena associated with the safety analyses of ACR are common, and the Candu industry standard tool-set of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design

  17. Digitized forensics: retaining a link between physical and digital crime scene traces using QR-codes

    Science.gov (United States)

    Hildebrandt, Mario; Kiltz, Stefan; Dittmann, Jana

    2013-03-01

    The digitization of physical traces from crime scenes in forensic investigations in effect creates a digital chain-of-custody and entrains the challenge of creating a link between the two or more representations of the same trace. In order to be forensically sound, especially the two security aspects of integrity and authenticity need to be maintained at all times. Especially the adherence to the authenticity using technical means proves to be a challenge at the boundary between the physical object and its digital representations. In this article we propose a new method of linking physical objects with its digital counterparts using two-dimensional bar codes and additional meta-data accompanying the acquired data for integration in the conventional documentation of collection of items of evidence (bagging and tagging process). Using the exemplary chosen QR-code as particular implementation of a bar code and a model of the forensic process, we also supply a means to integrate our suggested approach into forensically sound proceedings as described by Holder et al.1 We use the example of the digital dactyloscopy as a forensic discipline, where currently progress is being made by digitizing some of the processing steps. We show an exemplary demonstrator of the suggested approach using a smartphone as a mobile device for the verification of the physical trace to extend the chain-of-custody from the physical to the digital domain. Our evaluation of the demonstrator is performed towards the readability and the verification of its contents. We can read the bar code despite its limited size of 42 x 42 mm and rather large amount of embedded data using various devices. Furthermore, the QR-code's error correction features help to recover contents of damaged codes. Subsequently, our appended digital signature allows for detecting malicious manipulations of the embedded data.

  18. Assessment of CANDU physics codes using experimental data - II: CANDU core physics measurements

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Jeong, Chang Joon; Choi, Hang Bok

    2001-11-01

    Benchmark calculations of the advanced CANDU reactor analysis tools (WIMS-AECL, SHETAN and RFSP) and the Monte Carlo code MCNP-4B have been performed using Wolsong Units 2 and 3 Phase-B measurement data. In this study, the benchmark calculations have been done for the criticality, boron worth, reactivity device worth, reactivity coefficient, and flux scan. For the validation of the WIMS-AECL/SHETANRFSP code system, the lattice parameters of the fuel channel were generated by the WIMS-AECL code, and incremental cross sections of reactivity devices and structural material were generated by the SHETAN code. The results have shown that the criticality is under-predicted by -4 mk. The reactivity device worths are generally consistent with the measured data except for the strong absorbers such as shutoff rod and mechanical control absorber. The heat transport system temperature coefficient and flux distributions are in good agreement with the measured data. However, the moderator temperature coefficient has shown a relatively large error, which could be caused by the incremental cross-section generation methodology for the reactivity device. For the MCNP-4B benchmark calculation, cross section libraries were newly generated from ENDF/B-VI release 3 through the NJOY97.114 data processing system and a three-dimensional full core model was developed. The simulation results have shown that the criticality is estimated within 4 mk and the estimated reactivity worth of the control devices are generally consistent with the measurement data, which implies that the MCNP code is valid for CANDU core analysis. In the future, therefore, the MCNP code could be used as a reference tool to benchmark design and analysis codes for the advanced fuels for which experimental data are not available

  19. 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.

  20. Applications of FLUKA Monte Carlo code for nuclear and accelerator physics

    CERN Document Server

    Battistoni, Giuseppe; Brugger, Markus; Campanella, Mauro; Carboni, Massimo; Empl, Anton; Fasso, Alberto; Gadioli, Ettore; Cerutti, Francesco; Ferrari, Alfredo; Ferrari, Anna; Lantz, Matthias; Mairani, Andrea; Margiotta, M; Morone, Christina; Muraro, Silvia; Parodi, Katerina; Patera, Vincenzo; Pelliccioni, Maurizio; Pinsky, Lawrence; Ranft, Johannes; Roesler, Stefan; Rollet, Sofia; Sala, Paola R; Santana, Mario; Sarchiapone, Lucia; Sioli, Maximiliano; Smirnov, George; Sommerer, Florian; Theis, Christian; Trovati, Stefania; Villari, R; Vincke, Heinz; Vincke, Helmut; Vlachoudis, Vasilis; Vollaire, Joachim; Zapp, Neil

    2011-01-01

    FLUKA is a general purpose Monte Carlo code capable of handling all radiation components from thermal energies (for neutrons) or 1keV (for all other particles) to cosmic ray energies and can be applied in many different fields. Presently the code is maintained on Linux. The validity of the physical models implemented in FLUKA has been benchmarked against a variety of experimental data over a wide energy range, from accelerator data to cosmic ray showers in the Earth atmosphere. FLUKA is widely used for studies related both to basic research and to applications in particle accelerators, radiation protection and dosimetry, including the specific issue of radiation damage in space missions, radiobiology (including radiotherapy) and cosmic ray calculations. After a short description of the main features that make FLUKA valuable for these topics, the present paper summarizes some of the recent applications of the FLUKA Monte Carlo code in the nuclear as well high energy physics. In particular it addresses such top...

  1. Development of improved methods for the LWR lattice physics code EPRI-CELL

    International Nuclear Information System (INIS)

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

    1982-07-01

    A number of improvements have been made by ORNL to the lattice physics code EPRI-CELL (E-C) which is widely used by utilities for analysis of power reactors. The code modifications were made mainly in the thermal and epithermal routines and resulted in improved reactor physics approximations and more efficient running times. The improvements in the thermal flux calculation included implementation of a group-dependent rebalance procedure to accelerate the iterative process and a more rigorous calculation of interval-to-interval collision probabilities. The epithermal resonance shielding methods used in the code have been extensively studied to determine its major approximations and to examine the sensitivity of computed results to these approximations. The study has resulted in several improvements in the original methodology

  2. Effects of pulse frequency of input power on the physical and chemical properties of pulsed streamer discharge plasmas in water

    Science.gov (United States)

    Ruma; Lukes, P.; Aoki, N.; Spetlikova, E.; Hosseini, S. H. R.; Sakugawa, T.; Akiyama, H.

    2013-03-01

    A repetitive pulsed-power modulator, which employs a magnetic pulse compression circuit with a high-speed thyristor switch, was used to study the effects of the pulse repetition rate of input power on the physical and chemical properties of pulsed discharges in water. Positive high-voltage pulses of 20 kV with repetition rates of up to 1 kHz were used to generate a discharge in water using the point-to-plane electrode geometry. By varying the pulse repetition rate, two distinct modes of the discharge plasma were formed in water. The first mode was characterized by the formation of a corona-like discharge propagating through water in the form of streamer channels. The second mode was formed typically above 500 Hz, when the formation of streamer channels in water was suppressed and all plasmas occurred inside a spheroidal aggregate of very fine gas bubbles surrounding the tip of the high-voltage electrode. The production of hydrogen peroxide, degradation of organic dye Acid Orange 7 (AO7) and inactivation of bacteria Escherichia coli by the discharge in water were studied under different discharge plasma modes in dependence on the pulse repetition rate of input power. The efficiency of both chemical and biocidal processes induced by the plasma in water decreased significantly with pulse repetition rates above 500 Hz.

  3. Effects of pulse frequency of input power on the physical and chemical properties of pulsed streamer discharge plasmas in water

    International Nuclear Information System (INIS)

    Ruma; Aoki, N; Hosseini, S H R; Sakugawa, T; Akiyama, H; Lukes, P; Spetlikova, E

    2013-01-01

    A repetitive pulsed-power modulator, which employs a magnetic pulse compression circuit with a high-speed thyristor switch, was used to study the effects of the pulse repetition rate of input power on the physical and chemical properties of pulsed discharges in water. Positive high-voltage pulses of 20 kV with repetition rates of up to 1 kHz were used to generate a discharge in water using the point-to-plane electrode geometry. By varying the pulse repetition rate, two distinct modes of the discharge plasma were formed in water. The first mode was characterized by the formation of a corona-like discharge propagating through water in the form of streamer channels. The second mode was formed typically above 500 Hz, when the formation of streamer channels in water was suppressed and all plasmas occurred inside a spheroidal aggregate of very fine gas bubbles surrounding the tip of the high-voltage electrode. The production of hydrogen peroxide, degradation of organic dye Acid Orange 7 (AO7) and inactivation of bacteria Escherichia coli by the discharge in water were studied under different discharge plasma modes in dependence on the pulse repetition rate of input power. The efficiency of both chemical and biocidal processes induced by the plasma in water decreased significantly with pulse repetition rates above 500 Hz. (paper)

  4. Preliminary analyses for HTTR`s start-up physics tests by Monte Carlo code MVP

    Energy Technology Data Exchange (ETDEWEB)

    Nojiri, Naoki [Science and Technology Agency, Tokyo (Japan); Nakano, Masaaki; Ando, Hiroei; Fujimoto, Nozomu; Takeuchi, Mitsuo; Fujisaki, Shingo; Yamashita, Kiyonobu

    1998-08-01

    Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)

  5. Preliminary analyses for HTTR's start-up physics tests by Monte Carlo code MVP

    International Nuclear Information System (INIS)

    Nojiri, Naoki; Nakano, Masaaki; Ando, Hiroei; Fujimoto, Nozomu; Takeuchi, Mitsuo; Fujisaki, Shingo; Yamashita, Kiyonobu

    1998-08-01

    Analyses of start-up physics tests for High Temperature Engineering Test Reactor (HTTR) have been carried out by Monte Carlo code MVP based on continuous energy method. Heterogeneous core structures were modified precisely, such as the fuel compacts, fuel rods, coolant channels, burnable poisons, control rods, control rod insertion holes, reserved shutdown pellet insertion holes, gaps between graphite blocks, etc. Such precise modification of the core structures was difficult with diffusion calculation. From the analytical results, the followings were confirmed; The first criticality will be achieved around 16 fuel columns loaded. The reactivity at the first criticality can be controlled by only one control rod located at the center of the core with other fifteen control rods fully withdrawn. The excess reactivity, reactor shutdown margin and control rod criticality positions have been evaluated. These results were used for planning of the start-up physics tests. This report presents analyses of start-up physics tests for HTTR by MVP code. (author)

  6. Geometrical modification transfer between specific meshes of each coupled physical codes. Application to the Jules Horowitz research reactor experimental devices

    International Nuclear Information System (INIS)

    Duplex, B.

    2011-01-01

    The CEA develops and uses scientific software, called physical codes, in various physical disciplines to optimize installation and experimentation costs. During a study, several physical phenomena interact, so a code coupling and some data exchanges between different physical codes are required. Each physical code computes on a particular geometry, usually represented by a mesh composed of thousands to millions of elements. This PhD Thesis focuses on the geometrical modification transfer between specific meshes of each coupled physical code. First, it presents a physical code coupling method where deformations are computed by one of these codes. Next, it discusses the establishment of a model, common to different physical codes, grouping all the shared data. Finally, it covers the deformation transfers between meshes of the same geometry or adjacent geometries. Geometrical modifications are discrete data because they are based on a mesh. In order to permit every code to access deformations and to transfer them, a continuous representation is computed. Two functions are developed, one with a global support, and the other with a local support. Both functions combine a simplification method and a radial basis function network. A whole use case is dedicated to the Jules Horowitz reactor. The effect of differential dilatations on experimental device cooling is studied. (author) [fr

  7. Development of a three dimension multi-physics code for molten salt fast reactor

    International Nuclear Information System (INIS)

    Cheng Maosong; Dai Zhimin

    2014-01-01

    Molten Salt Reactor (MSR) was selected as one of the six innovative nuclear reactors by the Generation IV International Forum (GIF). The circulating-fuel in the can-type molten salt fast reactor makes the neutronics and thermo-hydraulics of the reactor strongly coupled and different from that of traditional solid-fuel reactors. In the present paper: a new coupling model is presented that physically describes the inherent relations between the neutron flux, the delayed neutron precursor, the heat transfer and the turbulent flow. Based on the model, integrating nuclear data processing, CAD modeling, structured and unstructured mesh technology, data analysis and visualization application, a three dimension steady state simulation code system (MSR3DS) for the can-type molten salt fast reactor is developed and validated. In order to demonstrate the ability of the code, the three dimension distributions of the velocity, the neutron flux, the delayed neutron precursor and the temperature were obtained for the simplified MOlten Salt Advanced Reactor Transmuter (MOSART) using this code. The results indicate that the MSR3DS code can provide a feasible description of multi-physical coupling phenomena in can-type molten salt fast reactor. Furthermore, the code can well predict the flow effect of fuel salt and the transport effect of the turbulent diffusion. (authors)

  8. Radiation physics and shielding codes and analyses applied to design-assist and safety analyses of CANDUR and ACRTM reactors

    International Nuclear Information System (INIS)

    Aydogdu, K.; Boss, C. R.

    2006-01-01

    This paper discusses the radiation physics and shielding codes and analyses applied in the design of CANDU and ACR reactors. The focus is on the types of analyses undertaken rather than the inputs supplied to the engineering disciplines. Nevertheless, the discussion does show how these analyses contribute to the engineering design. Analyses in radiation physics and shielding can be categorized as either design-assist or safety and licensing (accident) analyses. Many of the analyses undertaken are designated 'design-assist' where the analyses are used to generate recommendations that directly influence plant design. These recommendations are directed at mitigating or reducing the radiation hazard of the nuclear power plant with engineered systems and components. Thus the analyses serve a primary safety function by ensuring the plant can be operated with acceptable radiation hazards to the workers and public. In addition to this role of design assist, radiation physics and shielding codes are also deployed in safety and licensing assessments of the consequences of radioactive releases of gaseous and liquid effluents during normal operation and gaseous effluents following accidents. In the latter category, the final consequences of accident sequences, expressed in terms of radiation dose to members of the public, and inputs to accident analysis, e.g., decay heat in fuel following a loss-of-coolant accident, are also calculated. Another role of the analyses is to demonstrate that the design of the plant satisfies the principle of ALARA (as low as reasonably achievable) radiation doses. This principle is applied throughout the design process to minimize worker and public doses. The principle of ALARA is an inherent part of all design-assist recommendations and safety and licensing assessments. The main focus of an ALARA exercise at the design stage is to minimize the radiation hazards at the source. This exploits material selection and impurity specifications and relies

  9. Recent improvements and new features in the Westinghouse lattice physics codes

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  10. Basic physical and chemical information needed for development of Monte Carlo codes

    International Nuclear Information System (INIS)

    Inokuti, M.

    1993-01-01

    It is important to view track structure analysis as an application of a branch of theoretical physics (i.e., statistical physics and physical kinetics in the language of the Landau school). Monte Carlo methods and transport equation methods represent two major approaches. In either approach, it is of paramount importance to use as input the cross section data that best represent the elementary microscopic processes. Transport analysis based on unrealistic input data must be viewed with caution, because results can be misleading. Work toward establishing the cross section data, which demands a wide scope of knowledge and expertise, is being carried out through extensive international collaborations. In track structure analysis for radiation biology, the need for cross sections for the interactions of electrons with DNA and neighboring protein molecules seems to be especially urgent. Finally, it is important to interpret results of Monte Carlo calculations fully and adequately. To this end, workers should document input data as thoroughly as possible and report their results in detail in many ways. Workers in analytic transport theory are then likely to contribute to the interpretation of the results

  11. Physics Based Model for Cryogenic Chilldown and Loading. Part IV: Code Structure

    Science.gov (United States)

    Luchinsky, D. G.; Smelyanskiy, V. N.; Brown, B.

    2014-01-01

    This is the fourth report in a series of technical reports that describe separated two-phase flow model application to the cryogenic loading operation. In this report we present the structure of the code. The code consists of five major modules: (1) geometry module; (2) solver; (3) material properties; (4) correlations; and finally (5) stability control module. The two key modules - solver and correlations - are further divided into a number of submodules. Most of the physics and knowledge databases related to the properties of cryogenic two-phase flow are included into the cryogenic correlations module. The functional form of those correlations is not well established and is a subject of extensive research. Multiple parametric forms for various correlations are currently available. Some of them are included into correlations module as will be described in details in a separate technical report. Here we describe the overall structure of the code and focus on the details of the solver and stability control modules.

  12. QR-codes as a tool to increase physical activity level among school children during class hours

    DEFF Research Database (Denmark)

    Christensen, Jeanette Reffstrup; Kristensen, Allan; Bredahl, Thomas Viskum Gjelstrup

    the students physical activity level during class hours. Methods: A before-after study was used to examine 12 students physical activity level, measured with pedometers for six lessons. Three lessons of traditional teaching and three lessons, where QR-codes were used to make orienteering in school area...... as old fashioned. The students also felt positive about being physically active in teaching. Discussion and conclusion: QR-codes as a tool for teaching are usable for making students more physically active in teaching. The students were exited for using QR-codes and they experienced a good motivation......QR-codes as a tool to increase physical activity level among school children during class hours Introduction: Danish students are no longer fulfilling recommendations for everyday physical activity. Since August 2014, Danish students in public schools are therefore required to be physically active...

  13. 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.

  14. Evaluation of the HTR-10 Reactor as a Benchmark for Physics Code QA

    International Nuclear Information System (INIS)

    William K. Terry; Soon Sam Kim; Leland M. Montierth; Joshua J. Cogliati; Abderrafi M. Ougouag

    2006-01-01

    The HTR-10 is a small (10 MWt) pebble-bed research reactor intended to develop pebble-bed reactor (PBR) technology in China. It will be used to test and develop fuel, verify PBR safety features, demonstrate combined electricity production and co-generation of heat, and provide experience in PBR design, operation, and construction. As the only currently operating PBR in the world, the HTR-10 can provide data of great interest to everyone involved in PBR technology. In particular, if it yields data of sufficient quality, it can be used as a benchmark for assessing the accuracy of computer codes proposed for use in PBR analysis. This paper summarizes the evaluation for the International Reactor Physics Experiment Evaluation Project (IRPhEP) of data obtained in measurements of the HTR-10's initial criticality experiment for use as benchmarks for reactor physics codes

  15. Uniform physical theory of diffraction equivalent edge currents for implementation in general computer codes

    DEFF Research Database (Denmark)

    Johansen, Peter Meincke

    1996-01-01

    New uniform closed-form expressions for physical theory of diffraction equivalent edge currents are derived for truncated incremental wedge strips. In contrast to previously reported expressions, the new expressions are well-behaved for all directions of incidence and observation and take a finite...... value for zero strip length. Consequently, the new equivalent edge currents are, to the knowledge of the author, the first that are well-suited for implementation in general computer codes...

  16. Physical-layer Network Coding in Two-Way Heterogeneous Cellular Networks with Power Imbalance

    OpenAIRE

    Thampi, Ajay K; Liew, Soung Chang; Armour, Simon M D; Fan, Zhong; You, Lizhao; Kaleshi, Dritan

    2016-01-01

    The growing demand for high-speed data, quality of service (QoS) assurance and energy efficiency has triggered the evolution of 4G LTE-A networks to 5G and beyond. Interference is still a major performance bottleneck. This paper studies the application of physical-layer network coding (PNC), a technique that exploits interference, in heterogeneous cellular networks. In particular, we propose a rate-maximising relay selection algorithm for a single cell with multiple relays assuming the decode...

  17. SSYST-3. Input description

    International Nuclear Information System (INIS)

    Meyder, R.

    1983-12-01

    The code system SSYST-3 is designed to analyse the thermal and mechanical behaviour of a fuel rod during a LOCA. The report contains a complete input-list for all modules and several tested inputs for a LOCA analysis. (orig.)

  18. The bidimensional neutron transport code TWOTRAN-GG. Users manual and input data TWOTRAN-TRACA version; El codigo de transporte bidimensional TWOTRAN-GG. Manual de usuario y datos de entrada version TWOTRAN-TRACA

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C; Aragones, J M

    1981-07-01

    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.

  19. Computation of Thermodynamic Equilibria Pertinent to Nuclear Materials in Multi-Physics Codes

    Science.gov (United States)

    Piro, Markus Hans Alexander

    Nuclear energy plays a vital role in supporting electrical needs and fulfilling commitments to reduce greenhouse gas emissions. Research is a continuing necessity to improve the predictive capabilities of fuel behaviour in order to reduce costs and to meet increasingly stringent safety requirements by the regulator. Moreover, a renewed interest in nuclear energy has given rise to a "nuclear renaissance" and the necessity to design the next generation of reactors. In support of this goal, significant research efforts have been dedicated to the advancement of numerical modelling and computational tools in simulating various physical and chemical phenomena associated with nuclear fuel behaviour. This undertaking in effect is collecting the experience and observations of a past generation of nuclear engineers and scientists in a meaningful way for future design purposes. There is an increasing desire to integrate thermodynamic computations directly into multi-physics nuclear fuel performance and safety codes. A new equilibrium thermodynamic solver is being developed with this matter as a primary objective. This solver is intended to provide thermodynamic material properties and boundary conditions for continuum transport calculations. There are several concerns with the use of existing commercial thermodynamic codes: computational performance; limited capabilities in handling large multi-component systems of interest to the nuclear industry; convenient incorporation into other codes with quality assurance considerations; and, licensing entanglements associated with code distribution. The development of this software in this research is aimed at addressing all of these concerns. The approach taken in this work exploits fundamental principles of equilibrium thermodynamics to simplify the numerical optimization equations. In brief, the chemical potentials of all species and phases in the system are constrained by estimates of the chemical potentials of the system

  20. Assessment CANDU physics codes using experimental data - part 1: criticality measurement

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok; Jeong, Chang Joon

    2001-08-01

    In order to assess the applicability of MCNP-4B code to the heavy water moderated, light water cooled and pressure-tube type reactor, the MCNP-4B physics calculations has been carried out for the Deuterium Critical Assembly (DCA), and the results were compared with those of the experimental data. In this study, the key safety parameters like as the multiplication factor, void coefficient, local power peaking factor and bundle power distribution in the scattered core are simulated. In order to use the cross section data consistently for the fuels to be analyzed in the future, new MCNP libraries have been generated from ENDF/B-VI release 3. Generally, the MCNP-4B calculation results show a good agreement with experimental data of DCA core. After benchmarking MCNP-4B against available experimental data, it will be used as the reference tool to benchmark design and analysis codes for the advanced CANDU fuels

  1. 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...

  2. Verification of MVP-II and SRAC2006 code to the core physics vera benchmark problem

    International Nuclear Information System (INIS)

    Jati Susilo

    2014-01-01

    In this research, verification calculation for VERA core physics benchmark on the Zero Power Physical Test (ZPPT) of the nuclear reactor Watts Bar 1. The reactor is a 1000 MWe class of PWR designed by. Westinghouse, arranged from 193 unit of 17 x 17 fuel assembly consisting 3 type enrichment of UO2 that are 2.1wt%, 2.619wt% and 3.1wt%. Core power factor distribution and k-eff calculation has been done for the first cycle operation of the core at beginning of cycle (BOC) and hot zero power (HZP). In this calculation, MVP-II and CITATION module of SRAC2006 computer code has been used with ENDF/B-VII.0. cross section data library. Calculation result showed that differences value of k-eff for the core at controlled and uncontrolled condition between reference with MVP-II (-0,07% and -0,014%) and SRAC2006 (0,92% and 0,99%) are very small or below 1%. Differences value of radial power peaking factor at controlled and uncontrolled of the core between reference value with MVP-II are 0,38% and 1,53%, even though with SRAC2006 are 1,13% and -2,45%. It can be said that the calculation result by both computer code showing suitability with reference value. In order to determinate of criticality of the core, the calculation result using MVP-II code is more conservative compare with SRAC2006 code. (author)

  3. Integrated Numerical Experiments (INEX) and the Free-Electron Laser Physical Process Code (FELPPC)

    International Nuclear Information System (INIS)

    Thode, L.E.; Chan, K.C.D.; Schmitt, M.J.; McKee, J.; Ostic, J.; Elliott, C.J.; McVey, B.D.

    1990-01-01

    The strong coupling of subsystem elements, such as the accelerator, wiggler, and optics, greatly complicates the understanding and design of a free electron laser (FEL), even at the conceptual level. To address the strong coupling character of the FEL the concept of an Integrated Numerical Experiment (INEX) was proposed. Unique features of the INEX approach are consistency and numerical equivalence of experimental diagnostics. The equivalent numerical diagnostics mitigates the major problem of misinterpretation that often occurs when theoretical and experimental data are compared. The INEX approach has been applied to a large number of accelerator and FEL experiments. Overall, the agreement between INEX and the experiments is very good. Despite the success of INEX, the approach is difficult to apply to trade-off and initial design studies because of the significant manpower and computational requirements. On the other hand, INEX provides a base from which realistic accelerator, wiggler, and optics models can be developed. The Free Electron Laser Physical Process Code (FELPPC) includes models developed from INEX, provides coupling between the subsystem models, and incorporates application models relevant to a specific trade-off or design study. In other words, FELPPC solves the complete physical process model using realistic physics and technology constraints. Because FELPPC provides a detailed design, a good estimate for the FEL mass, cost, and size can be made from a piece-part count of the FEL. FELPPC requires significant accelerator and FEL expertise to operate. The code can calculate complex FEL configurations including multiple accelerator and wiggler combinations

  4. Fission product release from nuclear fuel I. Physical modelling in the ASTEC code

    International Nuclear Information System (INIS)

    Brillant, G.; Marchetto, C.; Plumecocq, W.

    2013-01-01

    Highlights: • Physical modeling of FP and SM release in ASTEC is presented. • The release is described as solid state diffusion within fuel for high volatile FP. • The release is described as FP vaporisation for semi volatile FP. • The release is described as fuel vaporisation for low volatile FP. • ASTEC validation is presented in the second paper. - Abstract: This article is the first of a series of two articles dedicated to the mechanisms of fission product release from a degraded core as they are modelled in the ASTEC code. The ASTEC code aims at simulating severe accidents in nuclear reactors from the initiating event up to the radiological consequences on the environment. This code is used for several applications such as nuclear plant safety evaluation including probabilistic studies and emergency preparedness. To cope with the requirements of robustness and low calculation time, the code is based on a semi-empirical approach and only the main limiting phenomena that govern the release from intact rods and from debris beds are considered. For solid fuel, fission products are classified into three groups, depending on their degree of volatility. The kinetics of volatile fission products release depend on the rate-limiting process of solid-state diffusion through fuel grains. For semi-volatile fission products, the release from the open fuel porosities is assumed to be governed by vaporisation and mass transfer processes. The key phenomenon for the release of low volatile fission products is supposed to be fuel volatilisation. A similar approach is used for the release of fission products from a rubble bed. An in-depth validation of the code including both analytical and integral experiments is the subject of the second article

  5. Experimental benchmark of non-local-thermodynamic-equilibrium plasma atomic physics codes

    International Nuclear Information System (INIS)

    Nagels-Silvert, V.

    2004-09-01

    The main purpose of this thesis is to get experimental data for the testing and validation of atomic physics codes dealing with non-local-thermodynamical-equilibrium plasmas. The first part is dedicated to the spectroscopic study of xenon and krypton plasmas that have been produced by a nanosecond laser pulse interacting with a gas jet. A Thomson scattering diagnostic has allowed us to measure independently plasma parameters such as electron temperature, electron density and the average ionisation state. We have obtained time integrated spectra in the range between 5 and 10 angstroms. We have identified about one hundred xenon rays between 8.6 and 9.6 angstroms via the use of the Relac code. We have discovered unknown rays for the krypton between 5.2 and 7.5 angstroms. In a second experiment we have extended the wavelength range to the X UV domain. The Averroes/Transpec code has been tested in the ranges from 9 to 15 angstroms and from 10 to 130 angstroms, the first range has been well reproduced while the second range requires a more complex data analysis. The second part is dedicated to the spectroscopic study of aluminium, selenium and samarium plasmas in femtosecond operating rate. We have designed an interferometry diagnostic in the frequency domain that has allowed us to measure the expanding speed of the target's backside. Via the use of an adequate isothermal model this parameter has led us to know the plasma electron temperature. Spectra and emission times of various rays from the aluminium and selenium plasmas have been computed satisfactorily with the Averroes/Transpec code coupled with Film and Multif hydrodynamical codes. (A.C.)

  6. Calculation codes in radiation protection, radiation physics and dosimetry; Codes de calcul en radioprotection, radiophysique et dosimetrie

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-07-01

    These scientific days had for objective to draw up the situation of calculation codes of radiation transport, of sources estimation, of radiation doses managements and to draw the future perspectives. (N.C.)

  7. Effects of physics change in Monte Carlo code on electron pencil beam dose distributions

    International Nuclear Information System (INIS)

    Toutaoui, Abdelkader; Khelassi-Toutaoui, Nadia; Brahimi, Zakia; Chami, Ahmed Chafik

    2012-01-01

    Pencil beam algorithms used in computerized electron beam dose planning are usually described using the small angle multiple scattering theory. Alternatively, the pencil beams can be generated by Monte Carlo simulation of electron transport. In a previous work, the 4th version of the Electron Gamma Shower (EGS) Monte Carlo code was used to obtain dose distributions from monoenergetic electron pencil beam, with incident energy between 1 MeV and 50 MeV, interacting at the surface of a large cylindrical homogeneous water phantom. In 2000, a new version of this Monte Carlo code has been made available by the National Research Council of Canada (NRC), which includes various improvements in its electron-transport algorithms. In the present work, we were interested to see if the new physics in this version produces pencil beam dose distributions very different from those calculated with oldest one. The purpose of this study is to quantify as well as to understand these differences. We have compared a series of pencil beam dose distributions scored in cylindrical geometry, for electron energies between 1 MeV and 50 MeV calculated with two versions of the Electron Gamma Shower Monte Carlo Code. Data calculated and compared include isodose distributions, radial dose distributions and fractions of energy deposition. Our results for radial dose distributions show agreement within 10% between doses calculated by the two codes for voxels closer to the pencil beam central axis, while the differences are up to 30% for longer distances. For fractions of energy deposition, the results of the EGS4 are in good agreement (within 2%) with those calculated by EGSnrc at shallow depths for all energies, whereas a slightly worse agreement (15%) is observed at deeper distances. These differences may be mainly attributed to the different multiple scattering for electron transport adopted in these two codes and the inclusion of spin effect, which produces an increase of the effective range of

  8. Theory and application of the RAZOR two-dimensional continuous energy lattice physics code

    International Nuclear Information System (INIS)

    Zerkle, M.L.; Abu-Shumays, I.K.; Ott, M.W.; Winwood, J.P.

    1997-01-01

    The theory and application of the RAZOR two-dimensional, continuous energy lattice physics code are discussed. RAZOR solves the continuous energy neutron transport equation in one- and two-dimensional geometries, and calculates equivalent few-group diffusion theory constants that rigorously account for spatial and spectral self-shielding effects. A dual energy resolution slowing down algorithm is used to reduce computer memory and disk storage requirements for the slowing down calculation. Results are presented for a 2D BWR pin cell depletion benchmark problem

  9. Physical-layer network coding for passive optical interconnect in datacenter networks.

    Science.gov (United States)

    Lin, Rui; Cheng, Yuxin; Guan, Xun; Tang, Ming; Liu, Deming; Chan, Chun-Kit; Chen, Jiajia

    2017-07-24

    We introduce physical-layer network coding (PLNC) technique in a passive optical interconnect (POI) architecture for datacenter networks. The implementation of the PLNC in the POI at 2.5 Gb/s and 10Gb/s have been experimentally validated while the gains in terms of network layer performances have been investigated by simulation. The results reveal that in order to realize negligible packet drop, the wavelengths usage can be reduced by half while a significant improvement in packet delay especially under high traffic load can be achieved by employing PLNC over POI.

  10. Relative validity of the pre-coded food diary used in the Danish National Survey of Diet and Physical Activity

    DEFF Research Database (Denmark)

    Knudsen, Vibeke Kildegaard; Gille, Maj-Britt; Nielsen, Trine Holmgaard

    2011-01-01

    Objective: To determine the relative validity of the pre-coded food diary applied in the Danish National Survey of Dietary Habits and Physical Activity. Design: A cross-over study among seventy-two adults (aged 20 to 69 years) recording diet by means of a pre-coded food diary over 4 d and a 4 d...

  11. Development of Off-take Model, Subcooled Boiling Model, and Radiation Heat Transfer Input Model into the MARS Code for a Regulatory Auditing of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, C.; Rhee, B. W.; Chung, B. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahn, S. H.; Kim, M. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2009-05-15

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to a lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use a vendor's code for a regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed the RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of the existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of the RELAP5/MOD3/CANDU code to the MARS code including a quality assurance of the developed models.

  12. Development of Off-take Model, Subcooled Boiling Model, and Radiation Heat Transfer Input Model into the MARS Code for a Regulatory Auditing of CANDU Reactors

    International Nuclear Information System (INIS)

    Yoon, C.; Rhee, B. W.; Chung, B. D.; Ahn, S. H.; Kim, M. W.

    2009-01-01

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to a lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use a vendor's code for a regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed the RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of the existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of the RELAP5/MOD3/CANDU code to the MARS code including a quality assurance of the developed models

  13. Methods, Devices and Computer Program Products Providing for Establishing a Model for Emulating a Physical Quantity Which Depends on at Least One Input Parameter, and Use Thereof

    DEFF Research Database (Denmark)

    2014-01-01

    The present invention proposes methods, devices and computer program products. To this extent, there is defined a set X including N distinct parameter values x_i for at least one input parameter x, N being an integer greater than or equal to 1, first measured the physical quantity Pm1 for each...

  14. Introduction to the simulation with MCNP Monte Carlo code and its applications in Medical Physics

    International Nuclear Information System (INIS)

    Parreno Z, F.; Paucar J, R.; Picon C, C.

    1998-01-01

    The simulation by Monte Carlo is tool which Medical Physics counts with it for the development of its research, the interest by this tool is growing, as we may observe in the main scientific journals for the years 1995-1997 where more than 27 % of the papers treat over Monte Carlo and/or its applications in the radiation transport.In the Peruvian Institute of Nuclear Energy we are implementing and making use of the MCNP4 and EGS4 codes. In this work are presented the general features of the Monte Carlo method and its more useful applications in Medical Physics. Likewise, it is made a simulation of the calculation of isodose curves in an interstitial treatment with Ir-192 wires in a mammary gland carcinoma. (Author)

  15. GENII-LIN: a Multipurpose Health Physics Code Built on GENII-1.485

    Directory of Open Access Journals (Sweden)

    Marco Sumini

    2006-10-01

    Full Text Available The aim of the GENII-LIN project was to develop an open source multipurpose health physics code running on Linux platform, for calculating radiation dose and risk from radionuclides released to the environment. The general features of the GENII-LIN system include [1] capabilities for calculating radiation dose both for acute and chronic releases, with options for annual dose, committed dose and accumulated dose [2] capabilities for evaluating exposure pathways including direct exposure via water (swimming, boating, fishing, soil (buried and surface sources and air (semi-infinite cloud and finite cloud model, inhalation pathways and ingestion pathways. The release scenarios considered are: - acute release to air, from ground level or elevated sources, or to water; - chronic release to air, from ground level or elevated sources, or to water; - initial contamination of soil or surfaces. Keywords: radiation protection, Linux, health physics, risk analysis.

  16. A statistical–mechanical view on source coding: physical compression and data compression

    International Nuclear Information System (INIS)

    Merhav, Neri

    2011-01-01

    We draw a certain analogy between the classical information-theoretic problem of lossy data compression (source coding) of memoryless information sources and the statistical–mechanical behavior of a certain model of a chain of connected particles (e.g. a polymer) that is subjected to a contracting force. The free energy difference pertaining to such a contraction turns out to be proportional to the rate-distortion function in the analogous data compression model, and the contracting force is proportional to the derivative of this function. Beyond the fact that this analogy may be interesting in its own right, it may provide a physical perspective on the behavior of optimum schemes for lossy data compression (and perhaps also an information-theoretic perspective on certain physical system models). Moreover, it triggers the derivation of lossy compression performance for systems with memory, using analysis tools and insights from statistical mechanics

  17. Benchmarking of epithermal methods in the lattice-physics code EPRI-CELL

    International Nuclear Information System (INIS)

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

    1982-01-01

    The epithermal cross section shielding methods used in the lattice physics code EPRI-CELL (E-C) have been extensively studied to determine its major approximations and to examine the sensitivity of computed results to these approximations. The study has resulted in several improvements in the original methodology. These include: treatment of the external moderator source with intermediate resonance (IR) theory, development of a new Dancoff factor expression to account for clad interactions, development of a new method for treating resonance interference, and application of a generalized least squares method to compute best-estimate values for the Bell factor and group-dependent IR parameters. The modified E-C code with its new ENDF/B-V cross section library is tested for several numerical benchmark problems. Integral parameters computed by EC are compared with those obtained with point-cross section Monte Carlo calculations, and E-C fine group cross sections are benchmarked against point-cross section descrete ordinates calculations. It is found that the code modifications improve agreement between E-C and the more sophisticated methods. E-C shows excellent agreement on the integral parameters and usually agrees within a few percent on fine-group, shielded cross sections

  18. Application of regional physically-based landslide early warning model: tuning of the input parameters and validation of the results

    Science.gov (United States)

    D'Ambrosio, Michele; Tofani, Veronica; Rossi, Guglielmo; Salvatici, Teresa; Tacconi Stefanelli, Carlo; Rosi, Ascanio; Benedetta Masi, Elena; Pazzi, Veronica; Vannocci, Pietro; Catani, Filippo; Casagli, Nicola

    2017-04-01

    The Aosta Valley region is located in North-West Alpine mountain chain. The geomorphology of the region is characterized by steep slopes, high climatic and altitude (ranging from 400 m a.s.l of Dora Baltea's river floodplain to 4810 m a.s.l. of Mont Blanc) variability. In the study area (zone B), located in Eastern part of Aosta Valley, heavy rainfall of about 800-900 mm per year is the main landslides trigger. These features lead to a high hydrogeological risk in all territory, as mass movements interest the 70% of the municipality areas (mainly shallow rapid landslides and rock falls). An in-depth study of the geotechnical and hydrological properties of hillslopes controlling shallow landslides formation was conducted, with the aim to improve the reliability of deterministic model, named HIRESS (HIgh REsolution Stability Simulator). In particular, two campaigns of on site measurements and laboratory experiments were performed. The data obtained have been studied in order to assess the relationships existing among the different parameters and the bedrock lithology. The analyzed soils in 12 survey points are mainly composed of sand and gravel, with highly variable contents of silt. The range of effective internal friction angle (from 25.6° to 34.3°) and effective cohesion (from 0 kPa to 9.3 kPa) measured and the median ks (10E-6 m/s) value are consistent with the average grain sizes (gravelly sand). The data collected contributes to generate input map of parameters for HIRESS (static data). More static data are: volume weight, residual water content, porosity and grain size index. In order to improve the original formulation of the model, the contribution of the root cohesion has been also taken into account based on the vegetation map and literature values. HIRESS is a physically based distributed slope stability simulator for analyzing shallow landslide triggering conditions in real time and in large areas using parallel computational techniques. The software

  19. Generation of initial geometries for the simulation of the physical system in the DualPHYsics code; Generacion de geometrias iniciales para la simulacion del sistema fisico en el codigo DualSPHysics

    Energy Technology Data Exchange (ETDEWEB)

    Segura Q, E.

    2013-07-01

    In the diverse research areas of the Instituto Nacional de Investigaciones Nucleares (ININ) are different activities related to science and technology, one of great interest is the study and treatment of the collection and storage of radioactive waste. Therefore at ININ the draft on the simulation of the pollutants diffusion in the soil through a porous medium (third stage) has this problem inherent aspects, hence a need for such a situation is to generate the initial geometry of the physical system For the realization of the simulation method is implemented smoothed particle hydrodynamics (SPH). This method runs in DualSPHysics code, which has great versatility and ability to simulate phenomena of any physical system where hydrodynamic aspects combine. In order to simulate a physical system DualSPHysics code, you need to preset the initial geometry of the system of interest, then this is included in the input file of the code. The simulation sets the initial geometry through regular geometric bodies positioned at different points in space. This was done through a programming language (Fortran, C + +, Java, etc..). This methodology will provide the basis to simulate more complex geometries future positions and form. (Author)

  20. A Development of a System Enables Character Input and PC Operation via Voice for a Physically Disabled Person with a Speech Impediment

    Science.gov (United States)

    Tanioka, Toshimasa; Egashira, Hiroyuki; Takata, Mayumi; Okazaki, Yasuhisa; Watanabe, Kenzi; Kondo, Hiroki

    We have designed and implemented a PC operation support system for a physically disabled person with a speech impediment via voice. Voice operation is an effective method for a physically disabled person with involuntary movement of the limbs and the head. We have applied a commercial speech recognition engine to develop our system for practical purposes. Adoption of a commercial engine reduces development cost and will contribute to make our system useful to another speech impediment people. We have customized commercial speech recognition engine so that it can recognize the utterance of a person with a speech impediment. We have restricted the words that the recognition engine recognizes and separated a target words from similar words in pronunciation to avoid misrecognition. Huge number of words registered in commercial speech recognition engines cause frequent misrecognition for speech impediments' utterance, because their utterance is not clear and unstable. We have solved this problem by narrowing the choice of input down in a small number and also by registering their ambiguous pronunciations in addition to the original ones. To realize all character inputs and all PC operation with a small number of words, we have designed multiple input modes with categorized dictionaries and have introduced two-step input in each mode except numeral input to enable correct operation with small number of words. The system we have developed is in practical level. The first author of this paper is physically disabled with a speech impediment. He has been able not only character input into PC but also to operate Windows system smoothly by using this system. He uses this system in his daily life. This paper is written by him with this system. At present, the speech recognition is customized to him. It is, however, possible to customize for other users by changing words and registering new pronunciation according to each user's utterance.

  1. Catalogue of nuclear fusion codes - 1976

    International Nuclear Information System (INIS)

    1976-10-01

    A catalogue is presented of the computer codes in nuclear fusion research developed by JAERI, Division of Thermonuclear Fusion Research and Division of Large Tokamak Development in particular. It contains a total of about 100 codes under the categories: Atomic Process, Data Handling, Experimental Data Processing, Engineering, Input and Output, Special Languages and Their Application, Mathematical Programming, Miscellaneous, Numerical Analysis, Nuclear Physics, Plasma Physics and Fusion Research, Plasma Simulation and Numerical Technique, Reactor Design, Solid State Physics, Statistics, and System Program. (auth.)

  2. Physical Processes and Applications of the Monte Carlo Radiative Energy Deposition (MRED) Code

    Science.gov (United States)

    Reed, Robert A.; Weller, Robert A.; Mendenhall, Marcus H.; Fleetwood, Daniel M.; Warren, Kevin M.; Sierawski, Brian D.; King, Michael P.; Schrimpf, Ronald D.; Auden, Elizabeth C.

    2015-08-01

    MRED is a Python-language scriptable computer application that simulates radiation transport. It is the computational engine for the on-line tool CRÈME-MC. MRED is based on c++ code from Geant4 with additional Fortran components to simulate electron transport and nuclear reactions with high precision. We provide a detailed description of the structure of MRED and the implementation of the simulation of physical processes used to simulate radiation effects in electronic devices and circuits. Extensive discussion and references are provided that illustrate the validation of models used to implement specific simulations of relevant physical processes. Several applications of MRED are summarized that demonstrate its ability to predict and describe basic physical phenomena associated with irradiation of electronic circuits and devices. These include effects from single particle radiation (including both direct ionization and indirect ionization effects), dose enhancement effects, and displacement damage effects. MRED simulations have also helped to identify new single event upset mechanisms not previously observed by experiment, but since confirmed, including upsets due to muons and energetic electrons.

  3. Code-Hopping Based Transmission Scheme for Wireless Physical-Layer Security

    Directory of Open Access Journals (Sweden)

    Liuguo Yin

    2018-01-01

    Full Text Available Due to the broadcast and time-varying natures of wireless channels, traditional communication systems that provide data encryption at the application layer suffer many challenges such as error diffusion. In this paper, we propose a code-hopping based secrecy transmission scheme that uses dynamic nonsystematic low-density parity-check (LDPC codes and automatic repeat-request (ARQ mechanism to jointly encode and encrypt source messages at the physical layer. In this scheme, secret keys at the transmitter and the legitimate receiver are generated dynamically upon the source messages that have been transmitted successfully. During the transmission, each source message is jointly encoded and encrypted by a parity-check matrix, which is dynamically selected from a set of LDPC matrices based on the shared dynamic secret key. As for the eavesdropper (Eve, the uncorrectable decoding errors prevent her from generating the same secret key as the legitimate parties. Thus she cannot select the correct LDPC matrix to recover the source message. We demonstrate that our scheme can be compatible with traditional cryptosystems and enhance the security without sacrificing the error-correction performance. Numerical results show that the bit error rate (BER of Eve approaches 0.5 as the number of transmitted source messages increases and the security gap of the system is small.

  4. Methods tuned on the physical problem. A way to improve numerical codes

    International Nuclear Information System (INIS)

    Ixaru, L.Gr.

    2010-01-01

    We consider the problem on how the numerical methods tuned on the physical problem can contribute to the enhancement of the performance of the codes. We illustrate this on two simple cases: solution of time independent one-dimensional Schroedinger equation, and the computation of integrals with oscillatory integrands. In both cases the tuned versions bring a massive gain in accuracy at negligible extra cost. We presented two simple problems where successive levels of tuning enhance significantly the accuracy at negligible extra cost. These problems should be seen as representing only some illustrations on how the codes can be improved but we must also mention that in many cases tuned versions still have to be developed. Just for a suggestion, quadrature formulae which involve the integrand and a number of successive derivatives of this exist, but no formula is available when some of these derivatives are missing, for example when we dispose of y and y'' but not of y'. A direct application will be on the case when the integrand involves the solution of the Schrodinger equation by the method of Numerov. (author)

  5. Joint Power Allocation for Multicast Systems with Physical-Layer Network Coding

    Directory of Open Access Journals (Sweden)

    Zhu Wei-Ping

    2010-01-01

    Full Text Available This paper addresses the joint power allocation issue in physical-layer network coding (PLNC of multicast systems with two sources and two destinations communicating via a large number of distributed relays. By maximizing the achievable system rate, a constrained optimization problem is first formulated to jointly allocate powers for the source and relay terminals. Due to the nonconvex nature of the cost function, an iterative algorithm with guaranteed convergence is developed to solve the joint power allocation problem. As an alternative, an upper bound of the achievable rate is also derived to modify the original cost function in order to obtain a convex optimization solution. This approximation is shown to be asymptotically optimal in the sense of maximizing the achievable rate. It is confirmed through Monte Carlo simulations that the proposed joint power allocation schemes are superior to the existing schemes in terms of achievable rate and cumulative distribution function (CDF.

  6. The physics of compensating calorimetry and the new CALOR89 code system

    International Nuclear Information System (INIS)

    Gabriel, T.A.; Brau, J.E.; Bishop, B.L.

    1989-03-01

    Much of the understanding of the physics of calorimetry has come from the use of excellent radiation transport codes. A new understanding of compensating calorimetry was introduced four years ago following detailed studies with a new CALOR system. Now, the CALOR system has again been revised to reflect a better comprehension of high energy nuclear collisions by incorporating a modified high energy fragmentation model from FLUKA87. This revision will allow for the accurate analysis of calorimeters at energies of 100's of GeV. Presented in this paper is a discussion of compensating calorimetry, the new CALOR system, the revisions to HETC, and recently generated calorimeter related data on modes of energy deposition and secondary neutron production (E < 50 MeV) in infinite iron and uranium blocks. 38 refs., 5 figs., 5 tabs

  7. Hotspot health physics codes used as a tool for managing excess risk on radiological emergencies

    International Nuclear Information System (INIS)

    Andrade, Edson Ramos de; Alves, Nelson Mendes; Rocha, Joao B.T.; Cruz, Ivana B. Manica da; Santos, Greice F. Fey dos; Machado, Michel Mansur; Rossato, Veronica Venturini; Bauermann, Liliane Freitas

    2008-01-01

    This work is aimed to use the Hotspot Health Physics codes in acute mode in order to estimate the immediate radiological impact associated with high acute radiation doses, which is applied to special target organs such as lung, small intestine wall, and red bone marrow. Organic compounds such as Diphenyl Diselenide (C 6 H 5 Se 2 C 6 H 5 ) and Ebselen (C 13 H 9 NOSe), an antioxidants selenium containing compounds, were used over irradiated phospholipids extracted from chicken yolk eggs, in vitro in order to reduce lipo-peroxidation. Experimental data were measured by Thiobarbituric Acid Reactive Substance (TBARS) assay which is able to measure the production of oxidative stress in the sample. Experimental data were extrapolated and applied as a reduction factors over equations for cancer excess risk calculation from BEIR V, for helping the decisonmaking process on Radiological Emergency Scenarios. (author)

  8. Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes

    International Nuclear Information System (INIS)

    Lindley, B.A.; Lillington, J.N.; Kotlyar, D.; Parks, G.T.; Petrovic, B.

    2016-01-01

    The majority of nuclear reactors operating in the world today and similarly the majority of near-term new build reactors will be LWRs. These currently accommodate traditional Zr clad UO_2/PuO_2 fuel designs which have an excellent performance record for normal operation. However, the events at Fukushima culminated in significant hydrogen production and hydrogen explosions, resulting from high temperature Zr/steam interaction following core uncovering for an extended period. These events have resulted in increased emphasis towards developing more accident tolerant fuels (ATFs)-clad systems, particularly for current and near-term build LWRs. R and D programmes are underway in the US and elsewhere to develop ATFs and the UK is engaging in these international programmes. Candidate advanced fuel materials include uranium nitride (UN) and uranium silicide (U_3Si_2). Candidate cladding materials include advanced stainless steel (FeCrAl) and silicon carbide. The UK has a long history in industrial fuel manufacture and fabrication for a wide range of reactor systems including LWRs. This is supported by a national infrastructure to perform experimental and theoretical R and D in fuel performance, fuel transient behaviour and reactor physics. In this paper, an analysis of the Integral Inherently Safe LWR design (I"2S-LWR), a reactor concept developed by an international collaboration led by the Georgia Institute of Technology, within a US DOE Nuclear Energy University Program (NEUP) Integrated Research Project (IRP) is considered. The analysis is performed using the ANSWERS reactor physics code WIMS and the EDF Energy core simulator PANTHER by researchers at the University of Cambridge. The I"2S-LWR is an advanced 2850 MWt integral PWR with inherent safety features. In order to enhance the safety features, the baseline fuel and cladding materials that were chosen for the I"2S-LWR design are U_3Si_2 and advanced stainless steel respectively. In addition, the I"2S-LWR design

  9. Synthesis of the scientific French speaking days on numerical codes in radiation protection, in radio physics and in dosimetry

    International Nuclear Information System (INIS)

    Paul, D.; Makovicka, L.; Ricard, M.

    2005-01-01

    Synthesis of the scientific French speaking days on numerical codes in radiation protection, in radio-physics and in dosimetry. The paper carries the title of 'French speaking' scientific days co-organized on October 2-3, 2003 in Sochaux by the SFRP, SFPM and FIRAM societies. It has for objective to establish the scientific balance sheet of this international event, to give the synthesis of current tendencies in the field of the development and of the use of the numerical codes in radiation protection, in radio-physics and in dosimetry. (author)

  10. Development of Legal Norms on Marriage and Divorce in Cambodia : The Civil Code Between Foreign Inputs and Local Growth (2)

    OpenAIRE

    KUONG, Teilee

    2016-01-01

    This second part of the research focuses on the development of law on divorce in Cambodia, aiming at identifying the continuation and changes in the legislative development of Cambodia in regulating divorce and post-divorce relationship. It starts with a brief overview of the legislative development from the early 20th century up till the latest codification of the 2007 Civil Code. Along this line of historical narratives, the research then examines two important elements in the legal effects...

  11. Apar-T: code, validation, and physical interpretation of particle-in-cell results

    Science.gov (United States)

    Melzani, Mickaël; Winisdoerffer, Christophe; Walder, Rolf; Folini, Doris; Favre, Jean M.; Krastanov, Stefan; Messmer, Peter

    2013-10-01

    We present the parallel particle-in-cell (PIC) code Apar-T and, more importantly, address the fundamental question of the relations between the PIC model, the Vlasov-Maxwell theory, and real plasmas. First, we present four validation tests: spectra from simulations of thermal plasmas, linear growth rates of the relativistic tearing instability and of the filamentation instability, and nonlinear filamentation merging phase. For the filamentation instability we show that the effective growth rates measured on the total energy can differ by more than 50% from the linear cold predictions and from the fastest modes of the simulation. We link these discrepancies to the superparticle number per cell and to the level of field fluctuations. Second, we detail a new method for initial loading of Maxwell-Jüttner particle distributions with relativistic bulk velocity and relativistic temperature, and explain why the traditional method with individual particle boosting fails. The formulation of the relativistic Harris equilibrium is generalized to arbitrary temperature and mass ratios. Both are required for the tearing instability setup. Third, we turn to the key point of this paper and scrutinize the question of what description of (weakly coupled) physical plasmas is obtained by PIC models. These models rely on two building blocks: coarse-graining, i.e., grouping of the order of p ~ 1010 real particles into a single computer superparticle, and field storage on a grid with its subsequent finite superparticle size. We introduce the notion of coarse-graining dependent quantities, i.e., quantities depending on p. They derive from the PIC plasma parameter ΛPIC, which we show to behave as ΛPIC ∝ 1/p. We explore two important implications. One is that PIC collision- and fluctuation-induced thermalization times are expected to scale with the number of superparticles per grid cell, and thus to be a factor p ~ 1010 smaller than in real plasmas, a fact that we confirm with

  12. Experimental benchmark of non-local-thermodynamic-equilibrium plasma atomic physics codes; Validation experimentale des codes de physique atomique des plasmas hors equilibre thermodynamique local

    Energy Technology Data Exchange (ETDEWEB)

    Nagels-Silvert, V

    2004-09-15

    The main purpose of this thesis is to get experimental data for the testing and validation of atomic physics codes dealing with non-local-thermodynamical-equilibrium plasmas. The first part is dedicated to the spectroscopic study of xenon and krypton plasmas that have been produced by a nanosecond laser pulse interacting with a gas jet. A Thomson scattering diagnostic has allowed us to measure independently plasma parameters such as electron temperature, electron density and the average ionisation state. We have obtained time integrated spectra in the range between 5 and 10 angstroms. We have identified about one hundred xenon rays between 8.6 and 9.6 angstroms via the use of the Relac code. We have discovered unknown rays for the krypton between 5.2 and 7.5 angstroms. In a second experiment we have extended the wavelength range to the X UV domain. The Averroes/Transpec code has been tested in the ranges from 9 to 15 angstroms and from 10 to 130 angstroms, the first range has been well reproduced while the second range requires a more complex data analysis. The second part is dedicated to the spectroscopic study of aluminium, selenium and samarium plasmas in femtosecond operating rate. We have designed an interferometry diagnostic in the frequency domain that has allowed us to measure the expanding speed of the target's backside. Via the use of an adequate isothermal model this parameter has led us to know the plasma electron temperature. Spectra and emission times of various rays from the aluminium and selenium plasmas have been computed satisfactorily with the Averroes/Transpec code coupled with Film and Multif hydrodynamical codes. (A.C.)

  13. Experimental benchmark of non-local-thermodynamic-equilibrium plasma atomic physics codes; Validation experimentale des codes de physique atomique des plasmas hors equilibre thermodynamique local

    Energy Technology Data Exchange (ETDEWEB)

    Nagels-Silvert, V

    2004-09-15

    The main purpose of this thesis is to get experimental data for the testing and validation of atomic physics codes dealing with non-local-thermodynamical-equilibrium plasmas. The first part is dedicated to the spectroscopic study of xenon and krypton plasmas that have been produced by a nanosecond laser pulse interacting with a gas jet. A Thomson scattering diagnostic has allowed us to measure independently plasma parameters such as electron temperature, electron density and the average ionisation state. We have obtained time integrated spectra in the range between 5 and 10 angstroms. We have identified about one hundred xenon rays between 8.6 and 9.6 angstroms via the use of the Relac code. We have discovered unknown rays for the krypton between 5.2 and 7.5 angstroms. In a second experiment we have extended the wavelength range to the X UV domain. The Averroes/Transpec code has been tested in the ranges from 9 to 15 angstroms and from 10 to 130 angstroms, the first range has been well reproduced while the second range requires a more complex data analysis. The second part is dedicated to the spectroscopic study of aluminium, selenium and samarium plasmas in femtosecond operating rate. We have designed an interferometry diagnostic in the frequency domain that has allowed us to measure the expanding speed of the target's backside. Via the use of an adequate isothermal model this parameter has led us to know the plasma electron temperature. Spectra and emission times of various rays from the aluminium and selenium plasmas have been computed satisfactorily with the Averroes/Transpec code coupled with Film and Multif hydrodynamical codes. (A.C.)

  14. Fuel Management Study for a CANDU reactor Using New Physics Codes Suite

    International Nuclear Information System (INIS)

    Kim, Won Young; Kim, Bong Ghi; Park, Joo Hwan

    2008-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. The primary reactivity control in a CANDU reactor is the on-power refueling on a daily basis and an additional reactivity control is provided through an individual reactivity device movement, which includes 21 adjusters, 6 liquid zone controllers, 4 mechanical control absorbers and 2 shutdown systems. The refueling in CANDU is carried out on power and this makes the in-core fuel management different from that in a reactor refueled during shutdowns. The objective of a fuel management is to determine a fuel loading and fuel replacement procedure which will result in a minimum total unit energy cost in a safe and reliable operation. In this article, the in-core fuel management for the CANDU reactor was studied by using the new physics code suite of WIMS-IST/DRAGON-IST/RFSP-IST with the model of Wolsong-1 NPP

  15. Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N. [Moscow Power Engineering Institute (Technical University), Moscow (Russian Federation)

    2007-07-01

    The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes.

  16. Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N.

    2007-01-01

    The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes

  17. A multi-physics code system based on ANC9, VIPRE-W and BOA for CIPS evaluation

    International Nuclear Information System (INIS)

    Zhang, B.; Sung, Y.; Secker, J.; Beard, C.; Hilton, P.; Wang, G.; Oelrich, R.; Karoutas, Z.; Sung, Y.

    2011-01-01

    This paper summarizes the development of a multi-physics code system for evaluation of Crud Induced Power Shift (CIPS) phenomenon experienced in some Pressurized Water Reactors (PWR). CIPS is an unexpected change in reactor core axial power distribution, caused by boron compounds in crud deposited in the high power fuel assemblies undergoing subcooled boiling. As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL) sponsored by the US Department of Energy (DOE), this paper describes the initial linkage and application of a multi-physics code system ANC9/VIPRE-W/BOA for evaluating changes in core power distributions due to boron deposited in crud. The initial linkage of the code system along with the application results will be the base for the future CASL development. (author)

  18. A multi-physics code system based on ANC9, VIPRE-W and BOA for CIPS evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, B.; Sung, Y.; Secker, J.; Beard, C.; Hilton, P.; Wang, G.; Oelrich, R.; Karoutas, Z.; Sung, Y. [Westinghouse Electric Company LLC, Pittsburgh (United States)

    2011-07-01

    This paper summarizes the development of a multi-physics code system for evaluation of Crud Induced Power Shift (CIPS) phenomenon experienced in some Pressurized Water Reactors (PWR). CIPS is an unexpected change in reactor core axial power distribution, caused by boron compounds in crud deposited in the high power fuel assemblies undergoing subcooled boiling. As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL) sponsored by the US Department of Energy (DOE), this paper describes the initial linkage and application of a multi-physics code system ANC9/VIPRE-W/BOA for evaluating changes in core power distributions due to boron deposited in crud. The initial linkage of the code system along with the application results will be the base for the future CASL development. (author)

  19. V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

    International Nuclear Information System (INIS)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

    2010-07-01

    V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

  20. V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

    Energy Technology Data Exchange (ETDEWEB)

    Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

    2010-07-15

    V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

  1. Gestures and multimodal input

    OpenAIRE

    Keates, Simeon; Robinson, Peter

    1999-01-01

    For users with motion impairments, the standard keyboard and mouse arrangement for computer access often presents problems. Other approaches have to be adopted to overcome this. In this paper, we will describe the development of a prototype multimodal input system based on two gestural input channels. Results from extensive user trials of this system are presented. These trials showed that the physical and cognitive loads on the user can quickly become excessive and detrimental to the interac...

  2. Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin Allen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, Sebastian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gleicher, Frederick Nathan [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, Mark David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition, this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.

  3. Applying physical input-output tables of energy to estimate the energy ecological footprint (EEF) of Galicia (NW Spain)

    International Nuclear Information System (INIS)

    Carballo Penela, Adolfo; Sebastian Villasante, Carlos

    2008-01-01

    Nowadays, the achievement of sustainable development constitutes an important constraint in the design of energy policies, being necessary the development of reliable indicators to obtain helpful information about the use of energy resources. The ecological footprint (EF) provides a referential framework for the analysis of human demand for bioproductivity, including energy issues. In this article, the theoretical bases of the footprint analysis are described by applying input-output tables of energy to estimate the Galician energy ecological footprint (EEF). It is concluded that the location of highly polluting industries in Galicia makes the Galician EEF quite higher than more developed regions of Spain. The relevance of the outer component of the Galician EEF is also studied. First, available information seems to indicate that the energy incorporated to the trading of manufactured goods would notably increase the Galician consumption of energy. On the other hand, the inclusion of electricity trade in the EEF analysis, including an adjustment, following the same philosophy as with manufactured goods is proposed. This adjustment would substantially reduce the Galician EEF, as the exported electricity widely exceeds the imported one

  4. Code development of total sensitivity and uncertainty analysis for reactor physics calculations

    International Nuclear Information System (INIS)

    Wan, C.; Cao, L.; Wu, H.; Zu, T.; Shen, W.

    2015-01-01

    Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)

  5. Code development of total sensitivity and uncertainty analysis for reactor physics calculations

    Energy Technology Data Exchange (ETDEWEB)

    Wan, C.; Cao, L.; Wu, H.; Zu, T., E-mail: chenghuiwan@stu.xjtu.edu.cn, E-mail: caolz@mail.xjtu.edu.cn, E-mail: hongchun@mail.xjtu.edu.cn, E-mail: tiejun@mail.xjtu.edu.cn [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Shen, W., E-mail: Wei.Shen@cnsc-ccsn.gc.ca [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

    2015-07-01

    Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)

  6. Effects of pulse frequency of input power on the physical and chemical properties of pulsed streamer discharge plasmas in water

    Czech Academy of Sciences Publication Activity Database

    Ruma, Ruma.; Lukeš, Petr; Aoki, N.; Doležalová, Eva; Hosseini, S.H.R.; Sakugawa, T.; Akiyama, H.

    2013-01-01

    Roč. 46, č. 12 (2013), s. 125202-125202 ISSN 0022-3727 R&D Projects: GA ČR(CZ) GD104/09/H080 Grant - others:Rada Programu interní podpory projektů mezinárodní spolupráce AV ČR(CZ) M100431203 Program:M Institutional support: RVO:61389021 Keywords : discharge in water * pulsed power * pulse frequency * hydrogen peroxide * organic dye * bacteria * generator * liquids Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.521, year: 2013 http://dx.doi.org/10.1088/0022-3727/46/12/125202

  7. Modeling the ionosphere-thermosphere response to a geomagnetic storm using physics-based magnetospheric energy input: OpenGGCM-CTIM results

    Directory of Open Access Journals (Sweden)

    Connor Hyunju Kim

    2016-01-01

    Full Text Available The magnetosphere is a major source of energy for the Earth’s ionosphere and thermosphere (IT system. Current IT models drive the upper atmosphere using empirically calculated magnetospheric energy input. Thus, they do not sufficiently capture the storm-time dynamics, particularly at high latitudes. To improve the prediction capability of IT models, a physics-based magnetospheric input is necessary. Here, we use the Open Global General Circulation Model (OpenGGCM coupled with the Coupled Thermosphere Ionosphere Model (CTIM. OpenGGCM calculates a three-dimensional global magnetosphere and a two-dimensional high-latitude ionosphere by solving resistive magnetohydrodynamic (MHD equations with solar wind input. CTIM calculates a global thermosphere and a high-latitude ionosphere in three dimensions using realistic magnetospheric inputs from the OpenGGCM. We investigate whether the coupled model improves the storm-time IT responses by simulating a geomagnetic storm that is preceded by a strong solar wind pressure front on August 24, 2005. We compare the OpenGGCM-CTIM results with low-earth-orbit satellite observations and with the model results of Coupled Thermosphere-Ionosphere-Plasmasphere electrodynamics (CTIPe. CTIPe is an up-to-date version of CTIM that incorporates more IT dynamics such as a low-latitude ionosphere and a plasmasphere, but uses empirical magnetospheric input. OpenGGCM-CTIM reproduces localized neutral density peaks at ~ 400 km altitude in the high-latitude dayside regions in agreement with in situ observations during the pressure shock and the early phase of the storm. Although CTIPe is in some sense a much superior model than CTIM, it misses these localized enhancements. Unlike the CTIPe empirical input models, OpenGGCM-CTIM more faithfully produces localized increases of both auroral precipitation and ionospheric electric fields near the high-latitude dayside region after the pressure shock and after the storm onset

  8. Optimization of reload of nuclear power plants using ACO together with the GENES reactor physics code

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto, E-mail: alan@lmp.ufrj.br, E-mail: andressa@lmp.ufrj.br, E-mail: schirru@lmp.ufrj.br, E-mail: ffreire@eletronuclear.gov.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil)

    2017-11-01

    The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10{sup 13} combinations and 10{sup 11} great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)

  9. Optimization of reload of nuclear power plants using ACO together with the GENES reactor physics code

    International Nuclear Information System (INIS)

    Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto

    2017-01-01

    The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10"1"3 combinations and 10"1"1 great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)

  10. Applications of the lots computer code to laser fusion systems and other physical optics problems

    International Nuclear Information System (INIS)

    Lawrence, G.; Wolfe, P.N.

    1979-01-01

    The Laser Optical Train Simulation (LOTS) code has been developed at the Optical Sciences Center, University of Arizona under contract to Los Alamos Scientific Laboratory (LASL). LOTS is a diffraction based code designed to beam quality and energy of the laser fusion system in an end-to-end calculation

  11. The observation of early childhood physical aggression: A psychometric study of the system for coding early physical aggression (SCEPA)

    NARCIS (Netherlands)

    Mesman, J.; Alink, L.R.A.; van Zeijl, J.; Stolk, M.N.; Bakermans-Kranenburg, M.J.; van IJzendoorn, M.H.; Juffer, F.; Koot, H.M.

    2008-01-01

    We investigated the reliability and (convergent and discriminant) validity of an observational measure of physical aggression in toddlers and preschoolers, originally developed by Keenan and Shaw [1994]. The observation instrument is based on a developmental definition of aggression. Physical

  12. The revised APTA code of ethics for the physical therapist and standards of ethical conduct for the physical therapist assistant: theory, purpose, process, and significance.

    Science.gov (United States)

    Swisher, Laura Lee; Hiller, Peggy

    2010-05-01

    In June 2009, the House of Delegates (HOD) of the American Physical Therapy Association (APTA) passed a major revision of the APTA Code of Ethics for physical therapists and the Standards of Ethical Conduct for the Physical Therapist Assistant. The revised documents will be effective July 1, 2010. The purposes of this article are: (1) to provide a historical, professional, and theoretical context for this important revision; (2) to describe the 4-year revision process; (3) to examine major features of the documents; and (4) to discuss the significance of the revisions from the perspective of the maturation of physical therapy as a doctoring profession. PROCESS OF REVISION: The process for revision is delineated within the context of history and the Bylaws of APTA. FORMAT, STRUCTURE, AND CONTENT OF REVISED CORE ETHICS DOCUMENTS: The revised documents represent a significant change in format, level of detail, and scope of application. Previous APTA Codes of Ethics and Standards of Ethical Conduct for the Physical Therapist Assistant have delineated very broad general principles, with specific obligations spelled out in the Ethics and Judicial Committee's Guide for Professional Conduct and Guide for Conduct of the Physical Therapist Assistant. In contrast to the current documents, the revised documents address all 5 roles of the physical therapist, delineate ethical obligations in organizational and business contexts, and align with the tenets of Vision 2020. The significance of this revision is discussed within historical parameters, the implications for physical therapists and physical therapist assistants, the maturation of the profession, societal accountability and moral community, potential regulatory implications, and the inclusive and deliberative process of moral dialogue by which changes were developed, revised, and approved.

  13. Recent Developments of JAEA's Monte Carlo Code MVP for Reactor Physics Applications

    Science.gov (United States)

    Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa

    2014-06-01

    This paper describes the recent development status of a Monte Carlo code MVP developed at Japan Atomic Energy Agency. The basic features and capabilities of MVP are overviewed. In addition, new capabilities useful for reactor analysis are also described.

  14. Benchmark of physics design of a proposed 30 MW Multi Purpose Research Reactor using a Monte Carlo code MCNP

    International Nuclear Information System (INIS)

    Singh, Tej; Kumar, Jainendra; Sharma, Archana; Singh, Kanchhi; Raina, V.K.; Srinivasan, P.

    2009-01-01

    At present Dhruva and Cirus reactors provide majority of research reactor based experimental/irradiation facilities to cater to various needs of the vast pool of researchers in the field of sciences research and development work for nuclear power plants and production of radioisotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 30 MWt Multi Purpose Research Reactor is proposed to be constructed. This paper describes some of the physics design features of this reactor using MCNP code to validate the deterministic methods. The criticality calculations for 100 material testing reactor (JHR) of France and 610 MW SAVANNAH thermal reactor were performed using MCNP computer codes to boost the confidence level in designing the physics design of reactor core. (author)

  15. FLUTAN input specifications

    International Nuclear Information System (INIS)

    Borgwaldt, H.; Baumann, W.; Willerding, G.

    1991-05-01

    FLUTAN is a highly vectorized computer code for 3-D fluiddynamic and thermal-hydraulic analyses in cartesian and cylinder coordinates. It is related to the family of COMMIX codes originally developed at Argonne National Laboratory, USA. To a large extent, FLUTAN relies on basic concepts and structures imported from COMMIX-1B and COMMIX-2 which were made available to KfK in the frame of cooperation contracts in the fast reactor safety field. While on the one hand not all features of the original COMMIX versions have been implemented in FLUTAN, the code on the other hand includes some essential innovative options like CRESOR solution algorithm, general 3-dimensional rebalacing scheme for solving the pressure equation, and LECUSSO-QUICK-FRAM techniques suitable for reducing 'numerical diffusion' in both the enthalphy and momentum equations. This report provides users with detailed input instructions, presents formulations of the various model options, and explains by means of comprehensive sample input, how to use the code. (orig.) [de

  16. Recent developments of JAEA’s Monte Carlo code MVP for reactor physics applications

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa

    2015-01-01

    Highlights: • This paper describes the recent development status of the Monte Carlo code MVP. • The basic features and capabilities of MVP are briefly described. • New capabilities useful for reactor analysis are also described. - Abstract: This paper describes the recent development status of a Monte Carlo code MVP developed at Japan Atomic Energy Agency. The basic features and capabilities of MVP are overviewed. In addition, new capabilities useful for reactor analysis are also described

  17. The Tile-map Based Vulnerability Assessment Code of a Physical Protection System: SAPE (Systematic Analysis of Protection Effectiveness)

    International Nuclear Information System (INIS)

    Jang, Sung Soon; Kwak, Sung Woo; Yoo, Ho Sik; Kim, Jung Soo; Yoon, Wan Ki

    2008-01-01

    Increasing threats on nuclear facilities demands stronger physical protection system (PPS) within the limited budget. For this reason we need an efficient physical protection system and before making an efficient PPS we need to evaluate it. This evaluation process should faithfully reflect real situation, reveal weak points and unnecessary protection elements, and give comparable quantitative values. Performance based analysis helps to build an efficient physical protection system. Instead of regulating the number of sensors and barriers, the performance based analysis evaluates a PPS fit to the situation of a facility. The analysis assesses delay (sensors) and detection (barriers) of a PPS against an intrusion, and judges whether a response force arrives before intruders complete their job. Performance based analysis needs complicated calculation and, hence, several assessment codes have been developed. A code called the estimation of adversary sequence interruption (EASI) was developed to analyze vulnerability along a single intrusion path. The systematic analysis of vulnerability to intrusion (SAVI) code investigates multi-paths to a valuable asset in an actual facility. SAVI uses adversary sequence diagram to describe multi-paths

  18. Carbon Inputs From Riparian Vegetation Limit Oxidation of Physically Bound Organic Carbon Via Biochemical and Thermodynamic Processes: OC Oxidation Processes Across Vegetation

    Energy Technology Data Exchange (ETDEWEB)

    Graham, Emily B. [Pacific Northwest National Laboratory, Richland WA USA; Tfaily, Malak M. [Environmental Molecular Sciences Laboratory, Richland WA USA; Crump, Alex R. [Pacific Northwest National Laboratory, Richland WA USA; Goldman, Amy E. [Pacific Northwest National Laboratory, Richland WA USA; Bramer, Lisa M. [Pacific Northwest National Laboratory, Richland WA USA; Arntzen, Evan [Pacific Northwest National Laboratory, Richland WA USA; Romero, Elvira [Pacific Northwest National Laboratory, Richland WA USA; Resch, C. Tom [Pacific Northwest National Laboratory, Richland WA USA; Kennedy, David W. [Pacific Northwest National Laboratory, Richland WA USA; Stegen, James C. [Pacific Northwest National Laboratory, Richland WA USA

    2017-12-01

    In light of increasing terrestrial carbon (C) transport across aquatic boundaries, the mechanisms governing organic carbon (OC) oxidation along terrestrial-aquatic interfaces are crucial to future climate predictions. Here, we investigate biochemistry, metabolic pathways, and thermodynamics corresponding to OC oxidation in the Columbia River corridor. We leverage natural vegetative differences to encompass variation in terrestrial C inputs. Our results suggest that decreases in terrestrial C deposition associated with diminished riparian vegetation induce oxidation of physically-bound (i.e., mineral and microbial) OC at terrestrial-aquatic interfaces. We also find that contrasting metabolic pathways oxidize OC in the presence and absence of vegetation and—in direct conflict with the concept of ‘priming’—that inputs of water-soluble and thermodynamically-favorable terrestrial OC protects bound-OC from oxidation. Based on our results, we propose a mechanistic conceptualization of OC oxidation along terrestrial-aquatic interfaces that can be used to model heterogeneous patterns of OC loss under changing land cover distributions.

  19. 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

  20. The multi-physics, user-friendly gas-dynamics code Visual Tsunami 2.0

    International Nuclear Information System (INIS)

    Debonnel, C. S.; Trubov, L.; Zeballos, C. A.; Peterson, P. F.

    2007-01-01

    Since the early 1990's, the series of simulation code known as TSUNAMI has been the main tool employed to explore gas dynamics phenomena in thick-liquid protected inertial fusion target chambers. The applicability and user-friendliness of the code was recently extended through a set of MATLAB pre- and post-processing tools and graphical user interfaces [1]. Geometry, initial, and boundary conditions can be specified from within AutoCAD through a set of in-house AutoLISP graphical user interfaces. A novel MATLAB core was recently developed and tested, and is now routinely used with the user-friendly pre- and post-processors [2]. An overview of Visual Tsunami 2.0, the latest version of the code, is presented here. (authors)

  1. FRESCO: fusion reactor simulation code for tokamaks

    International Nuclear Information System (INIS)

    Mantsinen, M.J.

    1995-03-01

    The study of the dynamics of tokamak fusion reactors, a zero-dimensional particle and power balance code FRESCO (Fusion Reactor Simulation Code) has been developed at the Department of Technical Physics of Helsinki University of Technology. The FRESCO code is based on zero-dimensional particle and power balance equations averaged over prescribed plasma profiles. In the report the data structure of the FRESCO code is described, including the description of the COMMON statements, program input, and program output. The general structure of the code is described, including the description of subprograms and functions. The physical model used and examples of the code performance are also included in the report. (121 tabs.) (author)

  2. 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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  4. The ZPIC educational code suite

    Science.gov (United States)

    Calado, R.; Pardal, M.; Ninhos, P.; Helm, A.; Mori, W. B.; Decyk, V. K.; Vieira, J.; Silva, L. O.; Fonseca, R. A.

    2017-10-01

    Particle-in-Cell (PIC) codes are used in almost all areas of plasma physics, such as fusion energy research, plasma accelerators, space physics, ion propulsion, and plasma processing, and many other areas. In this work, we present the ZPIC educational code suite, a new initiative to foster training in plasma physics using computer simulations. Leveraging on our expertise and experience from the development and use of the OSIRIS PIC code, we have developed a suite of 1D/2D fully relativistic electromagnetic PIC codes, as well as 1D electrostatic. These codes are self-contained and require only a standard laptop/desktop computer with a C compiler to be run. The output files are written in a new file format called ZDF that can be easily read using the supplied routines in a number of languages, such as Python, and IDL. The code suite also includes a number of example problems that can be used to illustrate several textbook and advanced plasma mechanisms, including instructions for parameter space exploration. We also invite contributions to this repository of test problems that will be made freely available to the community provided the input files comply with the format defined by the ZPIC team. The code suite is freely available and hosted on GitHub at https://github.com/zambzamb/zpic. Work partially supported by PICKSC.

  5. V.S.O.P.-computer code system for reactor physics and fuel cycle simulation

    International Nuclear Information System (INIS)

    Teuchert, E.; Hansen, U.; Haas, K.A.

    1980-03-01

    V.S.O.P. (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shutdown features, incore and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. A limitation of the storage requirement to 360 K-bites is achieved by an overlay structure. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. Beside its use in research and development work for the high temperature reactor the system has been applied successfully to LWR and Heavy Water Reactors. (orig.) [de

  6. 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...

  7. First experience with particle-in-cell plasma physics code on ARM-based HPC systems

    Science.gov (United States)

    Sáez, Xavier; Soba, Alejandro; Sánchez, Edilberto; Mantsinen, Mervi; Mateo, Sergi; Cela, José M.; Castejón, Francisco

    2015-09-01

    In this work, we will explore the feasibility of porting a Particle-in-cell code (EUTERPE) to an ARM multi-core platform from the Mont-Blanc project. The used prototype is based on a system-on-chip Samsung Exynos 5 with an integrated GPU. It is the first prototype that could be used for High-Performance Computing (HPC), since it supports double precision and parallel programming languages.

  8. Importance sampling implemented in the code PRIZMA for deep penetration and detection problems in reactor physics

    International Nuclear Information System (INIS)

    Kandiev, Y.Z.; Zatsepin, O.V.

    2013-01-01

    At RFNC-VNIITF, the PRIZMA code which has been developed for more than 30 years, is used to model radiation transport by the Monte Carlo method. The code implements individual and coupled tracking of neutrons, photons, electrons, positrons and ions in one dimensional (1D), 2D or 3D geometry. Attendance estimators are used for tallying, i.e., the estimators whose scores are only nonzero from particles which cross a region or surface of interest. Importance sampling is used to make deep penetration and detection calculations more effective. However, its application to reactor analysis appeared peculiar and required further development. The paper reviews methods used for deep penetration and detection calculations by PRIZMA. It describes in what these calculations differ when applied to reactor analysis and how we compute approximated importance functions and parameters for biased distributions. Methods to control the statistical weight of particles are also discussed. A number of test and applied calculations which were done for the purpose of verification are provided. They are shown to agree either with asymptotic solutions if exist, or with results of analog calculations or predictions by other codes. The applied calculations include the estimation of ex-core detector response from neutron sources arranged in the core, and the estimation of in-core detector response. (authors)

  9. Computation of thermodynamic equilibria of nuclear materials in multi-physics codes

    International Nuclear Information System (INIS)

    Piro, M.H.; Lewis, B.J.; Thompson, W.T.; Simunovic, S.; Besmann, T.M.

    2011-01-01

    A new equilibrium thermodynamic solver is being developed with the primary impetus of direct integration into nuclear fuel performance and safety codes to provide improved predictions of fuel behavior. This solver is intended to provide boundary conditions and material properties for continuum transport calculations. There are several legitimate concerns with the use of existing commercial thermodynamic codes: 1) licensing entanglements associated with code distribution, 2) computational performance, and 3) limited capabilities of handling large multi-component systems of interest to the nuclear industry. The development of this solver is specifically aimed at addressing these concerns. In support of this goal, a new numerical algorithm for computing chemical equilibria is presented which is not based on the traditional steepest descent method or 'Gibbs energy minimization' technique. This new approach exploits fundamental principles of equilibrium thermodynamics, which simplifies the optimization equations. The chemical potentials of all species and phases in the system are constrained by the system chemical potentials, and the objective is to minimize the residuals of the mass balance equations. Several numerical advantages are achieved through this simplification, as described in this paper. (author)

  10. Beamforming-Based Physical Layer Network Coding for Non-Regenerative Multi-Way Relaying

    Directory of Open Access Journals (Sweden)

    Klein Anja

    2010-01-01

    Full Text Available We propose non-regenerative multi-way relaying where a half-duplex multi-antenna relay station (RS assists multiple single-antenna nodes to communicate with each other. The required number of communication phases is equal to the number of the nodes, N. There are only one multiple-access phase, where the nodes transmit simultaneously to the RS, and broadcast (BC phases. Two transmission methods for the BC phases are proposed, namely, multiplexing transmission and analog network coded transmission. The latter is a cooperation method between the RS and the nodes to manage the interference in the network. Assuming that perfect channel state information is available, the RS performs transceive beamforming to the received signals and transmits simultaneously to all nodes in each BC phase. We address the optimum transceive beamforming maximising the sum rate of non-regenerative multi-way relaying. Due to the nonconvexity of the optimization problem, we propose suboptimum but practical signal processing schemes. For multiplexing transmission, we propose suboptimum schemes based on zero forcing, minimising the mean square error, and maximising the signal to noise ratio. For analog network coded transmission, we propose suboptimum schemes based on matched filtering and semidefinite relaxation of maximising the minimum signal to noise ratio. It is shown that analog network coded transmission outperforms multiplexing transmission.

  11. The database 'EDUD Base' for validation of neutron-physics codes used to analyze the WWER-440 cores

    International Nuclear Information System (INIS)

    Rocek, J.; Belac, J.; Miasnikov, A.

    2003-01-01

    The program and data system EDUDBase for validation of reactor computing codes was developed at NRI. It is designed for validation and evaluation of the precision of different computer codes used for WWER core analyses. The main goal of this database is to provide data for comparison with calculation results of tested codes and tools for statistical analysis or differences between the calculation results and the test data. The benchmark data sets are based on in-core measurements performed on WWER-440 reactors of Dukovany NPP. The initial data from NPP are verified, errors and inaccuracies are eliminated and data are transferred to a form, which is suitable for comparison with results of calculations. A special reduced operating history data set is created for each operating cycle ('Benchmark Operation History') to be used as an input data for calculation. It contains values of some integral quantities for each time point: effective time, integral thermal power, boron concentration, position of working group control assemblies (group 6) and inlet coolant temperature. At present, sets are available for all completed cycles up to: (unit/cycle) 1/17, 2/16, 3/15, 4/15. Power distribution is described for approx. 40 time steps during each operating cycle. 2D-power distributions are transferred into 60-degree core symmetry sector of reactor core. At present, such data sets are available only for later cycles starting with: (unit/cycle) 1/7, 2/6, 3/5, 4/5 (in other words last II cycles for each unit) (Authors)

  12. Statistical identification of effective input variables

    International Nuclear Information System (INIS)

    Vaurio, J.K.

    1982-09-01

    A statistical sensitivity analysis procedure has been developed for ranking the input data of large computer codes in the order of sensitivity-importance. The method is economical for large codes with many input variables, since it uses a relatively small number of computer runs. No prior judgemental elimination of input variables is needed. The sceening method is based on stagewise correlation and extensive regression analysis of output values calculated with selected input value combinations. The regression process deals with multivariate nonlinear functions, and statistical tests are also available for identifying input variables that contribute to threshold effects, i.e., discontinuities in the output variables. A computer code SCREEN has been developed for implementing the screening techniques. The efficiency has been demonstrated by several examples and applied to a fast reactor safety analysis code (Venus-II). However, the methods and the coding are general and not limited to such applications

  13. First experience with particle-in-cell plasma physics code on ARM-based HPC systems

    OpenAIRE

    Sáez, Xavier; Soba, Alejandro; Sánchez, Edilberto; Mantsinen, Mervi; Mateo, Sergio; Cela, José M.; Castejón, Francisco

    2015-01-01

    In this work, we will explore the feasibility of porting a Particle-in-cell code (EUTERPE) to an ARM multi-core platform from the Mont-Blanc project. The used prototype is based on a system-on-chip Samsung Exynos 5 with an integrated GPU. It is the first prototype that could be used for High-Performance Computing (HPC), since it supports double precision and parallel programming languages. The research leading to these results has received funding from the European Com- munity's Seventh...

  14. A TSTT integrated FronTier code and its applications in computational fluid physics

    International Nuclear Information System (INIS)

    Fix, Brian; Glimm, James; Li Xiaolin; Li Yuanhua; Liu Xinfeng; Samulyak, Roman; Xu Zhiliang

    2005-01-01

    We introduce the FronTier-Lite software package and its adaptation to the TSTT geometry and mesh entity data interface. This package is extracted from the original front tracking code for general purpose scientific and engineering applications. The package contains a static interface library and a dynamic front propagation library. It can be used in research of different scientific problems. We demonstrate the application of FronTier in the simulations of fuel injection jet, the fusion pellet injection and fluid mixing problems

  15. Assessing reactor physics codes capabilities to simulate fast reactors on the example of the BN-600 benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, Vladimir [Scientific and Engineering Centre for Nuclear and Radiation Safety (SES NRS), Moscow (Russian Federation); Bousquet, Jeremy [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    This work aims to assess the capabilities of reactor physics codes (initially validated for thermal reactors) to simulate fast sodium cooled reactors. The BFS-62-3A critical experiment from the BN-600 Hybrid Core Benchmark Analyses was chosen for the investigation. Monte-Carlo codes (KENO from SCALE and SERPENT 2.1.23) and the deterministic diffusion code DYN3D-MG are applied to calculate the neutronic parameters. It was found that the multiplication factor and reactivity effects calculated by KENO and SERPENT using the ENDF/B-VII.0 continuous energy library are in a good agreement with each other and with the measured benchmark values. Few-groups macroscopic cross sections, required for DYN3D-MG, were prepared in applying different methods implemented in SCALE and SERPENT. The DYN3D-MG results of a simplified benchmark show reasonable agreement with results from Monte-Carlo calculations and measured values. The former results are used to justify DYN3D-MG implementation for sodium cooled fast reactors coupled deterministic analysis.

  16. Improvements of Physical Models in TRITGO code for Tritium Behavior Analysis in VHTR

    International Nuclear Information System (INIS)

    Yoo, Jun Soo; Tak, Nam Il; Lim, Hong Sik

    2010-01-01

    Since tritium is radioactive material with 12.32 year of half-life and is generated by a ternary fission reaction in fuel as well as by neutron absorption reactions of impurities in Very High Temperature gas-cooled Reactor (VHTR) core, accurate prediction of tritium behavior and its concentration in product hydrogen is definitely important in terms of public safety for its construction. In this respect, TRITGO code was developed for estimating the tritium production and distribution in high temperature gas-cooled reactors by General Atomics (GA). However, some models in it are hard-wired to specific reactor type or too simplified, which makes the analysis results less applicable. Thus, major improvements need to be considered for better predictions. In this study, some of model improvements have been suggested and its effect is evaluated based on the analysis work against PMR600 design concept

  17. Proceedings of 5. French speaking scientific days on calculation codes for radioprotection, radio-physics and dosimetry

    International Nuclear Information System (INIS)

    Simon-Cornu, Marie; Mourlon, Christophe; Bordy, J.M.; Daures, J.; Dusiac, D.; Moignau, F.; Gouriou, J.; Million, M.; Moreno, B.; Chabert, I.; Lazaro, D.; Barat, E.; Dautremer, T.; Montagu, T.; Agelou, M.; De Carlan, L.; Patin, D.; Le Loirec, C.; Dupuis, P.; Gassa, F.; Guerin, L.; Batalla, A.; Leni, Pierre-Emmanuel; Laurent, Remy; Gschwind, Regine; Makovicka, Libor; Henriet, Julien; Salomon, Michel; Vivier, Alain; Lopez, Gerald; Dossat, C.; Pourrouquet, P.; Thomas, J.C.; Sarie, I.; Peyrard, P.F.; Chatry, N.; Lavielle, D.; Loze, R.; Brun, E.; Damian, F.; Diop, C.; Dumonteil, E.; Hugot, F.X.; Jouanne, C.; Lee, Y.K.; Malvagi, F.; Mazzolo, A.; Petit, O.; Trama, J.C.; Visonneau, T.; Zoia, A.; Courageot, Estelle; Gaillard-Lecanu, Emmanuelle; Kutschera, Reinald; Le Meur, Gaelle; Uzio, Fabien; De Conto, Celine; Gschwind, Regine; Makovicka, Libor; Farah, Jad; Martinetti, Florent; Sayah, Rima; Donadille, Laurent; Herault, Joel; Delacroix, Sabine; Nauraye, Catherine; Lee, Choonsik; Bolch, Wesley; Clairand, Isabelle; Horodynski, Jean-Michel; Pauwels, Nicolas; Robert, Pierre; VOLLAIRE, Joachim; Nicoletti, C.; Kitsos, S.; Tardy, M.; Marchaud, G.; Stankovskiy, Alexey; Van Den Eynde, Gert; Fiorito, Luca; Malambu, Edouard; Dreuil, Serge; Mougeot, X.; Be, M.M.; Bisch, C.; Villagrasa, C.; Dos Santos, M.; Clairand, I.; Karamitros, M.; Incerti, S.; Petitguillaume, Alice; Franck, Didier; Desbree, Aurelie; Bernardini, Michela; Labriolle-Vaylet, Claire de; Gnesin, Silvano; Leadermann, Jean-Pascal; Paterne, Loic; Bochud, Francois O.; Verdun, Francis R.; Baechler, Sebastien; Prior, John O.; Thomassin, Alain; Arial, Emmanuelle; Laget, Michael; Masse, Veronique; Saldarriaga Vargas, Clarita; Struelens, Lara; Vanhavere, Filip; Perier, Aurelien; Courageot, Estelle; Gaillard-Lecanu, Emmanuelle; Le-Meur, Gaelle; Monier, Catherine; Thers, Dominique; Le-Guen, Bernard; Blond, Serge; Cordier, Gerard; Le Roy, Maiwenn; De Carlan, Loic; Bordy, Jean-Marc; Caccia, Barbara; Andenna, Claudio; Charimadurai, Arun; Selvam, T Palani; Czarnecki, Damian; Zink, Klemens; Gschwind, Regine; Martin, Eric; Huot, Nicolas; Zoubair, Mariam; El Bardouni, Tarek; Lazaro, Delphine; Barat, Eric; Dautremer, Thomas; Montagu, Thierry; Chabert, Isabelle; Guerin, Lucie; Batalla, Alain; Moignier, C.; Huet, C.; Bassinet, C.; Baumann, M.; Barraux, V.; Sebe-Mercier, K.; Loiseau, C.; Batalla, A.; Makovicka, L.; Desnoyers, Yvon; Juhel, Gabriel; Mattera, Christophe; Tempier, Maryline

    2014-03-01

    These scientific days were organised by the 'technical protection' Section of the French Society of Radiation Protection (SFRP) in cooperation with the French society of medical physicists (SFPM), the Swiss Romandie association of radioprotection (ARRAD) and the associated laboratories of radio-physics and dosimetry (LARD). The objective of these days was to review the existing calculation codes used in radiation transport, source estimation and dose management, and to identify some future prospects. This document brings together the available presentations (slides) together with their corresponding abstracts (in French) and dealing with: 1 - Presentation of the conference days (L. De Carlan); 2 - Simulating radionuclide transfers in the environment: what calculation codes and for what? (C. Mourlon); 3 - Contribution of Monte-Carlo calculation to the theoretical foundation analysis of calibration procedures and dosemeters design for radioprotection photon dosimetry (J.M. Bordy); 4 - Use of calculation codes in R and D for the development of a new passive dosemeter for photons and beta radiations (B. Moreno); 5 - Development of a new virtual sources model for the Monte-Carlo prediction of EPID (Electronic Portal Imaging Device) images and implementation in PENELOPE (I. Chabert); 6 - Prediction of high-resolution EPID images for in-vivo dosimetry (D. Patin); 7 - 4D thorax modeling by artificial neural networks (P.E. Leni); 8 - Presentation of the calculation utilities of the book 'Calculation of ionizing radiations generated doses' (Vivier, Lopez, EDP Sciences 2012) (A. Vivier); 9 - RayXpert C : a 3D modeling and Monte-Carlo dose rate calculation software (C. Dossat); 10 - TRIPOLI-4 R Version 9 S Monte-Carlo code for radioprotection (F. Damian); 11 - Realistic radioprotection training with the digital school workshop (E. Courageot); 12 - Use of BEAMNRC code for dental prostheses influence evaluation in ENT cancers treatment by external radiotherapy (C. De Conto); 13

  18. Verification of results of core physics on-line simulation by NGFM code

    International Nuclear Information System (INIS)

    Zhao Yu; Cao Xinrong; Zhao Qiang

    2008-01-01

    Nodal Green's Function Method program NGFM/TNGFM has been trans- planted to windows system. The 2-D and 3-D benchmarks have been checked by this program. And the program has been used to check the results of QINSHAN-II reactor simulation. It is proved that the NGFM/TNGFM program is applicable for reactor core physics on-line simulation system. (authors)

  19. Advanced resonance self-shielding method for gray resonance treatment in lattice physics code GALAXY

    International Nuclear Information System (INIS)

    Koike, Hiroki; Yamaji, Kazuya; Kirimura, Kazuki; Sato, Daisuke; Matsumoto, Hideki; Yamamoto, Akio

    2012-01-01

    A new resonance self-shielding method based on the equivalence theory is developed for general application to the lattice physics calculations. The present scope includes commercial light water reactor (LWR) design applications which require both calculation accuracy and calculation speed. In order to develop the new method, all the calculation processes from cross-section library preparation to effective cross-section generation are reviewed and reframed by adopting the current enhanced methodologies for lattice calculations. The new method is composed of the following four key methods: (1) cross-section library generation method with a polynomial hyperbolic tangent formulation, (2) resonance self-shielding method based on the multi-term rational approximation for general lattice geometry and gray resonance absorbers, (3) spatially dependent gray resonance self-shielding method for generation of intra-pellet power profile and (4) integrated reaction rate preservation method between the multi-group and the ultra-fine-group calculations. From the various verifications and validations, applicability of the present resonance treatment is totally confirmed. As a result, the new resonance self-shielding method is established, not only by extension of a past concentrated effort in the reactor physics research field, but also by unification of newly developed unique and challenging techniques for practical application to the lattice physics calculations. (author)

  20. Probabilistic analysis of strength and thermal-physic WWER fuel rod characteristics using START-3 code

    International Nuclear Information System (INIS)

    Medvedev, A.; Bogatyr, S.; Khramtsov; Sokolov, F.

    2001-01-01

    During the last years probabilistic methods for evaluation of the influence of the fuel geometry and technology parameters on fuel operational reliability are widely used. In the present work the START-3 procedure is used to calculate the thermal physics and strength characteristics of WWER fuel rods behavior. The procedure is based on the Monte-Carlo method with the application of Sobol quasi-random sequences. This technique allows to treat the fuel rod technological and operating parameters as well as its strength and thermal physics characteristics as random variables. The work deals with a series of WWER-1000 fuel rod statistical tests and verification based on the PIE results. Also preliminary calculations are implemented with the aim to determine the design schema parameters. This should ensure the accuracy of the assessment of the parameters of WWER fuel rod characteristics distribution. The probability characteristics of fuel rod strength and thermal physics are assessed via the statistical analysis of the results of probability calculations

  1. Determination of the physical parameters of the nuclear subcritical assembly Chicago 9000 of the IPN using the Serpent code

    International Nuclear Information System (INIS)

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

    2013-10-01

    For the Serpent code was developed the three-dimensional model corresponding to the nuclear subcritical assembly (S A) Chicago 9000 of the Escuela Superior de Fisica y Matematicas del Instituto Politecnico Nacional (ESFM-IPN). The model includes: a) the core, formed by 312 aluminum pipes that contain 5 nuclear fuel rods (natural uranium in metallic form), b) the multi-perforated plates where they penetrate the inferior part of each pipe to be able to remain in vertical form, c) water, acting as moderator and reflector, and d) the recipient lodging to the core. The pipes arrangement is hexagonal although the transversal section of the recipient that lodges to the core is circular. The entrance file for the Serpent code was generated with the data provided by the manual of the S A use about the composition and density of the fuel rods and others obtained in direct form of the rods, as the interior and external diameter, mass and height. Of the obtained physical parameters, those more approached to that reported in the manual of the subcritical assembly are the effective multiplication factor and the reproduction factor η. The differences can be because the description of the fuel rods provided by the manual of the S A use do not correspond those that are physically in the S A core. This difference consists on the presence of a circular central channel of 1.245 diameter centimeters in each fuel rod. The fuel rods reported in the mentioned manual do not have that channel. Although the obtained results are encouraging, we want to continue improving the model to incorporate in this the detectors, defined this way by the Serpent code, which could determine the existent neutrons flux in diverse points of interest like the axial or radial aligned points and to compare these with those that are obtained in an experimental way when a generating neutrons source (Pu-Be) is introduced. Added to this effort the cross sections for each unitary cell will be determined, so that

  2. Modelling of the thermomechanical and physical processes in FR fuel pins using the GERMINAL code

    International Nuclear Information System (INIS)

    Roche, L.; Pelletier, M.

    2000-01-01

    In the frame of the R and D on Fast Reactor mixed oxide fuels, CEA/DEC has developed the computer code GERMINAL for studying fuel pin thermal and mechanical behaviour, both during steady-state and incidental conditions, up to high burn-up (25 at%). The first part of this paper is devoted to the description of the main models: fuel evolution (central hole and porosity evolution, Plutonium redistribution, O/M radial profile, transient gas swelling, melting fuel behaviour, minor actinides production), high burn-up models (fission gas, volatile fission products and JOG formation), fuel-cladding heat transfer, fuel-cladding mechanical interaction. The second part gives some examples of calculation results taken from the GERMINAL validation data base (more than 40 experiments from PHENIX, PFR, CABRI reactors), with special emphasis on: local fission gas retention and global release, fuel geometry evolution, radial redistribution of plutonium for high burn-up fuels, solid and annular fuel behaviour during power ramps including fuel melting, helium formation from MA (Am and Np) doped homogeneous fuels. (author)

  3. Qualification of the APOLLO2 lattice physics code of the NURISP platform for WWER hexagonal lattices

    International Nuclear Information System (INIS)

    Hegyi, G.; Kereszturi, A.; Tota, A.

    2011-01-01

    The experiments performed at the ZR-6 zero critical reactor by the Temporary International Collective and a numerical assembly burnup benchmark specified for depletion calculation of a WWER-440 assembly containing gadolinium burnable poison were used to qualify the APOLLO2 (APOLLO2.8-E3) code as a part of its ongoing validation activity. The work is part of the NURISP project, where KFKI Atomic Energy Research Institute undertook to develop and qualify some calculation schemes for hexagonal problems. Concerning the ZR-6 measurements, single cell, macro cell and two-dimensional calculations of selected regular and perturbed experiments are being used for the validation. In the two-dimensional cases the radial leakage is also taken into account in the calculations together with the axial leakage represented by the measured axial buckling. Criticality parameter and reaction rate comparisons are presented. Although various sets of the experiments have been selected for the validation, good agreement of the measured and calculated parameters could be found by using the different options offered by APOLLO2. An additional mathematical benchmark-presented in the paper - also attests for the reliability of APOLLO2. All the test results prove the reliability of APOLLO2 for WWER core calculations. (Authors)

  4. NSURE code

    International Nuclear Information System (INIS)

    Rattan, D.S.

    1993-11-01

    NSURE stands for Near-Surface Repository code. NSURE is a performance assessment code. developed for the safety assessment of near-surface disposal facilities for low-level radioactive waste (LLRW). Part one of this report documents the NSURE model, governing equations and formulation of the mathematical models, and their implementation under the SYVAC3 executive. The NSURE model simulates the release of nuclides from an engineered vault, their subsequent transport via the groundwater and surface water pathways tot he biosphere, and predicts the resulting dose rate to a critical individual. Part two of this report consists of a User's manual, describing simulation procedures, input data preparation, output and example test cases

  5. Medical Applications of the PHITS Code (3): User Assistance Program for Medical Physics Computation.

    Science.gov (United States)

    Furuta, Takuya; Hashimoto, Shintaro; Sato, Tatsuhiko

    2016-01-01

    DICOM2PHITS and PSFC4PHITS are user assistance programs for medical physics PHITS applications. DICOM2PHITS is a program to construct the voxel PHITS simulation geometry from patient CT DICOM image data by using a conversion table from CT number to material composition. PSFC4PHITS is a program to convert the IAEA phase-space file data to PHITS format to be used as a simulation source of PHITS. Both of the programs are useful for users who want to apply PHITS simulation to verification of the treatment planning of radiation therapy. We are now developing a program to convert dose distribution obtained by PHITS to DICOM RT-dose format. We also want to develop a program which is able to implement treatment information included in other DICOM files (RT-plan and RT-structure) as a future plan.

  6. Physical analysis and modelling of aerosols transport. implementation in a finite elements code. Experimental validation in laminar and turbulent flows

    International Nuclear Information System (INIS)

    Armand, Patrick

    1995-01-01

    The aim of this work consists in the Fluid Mechanics and aerosol Physics coupling. In the first part, the order of magnitude analysis of the particle dynamics is done. This particle is embedded in a non-uniform unsteady flow. Flow approximations around the inclusion are described. Corresponding aerodynamic drag formulae are expressed. Possible situations related to the problem data are extensively listed. In the second part, one studies the turbulent particles transport. Eulerian approach which is particularly well adapted to industrial codes is preferred in comparison with the Lagrangian methods. One chooses the two-fluid formalism in which career gas-particles slip is taken into account. Turbulence modelling gets through a k-epsilon modulated by the inclusions action on the flow. The model is implemented In a finite elements code. Finally, In the third part, one validates the modelling in laminar and turbulent cases. We compare simulations to various experiments (settling battery, inertial impaction in a bend, jets loaded with glass beads particles) which are taken in the literature or done by ourselves at the laboratory. The results are very close. It is a good point when it is thought of the particles transport model and associated software future use. (author) [fr

  7. Development of an anthropomorphic model and a Monte Carlo calculation code devoted to the physical reconstruction of a radiological accident

    International Nuclear Information System (INIS)

    Roux, A.

    2001-01-01

    The diversity of radiological accidents makes difficult the medical prognosis and the therapy choice from only clinical observations. To complete this information, it is important to know the global dose received by the organism and the dose distributions in depth in tissues. The dose estimation can be made by a physical reconstruction of the accident with the help of tools based on experimental techniques or on calculation. The software of the geometry construction (M.G.E.D.), associated to the Monte-Carlo code of photons and neutrons transport (M.O.R.S.E.) replies these constraints. An important result of this work is to determine the principal parameters to know in function of the accident type, as well as the precision level required for these parameters. (N.C.)

  8. The Rosslyn Code: Can Physics Explain a 500-Year Old Melody Etched in the Walls of a Scottish Chapel?

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Chris [State Magazine

    2011-10-19

    For centuries, historians have puzzled over a series of 213 symbols carved into the stone of Scotland’s Rosslyn Chapel. (Disclaimer: You may recognize this chapel from The Da Vinci Code, but this is real and unrelated!) Several years ago, a composer and science enthusiast noticed that the symbols bore a striking similarity to Chladni patterns, the elegant images that form on a two- dimensional surface when it vibrates at certain frequencies. This man’s theory: A 500-year-old melody was inscribed in the chapel using the language of physics. But not everyone is convinced. Slate senior editor Chris Wilson travelled to Scotland to investigate the claims and listen to this mysterious melody, whatever it is. Come find out what he discovered, including images of the patterns and audio of the music they inspired.

  9. Quasi-optical converters for high-power gyrotrons: a brief review of physical models, numerical methods and computer codes

    International Nuclear Information System (INIS)

    Sabchevski, S; Zhelyazkov, I; Benova, E; Atanassov, V; Dankov, P; Thumm, M; Arnold, A; Jin, J; Rzesnicki, T

    2006-01-01

    Quasi-optical (QO) mode converters are used to transform electromagnetic waves of complex structure and polarization generated in gyrotron cavities into a linearly polarized, Gaussian-like beam suitable for transmission. The efficiency of this conversion as well as the maintenance of low level of diffraction losses are crucial for the implementation of powerful gyrotrons as radiation sources for electron-cyclotron-resonance heating of fusion plasmas. The use of adequate physical models, efficient numerical schemes and up-to-date computer codes may provide the high accuracy necessary for the design and analysis of these devices. In this review, we briefly sketch the most commonly used QO converters, the mathematical base they have been treated on and the basic features of the numerical schemes used. Further on, we discuss the applicability of several commercially available and free software packages, their advantages and drawbacks, for solving QO related problems

  10. Advancements in reactor physics modelling methodology of Monte Carlo Burnup Code MCB dedicated to higher simulation fidelity of HTR cores

    International Nuclear Information System (INIS)

    Cetnar, Jerzy

    2014-01-01

    The recent development of MCB - Monte Carlo Continuous Energy Burn-up code is directed towards advanced description of modern reactors, including double heterogeneity structures that exist in HTR-s. In this, we exploit the advantages of MCB methodology in integrated approach, where physics, neutronics, burnup, reprocessing, non-stationary process modeling (control rod operation) and refined spatial modeling are carried in a single flow. This approach allows for implementations of advanced statistical options like analysis of error propagation, perturbation in time domain, sensitivity and source convergence analyses. It includes statistical analysis of burnup process, emitted particle collection, thermal-hydraulic coupling, automatic power profile calculations, advanced procedures of burnup step normalization and enhanced post processing capabilities. (author)

  11. DLLExternalCode

    Energy Technology Data Exchange (ETDEWEB)

    2014-05-14

    DLLExternalCode is the a general dynamic-link library (DLL) interface for linking GoldSim (www.goldsim.com) with external codes. The overall concept is to use GoldSim as top level modeling software with interfaces to external codes for specific calculations. The DLLExternalCode DLL that performs the linking function is designed to take a list of code inputs from GoldSim, create an input file for the external application, run the external code, and return a list of outputs, read from files created by the external application, back to GoldSim. Instructions for creating the input file, running the external code, and reading the output are contained in an instructions file that is read and interpreted by the DLL.

  12. Wind-US Code Physical Modeling Improvements to Complement Hypersonic Testing and Evaluation

    Science.gov (United States)

    Georgiadis, Nicholas J.; Yoder, Dennis A.; Towne, Charles S.; Engblom, William A.; Bhagwandin, Vishal A.; Power, Greg D.; Lankford, Dennis W.; Nelson, Christopher C.

    2009-01-01

    This report gives an overview of physical modeling enhancements to the Wind-US flow solver which were made to improve the capabilities for simulation of hypersonic flows and the reliability of computations to complement hypersonic testing. The improvements include advanced turbulence models, a bypass transition model, a conjugate (or closely coupled to vehicle structure) conduction-convection heat transfer capability, and an upgraded high-speed combustion solver. A Mach 5 shock-wave boundary layer interaction problem is used to investigate the benefits of k- s and k-w based explicit algebraic stress turbulence models relative to linear two-equation models. The bypass transition model is validated using data from experiments for incompressible boundary layers and a Mach 7.9 cone flow. The conjugate heat transfer method is validated for a test case involving reacting H2-O2 rocket exhaust over cooled calorimeter panels. A dual-mode scramjet configuration is investigated using both a simplified 1-step kinetics mechanism and an 8-step mechanism. Additionally, variations in the turbulent Prandtl and Schmidt numbers are considered for this scramjet configuration.

  13. PHYSICS

    CERN Multimedia

    P. Sphicas

    The CPT project came to an end in December 2006 and its original scope is now shared among three new areas, namely Computing, Offline and Physics. In the physics area the basic change with respect to the previous system (where the PRS groups were charged with detector and physics object reconstruction and physics analysis) was the split of the detector PRS groups (the old ECAL-egamma, HCAL-jetMET, Tracker-btau and Muons) into two groups each: a Detector Performance Group (DPG) and a Physics Object Group. The DPGs are now led by the Commissioning and Run Coordinator deputy (Darin Acosta) and will appear in the correspond¬ing column in CMS bulletins. On the physics side, the physics object groups are charged with the reconstruction of physics objects, the tuning of the simulation (in collaboration with the DPGs) to reproduce the data, the provision of code for the High-Level Trigger, the optimization of the algorithms involved for the different physics analyses (in collaboration with the analysis gr...

  14. Physical models, cross sections, and numerical approximations used in MCNP and GEANT4 Monte Carlo codes for photon and electron absorbed fraction calculation.

    Science.gov (United States)

    Yoriyaz, Hélio; Moralles, Maurício; Siqueira, Paulo de Tarso Dalledone; Guimarães, Carla da Costa; Cintra, Felipe Belonsi; dos Santos, Adimir

    2009-11-01

    Radiopharmaceutical applications in nuclear medicine require a detailed dosimetry estimate of the radiation energy delivered to the human tissues. Over the past years, several publications addressed the problem of internal dose estimate in volumes of several sizes considering photon and electron sources. Most of them used Monte Carlo radiation transport codes. Despite the widespread use of these codes due to the variety of resources and potentials they offered to carry out dose calculations, several aspects like physical models, cross sections, and numerical approximations used in the simulations still remain an object of study. Accurate dose estimate depends on the correct selection of a set of simulation options that should be carefully chosen. This article presents an analysis of several simulation options provided by two of the most used codes worldwide: MCNP and GEANT4. For this purpose, comparisons of absorbed fraction estimates obtained with different physical models, cross sections, and numerical approximations are presented for spheres of several sizes and composed as five different biological tissues. Considerable discrepancies have been found in some cases not only between the different codes but also between different cross sections and algorithms in the same code. Maximum differences found between the two codes are 5.0% and 10%, respectively, for photons and electrons. Even for simple problems as spheres and uniform radiation sources, the set of parameters chosen by any Monte Carlo code significantly affects the final results of a simulation, demonstrating the importance of the correct choice of parameters in the simulation.

  15. Arsenic in New Jersey Coastal Plain streams, sediments, and shallow groundwater: effects from different geologic sources and anthropogenic inputs on biogeochemical and physical mobilization processes

    Science.gov (United States)

    Barringer, Julia L.; Reilly, Pamela A.; Eberl, Dennis D.; Mumford, Adam C.; Benzel, William M.; Szabo, Zoltan; Shourds, Jennifer L.; Young, Lily Y.

    2013-01-01

    Arsenic (As) concentrations in New Jersey Coastal Plain streams generally exceed the State Surface Water Quality Standard (0.017 micrograms per liter (µg/L)), but concentrations seldom exceed 1 µg/L in filtered stream-water samples, regardless of geologic contributions or anthropogenic inputs. Nevertheless, As concentrations in unfiltered stream water indicate substantial variation because of particle inputs from soils and sediments with differing As contents, and because of discharges from groundwater of widely varying chemistry.

  16. Implementing a modular system of computer codes

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.

    1983-07-01

    A modular computation system has been developed for nuclear reactor core analysis. The codes can be applied repeatedly in blocks without extensive user input data, as needed for reactor history calculations. The primary control options over the calculational paths and task assignments within the codes are blocked separately from other instructions, admitting ready access by user input instruction or directions from automated procedures and promoting flexible and diverse applications at minimum application cost. Data interfacing is done under formal specifications with data files manipulated by an informed manager. This report emphasizes the system aspects and the development of useful capability, hopefully informative and useful to anyone developing a modular code system of much sophistication. Overall, this report in a general way summarizes the many factors and difficulties that are faced in making reactor core calculations, based on the experience of the authors. It provides the background on which work on HTGR reactor physics is being carried out

  17. The arbitrary order design code Tlie 1.0

    International Nuclear Information System (INIS)

    Zeijts, J. van; Neri, Filippo

    1993-01-01

    We describe the arbitrary order charged particle transfer map code TLIE. This code is a general 6D relativistic design code with a MAD compatible input language and among others implements user defined functions and subroutines and nested fitting and optimization. First we describe the mathematics and physics in the code. Aside from generating maps for all the standard accelerator elements we describe an efficient method for generating nonlinear transfer maps for realistic magnet models. We have implemented the method to arbitrary order in our accelerator design code for cylindrical current sheet magnets. We also have implemented a self-consistent space-charge approach as in CHARLIE. Subsequently we give a description of the input language and finally, we give several examples from productions run, such as cases with stacked multipoles with overlapping fringe fields. (Author)

  18. Langmuir probe-based observables for plasma-turbulence code validation and application to the TORPEX basic plasma physics experiment

    International Nuclear Information System (INIS)

    Ricci, Paolo; Theiler, C.; Fasoli, A.; Furno, I.; Labit, B.; Mueller, S. H.; Podesta, M.; Poli, F. M.

    2009-01-01

    The methodology for plasma-turbulence code validation is discussed, with focus on the quantities to use for the simulation-experiment comparison, i.e., the validation observables, and application to the TORPEX basic plasma physics experiment [A. Fasoli et al., Phys. Plasmas 13, 055902 (2006)]. The considered validation observables are deduced from Langmuir probe measurements and are ordered into a primacy hierarchy, according to the number of model assumptions and to the combinations of measurements needed to form each of them. The lowest levels of the primacy hierarchy correspond to observables that require the lowest number of model assumptions and measurement combinations, such as the statistical and spectral properties of the ion saturation current time trace, while at the highest levels, quantities such as particle transport are considered. The comparison of the observables at the lowest levels in the hierarchy is more stringent than at the highest levels. Examples of the use of the proposed observables are applied to a specific TORPEX plasma configuration characterized by interchange-driven turbulence.

  19. NASA low-speed centrifugal compressor for 3-D viscous code assessment and fundamental flow physics research

    Science.gov (United States)

    Hathaway, M. D.; Wood, J. R.; Wasserbauer, C. A.

    1991-01-01

    A low speed centrifugal compressor facility recently built by the NASA Lewis Research Center is described. The purpose of this facility is to obtain detailed flow field measurements for computational fluid dynamic code assessment and flow physics modeling in support of Army and NASA efforts to advance small gas turbine engine technology. The facility is heavily instrumented with pressure and temperature probes, both in the stationary and rotating frames of reference, and has provisions for flow visualization and laser velocimetry. The facility will accommodate rotational speeds to 2400 rpm and is rated at pressures to 1.25 atm. The initial compressor stage being tested is geometrically and dynamically representative of modern high-performance centrifugal compressor stages with the exception of Mach number levels. Preliminary experimental investigations of inlet and exit flow uniformly and measurement repeatability are presented. These results demonstrate the high quality of the data which may be expected from this facility. The significance of synergism between computational fluid dynamic analysis and experimentation throughout the development of the low speed centrifugal compressor facility is demonstrated.

  20. ANIMAL code

    International Nuclear Information System (INIS)

    Lindemuth, I.R.

    1979-01-01

    This report describes ANIMAL, a two-dimensional Eulerian magnetohydrodynamic computer code. ANIMAL's physical model also appears. Formulated are temporal and spatial finite-difference equations in a manner that facilitates implementation of the algorithm. Outlined are the functions of the algorithm's FORTRAN subroutines and variables

  1. The Intercomparison of 3D Radiation Codes (I3RC): Showcasing Mathematical and Computational Physics in a Critical Atmospheric Application

    Science.gov (United States)

    Davis, A. B.; Cahalan, R. F.

    2001-05-01

    The Intercomparison of 3D Radiation Codes (I3RC) is an on-going initiative involving an international group of over 30 researchers engaged in the numerical modeling of three-dimensional radiative transfer as applied to clouds. Because of their strong variability and extreme opacity, clouds are indeed a major source of uncertainty in the Earth's local radiation budget (at GCM grid scales). Also 3D effects (at satellite pixel scales) invalidate the standard plane-parallel assumption made in the routine of cloud-property remote sensing at NASA and NOAA. Accordingly, the test-cases used in I3RC are based on inputs and outputs which relate to cloud effects in atmospheric heating rates and in real-world remote sensing geometries. The main objectives of I3RC are to (1) enable participants to improve their models, (2) publish results as a community, (3) archive source code, and (4) educate. We will survey the status of I3RC and its plans for the near future with a special emphasis on the mathematical models and computational approaches. We will also describe some of the prime applications of I3RC's efforts in climate models, cloud-resolving models, and remote-sensing observations of clouds, or that of the surface in their presence. In all these application areas, computational efficiency is the main concern and not accuracy. One of I3RC's main goals is to document the performance of as wide a variety as possible of three-dimensional radiative transfer models for a small but representative number of ``cases.'' However, it is dominated by modelers working at the level of linear transport theory (i.e., they solve the radiative transfer equation) and an overwhelming majority of these participants use slow-but-robust Monte Carlo techniques. This means that only a small portion of the efficiency vs. accuracy vs. flexibility domain is currently populated by I3RC participants. To balance this natural clustering the present authors have organized a systematic outreach towards

  2. Comparison of computer codes related to the sodium oxide aerosol behavior in a containment building

    International Nuclear Information System (INIS)

    Fermandjian, J.

    1984-09-01

    In order to ensure that the problems of describing the physical behavior of sodium aerosols, during hypothetical fast reactor accidents, were adequately understood, a comparison of the computer codes (ABC/INTG, PNC, Japan; AEROSIM, UKAEA/SRD, United Kingdom; PARDISEKO IIIb, KfK, Germany; AEROSOLS/A2 and AEROSOLS/B1, CEA France) was undertaken in the frame of the CEC: exercise in which code users have run their own codes with a prearranged input

  3. User's manual for the code STAPRE as implemented at Lawrence Livermore National Laboratory

    International Nuclear Information System (INIS)

    Vonach, H.

    1982-01-01

    This report gives a detailed description of the input and output of the statistical model code STAPRE for compound-nucleus reactions including a special section on the various level density options of the code. It is to be used in conjunction with the report IRK 76/01 + Add 76 + Add 78 by B. Strohmaier and M. Uhl which describes in detail the physical models on which the code is based and its general organization and structure

  4. TORBEAM 2.0, a paraxial beam tracing code for electron-cyclotron beams in fusion plasmas for extended physics applications

    Science.gov (United States)

    Poli, E.; Bock, A.; Lochbrunner, M.; Maj, O.; Reich, M.; Snicker, A.; Stegmeir, A.; Volpe, F.; Bertelli, N.; Bilato, R.; Conway, G. D.; Farina, D.; Felici, F.; Figini, L.; Fischer, R.; Galperti, C.; Happel, T.; Lin-Liu, Y. R.; Marushchenko, N. B.; Mszanowski, U.; Poli, F. M.; Stober, J.; Westerhof, E.; Zille, R.; Peeters, A. G.; Pereverzev, G. V.

    2018-04-01

    The paraxial WKB code TORBEAM (Poli, 2001) is widely used for the description of electron-cyclotron waves in fusion plasmas, retaining diffraction effects through the solution of a set of ordinary differential equations. With respect to its original form, the code has undergone significant transformations and extensions, in terms of both the physical model and the spectrum of applications. The code has been rewritten in Fortran 90 and transformed into a library, which can be called from within different (not necessarily Fortran-based) workflows. The models for both absorption and current drive have been extended, including e.g. fully-relativistic calculation of the absorption coefficient, momentum conservation in electron-electron collisions and the contribution of more than one harmonic to current drive. The code can be run also for reflectometry applications, with relativistic corrections for the electron mass. Formulas that provide the coupling between the reflected beam and the receiver have been developed. Accelerated versions of the code are available, with the reduced physics goal of inferring the location of maximum absorption (including or not the total driven current) for a given setting of the launcher mirrors. Optionally, plasma volumes within given flux surfaces and corresponding values of minimum and maximum magnetic field can be provided externally to speed up the calculation of full driven-current profiles. These can be employed in real-time control algorithms or for fast data analysis.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  6. DISP1 code

    International Nuclear Information System (INIS)

    Vokac, P.

    1999-12-01

    DISP1 code is a simple tool for assessment of the dispersion of the fission product cloud escaping from a nuclear power plant after an accident. The code makes it possible to tentatively check the feasibility of calculations by more complex PSA3 codes and/or codes for real-time dispersion calculations. The number of input parameters is reasonably low and the user interface is simple enough to allow a rapid processing of sensitivity analyses. All input data entered through the user interface are stored in the text format. Implementation of dispersion model corrections taken from the ARCON96 code enables the DISP1 code to be employed for assessment of the radiation hazard within the NPP area, in the control room for instance. (P.A.)

  7. Comparison of sodium aerosol codes

    International Nuclear Information System (INIS)

    Dunbar, I.H.; Fermandjian, J.; Bunz, H.; L'homme, A.; Lhiaubet, G.; Himeno, Y.; Kirby, C.R.; Mitsutsuka, N.

    1984-01-01

    Although hypothetical fast reactor accidents leading to severe core damage are very low probability events, their consequences are to be assessed. During such accidents, one can envisage the ejection of sodium, mixed with fuel and fission products, from the primary circuit into the secondary containment. Aerosols can be formed either by mechanical dispersion of the molten material or as a result of combustion of the sodium in the mixture. Therefore considerable effort has been devoted to study the different sodium aerosol phenomena. To ensure that the problems of describing the physical behaviour of sodium aerosols were adequately understood, a comparison of the codes being developed to describe their behaviour was undertaken. The comparison consists of two parts. The first is a comparative study of the computer codes used to predict aerosol behaviour during a hypothetical accident. It is a critical review of documentation available. The second part is an exercise in which code users have run their own codes with a pre-arranged input. For the critical comparative review of the computer models, documentation has been made available on the following codes: AEROSIM (UK), MAEROS (USA), HAARM-3 (USA), AEROSOLS/A2 (France), AEROSOLS/B1 (France), and PARDISEKO-IIIb (FRG)

  8. Gravity inversion code

    International Nuclear Information System (INIS)

    Burkhard, N.R.

    1979-01-01

    The gravity inversion code applies stabilized linear inverse theory to determine the topography of a subsurface density anomaly from Bouguer gravity data. The gravity inversion program consists of four source codes: SEARCH, TREND, INVERT, and AVERAGE. TREND and INVERT are used iteratively to converge on a solution. SEARCH forms the input gravity data files for Nevada Test Site data. AVERAGE performs a covariance analysis on the solution. This document describes the necessary input files and the proper operation of the code. 2 figures, 2 tables

  9. Some questions of using the algebraic coding theory for construction of special-purpose processors in high energy physics spectrometers

    International Nuclear Information System (INIS)

    Nikityuk, N.M.

    1989-01-01

    The results of investigations of using the algebraic coding theory for the creation of parallel encoders, majority coincidence schemes and coordinate processors for the first and second trigger levels are described. Concrete examples of calculation and structure of special-purpose processor using the table arithmetic method are given for multiplicity t ≤ 5. The question of using parallel and sequential syndrome coding methods for the registration of events with clusters is discussed. 30 refs.; 10 figs

  10. The Development of a Consumer Input Program for the National Library Service for the Blind and Physically Handicapped (NLS/BPH) and Network Libraries. Final Report.

    Science.gov (United States)

    Cavenaugh, David

    This document presents a review of the current consumer relations activites of the National Library Service (NLS) for the Blind and Physically Handicapped of the Library of Congress, and an overall plan to improve NLS receipt of user suggestions, comments, opinions, or complaints through libraries which form the nationwide NLS distribution system.…

  11. COSY INFINITY, a new arbitrary order optics code

    International Nuclear Information System (INIS)

    Berz, M.

    1990-01-01

    The new arbitrary order particle optics and beam dynamics code COSY INFINITY is presented. The code is based on differential algebraic (DA) methods. COSY INFINITY has a full structured object oriented language environment. This provides a simple interface for the casual or novice user. At the same time, it offers the advanced user a very flexible and powerful tool for the utilization of DA. The power and generality of the environment is perhaps best demonstrated by the fact that the physics routines of COSY INFINITY are written in its own input language. The approach also facilitates the implementation of new features because special code generated by a user can be readily adopted to the source code. Besides being very compact in size, the code is also very fast, thanks to efficiently programmed elementary DA operations. For simple low order problems, which can be handled by conventional codes, the speed of COSY INFINITY is comparable and in certain cases even higher

  12. IVA3: Computer code for modelling of transient three dimensional three phase flow in complicated geometry

    International Nuclear Information System (INIS)

    Kolev, N.I.

    1991-12-01

    This report describes the input and output ov IVA3 computer code and the procedure how to compile, link, and run the code. The common blocs recorded for restarts files and post processing are described in detail as well as the IVA3 interface for thermodynamic and thermo physical properties. Some recommendations for the input preparation together with some detailed comments on some architectural and functional features of the code are given in order to give some insight of the caused actions by changing some control parameters. (orig.) [de

  13. BOOK REVIEW Cracking the Einstein Code: Relativity and the Birth of Black Hole Physics With an Afterword by Roy Kerr Cracking the Einstein Code: Relativity and the Birth of Black Hole Physics With an Afterword by Roy Kerr

    Science.gov (United States)

    Carr, Bernard

    2011-02-01

    General relativity is arguably the most beautiful scientific theory ever conceived but its status within mainstream physics has vacillated since it was proposed in 1915. It began auspiciously with the successful explanation of the precession of Mercury and the dramatic confirmation of light-bending in the 1919 solar eclipse expedition, which turned Einstein into an overnight celebrity. Though little noticed at the time, there was also Karl Schwarzschild's discovery of the spherically symmetric solution in 1916 (later used to predict the existence of black holes) and Alexander Friedmann's discovery of the cosmological solution in 1922 (later confirmed by the discovery of the cosmic expansion). Then for 40 years the theory was more or less forgotten, partly because most physicists were turning their attention to the even more radical developments of quantum theory but also because the equations were too complicated to solve except in situations involving special symmetries or very weak gravitational fields (where general relativity is very similar to Newtonian theory). Furthermore, it was not clear that strong gravitational fields would ever arise in the real universe and, even if they did, it seemed unlikely that Einstein's equations could then be solved. So research in relativity became a quiet backwater as mainstream physics swept forward in other directions. Even Einstein lost interest, turning his attention to the search for a unified field theory. This book tells the remarkable story of how the tide changed in 1963, when the 28-year-old New Zealand mathematician Roy Kerr discovered an exact solution of Einstein's equations which represents a rotating black hole, thereby cracking the code of the title. The paper was just a few pages long, it being left for others to fill in the extensive beautiful mathematics which underlay the result, but it ushered in a golden age of relativity and is now one of the most cited works in physics. Coincidentally, Kerr

  14. Efficacy of physical activity interventions in post-natal populations: systematic review, meta-analysis and content coding of behaviour change techniques.

    Science.gov (United States)

    Gilinsky, Alyssa Sara; Dale, Hannah; Robinson, Clare; Hughes, Adrienne R; McInnes, Rhona; Lavallee, David

    2015-01-01

    This systematic review and meta-analysis reports the efficacy of post-natal physical activity change interventions with content coding of behaviour change techniques (BCTs). Electronic databases (MEDLINE, CINAHL and PsychINFO) were searched for interventions published from January 1980 to July 2013. Inclusion criteria were: (i) interventions including ≥1 BCT designed to change physical activity behaviour, (ii) studies reporting ≥1 physical activity outcome, (iii) interventions commencing later than four weeks after childbirth and (iv) studies including participants who had given birth within the last year. Controlled trials were included in the meta-analysis. Interventions were coded using the 40-item Coventry, Aberdeen & London - Refined (CALO-RE) taxonomy of BCTs and study quality assessment was conducted using Cochrane criteria. Twenty studies were included in the review (meta-analysis: n = 14). Seven were interventions conducted with healthy inactive post-natal women. Nine were post-natal weight management studies. Two studies included women with post-natal depression. Two studies focused on improving general well-being. Studies in healthy populations but not for weight management successfully changed physical activity. Interventions increased frequency but not volume of physical activity or walking behaviour. Efficacious interventions always included the BCTs 'goal setting (behaviour)' and 'prompt self-monitoring of behaviour'.

  15. Physical models and codes for prediction of activity release from defective fuel rods under operation conditions and in leakage tests during refuelling

    International Nuclear Information System (INIS)

    Likhanskii, V.; Evdokimov, I.; Khoruzhii, O.; Sorokin, A.; Novikov, V.

    2003-01-01

    It is appropriate to use the dependences, based on physical models, in the design-analytical codes for improving of reliability of defective fuel rod detection and for determination of defect characteristics by activity measuring in the primary coolant. In the paper the results on development of some physical models and integral mechanistic codes, assigned for prediction of defective fuel rod behaviour are presented. The analysis of mass transfer and mass exchange between fuel rod and coolant showed that the rates of these processes depends on many factors, such as coolant turbulent flow, pressure, effective hydraulic diameter of defect, fuel rod geometric parameters. The models, which describe these dependences, have been created. The models of thermomechanical fuel behaviour, stable gaseous FP release were modified and new computer code RTOP-CA was created thereupon for description of defective fuel rod behaviour and activity release into the primary coolant. The model of fuel oxidation in in-pile conditions, which includes radiolysis and RTOP-LT after validation of physical models are planned to be used for prediction of defective fuel rods behaviour

  16. Periodic Boundary Conditions in the ALEGRA Finite Element Code

    International Nuclear Information System (INIS)

    Aidun, John B.; Robinson, Allen C.; Weatherby, Joe R.

    1999-01-01

    This document describes the implementation of periodic boundary conditions in the ALEGRA finite element code. ALEGRA is an arbitrary Lagrangian-Eulerian multi-physics code with both explicit and implicit numerical algorithms. The periodic boundary implementation requires a consistent set of boundary input sets which are used to describe virtual periodic regions. The implementation is noninvasive to the majority of the ALEGRA coding and is based on the distributed memory parallel framework in ALEGRA. The technique involves extending the ghost element concept for interprocessor boundary communications in ALEGRA to additionally support on- and off-processor periodic boundary communications. The user interface, algorithmic details and sample computations are given

  17. Development of a modular systems code to analyse the implications of physics assumptions on the design of a demonstration fusion power plant

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Tobias

    2013-07-03

    The successful development and operation of a demonstration power plant (DEMO) is the next important step on roadmaps for fusion energy after ITER that is currently constructed in France. In the first phase of the development process for such devices, the conceptual design phase, the primary aim is to identify coherent designs that are composed of self-consistent sets of values for all key parameters like machine size, plasma current or magnetic field strength. This multidimensional parameter space can be explored with systems codes in order to identify areas that seem to be suited for more detailed investigation. Systems codes are composed of simplified models for all crucial systems of fusion devices that take into account all requirements and constraints of each component. This thesis is about the development of a new systems code called TREND (Tokamak Reactor code for the Evaluation of Next-step Devices). TREND is implemented with modular code architecture and consists of modules for geometry, core plasma physics, divertor, power flow, technology and costing. The main focus has been on the core physics module, since the development of TREND was done in parallel to work on physics design guidelines for DEMO. Moreover, the validation of TREND in terms of benchmarks with other European and Japanese systems codes is discussed. For these benchmarks, specific parameter sets were selected and the observed deviations were traced back to differences concerning the individual modellings. One of these parameter sets constitutes also the basis for parameter studies that were conducted with TREND. The general idea behind these studies is the analysis of implications that arise from specific assumptions on selected key parameters. Besides constant fusion power and constant additional heating power, the plasma density is fixed with respect to the Greenwald limit. The benchmarks helped particularly to detect shortages in the modellings of all involved systems codes

  18. Development of a modular systems code to analyse the implications of physics assumptions on the design of a demonstration fusion power plant

    International Nuclear Information System (INIS)

    Hartmann, Tobias

    2013-01-01

    The successful development and operation of a demonstration power plant (DEMO) is the next important step on roadmaps for fusion energy after ITER that is currently constructed in France. In the first phase of the development process for such devices, the conceptual design phase, the primary aim is to identify coherent designs that are composed of self-consistent sets of values for all key parameters like machine size, plasma current or magnetic field strength. This multidimensional parameter space can be explored with systems codes in order to identify areas that seem to be suited for more detailed investigation. Systems codes are composed of simplified models for all crucial systems of fusion devices that take into account all requirements and constraints of each component. This thesis is about the development of a new systems code called TREND (Tokamak Reactor code for the Evaluation of Next-step Devices). TREND is implemented with modular code architecture and consists of modules for geometry, core plasma physics, divertor, power flow, technology and costing. The main focus has been on the core physics module, since the development of TREND was done in parallel to work on physics design guidelines for DEMO. Moreover, the validation of TREND in terms of benchmarks with other European and Japanese systems codes is discussed. For these benchmarks, specific parameter sets were selected and the observed deviations were traced back to differences concerning the individual modellings. One of these parameter sets constitutes also the basis for parameter studies that were conducted with TREND. The general idea behind these studies is the analysis of implications that arise from specific assumptions on selected key parameters. Besides constant fusion power and constant additional heating power, the plasma density is fixed with respect to the Greenwald limit. The benchmarks helped particularly to detect shortages in the modellings of all involved systems codes

  19. Application of the new GeN-Foam multi-physics solver to the European sodium fast reactor and verification against available codes - 15226

    International Nuclear Information System (INIS)

    Fiorina, C.; Mikityuk, K.

    2015-01-01

    A new multi-physics solver for nuclear reactor analysis, named GeN-Foam (Generalized Nuclear Foam), has been developed by the FAST group at the Paul Scherrer Institut. It is based on OpenFOAM and has been developed for the multi-physics transient analyses of pin-based (e.g., liquid metal Fast Reactors, Light Water Reactors) or homogeneous (e.g., fast spectrum Molten Salt Reactors) nuclear reactors. It includes solutions of coarse or fine mesh thermal-hydraulics, thermal-mechanics and neutron diffusion. In particular, thermal-hydraulics solution can combine on the same mesh both a traditional RANS model and a porous medium model, depending on the desired degree of approximation for each region. In case the active reactor core is modeled as a porous medium, a simple sub-solver computes the sub-scale radial temperature profiles in fuel and cladding. The mesh used for neutronics calculations is deformed according to the displacement field predicted by the thermal-mechanics solver, thus allowing for a direct prediction of expansion-related feedback effects in Fast Reactors. To limit computational requirements, GeN-Foam permits the use of three different unstructured meshes for thermal-hydraulics, thermal-mechanics and neutron diffusion. For the same reason, an adaptive time step is employed. The different equations can be solved altogether or selectively included. In this work, GeN-Foam is applied to the analysis of the European Sodium Fast Reactor (ESFR). In particular, a 3-D model of the ESFR core is set up employing a coarse-mesh porous-medium approach for the thermal-hydraulics. The reactor steady-state and different accidental transients are investigated to offer an overview of GeN-Foam use and capabilities, as well as to preliminarily investigate the impact of a relatively accurate thermal-mechanic treatment on the predicted ESFR behavior. A code-to-code benchmark against the TRACE system code is performed to verify the adequacy of the results provided by the new

  20. Contribution of the Nea data bank in the field of calculation codes in radiation protection, radio physics and dosimetry

    International Nuclear Information System (INIS)

    Kodeli, I.; Sartori, E.

    2003-01-01

    The Nuclear energy agency is a specialised agency of OECD (organization economic co-operation and development). These missions are to help its members to keep and improve by international cooperation, the scientific, technological and legal bases necessary to a peaceful use of nuclear energy. Nea includes twenty eight countries. Nea works in collaboration with IAEA. The field of activities concerns the acquisition, validation and distribution of nuclear data, calculation codes and experiments. To help users, it organises conferences and training about the calculation codes that it shares out. (N.C.)

  1. Colour-coded three-dimensional reconstruction from spiral CT data sets: Improvement from the physical point of view

    International Nuclear Information System (INIS)

    Wunderlich, A.P.; Lenz, M.; Kirsten, R.; Gerhardt, P.

    1993-01-01

    This paper demonstrates the possibility of improving the spatial depth impression of colour-coded three-dimensional reconstructions by modulation of colour saturation. Patients were observed with spiral computed tomography (slice thickness 10 mm, table feed 10 mm/s, reconstruction of overlapping axial images at 2 mm increment). Interesting anatomical and pathological objects (vessels, organs, tumours, metastases) were segmented, colour-coded, and reconstructed three-dimensionally. Spatial depth impression of the coloured objects could be improved by modulating not only the brightness, but also the colour saturation. (orig.) [de

  2. Post-BEMUSE Reflood Model Input Uncertainty Methods (PREMIUM) Benchmark Phase II: Identification of Influential Parameters

    International Nuclear Information System (INIS)

    Kovtonyuk, A.; Petruzzi, A.; D'Auria, F.

    2015-01-01

    The objective of the Post-BEMUSE Reflood Model Input Uncertainty Methods (PREMIUM) benchmark is to progress on the issue of the quantification of the uncertainty of the physical models in system thermal-hydraulic codes by considering a concrete case: the physical models involved in the prediction of core reflooding. The PREMIUM benchmark consists of five phases. This report presents the results of Phase II dedicated to the identification of the uncertain code parameters associated with physical models used in the simulation of reflooding conditions. This identification is made on the basis of the Test 216 of the FEBA/SEFLEX programme according to the following steps: - identification of influential phenomena; - identification of the associated physical models and parameters, depending on the used code; - quantification of the variation range of identified input parameters through a series of sensitivity calculations. A procedure for the identification of potentially influential code input parameters has been set up in the Specifications of Phase II of PREMIUM benchmark. A set of quantitative criteria has been as well proposed for the identification of influential IP and their respective variation range. Thirteen participating organisations, using 8 different codes (7 system thermal-hydraulic codes and 1 sub-channel module of a system thermal-hydraulic code) submitted Phase II results. The base case calculations show spread in predicted cladding temperatures and quench front propagation that has been characterized. All the participants, except one, predict a too fast quench front progression. Besides, the cladding temperature time trends obtained by almost all the participants show oscillatory behaviour which may have numeric origins. Adopted criteria for identification of influential input parameters differ between the participants: some organisations used the set of criteria proposed in Specifications 'as is', some modified the quantitative thresholds

  3. Input-output supervisor

    International Nuclear Information System (INIS)

    Dupuy, R.

    1970-01-01

    The input-output supervisor is the program which monitors the flow of informations between core storage and peripheral equipments of a computer. This work is composed of three parts: 1 - Study of a generalized input-output supervisor. With sample modifications it looks like most of input-output supervisors which are running now on computers. 2 - Application of this theory on a magnetic drum. 3 - Hardware requirement for time-sharing. (author) [fr

  4. A review on the CIRCE methodology to quantify the uncertainty of the physical models of a code

    International Nuclear Information System (INIS)

    Jeon, Seong Su; Hong, Soon Joon; Bang, Young Seok

    2012-01-01

    In the field of nuclear engineering, recent regulatory audit calculations of large break loss of coolant accident (LBLOCA) have been performed with the best estimate code such as MARS, RELAP5 and CATHARE. Since the credible regulatory audit calculation is very important in the evaluation of the safety of the nuclear power plant (NPP), there have been many researches to develop rules and methodologies for the use of best estimate codes. One of the major points is to develop the best estimate plus uncertainty (BEPU) method for uncertainty analysis. As a representative BEPU method, NRC proposes the CSAU (Code scaling, applicability and uncertainty) methodology, which clearly identifies the different steps necessary for an uncertainty analysis. The general idea is 1) to determine all the sources of uncertainty in the code, also called basic uncertainties, 2) quantify them and 3) combine them in order to obtain the final uncertainty for the studied application. Using the uncertainty analysis such as CSAU methodology, an uncertainty band for the code response (calculation result), important from the safety point of view is calculated and the safety margin of the NPP is quantified. An example of such a response is the peak cladding temperature (PCT) for a LBLOCA. However, there is a problem in the uncertainty analysis with the best estimate codes. Generally, it is very difficult to determine the uncertainties due to the empiricism of closure laws (also called correlations or constitutive relationships). So far the only proposed approach is based on the expert judgment. For this case, the uncertainty range of important parameters can be wide and inaccurate so that the confidence level of the BEPU calculation results can be decreased. In order to solve this problem, recently CEA (France) proposes a statistical method of data analysis, called CIRCE. The CIRCE method is intended to quantify the uncertainties of the correlations of a code. It may replace the expert judgment

  5. Code Cactus; Code Cactus

    Energy Technology Data Exchange (ETDEWEB)

    Fajeau, M; Nguyen, L T; Saunier, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1966-09-01

    This code handles the following problems: -1) Analysis of thermal experiments on a water loop at high or low pressure; steady state or transient behavior; -2) Analysis of thermal and hydrodynamic behavior of water-cooled and moderated reactors, at either high or low pressure, with boiling permitted; fuel elements are assumed to be flat plates: - Flowrate in parallel channels coupled or not by conduction across plates, with conditions of pressure drops or flowrate, variable or not with respect to time is given; the power can be coupled to reactor kinetics calculation or supplied by the code user. The code, containing a schematic representation of safety rod behavior, is a one dimensional, multi-channel code, and has as its complement (FLID), a one-channel, two-dimensional code. (authors) [French] Ce code permet de traiter les problemes ci-dessous: 1. Depouillement d'essais thermiques sur boucle a eau, haute ou basse pression, en regime permanent ou transitoire; 2. Etudes thermiques et hydrauliques de reacteurs a eau, a plaques, a haute ou basse pression, ebullition permise: - repartition entre canaux paralleles, couples on non par conduction a travers plaques, pour des conditions de debit ou de pertes de charge imposees, variables ou non dans le temps; - la puissance peut etre couplee a la neutronique et une representation schematique des actions de securite est prevue. Ce code (Cactus) a une dimension d'espace et plusieurs canaux, a pour complement Flid qui traite l'etude d'un seul canal a deux dimensions. (auteurs)

  6. BAR-CODE BASED WEIGHT MEASUREMENT STATION FOR PHYSICAL INVENTORY TAKING OF PLUTONIUM OXIDE CONTAINERS AT THE MINING AND CHEMICAL COMBINE RADIOCHEMICAL REPROCESSING PLANT NEAR KRASNOYARSK, SIBERIA

    International Nuclear Information System (INIS)

    SUDA, S.

    1999-01-01

    This paper describes the technical tasks being implemented to computerize the physical inventory taking (PIT) at the Mining and Chemical Combine (Gorno-Khimichesky Kombinat, GKhK) radiochemical plant under the US/Russian cooperative nuclear material protection, control, and accounting (MPC and A) program. Under the MPC and A program, Lab-to-Lab task agreements with GKhK were negotiated that involved computerized equipment for item verification and confirmatory measurement of the Pu containers. Tasks under Phase I cover the work for demonstrating the plan and procedures for carrying out the comparison of the Pu container identification on the container with the computerized inventory records. In addition to the records validation, the verification procedures include the application of bar codes and bar coded TIDs to the Pu containers. Phase II involves the verification of the Pu content. A plan and procedures are being written for carrying out confirmatory measurements on the Pu containers

  7. MDS MIC Catalog Inputs

    Science.gov (United States)

    Johnson-Throop, Kathy A.; Vowell, C. W.; Smith, Byron; Darcy, Jeannette

    2006-01-01

    This viewgraph presentation reviews the inputs to the MDS Medical Information Communique (MIC) catalog. The purpose of the group is to provide input for updating the MDS MIC Catalog and to request that MMOP assign Action Item to other working groups and FSs to support the MITWG Process for developing MIC-DDs.

  8. A PC version of the Monte Carlo criticality code OMEGA

    International Nuclear Information System (INIS)

    Seifert, E.

    1996-05-01

    A description of the PC version of the Monte Carlo criticality code OMEGA is given. The report contains a general description of the code together with a detailed input description. Furthermore, some examples are given illustrating the generation of an input file. The main field of application is the calculation of the criticality of arrangements of fissionable material. Geometrically complicated arrangements that often appear inside and outside a reactor, e.g. in a fuel storage or transport container, can be considered essentially without geometrical approximations. For example, the real geometry of assemblies containing hexagonal or square lattice structures can be described in full detail. Moreover, the code can be used for special investigations in the field of reactor physics and neutron transport. Many years of practical experience and comparison with reference cases have shown that the code together with the built-in data libraries gives reliable results. OMEGA is completely independent on other widely used criticality codes (KENO, MCNP, etc.), concerning programming and the data base. It is a good practice to run difficult criticality safety problems by different independent codes in order to mutually verify the results. In this way, OMEGA can be used as a redundant code within the family of criticality codes. An advantage of OMEGA is the short calculation time: A typical criticality safety application takes only a few minutes on a Pentium PC. Therefore, the influence of parameter variations can simply be investigated by running many variants of a problem. (orig.)

  9. Hypoxia in the St. Lawrence Estuary: How a Coding Error Led to the Belief that "Physics Controls Spatial Patterns".

    Directory of Open Access Journals (Sweden)

    Daniel Bourgault

    Full Text Available Two fundamental sign errors were found in a computer code used for studying the oxygen minimum zone (OMZ and hypoxia in the Estuary and Gulf of St. Lawrence. These errors invalidate the conclusions drawn from the model, and call into question a proposed mechanism for generating OMZ that challenges classical understanding. The study in question is being cited frequently, leading the discipline in the wrong direction.

  10. Compensation crisis related to the onsite adequacy evaluation during FNA procedures-Urgent proactive input from cytopathology community is critical to establish appropriate reimbursement for CPT code 88172 (or its new counterpart if introduced in the future

    Directory of Open Access Journals (Sweden)

    Dhillon Inderpreet

    2010-01-01

    Full Text Available The confusion centered around appropriate use of the CPT billing code 88172 is addressed in the commentary from the Economic and Government Affairs Committee of the American Society of Cytopathology (ASC who have written a timely commentary in this issue of Cytojournal, "Adequate Reimbursement is Crucial to Support Cost-Effective Rapid Onsite Cytopathology Evaluations". Currently, lack of standardized use within and between pathology departments is stirring unhealthy practices of denying reimbursements for this critical and legitimate cytopathology service. This editorial discusses the important concerns raised in this commentary and recommends immediate corrective action. (See also Al-Abbadi MA, et al. Adequate reimbursement is crucial to support cost-effective rapid on-site cytopathology evaluations. CytoJournal 2010;7:22

  11. S3C: EBT Steady-State Shooting code description and user's guide

    International Nuclear Information System (INIS)

    Downum, W.B.

    1983-09-01

    The Oak Ridge National Laboratory (ORNL) one-dimensional (1-D) Steady-State Shooting code (S3C) for ELMO Bumpy Torus (EBT) plasmas is described. Benchmark calculations finding the steady-state density and electron and ion temperature profiles for a known neutral density profile and known external energy sources are carried out. Good agreement is obtained with results from the ORNL Radially Resolved Time Dependent 1-D Transport code for an EBT-Q type reactor. The program logic is described, along with the physics models in each code block and the variable names used. Sample input and output files are listed, along with the main code

  12. Physical, taxonomic code, and other data from current meter and other instruments in New York Bight from DOLPHIN and other platforms; 14 March 1971 to 03 August 1975 (NODC Accession 7601385)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Physical, taxonomic code, and other data were collected using current meter and other instruments from DOLPHIN and other platforms in New York Bight. Data were...

  13. Taxonomic code, physical, and other data collected from NOAA Ship DELAWARE II and other platforms in New York Bight from net casts and other instruments; 1973-02-20 to 1975-12-16 (NODC Accession 7601402)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Taxonomic Code, physical, and other data were collected using net casts and other instruments in the New York Bight from NOAA Ship DELAWARE II and other platforms....

  14. PLEXOS Input Data Generator

    Energy Technology Data Exchange (ETDEWEB)

    2017-02-01

    The PLEXOS Input Data Generator (PIDG) is a tool that enables PLEXOS users to better version their data, automate data processing, collaborate in developing inputs, and transfer data between different production cost modeling and other power systems analysis software. PIDG can process data that is in a generalized format from multiple input sources, including CSV files, PostgreSQL databases, and PSS/E .raw files and write it to an Excel file that can be imported into PLEXOS with only limited manual intervention.

  15. ColloInputGenerator

    DEFF Research Database (Denmark)

    2013-01-01

    This is a very simple program to help you put together input files for use in Gries' (2007) R-based collostruction analysis program. It basically puts together a text file with a frequency list of lexemes in the construction and inserts a column where you can add the corpus frequencies. It requires...... it as input for basic collexeme collostructional analysis (Stefanowitsch & Gries 2003) in Gries' (2007) program. ColloInputGenerator is, in its current state, based on programming commands introduced in Gries (2009). Projected updates: Generation of complete work-ready frequency lists....

  16. INTRA/Mod3.2. Manual and code description. Volume 2 - User's manual

    International Nuclear Information System (INIS)

    Andersson, Jenny; Edlund, O.; Hermann, J.; Johansson, Lise-Lotte

    1999-01-01

    The INTRA Manual consists of two volumes. Volume I of the manual is a thorough description of the code INTRA, the physical modelling of INTRA and the ruling numerics, and volume II, the User's Manual is an input description. This document, the User's Manual, Volume II, contains a detailed description of how to use INTRA, how to set up an input file, how to run INTRA and also post-processing

  17. INTRA/Mod3.2. Manual and code description. Volume 2 - User`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Jenny; Edlund, O.; Hermann, J.; Johansson, Lise-Lotte

    1999-01-01

    The INTRA Manual consists of two volumes. Volume I of the manual is a thorough description of the code INTRA, the physical modelling of INTRA and the ruling numerics, and volume II, the User`s Manual is an input description. This document, the User`s Manual, Volume II, contains a detailed description of how to use INTRA, how to set up an input file, how to run INTRA and also post-processing

  18. NAGRADATA. Code key. Geology

    International Nuclear Information System (INIS)

    Mueller, W.H.; Schneider, B.; Staeuble, J.

    1984-01-01

    This reference manual provides users of the NAGRADATA system with comprehensive keys to the coding/decoding of geological and technical information to be stored in or retreaved from the databank. Emphasis has been placed on input data coding. When data is retreaved the translation into plain language of stored coded information is done automatically by computer. Three keys each, list the complete set of currently defined codes for the NAGRADATA system, namely codes with appropriate definitions, arranged: 1. according to subject matter (thematically) 2. the codes listed alphabetically and 3. the definitions listed alphabetically. Additional explanation is provided for the proper application of the codes and the logic behind the creation of new codes to be used within the NAGRADATA system. NAGRADATA makes use of codes instead of plain language for data storage; this offers the following advantages: speed of data processing, mainly data retrieval, economies of storage memory requirements, the standardisation of terminology. The nature of this thesaurian type 'key to codes' makes it impossible to either establish a final form or to cover the entire spectrum of requirements. Therefore, this first issue of codes to NAGRADATA must be considered to represent the current state of progress of a living system and future editions will be issued in a loose leave ringbook system which can be updated by an organised (updating) service. (author)

  19. XSOR codes users manual

    International Nuclear Information System (INIS)

    Jow, Hong-Nian; Murfin, W.B.; Johnson, J.D.

    1993-11-01

    This report describes the source term estimation codes, XSORs. The codes are written for three pressurized water reactors (Surry, Sequoyah, and Zion) and two boiling water reactors (Peach Bottom and Grand Gulf). The ensemble of codes has been named ''XSOR''. The purpose of XSOR codes is to estimate the source terms which would be released to the atmosphere in severe accidents. A source term includes the release fractions of several radionuclide groups, the timing and duration of releases, the rates of energy release, and the elevation of releases. The codes have been developed by Sandia National Laboratories for the US Nuclear Regulatory Commission (NRC) in support of the NUREG-1150 program. The XSOR codes are fast running parametric codes and are used as surrogates for detailed mechanistic codes. The XSOR codes also provide the capability to explore the phenomena and their uncertainty which are not currently modeled by the mechanistic codes. The uncertainty distributions of input parameters may be used by an. XSOR code to estimate the uncertainty of source terms

  20. Parameter definition for reactor physics calculation of Obrigheim KWO PWR type reactor using the Gels and Erebus codes

    International Nuclear Information System (INIS)

    Faya, A.G.; Nakata, H.; Rodrigues, V.G.; Oosterkamp, W.J.

    1974-01-01

    The main variables for Obrigheim Reactor - KWO diffusion theory calculations, using the EREBUS code were defined. The variables under consideration were: mesh spacing for reactor description, time-step in burn-up calculation, and the temperature in both the moderator and the fuel. The best mesh spacing and time-step were defined considering the relative deviations and the computer time expended in each case. It has been verified that the error involved in the mean fuel temperature calculation (1317 0 K as given by SIEMENS and 1028 0 K as calculated by Dr. Penndorf) does not change substancially the calculation results

  1. Input measurements in reprocessing plants

    International Nuclear Information System (INIS)

    Trincherini, P.R.; Facchetti, S.

    1980-01-01

    The aim of this work is to give a review of the methods and the problems encountered in measurements in 'input accountability tanks' of irradiated fuel treatment plants. This study was prompted by the conviction that more and more precise techniques and methods should be at the service of safeguards organizations and that ever greater efforts should be directed towards promoting knowledge of them among operators and all those general area of interest includes the nuclear fuel cycle. The overall intent is to show the necessity of selecting methods which produce measurements which are not only more precise but are absolutely reliable both for routine plant operation and for safety checks in the input area. A description and a critical evaluation of the most common physical and chemical methods are provided, together with an estimate of the precision and accuracy obtained in real operating conditions

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

    International Nuclear Information System (INIS)

    Antunes, Alberi

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Trambauer, K. [GRS, Garching (Germany)

    1997-07-01

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

  4. Bar code usage in nuclear materials accountability

    International Nuclear Information System (INIS)

    Mee, W.T.

    1983-01-01

    The age old method of physically taking an inventory of materials by listing each item's identification number has lived beyond its usefulness. In this age of computerization, which offers the local grocery store a quick, sure, and easy means to inventory, it is time for nuclear materials facilities to automate accountability activities. The Oak Ridge Y-12 Plant began investigating the use of automated data collection devices in 1979. At that time, bar code and optical-character-recognition (OCR) systems were reviewed with the purpose of directly entering data into DYMCAS (Dynamic Special Nuclear Materials Control and Accountability System). Both of these systems appeared applicable; however, other automated devices already employed for production control made implementing the bar code and OCR seem improbable. However, the DYMCAS was placed on line for nuclear material accountability, a decision was made to consider the bar code for physical inventory listings. For the past several months a development program has been underway to use a bar code device to collect and input data to the DYMCAS on the uranium recovery operations. Programs have been completed and tested, and are being employed to ensure that data will be compatible and useful. Bar code implementation and expansion of its use for all nuclear material inventory activity in Y-12 is presented

  5. Input description for BIOPATH

    International Nuclear Information System (INIS)

    Marklund, J.E.; Bergstroem, U.; Edlund, O.

    1980-01-01

    The computer program BIOPATH describes the flow of radioactivity within a given ecosystem after a postulated release of radioactive material and the resulting dose for specified population groups. The present report accounts for the input data necessary to run BIOPATH. The report also contains descriptions of possible control cards and an input example as well as a short summary of the basic theory.(author)

  6. Input and execution

    International Nuclear Information System (INIS)

    Carr, S.; Lane, G.; Rowling, G.

    1986-11-01

    This document describes the input procedures, input data files and operating instructions for the SYVAC A/C 1.03 computer program. SYVAC A/C 1.03 simulates the groundwater mediated movement of radionuclides from underground facilities for the disposal of low and intermediate level wastes to the accessible environment, and provides an estimate of the subsequent radiological risk to man. (author)

  7. User manual of UNF code

    International Nuclear Information System (INIS)

    Zhang Jingshang

    2001-01-01

    The UNF code (2001 version) written in FORTRAN-90 is developed for calculating fast neutron reaction data of structure materials with incident energies from about 1 Kev up to 20 Mev. The code consists of the spherical optical model, the unified Hauser-Feshbach and exciton model. The man nal of the UNF code is available for users. The format of the input parameter files and the output files, as well as the functions of flag used in UNF code, are introduced in detail, and the examples of the format of input parameters files are given

  8. Converter of a continuous code into the Grey code

    International Nuclear Information System (INIS)

    Gonchar, A.I.; TrUbnikov, V.R.

    1979-01-01

    Described is a converter of a continuous code into the Grey code used in a 12-charged precision amplitude-to-digital converter to decrease the digital component of spectrometer differential nonlinearity to +0.7% in the 98% range of the measured band. To construct the converter of a continuous code corresponding to the input signal amplitude into the Grey code used is the regularity in recycling of units and zeroes in each discharge of the Grey code in the case of a continuous change of the number of pulses of a continuous code. The converter is constructed on the elements of 155 series, the frequency of continuous code pulse passing at the converter input is 25 MHz

  9. Evaluation of speedup of Monte Carlo calculations of two simple reactor physics problems coded for the GPU/CUDA environment

    International Nuclear Information System (INIS)

    Ding, Aiping; Liu, Tianyu; Liang, Chao; Ji, Wei; Shephard, Mark S.; Xu, X George; Brown, Forrest B.

    2011-01-01

    Monte Carlo simulation is ideally suited for solving Boltzmann neutron transport equation in inhomogeneous media. However, routine applications require the computation time to be reduced to hours and even minutes in a desktop system. The interest in adopting GPUs for Monte Carlo acceleration is rapidly mounting, fueled partially by the parallelism afforded by the latest GPU technologies and the challenge to perform full-size reactor core analysis on a routine basis. In this study, Monte Carlo codes for a fixed-source neutron transport problem and an eigenvalue/criticality problem were developed for CPU and GPU environments, respectively, to evaluate issues associated with computational speedup afforded by the use of GPUs. The results suggest that a speedup factor of 30 in Monte Carlo radiation transport of neutrons is within reach using the state-of-the-art GPU technologies. However, for the eigenvalue/criticality problem, the speedup was 8.5. In comparison, for a task of voxelizing unstructured mesh geometry that is more parallel in nature, the speedup of 45 was obtained. It was observed that, to date, most attempts to adopt GPUs for Monte Carlo acceleration were based on naïve implementations and have not yielded the level of anticipated gains. Successful implementation of Monte Carlo schemes for GPUs will likely require the development of an entirely new code. Given the prediction that future-generation GPU products will likely bring exponentially improved computing power and performances, innovative hardware and software solutions may make it possible to achieve full-core Monte Carlo calculation within one hour using a desktop computer system in a few years. (author)

  10. 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

  11. Improving the physical layer security of wireless communication networks using spread spectrum coding and artificial noise approach

    CSIR Research Space (South Africa)

    Adedeji, K

    2016-09-01

    Full Text Available at the application layer to protect the messages against eavesdropping. However, the evolution of strong deciphering mechanisms has made conventional cryptography-based security techniques ineffective against attacks from an intruder. Figure 1: Layer protocol... communication networks with passive and active eavesdropper,” IEEE Globecom; Wireless Communication System, pp. 4868-4873, 2012. [9] Y. Zou, X. Wang and W. Shen, “Optimal relay selection for physical layer security in cooperative wireless networks,” IEEE...

  12. 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)

  13. Rate-Compatible LDPC Codes with Linear Minimum Distance

    Science.gov (United States)

    Divsalar, Dariush; Jones, Christopher; Dolinar, Samuel

    2009-01-01

    A recently developed method of constructing protograph-based low-density parity-check (LDPC) codes provides for low iterative decoding thresholds and minimum distances proportional to block sizes, and can be used for various code rates. A code constructed by this method can have either fixed input block size or fixed output block size and, in either case, provides rate compatibility. The method comprises two submethods: one for fixed input block size and one for fixed output block size. The first mentioned submethod is useful for applications in which there are requirements for rate-compatible codes that have fixed input block sizes. These are codes in which only the numbers of parity bits are allowed to vary. The fixed-output-blocksize submethod is useful for applications in which framing constraints are imposed on the physical layers of affected communication systems. An example of such a system is one that conforms to one of many new wireless-communication standards that involve the use of orthogonal frequency-division modulation

  14. Statistical theory applications and associated computer codes

    International Nuclear Information System (INIS)

    Prince, A.

    1980-01-01

    The general format is along the same lines as that used in the O.M. Session, i.e. an introduction to the nature of the physical problems and methods of solution based on the statistical model of the nucleus. Both binary and higher multiple reactions are considered. The computer codes used in this session are a combination of optical model and statistical theory. As with the O.M. sessions, the preparation of input and analysis of output are thoroughly examined. Again, comparison with experimental data serves to demonstrate the validity of the results and possible areas for improvement. (author)

  15. Present status of reactor physics in the United States and Japan-I. 5. Development of the MVP Monte Carlo Code at JAERI

    International Nuclear Information System (INIS)

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

    2001-01-01

    The MVP general-purpose continuous-energy Monte Carlo code for neutron and photon transport calculations, together with its multigroup version GMVP, has been developed since the late 1980's at Japan Atomic Energy Research Institute (JAERI). These two codes were designed for vector supercomputers at the first stage, and then a parallel processing capability was added for several computers including workstation clusters. The first versions of the codes were released for domestic use in 1994, with cross-section libraries based on JENDL, ENDF/B, etc. Since then, many functions have been added for production use. Special features and main capabilities are as follows: 1. vectorization and parallelization; 2. combinatorial geometry with multiple-lattice capability and the statistical geometry model; 3. the probability table method for unresolved resonance; 4. realistic calculations of power reactors at arbitrary temperatures; 5. depletion calculations; 6. perturbation calculations for an eigenvalue (k eff ) problem; 7. useful tallies for improvement of the multigroup method such as effective macroscopic and/or microscopic cross sections, and so on. The MVP code is widely used in Japan, especially in the field of reactor physics analyses. Recently, the development work has concentrated on capabilities of applying the code to accelerator-driven subcritical reactors. For this purpose, we have been adding functions for the high-energy particle transport capability and simulations of the Feynman-α experiment (noise analysis). As a first step of extension of energy range and particles treated in MVP, the physics model of neutron reactions was modified to treat the (z, anything) reaction (MT = 5) in the ENDF-6 format. For a benchmark test of the modified MVP code, the TIARA shielding experiment on iron with quasi-mono-energetic p- 7 Li neutrons for E p 5 68 MeV was analyzed by using the LA- 150 cross-section library. In all the calculations, the measured spectrum of the source

  16. Group-Orthogonal Code-Division Multiplex: A Physical-Layer Enhancement for IEEE 802.11n Networks

    Directory of Open Access Journals (Sweden)

    Felip Riera-Palou

    2010-01-01

    Full Text Available The new standard for wireless local area networks (WLANs, named IEEE 802.11n, has been recently released. This new norm builds upon and remains compatible with the previous WLANs standards IEEE 802.11a/g while it is able to achieve transmission rates of up to 600 Mbps. These increased data rates are mainly a consequence of two important new features: (1 multiple antenna technology at transmission and reception, and (2 optional doubling of the system bandwidth thanks to the availability of an additional 20 MHz band. This paper proposes the use of Group-Orthogonal Code Division Multiplex (GO-CDM as a means to improve the performance of the 802.11n standard by further exploiting the inherent frequency diversity. It is explained why GO-CDM synergistically matches with the two aforementioned new features and the performance gains it can offer under different configurations is illustrated. Furthermore, the effects that group-orthogonal has on key implementation issues such as channel estimation, carrier frequency offset, and peak-to-average power ratio (PAPR are also considered.

  17. 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

  18. Nuclear reaction inputs based on effective interactions

    Energy Technology Data Exchange (ETDEWEB)

    Hilaire, S.; Peru, S.; Dubray, N.; Dupuis, M.; Bauge, E. [CEA, DAM, DIF, Arpajon (France); Goriely, S. [Universite Libre de Bruxelles, Institut d' Astronomie et d' Astrophysique, CP-226, Brussels (Belgium)

    2016-11-15

    Extensive nuclear structure studies have been performed for decades using effective interactions as sole input. They have shown a remarkable ability to describe rather accurately many types of nuclear properties. In the early 2000 s, a major effort has been engaged to produce nuclear reaction input data out of the Gogny interaction, in order to challenge its quality also with respect to nuclear reaction observables. The status of this project, well advanced today thanks to the use of modern computers as well as modern nuclear reaction codes, is reviewed and future developments are discussed. (orig.)

  19. Physics study of microbeam radiation therapy with PSI-version of Monte Carlo code GEANT as a new computational tool

    CERN Document Server

    Stepanek, J; Laissue, J A; Lyubimova, N; Di Michiel, F; Slatkin, D N

    2000-01-01

    Microbeam radiation therapy (MRT) is a currently experimental method of radiotherapy which is mediated by an array of parallel microbeams of synchrotron-wiggler-generated X-rays. Suitably selected, nominally supralethal doses of X-rays delivered to parallel microslices of tumor-bearing tissues in rats can be either palliative or curative while causing little or no serious damage to contiguous normal tissues. Although the pathogenesis of MRT-mediated tumor regression is not understood, as in all radiotherapy such understanding will be based ultimately on our understanding of the relationships among the following three factors: (1) microdosimetry, (2) damage to normal tissues, and (3) therapeutic efficacy. Although physical microdosimetry is feasible, published information on MRT microdosimetry to date is computational. This report describes Monte Carlo-based computational MRT microdosimetry using photon and/or electron scattering and photoionization cross-section data in the 1 e V through 100 GeV range distrib...

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

    International Nuclear Information System (INIS)

    Chiba, Go; Kawamoto, Yosuke; Narabayashi, Tadashi

    2016-01-01

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

  1. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    Energy Technology Data Exchange (ETDEWEB)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

  2. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    International Nuclear Information System (INIS)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions

  3. ADLIB: A simple database framework for beamline codes

    International Nuclear Information System (INIS)

    Mottershead, C.T.

    1993-01-01

    There are many well developed codes available for beamline design and analysis. A significant fraction of each of these codes is devoted to processing its own unique input language for describing the problem. None of these large, complex, and powerful codes does everything. Adding a new bit of specialized physics can be a difficult task whose successful completion makes the code even larger and more complex. This paper describes an attempt to move in the opposite direction, toward a family of small, simple, single purpose physics and utility modules, linked by an open, portable, public domain database framework. These small specialized physics codes begin with the beamline parameters already loaded in the database, and accessible via the handful of subroutines that constitute ADLIB. Such codes are easier to write, and inherently organized in a manner suitable for incorporation in model based control system algorithms. Examples include programs for analyzing beamline misalignment sensitivities, for simulating and fitting beam steering data, and for translating among MARYLIE, TRANSPORT, and TRACE3D formats

  4. Bar code usage in nuclear materials accountability

    International Nuclear Information System (INIS)

    Mee, W.T.

    1983-01-01

    The Oak Ridge Y-12 Plant began investigating the use of automated data collection devices in 1979. At this time, bar code and optical-character-recognition (OCR) systems were reviewed with the purpose of directly entering data into DYMCAS (Dynamic Special Nuclear Materials Control and Accountability System). Both of these systems appeared applicable, however, other automated devices already employed for production control made implementing the bar code and OCR seem improbable. However, the DYMCAS was placed on line for nuclear material accountability, a decision was made to consider the bar code for physical inventory listings. For the past several months a development program has been underway to use a bar code device to collect and input data to the DYMCAS on the uranium recovery operations. Programs have been completed and tested, and are being employed to ensure that data will be compatible and useful. Bar code implementation and expansion of its use for all nuclear material inventory activity in Y-12 is presented

  5. Towards an affordable alternative educational video game input device

    CSIR Research Space (South Africa)

    Smith, Andrew C

    2008-05-01

    Full Text Available The authors present the prototype design results of an alternative physical educational video gaming input device. The device elicits increased physical activity from the players as compared to the compact gaming controller. Complicated...

  6. Summary - COG: A new point-wise Monte Carlo code for burnup credit analysis

    International Nuclear Information System (INIS)

    Alesso, H.P.

    1989-01-01

    COG, a new point-wise Monte Carlo code being developed and tested at Lawrence Livermore National Laboratory (LLNL) for the Cray-1, solves the Boltzmann equation for the transport of neutrons, photons, and (in future versions) other particles. Techniques included in the code for modifying the random walk of particles make COG most suitable for solving deep-penetration (shielding) problems and a wide variety of criticality problems. COG is similar to a number of other computer codes used in the shielding community. Each code is a little different in its geometry input and its random-walk modification options. COG is a Monte Carlo code specifically designed for the CRAY (in 1986) to be as precise as the current state of physics knowledge. It has been extensively benchmarked and used as a shielding code at LLNL since 1986, and has recently been extended to accomplish criticality calculations. It will make an excellent tool for future shipping cask studies

  7. Lysimeter data as input to performance assessment source term codes

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Rogers, R.D.; Sullivan, T.

    1992-01-01

    The Field Lysimeter Investigation: Low-Level Waste Data Base Development Program is obtaining information on the performance of radioactive waste in a disposal environment. Waste forms fabricated using ion-exchange resins from EPICOR-II c prefilters employed in the cleanup of the Three Mile Island (TMI) Nuclear Power Station are being tested to develop a low-level waste data base and to obtain information on survivability of waste forms in a disposal environment. In this paper, radionuclide releases from waste forms in the first seven years of sampling are presented and discussed. Application of lysimeter data to be used in performance assessment source term models is presented. Initial results from use of data in two models are discussed

  8. Automation of Geometry Input for Building Code Compliance Check

    DEFF Research Database (Denmark)

    Petrova, Ekaterina Aleksandrova; Johansen, Peter Lind; Jensen, Rasmus Lund

    2017-01-01

    . That has left the industry in constant pursuit of possibilities for integration of the tool within the Building Information Modelling environment so that the potential provided by the latter can be harvested and the processed can be optimized. This paper presents a solution for automated data extraction...

  9. UNSPEC: revisited (semaphore code)

    International Nuclear Information System (INIS)

    Neifert, R.D.

    1981-01-01

    The UNSPEC code is used to solve the problem of unfolding an observed x-ray spectrum given the response matrix of the measuring system and the measured signal values. UNSPEC uses an iterative technique to solve the unfold problem. Due to experimental errors in the measured signal values and/or computer round-off errors, discontinuities and oscillatory behavior may occur in the iterated spectrum. These can be suppressed by smoothing the results after each iteration. Input/output options and control cards are explained; sample input and output are provided

  10. RAYIC - a numerical code for the study of ion cyclotron heating of large Tokamak plasmas

    International Nuclear Information System (INIS)

    Brambilla, M.

    1984-02-01

    The code RAYIC models coupling, propagation and absorption of e.m. waves in large axisymmetric plasmas in the ion cyclotron frequency domain. It can be used both to investigate the waves behaviour, and as a source of the power deposition profiles for use in transport codes. The present user manual, after a brief summary of the physical model, presents the structure of RAYIC, the complete list of input-output variables (calling sequence), and some examples of the output which can be obtained from the code. (orig.)

  11. FLUTAN 2.0. Input specifications

    International Nuclear Information System (INIS)

    Willerding, G.; Baumann, W.

    1996-05-01

    FLUTAN is a highly vectorized computer code for 3D fluiddynamic and thermal-hydraulic analyses in Cartesian or cylinder coordinates. It is related to the family of COMMIX codes originally developed at Argonne National Laboratory, USA, and particularly to COMMIX-1A and COMMIX-1B, which were made available to FZK in the frame of cooperation contracts within the fast reactor safety field. FLUTAN 2.0 is an improved version of the FLUTAN code released in 1992. It offers some additional innovations, e.g. the QUICK-LECUSSO-FRAM techniques for reducing numerical diffusion in the k-ε turbulence model equations; a higher sophisticated wall model for specifying a mass flow outside the surface walls together with its flow path and its associated inlet and outlet flow temperatures; and a revised and upgraded pressure boundary condition to fully include the outlet cells in the solution process of the conservation equations. Last but not least, a so-called visualization option based on VISART standards has been provided. This report contains detailed input instructions, presents formulations of the various model options, and explains how to use the code by means of comprehensive sample input. (orig.) [de

  12. PHYSICS

    CERN Multimedia

    P. Sphicas

    There have been three physics meetings since the last CMS week: “physics days” on March 27-29, the Physics/ Trigger week on April 23-27 and the most recent physics days on May 22-24. The main purpose of the March physics days was to finalize the list of “2007 analyses”, i.e. the few topics that the physics groups will concentrate on for the rest of this calendar year. The idea is to carry out a full physics exercise, with CMSSW, for select physics channels which test key features of the physics objects, or represent potential “day 1” physics topics that need to be addressed in advance. The list of these analyses was indeed completed and presented in the plenary meetings. As always, a significant amount of time was also spent in reviewing the status of the physics objects (reconstruction) as well as their usage in the High-Level Trigger (HLT). The major event of the past three months was the first “Physics/Trigger week” in Apri...

  13. Integrated Fuel-Coolant Interaction (IFCI 6.0) code

    International Nuclear Information System (INIS)

    Davis, F.J.; Young, M.F.

    1994-04-01

    The integrated Fuel-Coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, four-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a product of the effort to generate a stand-alone version of IFCI, IFCI 6.0. The User's Manual describes in detail the hydrodynamic method and physical models used in IFCI 6.0. Appendix A is an input manual, provided for the creation of working decks

  14. Coding Partitions

    Directory of Open Access Journals (Sweden)

    Fabio Burderi

    2007-05-01

    Full Text Available Motivated by the study of decipherability conditions for codes weaker than Unique Decipherability (UD, we introduce the notion of coding partition. Such a notion generalizes that of UD code and, for codes that are not UD, allows to recover the ``unique decipherability" at the level of the classes of the partition. By tacking into account the natural order between the partitions, we define the characteristic partition of a code X as the finest coding partition of X. This leads to introduce the canonical decomposition of a code in at most one unambiguouscomponent and other (if any totally ambiguouscomponents. In the case the code is finite, we give an algorithm for computing its canonical partition. This, in particular, allows to decide whether a given partition of a finite code X is a coding partition. This last problem is then approached in the case the code is a rational set. We prove its decidability under the hypothesis that the partition contains a finite number of classes and each class is a rational set. Moreover we conjecture that the canonical partition satisfies such a hypothesis. Finally we consider also some relationships between coding partitions and varieties of codes.

  15. GARFEM input deck description

    Energy Technology Data Exchange (ETDEWEB)

    Zdunek, A.; Soederberg, M. (Aeronautical Research Inst. of Sweden, Bromma (Sweden))

    1989-01-01

    The input card deck for the finite element program GARFEM version 3.2 is described in this manual. The program includes, but is not limited to, capabilities to handle the following problems: * Linear bar and beam element structures, * Geometrically non-linear problems (bar and beam), both static and transient dynamic analysis, * Transient response dynamics from a catalog of time varying external forcing function types or input function tables, * Eigenvalue solution (modes and frequencies), * Multi point constraints (MPC) for the modelling of mechanisms and e.g. rigid links. The MPC definition is used only in the geometrically linearized sense, * Beams with disjunct shear axis and neutral axis, * Beams with rigid offset. An interface exist that connects GARFEM with the program GAROS. GAROS is a program for aeroelastic analysis of rotating structures. Since this interface was developed GARFEM now serves as a preprocessor program in place of NASTRAN which was formerly used. Documentation of the methods applied in GARFEM exists but is so far limited to the capacities in existence before the GAROS interface was developed.

  16. Input or intimacy

    Directory of Open Access Journals (Sweden)

    Judit Navracsics

    2014-01-01

    Full Text Available According to the critical period hypothesis, the earlier the acquisition of a second language starts, the better. Owing to the plasticity of the brain, up until a certain age a second language can be acquired successfully according to this view. Early second language learners are commonly said to have an advantage over later ones especially in phonetic/phonological acquisition. Native-like pronunciation is said to be most likely to be achieved by young learners. However, there is evidence of accentfree speech in second languages learnt after puberty as well. Occasionally, on the other hand, a nonnative accent may appear even in early second (or third language acquisition. Cross-linguistic influences are natural in multilingual development, and we would expect the dominant language to have an impact on the weaker one(s. The dominant language is usually the one that provides the largest amount of input for the child. But is it always the amount that counts? Perhaps sometimes other factors, such as emotions, ome into play? In this paper, data obtained from an EnglishPersian-Hungarian trilingual pair of siblings (under age 4 and 3 respectively is analyzed, with a special focus on cross-linguistic influences at the phonetic/phonological levels. It will be shown that beyond the amount of input there are more important factors that trigger interference in multilingual development.

  17. WDEC: A Code for Modeling White Dwarf Structure and Pulsations

    Science.gov (United States)

    Bischoff-Kim, Agnès; Montgomery, Michael H.

    2018-05-01

    The White Dwarf Evolution Code (WDEC), written in Fortran, makes models of white dwarf stars. It is fast, versatile, and includes the latest physics. The code evolves hot (∼100,000 K) input models down to a chosen effective temperature by relaxing the models to be solutions of the equations of stellar structure. The code can also be used to obtain g-mode oscillation modes for the models. WDEC has a long history going back to the late 1960s. Over the years, it has been updated and re-packaged for modern computer architectures and has specifically been used in computationally intensive asteroseismic fitting. Generations of white dwarf astronomers and dozens of publications have made use of the WDEC, although the last true instrument paper is the original one, published in 1975. This paper discusses the history of the code, necessary to understand why it works the way it does, details the physics and features in the code today, and points the reader to where to find the code and a user guide.

  18. Development of a simulation program to study error propagation in the reprocessing input accountancy measurements

    International Nuclear Information System (INIS)

    Sanfilippo, L.

    1987-01-01

    A physical model and a computer program have been developed to simulate all the measurement operations involved with the Isotopic Dilution Analysis technique currently applied in the Volume - Concentration method for the Reprocessing Input Accountancy, together with their errors or uncertainties. The simulator is apt to easily solve a number of problems related to the measurement sctivities of the plant operator and the inspector. The program, written in Fortran 77, is based on a particular Montecarlo technique named ''Random Sampling''; a full description of the code is reported

  19. GAROS input deck description

    Energy Technology Data Exchange (ETDEWEB)

    Vollan, A.; Soederberg, M. (Aeronautical Research Inst. of Sweden, Bromma (Sweden))

    1989-01-01

    This report describes the input for the programs GAROS1 and GAROS2, version 5.8 and later, February 1988. The GAROS system, developed by Arne Vollan, Omega GmbH, is used for the analysis of the mechanical and aeroelastic properties for general rotating systems. It has been specially designed to meet the requirements of aeroelastic stability and dynamic response of horizontal axis wind energy converters. Some of the special characteristics are: * The rotor may have one or more blades. * The blades may be rigidly attached to the hub, or they may be fully articulated. * The full elastic properties of the blades, the hub, the machine house and the tower are taken into account. * With the same basic model, a number of different analyses can be performed: Snap-shot analysis, Floquet method, transient response analysis, frequency response analysis etc.

  20. Access to Research Inputs

    DEFF Research Database (Denmark)

    Czarnitzki, Dirk; Grimpe, Christoph; Pellens, Maikel

    2015-01-01

    The viability of modern open science norms and practices depends on public disclosure of new knowledge, methods, and materials. However, increasing industry funding of research can restrict the dissemination of results and materials. We show, through a survey sample of 837 German scientists in life...... sciences, natural sciences, engineering, and social sciences, that scientists who receive industry funding are twice as likely to deny requests for research inputs as those who do not. Receiving external funding in general does not affect denying others access. Scientists who receive external funding...... of any kind are, however, 50 % more likely to be denied access to research materials by others, but this is not affected by being funded specifically by industry...

  1. Access to Research Inputs

    DEFF Research Database (Denmark)

    Czarnitzki, Dirk; Grimpe, Christoph; Pellens, Maikel

    The viability of modern open science norms and practices depend on public disclosure of new knowledge, methods, and materials. However, increasing industry funding of research can restrict the dissemination of results and materials. We show, through a survey sample of 837 German scientists in life...... sciences, natural sciences, engineering, and social sciences, that scientists who receive industry funding are twice as likely to deny requests for research inputs as those who do not. Receiving external funding in general does not affect denying others access. Scientists who receive external funding...... of any kind are, however, 50% more likely to be denied access to research materials by others, but this is not affected by being funded specifically by industry....

  2. Easy web interfaces to IDL code for NSTX Data Analysis

    International Nuclear Information System (INIS)

    Davis, W.M.

    2012-01-01

    Highlights: ► Web interfaces to IDL code can be developed quickly. ► Dozens of Web Tools are used effectively on NSTX for Data Analysis. ► Web interfaces are easier to use than X-window applications. - Abstract: Reusing code is a well-known Software Engineering practice to substantially increase the efficiency of code production, as well as to reduce errors and debugging time. A variety of “Web Tools” for the analysis and display of raw and analyzed physics data are in use on NSTX [1], and new ones can be produced quickly from existing IDL [2] code. A Web Tool with only a few inputs, and which calls an IDL routine written in the proper style, can be created in less than an hour; more typical Web Tools with dozens of inputs, and the need for some adaptation of existing IDL code, can be working in a day or so. Efficiency is also increased for users of Web Tools because of the familiar interface of the web browser, and not needing X-windows, or accounts and passwords, when used within our firewall. Web Tools were adapted for use by PPPL physicists accessing EAST data stored in MDSplus with only a few man-weeks of effort; adapting to additional sites should now be even easier. An overview of Web Tools in use on NSTX, and a list of the most useful features, is also presented.

  3. PHYSICS

    CERN Multimedia

    D. Acosta

    2010-01-01

    A remarkable amount of progress has been made in Physics since the last CMS Week in June given the exponential growth in the delivered LHC luminosity. The first major milestone was the delivery of a variety of results to the ICHEP international conference held in Paris this July. For this conference, CMS prepared 15 Physics Analysis Summaries on physics objects and 22 Summaries on new and interesting physics measurements that exploited the luminosity recorded by the CMS detector. The challenge was incorporating the largest batch of luminosity that was delivered only days before the conference (300 nb-1 total). The physics covered from this initial running period spanned hadron production measurements, jet production and properties, electroweak vector boson production, and even glimpses of the top quark. Since then, the accumulated integrated luminosity has increased by a factor of more than 100, and all groups have been working tremendously hard on analysing this dataset. The September Physics Week was held ...

  4. PHYSICS

    CERN Multimedia

    J. Incandela

    There have been numerous developments in the physics area since the September CMS week. The biggest single event was the Physics/Trigger week in the end of Octo¬ber, whereas in terms of ongoing activities the “2007 analyses” went into high gear. This was in parallel with participation in CSA07 by the physics groups. On the or¬ganizational side, the new conveners of the physics groups have been selected, and a new database for man¬aging physics analyses has been deployed. Physics/Trigger week The second Physics-Trigger week of 2007 took place during the week of October 22-26. The first half of the week was dedicated to working group meetings. The ple¬nary Joint Physics-Trigger meeting took place on Wednesday afternoon and focused on the activities of the new Trigger Studies Group (TSG) and trigger monitoring. Both the Physics and Trigger organizations are now focused on readiness for early data-taking. Thus, early trigger tables and preparations for calibr...

  5. Integrating bar-code devices with computerized MC and A systems

    International Nuclear Information System (INIS)

    Anderson, L.K.; Boor, M.G.; Hurford, J.M.

    1998-01-01

    Over the past seven years, Los Alamos National Laboratory developed several generations of computerized nuclear materials control and accountability (MC and A) systems for tracking and reporting the storage, movement, and management of nuclear materials at domestic and international facilities. During the same period, Oak Ridge National Laboratory was involved with automated data acquisition (ADA) equipment, including installation of numerous bar-code scanning stations at various facilities to serve as input devices to computerized systems. Bar-code readers, as well as other ADA devices, reduce input errors, provide faster input, and allow the capture of data in remote areas where workstations do not exist. Los Alamos National Laboratory and Oak Ridge National Laboratory teamed together to implement the integration of bar-code hardware technology with computerized MC and A systems. With the expertise of both sites, the two technologies were successfully merged with little difficulty. Bar-code input is now available with several functions of the MC and A systems: material movements within material balance areas (MBAs), material movements between MBAs, and physical inventory verification. This paper describes the various components required for the integration of these MC and A systems with the installed bar-code reader devices and the future directions for these technologies

  6. Investigation on iterative multiuser detection physical layer network coding in two-way relay free-space optical links with turbulences and pointing errors.

    Science.gov (United States)

    Abu-Almaalie, Zina; Ghassemlooy, Zabih; Bhatnagar, Manav R; Le-Minh, Hoa; Aslam, Nauman; Liaw, Shien-Kuei; Lee, It Ee

    2016-11-20

    Physical layer network coding (PNC) improves the throughput in wireless networks by enabling two nodes to exchange information using a minimum number of time slots. The PNC technique is proposed for two-way relay channel free space optical (TWR-FSO) communications with the aim of maximizing the utilization of network resources. The multipair TWR-FSO is considered in this paper, where a single antenna on each pair seeks to communicate via a common receiver aperture at the relay. Therefore, chip interleaving is adopted as a technique to separate the different transmitted signals at the relay node to perform PNC mapping. Accordingly, this scheme relies on the iterative multiuser technique for detection of users at the receiver. The bit error rate (BER) performance of the proposed system is examined under the combined influences of atmospheric loss, turbulence-induced channel fading, and pointing errors (PEs). By adopting the joint PNC mapping with interleaving and multiuser detection techniques, the BER results show that the proposed scheme can achieve a significant performance improvement against the degrading effects of turbulences and PEs. It is also demonstrated that a larger number of simultaneous users can be supported with this new scheme in establishing a communication link between multiple pairs of nodes in two time slots, thereby improving the channel capacity.

  7. CITOPP, CITMOD, CITWI, Processing codes for CITATION Code

    International Nuclear Information System (INIS)

    Albarhoum, M.

    2008-01-01

    Description of program or function: CITOPP processes the output file of the CITATION 3-D diffusion code. The program can plot axial, radial and circumferential flux distributions (in cylindrical geometry) in addition to the multiplication factor convergence. The flux distributions can be drawn for each group specified by the program and visualized on the screen. CITMOD processes both the output and the input files of the CITATION 3-D diffusion code. CITMOD can visualize both axial, and radial-angular models of the reactor described by CITATION input/output files. CITWI processes the input file (CIT.INP) of CITATION 3-D diffusion code. CIT.INP is processed to deduce the dimensions of the cell whose cross sections can be representative of the homonym reactor component in section 008 of CIT.INP

  8. The CORSYS neutronics code system

    International Nuclear Information System (INIS)

    Caner, M.; Krumbein, A.D.; Saphier, D.; Shapira, M.

    1994-01-01

    The purpose of this work is to assemble a code package for LWR core physics including coupled neutronics, burnup and thermal hydraulics. The CORSYS system is built around the cell code WIMS (for group microscopic cross section calculations) and 3-dimension diffusion code CITATION (for burnup and fuel management). We are implementing such a system on an IBM RS-6000 workstation. The code was rested with a simplified model of the Zion Unit 2 PWR. (authors). 6 refs., 8 figs., 1 tabs

  9. TRACMAB. A computer code to form part of the link between the codes TRAC and MABEL

    International Nuclear Information System (INIS)

    Newbon, S.

    1982-05-01

    This report describes the function of the link program TRACMAB and provides a guide for users. The program is required to convert the thermal disequilibrium data output by the transient code TRAC into equilibrium data in a format compatible with the input data required by the code CAIN which in turn produces input data for MABEL. (author)

  10. Applications guide to the MORSE Monte Carlo code

    International Nuclear Information System (INIS)

    Cramer, S.N.

    1985-08-01

    A practical guide for the implementation of the MORESE-CG Monte Carlo radiation transport computer code system is presented. The various versions of the MORSE code are compared and contrasted, and the many references dealing explicitly with the MORSE-CG code are reviewed. The treatment of angular scattering is discussed, and procedures for obtaining increased differentiality of results in terms of reaction types and nuclides from a multigroup Monte Carlo code are explained in terms of cross-section and geometry data manipulation. Examples of standard cross-section data input and output are shown. Many other features of the code system are also reviewed, including (1) the concept of primary and secondary particles, (2) fission neutron generation, (3) albedo data capability, (4) DOMINO coupling, (5) history file use for post-processing of results, (6) adjoint mode operation, (7) variance reduction, and (8) input/output. In addition, examples of the combinatorial geometry are given, and the new array of arrays geometry feature (MARS) and its three-dimensional plotting code (JUNEBUG) are presented. Realistic examples of user routines for source, estimation, path-length stretching, and cross-section data manipulation are given. A deatiled explanation of the coupling between the random walk and estimation procedure is given in terms of both code parameters and physical analogies. The operation of the code in the adjoint mode is covered extensively. The basic concepts of adjoint theory and dimensionality are discussed and examples of adjoint source and estimator user routines are given for all common situations. Adjoint source normalization is explained, a few sample problems are given, and the concept of obtaining forward differential results from adjoint calculations is covered. Finally, the documentation of the standard MORSE-CG sample problem package is reviewed and on-going and future work is discussed

  11. Criticality benchmarks for COG: A new point-wise Monte Carlo code

    International Nuclear Information System (INIS)

    Alesso, H.P.; Pearson, J.; Choi, J.S.

    1989-01-01

    COG is a new point-wise Monte Carlo code being developed and tested at LLNL for the Cray computer. It solves the Boltzmann equation for the transport of neutrons, photons, and (in future versions) charged particles. Techniques included in the code for modifying the random walk of particles make COG most suitable for solving deep-penetration (shielding) problems. However, its point-wise cross-sections also make it effective for a wide variety of criticality problems. COG has some similarities to a number of other computer codes used in the shielding and criticality community. These include the Lawrence Livermore National Laboratory (LLNL) codes TART and ALICE, the Los Alamos National Laboratory code MCNP, the Oak Ridge National Laboratory codes 05R, 06R, KENO, and MORSE, the SACLAY code TRIPOLI, and the MAGI code SAM. Each code is a little different in its geometry input and its random-walk modification options. Validating COG consists in part of running benchmark calculations against critical experiments as well as other codes. The objective of this paper is to present calculational results of a variety of critical benchmark experiments using COG, and to present the resulting code bias. Numerous benchmark calculations have been completed for a wide variety of critical experiments which generally involve both simple and complex physical problems. The COG results, which they report in this paper, have been excellent

  12. A description of the two-dimensional combustion code FLARE

    International Nuclear Information System (INIS)

    Martin, D.

    1986-07-01

    This report gives details of the computer code FLARE. The model used for the turbulent combustion of premixed gases is described. Details of the numerical scheme used to solve the resulting equations are discussed. The input and output for the code are also described. Details of the coding are given in the Appendices together with sample input and output. (author)

  13. The PASC-3 code system and the UNIPASC environment

    International Nuclear Information System (INIS)

    Pijlgroms, B.J.; Oppe, J.; Oudshoorn, H.

    1991-08-01

    A brief description is given of the PASC-3 (Petten-AMPX-SCALE) Reactor Physics code system and its associated UNIPASC work environment. The PASC-3 code system is used for criticality and reactor calculations and consists of a selection from the Oak Ridge National Laboratory AMPX-SCALE-3 code collection complemented with a number of additional codes and nuclear data bases. The original codes have been adapted to run under the UNIX operating system. The recommended nuclear data base is a complete 219 group cross section library derived from JEF-1 of which some benchmark results are presented. By the addition of the UNIPASC work environment the usage of the code system is greatly simplified, Complex chains of programs can easily be coupled together to form a single job. In addition, the model parameters can be represented by variables instead of literal values which enhances the readability and may improve the integrity of the code inputs. (author). 8 refs.; 6 figs.; 1 tab

  14. Overview of the ArbiTER edge plasma eigenvalue code

    Science.gov (United States)

    Baver, Derek; Myra, James; Umansky, Maxim

    2011-10-01

    The Arbitrary Topology Equation Reader, or ArbiTER, is a flexible eigenvalue solver that is currently under development for plasma physics applications. The ArbiTER code builds on the equation parser framework of the existing 2DX code, extending it to include a topology parser. This will give the code the capability to model problems with complicated geometries (such as multiple X-points and scrape-off layers) or model equations with arbitrary numbers of dimensions (e.g. for kinetic analysis). In the equation parser framework, model equations are not included in the program's source code. Instead, an input file contains instructions for building a matrix from profile functions and elementary differential operators. The program then executes these instructions in a sequential manner. These instructions may also be translated into analytic form, thus giving the code transparency as well as flexibility. We will present an overview of how the ArbiTER code is to work, as well as preliminary results from early versions of this code. Work supported by the U.S. DOE.

  15. Modeling and generating input processes

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, M.E.

    1987-01-01

    This tutorial paper provides information relevant to the selection and generation of stochastic inputs to simulation studies. The primary area considered is multivariate but much of the philosophy at least is relevant to univariate inputs as well. 14 refs.

  16. PHYSICS

    CERN Multimedia

    Submitted by

    Physics Week: plenary meeting on physics groups plans for startup (14–15 May 2008) The Physics Objects (POG) and Physics Analysis (PAG) Groups presented their latest developments at the plenary meeting during the Physics Week. In the presentations particular attention was given to startup plans and readiness for data-taking. Many results based on the recent cosmic run were shown. A special Workshop on SUSY, described in a separate section, took place the day before the plenary. At the meeting, we had also two special DPG presentations on “Tracker and Muon alignment with CRAFT” (Ernesto Migliore) and “Calorimeter studies with CRAFT” (Chiara Rovelli). We had also a report from Offline (Andrea Rizzi) and Computing (Markus Klute) on the San Diego Workshop, described elsewhere in this bulletin. Tracking group (Boris Mangano). The level of sophistication of the tracking software increased significantly over the last few months: V0 (K0 and Λ) reconstr...

  17. Reprocessing input data validation

    International Nuclear Information System (INIS)

    Persiani, P.J.; Bucher, R.G.; Pond, R.B.; Cornella, R.J.

    1990-01-01

    The Isotope Correlation Technique (ICT), in conjunction with the gravimetric (Pu/U ratio) method for mass determination, provides an independent verification of the input accountancy at the dissolver or accountancy stage of the reprocessing plant. The Isotope Correlation Technique has been applied to many classes of domestic and international reactor systems (light-water, heavy-water, graphite, and liquid-metal) operating in a variety of modes (power, research, production, and breeder), and for a variety of reprocessing fuel cycle management strategies. Analysis of reprocessing operations data based on isotopic correlations derived for assemblies in a PWR environment and fuel management scheme, yielded differences between the measurement-derived and ICT-derived plutonium mass determinations of (-0.02 ± 0.23)% for the measured U-235 and (+0.50 ± 0.31)% for the measured Pu-239, for a core campaign. The ICT analyses has been implemented for the plutonium isotopics in a depleted uranium assembly in a heavy-water, enriched uranium system and for the uranium isotopes in the fuel assemblies in light-water, highly-enriched systems. 7 refs., 5 figs., 4 tabs

  18. KENO-V code

    International Nuclear Information System (INIS)

    Cramer, S.N.

    1984-01-01

    The KENO-V code is the current release of the Oak Ridge multigroup Monte Carlo criticality code development. The original KENO, with 16 group Hansen-Roach cross sections and P 1 scattering, was one ot the first multigroup Monte Carlo codes and it and its successors have always been a much-used research tool for criticality studies. KENO-V is able to accept large neutron cross section libraries (a 218 group set is distributed with the code) and has a general P/sub N/ scattering capability. A supergroup feature allows execution of large problems on small computers, but at the expense of increased calculation time and system input/output operations. This supergroup feature is activated automatically by the code in a manner which utilizes as much computer memory as is available. The primary purpose of KENO-V is to calculate the system k/sub eff/, from small bare critical assemblies to large reflected arrays of differing fissile and moderator elements. In this respect KENO-V neither has nor requires the many options and sophisticated biasing techniques of general Monte Carlo codes

  19. Simplified modeling and code usage in the PASC-3 code system by the introduction of a programming environment

    International Nuclear Information System (INIS)

    Pijlgroms, B.J.; Oppe, J.; Oudshoorn, H.L.; Slobben, J.

    1991-06-01

    A brief description is given of the PASC-3 (Petten-AMPX-SCALE) Reactor Physics code system and associated UNIPASC work environment. The PASC-3 code system is used for criticality and reactor calculations and consists of a selection from the Oak Ridge National Laboratory AMPX-SCALE-3 code collection complemented with a number of additional codes and nuclear data bases. The original codes have been adapted to run under the UNIX operating system. The recommended nuclear data base is a complete 219 group cross section library derived from JEF-1 of which some benchmark results are presented. By the addition of the UNIPASC work environment the usage of the code system is greatly simplified. Complex chains of programs can easily be coupled together to form a single job. In addition, the model parameters can be represented by variables instead of literal values which enhances the readability and may improve the integrity of the code inputs. (author). 8 refs.; 6 figs.; 1 tab

  20. Revised SRAC code system

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro; Ishiguro, Yukio; Kaneko, Kunio; Ido, Masaru.

    1986-09-01

    Since the publication of JAERI-1285 in 1983 for the preliminary version of the SRAC code system, a number of additions and modifications to the functions have been made to establish an overall neutronics code system. Major points are (1) addition of JENDL-2 version of data library, (2) a direct treatment of doubly heterogeneous effect on resonance absorption, (3) a generalized Dancoff factor, (4) a cell calculation based on the fixed boundary source problem, (5) the corresponding edit required for experimental analysis and reactor design, (6) a perturbation theory calculation for reactivity change, (7) an auxiliary code for core burnup and fuel management, etc. This report is a revision of the users manual which consists of the general description, input data requirements and their explanation, detailed information on usage, mathematics, contents of libraries and sample I/O. (author)

  1. WORM: A general-purpose input deck specification language

    International Nuclear Information System (INIS)

    Jones, T.

    1999-01-01

    Using computer codes to perform criticality safety calculations has become common practice in the industry. The vast majority of these codes use simple text-based input decks to represent the geometry, materials, and other parameters that describe the problem. However, the data specified in input files are usually processed results themselves. For example, input decks tend to require the geometry specification in linear dimensions and materials in atom or weight fractions, while the parameter of interest might be mass or concentration. The calculations needed to convert from the item of interest to the required parameter in the input deck are usually performed separately and then incorporated into the input deck. This process of calculating, editing, and renaming files to perform a simple parameter study is tedious at best. In addition, most computer codes require dimensions to be specified in centimeters, while drawings or other materials used to create the input decks might be in other units. This also requires additional calculation or conversion prior to composition of the input deck. These additional calculations, while extremely simple, introduce a source for error in both the calculations and transcriptions. To overcome these difficulties, WORM (Write One, Run Many) was created. It is an easy-to-use programming language to describe input decks and can be used with any computer code that uses standard text files for input. WORM is available, via the Internet, at worm.lanl.gov. A user's guide, tutorials, example models, and other WORM-related materials are also available at this Web site. Questions regarding WORM should be directed to wormatlanl.gov

  2. Users manual for CAFE-3D : a computational fluid dynamics fire code

    International Nuclear Information System (INIS)

    Khalil, Imane; Lopez, Carlos; Suo-Anttila, Ahti Jorma

    2005-01-01

    The Container Analysis Fire Environment (CAFE) computer code has been developed to model all relevant fire physics for predicting the thermal response of massive objects engulfed in large fires. It provides realistic fire thermal boundary conditions for use in design of radioactive material packages and in risk-based transportation studies. The CAFE code can be coupled to commercial finite-element codes such as MSC PATRAN/THERMAL and ANSYS. This coupled system of codes can be used to determine the internal thermal response of finite element models of packages to a range of fire environments. This document is a user manual describing how to use the three-dimensional version of CAFE, as well as a description of CAFE input and output parameters. Since this is a user manual, only a brief theoretical description of the equations and physical models is included

  3. TASS code topical report. V.1 TASS code technical manual

    International Nuclear Information System (INIS)

    Sim, Suk K.; Chang, W. P.; Kim, K. D.; Kim, H. C.; Yoon, H. Y.

    1997-02-01

    TASS 1.0 code has been developed at KAERI for the initial and reload non-LOCA safety analysis for the operating PWRs as well as the PWRs under construction in Korea. TASS code will replace various vendor's non-LOCA safety analysis codes currently used for the Westinghouse and ABB-CE type PWRs in Korea. This can be achieved through TASS code input modifications specific to each reactor type. The TASS code can be run interactively through the keyboard operation. A simimodular configuration used in developing the TASS code enables the user easily implement new models. TASS code has been programmed using FORTRAN77 which makes it easy to install and port for different computer environments. The TASS code can be utilized for the steady state simulation as well as the non-LOCA transient simulations such as power excursions, reactor coolant pump trips, load rejections, loss of feedwater, steam line breaks, steam generator tube ruptures, rod withdrawal and drop, and anticipated transients without scram (ATWS). The malfunctions of the control systems, components, operator actions and the transients caused by the malfunctions can be easily simulated using the TASS code. This technical report describes the TASS 1.0 code models including reactor thermal hydraulic, reactor core and control models. This TASS code models including reactor thermal hydraulic, reactor core and control models. This TASS code technical manual has been prepared as a part of the TASS code manual which includes TASS code user's manual and TASS code validation report, and will be submitted to the regulatory body as a TASS code topical report for a licensing non-LOCA safety analysis for the Westinghouse and ABB-CE type PWRs operating and under construction in Korea. (author). 42 refs., 29 tabs., 32 figs

  4. A guidance on MELCOR input preparation : An input deck for Ul-Chin 3 and 4 Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Song Won

    1997-02-01

    The objective of this study is to enhance the capability of assessing the severe accident sequence analyses and the containment behavior using MELCOR computer code and to provide the guideline of its efficient use. This report shows the method of the input deck preparation as well as the assessment strategy for the MELCOR code. MELCOR code is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. The code is being developed at Sandia National Laboratories for the U.S. NRC as a second generation plant risk assessment tool and the successor to the source term code package. The accident sequence of the reference input deck prepared in this study for Ulchin unit 3 and 4 nuclear power plants, is the total loss of feedwater (TLOFW) without any success of safety systems, which is similar to station blackout (TLMB). It is very useful to simulate a well-known sequence through the best estimated code or experiment, because the results of the simulation before core melt can be compared with the FSAR, but no data is available after core melt. The precalculation for the TLOFW using the reference input deck is performed successfully as expected. The other sequences will be carried out with minor changes in the reference input. This input deck will be improved continually by the adding of the safety systems not included in this input deck, and also through the sensitivity and uncertainty analyses. (author). 19 refs., 10 tabs., 55 figs.

  5. A guidance on MELCOR input preparation : An input deck for Ul-Chin 3 and 4 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Cho, Song Won.

    1997-02-01

    The objective of this study is to enhance the capability of assessing the severe accident sequence analyses and the containment behavior using MELCOR computer code and to provide the guideline of its efficient use. This report shows the method of the input deck preparation as well as the assessment strategy for the MELCOR code. MELCOR code is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. The code is being developed at Sandia National Laboratories for the U.S. NRC as a second generation plant risk assessment tool and the successor to the source term code package. The accident sequence of the reference input deck prepared in this study for Ulchin unit 3 and 4 nuclear power plants, is the total loss of feedwater (TLOFW) without any success of safety systems, which is similar to station blackout (TLMB). It is very useful to simulate a well-known sequence through the best estimated code or experiment, because the results of the simulation before core melt can be compared with the FSAR, but no data is available after core melt. The precalculation for the TLOFW using the reference input deck is performed successfully as expected. The other sequences will be carried out with minor changes in the reference input. This input deck will be improved continually by the adding of the safety systems not included in this input deck, and also through the sensitivity and uncertainty analyses. (author). 19 refs., 10 tabs., 55 figs

  6. The DarkStars code: a publicly available dark stellar evolution package

    CERN Document Server

    Scott, Pat; Fairbairn, Malcolm

    2009-01-01

    We announce the public release of the 'dark' stellar evolution code DarkStars. The code simultaneously solves the equations of WIMP capture and annihilation in a star with those of stellar evolution assuming approximate hydrostatic equilibrium. DarkStars includes the most extensive WIMP microphysics of any dark evolution code to date. The code employs detailed treatments of the capture process from a range of WIMP velocity distributions, as well as composite WIMP distribution and conductive energy transport schemes based on the WIMP mean-free path in the star. We give a brief description of the input physics and practical usage of the code, as well as examples of its application to dark stars at the Galactic centre.

  7. PHYSICS

    CERN Multimedia

    D. Futyan

    A lot has transpired on the “Physics” front since the last CMS Bulletin. The summer was filled with preparations of new Monte Carlo samples based on CMSSW_3, the finalization of all the 10 TeV physics analyses [in total 50 analyses were approved] and the preparations for the Physics Week in Bologna. A couple weeks later, the “October Exercise” commenced and ran through an intense two-week period. The Physics Days in October were packed with a number of topics that are relevant to data taking, in a number of “mini-workshops”: the luminosity measurement, the determination of the beam spot and the measurement of the missing transverse energy (MET) were the three main topics.  Physics Week in Bologna The second physics week in 2009 took place in Bologna, Italy, on the week of Sep 7-11. The aim of the week was to review and establish how ready we are to do physics with the early collisions at the LHC. The agenda of the week was thus pac...

  8. PHYSICS

    CERN Multimedia

    D. Futyan

    A lot has transpired on the “Physics” front since the last CMS Bulletin. The summer was filled with preparations of new Monte Carlo samples based on CMSSW_3, the finalization of all the 10 TeV physics analyses [in total 50 analyses were approved] and the preparations for the Physics Week in Bologna. A couple weeks later, the “October Exercise” commenced and ran through an intense two-week period. The Physics Days in October were packed with a number of topics that are relevant to data taking, in a number of “mini-workshops”: the luminosity measurement, the determination of the beam spot and the measurement of the missing transverse energy (MET) were the three main topics.   Physics Week in Bologna The second physics week in 2009 took place in Bologna, Italy, on the week of Sep 7-11. The aim of the week was to review and establish (we hoped) the readiness of CMS to do physics with the early collisions at the LHC. The agenda of the...

  9. RFQ simulation code

    International Nuclear Information System (INIS)

    Lysenko, W.P.

    1984-04-01

    We have developed the RFQLIB simulation system to provide a means to systematically generate the new versions of radio-frequency quadrupole (RFQ) linac simulation codes that are required by the constantly changing needs of a research environment. This integrated system simplifies keeping track of the various versions of the simulation code and makes it practical to maintain complete and up-to-date documentation. In this scheme, there is a certain standard version of the simulation code that forms a library upon which new versions are built. To generate a new version of the simulation code, the routines to be modified or added are appended to a standard command file, which contains the commands to compile the new routines and link them to the routines in the library. The library itself is rarely changed. Whenever the library is modified, however, this modification is seen by all versions of the simulation code, which actually exist as different versions of the command file. All code is written according to the rules of structured programming. Modularity is enforced by not using COMMON statements, simplifying the relation of the data flow to a hierarchy diagram. Simulation results are similar to those of the PARMTEQ code, as expected, because of the similar physical model. Different capabilities, such as those for generating beams matched in detail to the structure, are available in the new code for help in testing new ideas in designing RFQ linacs

  10. Computer Generated Inputs for NMIS Processor Verification

    International Nuclear Information System (INIS)

    J. A. Mullens; J. E. Breeding; J. A. McEvers; R. W. Wysor; L. G. Chiang; J. R. Lenarduzzi; J. T. Mihalczo; J. K. Mattingly

    2001-01-01

    Proper operation of the Nuclear Identification Materials System (NMIS) processor can be verified using computer-generated inputs [BIST (Built-In-Self-Test)] at the digital inputs. Preselected sequences of input pulses to all channels with known correlation functions are compared to the output of the processor. These types of verifications have been utilized in NMIS type correlation processors at the Oak Ridge National Laboratory since 1984. The use of this test confirmed a malfunction in a NMIS processor at the All-Russian Scientific Research Institute of Experimental Physics (VNIIEF) in 1998. The NMIS processor boards were returned to the U.S. for repair and subsequently used in NMIS passive and active measurements with Pu at VNIIEF in 1999

  11. Language Recognition via Sparse Coding

    Science.gov (United States)

    2016-09-08

    explanation is that sparse coding can achieve a near-optimal approximation of much complicated nonlinear relationship through local and piecewise linear...training examples, where x(i) ∈ RN is the ith example in the batch. Optionally, X can be normalized and whitened before sparse coding for better result...normalized input vectors are then ZCA- whitened [20]. Em- pirically, we choose ZCA- whitening over PCA- whitening , and there is no dimensionality reduction

  12. PHYSICS

    CERN Multimedia

    J. Incandela

    The all-plenary format of the CMS week in Cyprus gave the opportunity to the conveners of the physics groups to present the plans of each physics analysis group for tackling early physics analyses. The presentations were complete, so all are encouraged to browse through them on the Web. There is a wealth of information on what is going on, by whom and on what basis and priority. The CMS week was followed by two CMS “physics events”, the ICHEP08 days and the physics days in July. These were two weeks dedicated to either the approval of all the results that would be presented at ICHEP08, or to the review of all the other Monte-Carlo based analyses that were carried out in the context of our preparations for analysis with the early LHC data (the so-called “2008 analyses”). All this was planned in the context of the beginning of a ramp down of these Monte Carlo efforts, in anticipation of data.  The ICHEP days are described below (agenda and talks at: http://indic...

  13. PHYSICS

    CERN Multimedia

    Joe Incandela

    There have been two plenary physics meetings since the December CMS week. The year started with two workshops, one on the measurements of the Standard Model necessary for “discovery physics” as well as one on the Physics Analysis Toolkit (PAT). Meanwhile the tail of the “2007 analyses” is going through the last steps of approval. It is expected that by the end of January all analyses will have converted to using the data from CSA07 – which include the effects of miscalibration and misalignment. January Physics Days The first Physics Days of 2008 took place on January 22-24. The first two days were devoted to comprehensive re¬ports from the Detector Performance Groups (DPG) and Physics Objects Groups (POG) on their planning and readiness for early data-taking followed by approvals of several recent studies. Highlights of POG presentations are included below while the activities of the DPGs are covered elsewhere in this bulletin. January 24th was devo...

  14. Speaking Code

    DEFF Research Database (Denmark)

    Cox, Geoff

    Speaking Code begins by invoking the “Hello World” convention used by programmers when learning a new language, helping to establish the interplay of text and code that runs through the book. Interweaving the voice of critical writing from the humanities with the tradition of computing and software...

  15. A MHD equilibrium code 'EQUCIR version 2' applicable to up-down asymmetric toroidal plasma

    International Nuclear Information System (INIS)

    Shinya, Kichiro; Ninomiya, Hiromasa

    1981-01-01

    Computer code EQUCIR version 2, which can analyse tokamak plasma equilibrium without assuming up-down symmetry with respect to the mid-plane, has been developed. This code is essentially the same as EQUCIR version 1 which has already been reported and can deal with only symmetrical plasma with respect to the mid-plane. Because data input stream is slightly different from version 1 physical background of the change and the method of calculation are explained. Data input manual for the different points is also summarized. The code has been applied to the analysis of INTOR single-null divertor plasmas and to the design of hybrid poloidal coils resulting in useful and powerful means for the design. (author)

  16. DUSTMS-D: DISPOSAL UNIT SOURCE TERM - MULTIPLE SPECIES - DISTRIBUTED FAILURE DATA INPUT GUIDE.

    Energy Technology Data Exchange (ETDEWEB)

    SULLIVAN, T.M.

    2006-01-01

    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 source term). The focus of this work is to develop a methodology for calculating the source term. In general, the source term is influenced by the radionuclide inventory, the wasteforms and containers used to dispose of the inventory, and the physical processes that lead to release from the facility (fluid flow, container degradation, wasteform leaching, and radionuclide transport). Many of these physical processes are influenced by the design of the disposal facility (e.g., how the engineered barriers control infiltration of water). The complexity of the problem and the absence of appropriate data prevent development of an entirely mechanistic representation of radionuclide release from a disposal facility. Typically, a number of assumptions, based on knowledge of the disposal system, are used to simplify the problem. This has been done and the resulting models have been incorporated into the computer code DUST-MS (Disposal Unit Source Term-Multiple Species). The DUST-MS computer code is designed to model water flow, container degradation, release of contaminants from the wasteform to the contacting solution and transport through the subsurface media. Water flow through the facility over time is modeled using tabular input. Container degradation models include three types of failure rates: (a) instantaneous (all containers in a control volume fail at once), (b) uniformly distributed failures (containers fail at a linear rate between a specified starting and ending time), and (c) gaussian failure rates (containers fail at a rate determined by a mean failure time, standard deviation and gaussian distribution). Wasteform release models include four release mechanisms: (a) rinse with partitioning (inventory is released instantly upon container failure subject to equilibrium partitioning (sorption) with

  17. Physics

    CERN Document Server

    Cullen, Katherine

    2005-01-01

    Defined as the scientific study of matter and energy, physics explains how all matter behaves. Separated into modern and classical physics, the study attracts both experimental and theoretical physicists. From the discovery of the process of nuclear fission to an explanation of the nature of light, from the theory of special relativity to advancements made in particle physics, this volume profiles 10 pioneers who overcame tremendous odds to make significant breakthroughs in this heavily studied branch of science. Each chapter contains relevant information on the scientist''s childhood, research, discoveries, and lasting contributions to the field and concludes with a chronology and a list of print and Internet references specific to that individual.

  18. Six axis force feedback input device

    Science.gov (United States)

    Ohm, Timothy (Inventor)

    1998-01-01

    The present invention is a low friction, low inertia, six-axis force feedback input device comprising an arm with double-jointed, tendon-driven revolute joints, a decoupled tendon-driven wrist, and a base with encoders and motors. The input device functions as a master robot manipulator of a microsurgical teleoperated robot system including a slave robot manipulator coupled to an amplifier chassis, which is coupled to a control chassis, which is coupled to a workstation with a graphical user interface. The amplifier chassis is coupled to the motors of the master robot manipulator and the control chassis is coupled to the encoders of the master robot manipulator. A force feedback can be applied to the input device and can be generated from the slave robot to enable a user to operate the slave robot via the input device without physically viewing the slave robot. Also, the force feedback can be generated from the workstation to represent fictitious forces to constrain the input device's control of the slave robot to be within imaginary predetermined boundaries.

  19. Documentation of CATHENA input files for the APOLLO computer

    International Nuclear Information System (INIS)

    1988-06-01

    Input files created for the VAX version of the CATHENA two-fluid code have been modified and documented for simulation on the AECB's APOLLO computer system. The input files describe the RD-14 thermalhydraulic loop, the RD-14 steam generator, the RD-12 steam generator blowdown test facility, the Stern Laboratories Cold Water Injection Facility (CWIT), and a CANDU 600 reactor. Sample CATHENA predictions are given and compared with experimental results where applicable. 24 refs

  20. Global sensitivity analysis of computer models with functional inputs

    International Nuclear Information System (INIS)

    Iooss, Bertrand; Ribatet, Mathieu

    2009-01-01

    Global sensitivity analysis is used to quantify the influence of uncertain model inputs on the response variability of a numerical model. The common quantitative methods are appropriate with computer codes having scalar model inputs. This paper aims at illustrating different variance-based sensitivity analysis techniques, based on the so-called Sobol's indices, when some model inputs are functional, such as stochastic processes or random spatial fields. In this work, we focus on large cpu time computer codes which need a preliminary metamodeling step before performing the sensitivity analysis. We propose the use of the joint modeling approach, i.e., modeling simultaneously the mean and the dispersion of the code outputs using two interlinked generalized linear models (GLMs) or generalized additive models (GAMs). The 'mean model' allows to estimate the sensitivity indices of each scalar model inputs, while the 'dispersion model' allows to derive the total sensitivity index of the functional model inputs. The proposed approach is compared to some classical sensitivity analysis methodologies on an analytical function. Lastly, the new methodology is applied to an industrial computer code that simulates the nuclear fuel irradiation.

  1. Modifications to the Monte Carlo neutronics code MONK

    International Nuclear Information System (INIS)

    Hutton, J.L.

    1979-09-01

    The Monte Carlo neutronics code MONK has been widely used for criticality calculations, and is one of the standard methods for assessing the safety of transport flasks and fuel storage facilities in the UK. Recently, attempts have been made to extend the range of applications of this calculational technique. In particular studies have been carried out using Monte Carlo to analyse reactor physics experiments. In these applications various shortcomings of the standard version MONK5 became apparent. The basic data library was found to be inadequate and additional estimates of parameters (eg power distribution) not normally included in criticality studies were required. These features which required improvement, primarily in the context of using the code for reactor physics calculations, are enumerated. To facilitate the use of the code as a reactor physics calculational tool a series of modifications have been carried out. The code has been modified so that the user can use group data tabulations of the cross sections instead of the present 'point' data values. The code can now interface with a number of reactor physics group data preparation schemes but in particular it can use WIMS-E interfaces as a source of group data. Details of the changes are outlined and a new version of MONK incorporating these modifications has been created. This version is called MONK5W. This paper provides a guide to the use of this version. The data input is described along with other details required to use this code on the Harwell IBM 3033. To aid the user, examples of calculations using the new facilities incorporated in MONK5W are given. (UK)

  2. Tokamak simulation code manual

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  3. Development of MIDAS/SMR Input Deck for SMART

    International Nuclear Information System (INIS)

    Cho, S. W.; Oh, H. K.; Lee, J. M.; Lee, J. H.; Yoo, K. J.; Kwun, S. K.; Hur, H.

    2010-01-01

    The objective of this study is to develop MIDAS/SMR code basic input deck for the severe accidents by simulating the steady state for the SMART plant. SMART plant is an integrated reactor developed by KAERI. For the assessment of reactor safety and severe accident management strategy, it is necessary to simulate severe accidents using the MIDAS/SMR code which is being developed by KAERI. The input deck of the MIDAS/SMR code for the SMART plant is prepared to simulate severe accident sequences for the users who are not familiar with the code. A steady state is obtained and the results are compared with design values. The input deck will be improved through the simulation of the DBAs and severe accidents. The base input deck of the MIDAS/SMR code can be used to simulate severe accident scenarios after improvement. Source terms and hydrogen generation can be analyzed through the simulation of the severe accident. The information gained from analyses of severe accidents is expected to be helpful to develop the severe accident management strategy

  4. PHYSICS

    CERN Multimedia

    Guenther Dissertori

    The time period between the last CMS week and this June was one of intense activity with numerous get-together targeted at addressing specific issues on the road to data-taking. The two series of workshops, namely the “En route to discoveries” series and the “Vertical Integration” meetings continued.   The first meeting of the “En route to discoveries” sequence (end 2007) had covered the measurements of the Standard Model signals as necessary prerequisite to any claim of signals beyond the Standard Model. The second meeting took place during the Feb CMS week and concentrated on the commissioning of the Physics Objects, whereas the third occurred during the April Physics Week – and this time the theme was the strategy for key new physics signatures. Both of these workshops are summarized below. The vertical integration meetings also continued, with two DPG-physics get-togethers on jets and missing ET and on electrons and photons. ...

  5. PHYSICS

    CERN Multimedia

    Chris Hill

    2012-01-01

    The months that have passed since the last CMS Bulletin have been a very busy and exciting time for CMS physics. We have gone from observing the very first 8TeV collisions produced by the LHC to collecting a dataset of the collisions that already exceeds that recorded in all of 2011. All in just a few months! Meanwhile, the analysis of the 2011 dataset and publication of the subsequent results has continued. These results come from all the PAGs in CMS, including searches for the Higgs boson and other new phenomena, that have set the most stringent limits on an ever increasing number of models of physics beyond the Standard Model including dark matter, Supersymmetry, and TeV-scale gravity scenarios, top-quark physics where CMS has overtaken the Tevatron in the precision of some measurements, and bottom-quark physics where CMS made its first discovery of a new particle, the Ξ*0b baryon (candidate event pictured below). Image 2:  A Ξ*0b candidate event At the same time POGs and PAGs...

  6. PHYSICS

    CERN Multimedia

    D. Acosta

    2011-01-01

    Since the last CMS Week, all physics groups have been extremely active on analyses based on the full 2010 dataset, with most aiming for a preliminary measurement in time for the winter conferences. Nearly 50 analyses were approved in a “marathon” of approval meetings during the first two weeks of March, and the total number of approved analyses reached 90. The diversity of topics is very broad, including precision QCD, Top, and electroweak measurements, the first observation of single Top production at the LHC, the first limits on Higgs production at the LHC including the di-tau final state, and comprehensive searches for new physics in a wide range of topologies (so far all with null results unfortunately). Most of the results are based on the full 2010 pp data sample, which corresponds to 36 pb-1 at √s = 7 TeV. This report can only give a few of the highlights of a very rich physics program, which is listed below by physics group...

  7. Code system BCG for gamma-ray skyshine calculation

    International Nuclear Information System (INIS)

    Ryufuku, Hiroshi; Numakunai, Takao; Miyasaka, Shun-ichi; Minami, Kazuyoshi.

    1979-03-01

    A code system BCG has been developed for calculating conveniently and efficiently gamma-ray skyshine doses using the transport calculation codes ANISN and DOT and the point-kernel calculation codes G-33 and SPAN. To simplify the input forms to the system, the forms for these codes are unified, twelve geometric patterns are introduced to give material regions, and standard data are available as a library. To treat complex arrangements of source and shield, it is further possible to use successively the code such that the results from one code may be used as input data to the same or other code. (author)

  8. Radioactive releases of nuclear power plants: the code ASTEC

    International Nuclear Information System (INIS)

    Sdouz, G.; Pachole, M.

    1999-11-01

    In order to adopt potential countermeasures to protect the population during the course of an accident in a nuclear power plant a fast prediction of the radiation exposure is necessary. The basic input value for such a dispersion calculation is the source term, which is the description of the physical and chemical behavior of the released radioactive nuclides. Based on a source term data base a pilot system has been developed to determine a relevant source term and to generate the input file for the dispersion code TAMOS of the Zentralanstalt fuer Meteorologie und Geodynamik (ZAMG). This file can be sent directly as an attachment of e-mail to the TAMOS user for further processing. The source terms for 56 European nuclear power plant units are included in the pilot version of the code ASTEC (Austrian Source Term Estimation Code). The use of the system is demonstrated in an example based on an accident in the unit TEMELIN-1. In order to calculate typical core inventories for the data bank the international computer code OBIGEN 2.1 was installed and applied. The report has been completed with a discussion on the optimal data transfer. (author)

  9. APR1400 Containment Simulation with CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-05-15

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

  10. APR1400 Containment Simulation with CONTAIN code

    International Nuclear Information System (INIS)

    Hwang, Moon Kyu; Chung, Bub Dong

    2010-01-01

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

  11. ORIGNATE: PC input processor for ORIGEN-S

    International Nuclear Information System (INIS)

    Bowman, S.M.

    1992-01-01

    ORIGNATE is a personal computer program that serves as a user- friendly interface for the ORIGEN-S isotopic generation and depletion code. It is designed to assist an ORIGEN-S user in preparing an input file for execution of light-water-reactor fuel depletion and decay cases. Output from ORIGNATE is a card-image input file that may be uploaded to a mainframe computer to execute ORIGEN-S in SCALE-4. ORIGNATE features a pulldown menu system that accesses sophisticated data entry screens. The program allows the user to quickly set up an ORIGEN-S input file and perform error checking

  12. An Exploration of Input Conditions for Virtual Teleportation

    DEFF Research Database (Denmark)

    Høeg, Emil Rosenlund; Ruder, Kevin Vignola; Nilsson, Niels Chr.

    2017-01-01

    This poster describes a within-groups study (n=17) comparing participants' experience of three different input conditions for instigating virtual teleportation (button clicking, physical jumping, and fist clenching). The results indicated that teleportation by clicking a button generally required...

  13. PHYSICS

    CERN Multimedia

    Darin Acosta

    2010-01-01

    The collisions last year at 900 GeV and 2.36 TeV provided the long anticipated collider data to the CMS physics groups. Quite a lot has been accomplished in a very short time. Although the delivered luminosity was small, CMS was able to publish its first physics paper (with several more in preparation), and commence the commissioning of physics objects for future analyses. Many new performance results have been approved in advance of this CMS Week. One remarkable outcome has been the amazing agreement between out-of-the-box data with simulation at these low energies so early in the commissioning of the experiment. All of this is testament to the hard work and preparation conducted beforehand by many people in CMS. These analyses could not have happened without the dedicated work of the full collaboration on building and commissioning the detector, computing, and software systems combined with the tireless work of many to collect, calibrate and understand the data and our detector. To facilitate the efficien...

  14. PHYSICS

    CERN Multimedia

    D. Acosta

    2010-01-01

    The Physics Groups are actively engaged on analyses of the first data from the LHC at 7 TeV, targeting many results for the ICHEP conference taking place in Paris this summer. The first large batch of physics approvals is scheduled for this CMS Week, to be followed by four more weeks of approvals and analysis updates leading to the start of the conference in July. Several high priority analysis areas were organized into task forces to ensure sufficient coverage from the relevant detector, object, and analysis groups in the preparation of these analyses. Already some results on charged particle correlations and multiplicities in 7 TeV minimum bias collisions have been approved. Only one small detail remains before ICHEP: further integrated luminosity delivered by the LHC! Beyond the Standard Model measurements that can be done with these data, the focus changes to the search for new physics at the TeV scale and for the Higgs boson in the period after ICHEP. Particle Flow The PFT group is focusing on the ...

  15. PHYSICS

    CERN Multimedia

    the PAG conveners

    2011-01-01

    The delivered LHC integrated luminosity of more than 1 inverse femtobarn by summer and more than 5 by the end of 2011 has been a gold mine for the physics groups. With 2011 data, we have submitted or published 14 papers, 7 others are in collaboration-wide review, and 75 Physics Analysis Summaries have been approved already. They add to the 73 papers already published based on the 2010 and 2009 datasets. Highlights from each physics analysis group are described below. Heavy ions Many important results have been obtained from the first lead-ion collision run in 2010. The published measurements include the first ever indications of Υ excited state suppression (PRL synopsis), long-range correlation in PbPb, and track multiplicity over a wide η range. Preliminary results include the first ever measurement of isolated photons (showing no modification), J/ψ suppression including the separation of the non-prompt component, further study of jet fragmentation, nuclear modification factor...

  16. PHYSICS

    CERN Multimedia

    L. Demortier

    Physics-wise, the CMS week in December was dominated by discussions of the analyses that will be carried out in the “next six months”, i.e. while waiting for the first LHC collisions.  As presented in December, analysis approvals based on Monte Carlo simulation were re-opened, with the caveat that for this work to be helpful to the goals of CMS, it should be carried out using the new software (CMSSW_2_X) and associated samples.  By the end of the week, the goal for the physics groups was set to be the porting of our physics commissioning methods and plans, as well as the early analyses (based an integrated luminosity in the range 10-100pb-1) into this new software. Since December, the large data samples from CMSSW_2_1 were completed. A big effort by the production group gave a significant number of events over the end-of-year break – but also gave out the first samples with the fast simulation. Meanwhile, as mentioned in December, the arrival of 2_2 meant that ...

  17. PHYSICS

    CERN Multimedia

    C. Hill

    2012-01-01

      2012 has started off as a very busy year for the CMS Physics Groups. Planning for the upcoming higher luminosity/higher energy (8 TeV) operation of the LHC and relatively early Rencontres de Moriond are the high-priority activities for the group at the moment. To be ready for the coming 8-TeV data, CMS has made a concerted effort to perform and publish analyses on the 5 fb−1 dataset recorded in 2011. This has resulted in the submission of 16 papers already, including nine on the search for the Higgs boson. In addition, a number of preliminary results on the 2011 dataset have been released to the public. The Exotica and SUSY groups approved several searches for new physics in January, such as searches for W′ and exotic highly ionising particles. These were highlighted at a CERN seminar given on 24th  January. Many more analyses, from all the PAGs, including the newly formed SMP (Standard Model Physics) and FSQ (Forward and Small-x QCD), were approved in February. The ...

  18. PHYSICS

    CERN Document Server

    C. Hill

    2012-01-01

      The period since the last CMS Bulletin has been historic for CMS Physics. The pinnacle of our physics programme was an observation of a new particle – a strong candidate for a Higgs boson – which has captured worldwide interest and made a profound impact on the very field of particle physics. At the time of the discovery announcement on 4 July, 2012, prominent signals were observed in the high-resolution H→γγ and H→ZZ(4l) modes. Corroborating excess was observed in the H→W+W– mode as well. The fermionic channel analyses (H→bb, H→ττ), however, yielded less than the Standard Model (SM) expectation. Collectively, the five channels established the signal with a significance of five standard deviations. With the exception of the diphoton channel, these analyses have all been updated in the last months and several new channels have been added. With improved analyses and more than twice the i...

  19. Coding Labour

    Directory of Open Access Journals (Sweden)

    Anthony McCosker

    2014-03-01

    Full Text Available As well as introducing the Coding Labour section, the authors explore the diffusion of code across the material contexts of everyday life, through the objects and tools of mediation, the systems and practices of cultural production and organisational management, and in the material conditions of labour. Taking code beyond computation and software, their specific focus is on the increasingly familiar connections between code and labour with a focus on the codification and modulation of affect through technologies and practices of management within the contemporary work organisation. In the grey literature of spreadsheets, minutes, workload models, email and the like they identify a violence of forms through which workplace affect, in its constant flux of crisis and ‘prodromal’ modes, is regulated and governed.

  20. [Prosody, speech input and language acquisition].

    Science.gov (United States)

    Jungheim, M; Miller, S; Kühn, D; Ptok, M

    2014-04-01

    In order to acquire language, children require speech input. The prosody of the speech input plays an important role. In most cultures adults modify their code when communicating with children. Compared to normal speech this code differs especially with regard to prosody. For this review a selective literature search in PubMed and Scopus was performed. Prosodic characteristics are a key feature of spoken language. By analysing prosodic features, children gain knowledge about underlying grammatical structures. Child-directed speech (CDS) is modified in a way that meaningful sequences are highlighted acoustically so that important information can be extracted from the continuous speech flow more easily. CDS is said to enhance the representation of linguistic signs. Taking into consideration what has previously been described in the literature regarding the perception of suprasegmentals, CDS seems to be able to support language acquisition due to the correspondence of prosodic and syntactic units. However, no findings have been reported, stating that the linguistically reduced CDS could hinder first language acquisition.

  1. MACRO1: a code to test a methodology for analyzing nuclear-waste management systems

    International Nuclear Information System (INIS)

    Edwards, L.L.

    1979-01-01

    The code is primarily a manager of probabilistic data and deterministic mathematical models. The user determines the desired aggregation of the available models into a composite model of a physical system. MACRO1 then propagates the finite probability distributions of the inputs to the model to finite probability distributions over the outputs. MACRO1 has been applied to a sample analysis of a nuclear-waste repository, and its results compared satisfactorily with previously obtained Monte Carlo statistics

  2. The SWAN coupling code: user's guide

    International Nuclear Information System (INIS)

    Litaudon, X.; Moreau, D.

    1988-11-01

    Coupling of slow waves in a plasma near the lower hybrid frequency is well known and linear theory with density step followed by a constant gradient can be used with some confidence. With the aid of the computer code SWAN, which stands for 'Slow Wave Antenna', the following parameters can be numerically calculated: n parallel power spectrum, directivity (weighted by the current drive efficiency), reflection coefficients (amplitude and phase) both before and after the E-plane junctions, scattering matrix at the plasma interface, scattering matrix at the E-plane junctions, maximum electric fields in secondary waveguides and location where it occurs, effect of passive waveguides on each side of the antenna, and the effect of a finite magnetic field in front of the antenna (for homogeneous plasma). This manual gives the basic information on the main assumptions of the coupling theory and on the use and general structure of the code itself. It answers the questions what are the main assumptions of the physical model? how to execute a job? what are the input parameters of the code? and what are the output results and where are they written? (author)

  3. GOC: General Orbit Code

    International Nuclear Information System (INIS)

    Maddox, L.B.; McNeilly, G.S.

    1979-08-01

    GOC (General Orbit Code) is a versatile program which will perform a variety of calculations relevant to isochronous cyclotron design studies. In addition to the usual calculations of interest (e.g., equilibrium and accelerated orbits, focusing frequencies, field isochronization, etc.), GOC has a number of options to calculate injections with a charge change. GOC provides both printed and plotted output, and will follow groups of particles to allow determination of finite-beam properties. An interactive PDP-10 program called GIP, which prepares input data for GOC, is available. GIP is a very easy and convenient way to prepare complicated input data for GOC. Enclosed with this report are several microfiche containing source listings of GOC and other related routines and the printed output from a multiple-option GOC run

  4. Enhanced Input in LCTL Pedagogy

    Directory of Open Access Journals (Sweden)

    Marilyn S. Manley

    2009-08-01

    Full Text Available Language materials for the more-commonly-taught languages (MCTLs often include visual input enhancement (Sharwood Smith 1991, 1993 which makes use of typographical cues like bolding and underlining to enhance the saliency of targeted forms. For a variety of reasons, this paper argues that the use of enhanced input, both visual and oral, is especially important as a tool for the lesscommonly-taught languages (LCTLs. As there continues to be a scarcity of teaching resources for the LCTLs, individual teachers must take it upon themselves to incorporate enhanced input into their own self-made materials. Specific examples of how to incorporate both visual and oral enhanced input into language teaching are drawn from the author’s own experiences teaching Cuzco Quechua. Additionally, survey results are presented from the author’s Fall 2010 semester Cuzco Quechua language students, supporting the use of both visual and oral enhanced input.

  5. Enhanced Input in LCTL Pedagogy

    Directory of Open Access Journals (Sweden)

    Marilyn S. Manley

    2010-08-01

    Full Text Available Language materials for the more-commonly-taught languages (MCTLs often include visual input enhancement (Sharwood Smith 1991, 1993 which makes use of typographical cues like bolding and underlining to enhance the saliency of targeted forms. For a variety of reasons, this paper argues that the use of enhanced input, both visual and oral, is especially important as a tool for the lesscommonly-taught languages (LCTLs. As there continues to be a scarcity of teaching resources for the LCTLs, individual teachers must take it upon themselves to incorporate enhanced input into their own self-made materials. Specific examples of how to incorporate both visual and oral enhanced input into language teaching are drawn from the author’s own experiences teaching Cuzco Quechua. Additionally, survey results are presented from the author’s Fall 2010 semester Cuzco Quechua language students, supporting the use of both visual and oral enhanced input.

  6. PREP-45, Input Preparation for CITATION-2

    International Nuclear Information System (INIS)

    Ramalho Carlos, C.A.

    1995-01-01

    1 - Description of program or function: A Fortran program has been created, which saves much effort in preparing sections 004 (intervals in the coordinates) and 005 (zone numbers) of the input data file for the multigroup theory code CITATION (version CITATION-2, NESC0387/09), particularly when a thin complicated mesh is used. 2 - Method of solution: A domain is defined for CITATION calculations through specifying its sub-domains (e.g. graphite, lead, beryllium, water and fuel sub-domains) in a compact and simple way. An independent and previous geometrical specification is made of the various types of elements which are envisaged to constitute the contents of the reactor core grid positions. Then the load table for the configuration is input and scanned throughout, thus enabling the geometric mesh description to be produced (section 004). Also the zone placement (section 005) is achieved by means of element description subroutines for the different types of element (which may require appropriate but simple changes in the actual cases). The output of PREP45 is directly obtained in a format which is compatible with CITATION-2 input. 3 - Restrictions on the complexity of the problem: Only rectangular two-dimensional Cartesian coordinates are considered. A maximum of 12 sub-domains in the x direction (18 in the y direction) and up to 8 distinct element types are considered in this version. Other limitations exist which can nevertheless be overcome with simple changes in the source program

  7. 77 FR 34020 - National Fire Codes: Request for Public Input for Revision of Codes and Standards

    Science.gov (United States)

    2012-06-08

    .... (CO) Detection and Warning Equipment. NFPA 790--2012 Standard for Competency of Third-Party Field 6/22... Health-Related Fitness Programs for 1/4/2013. Fire Department Members. NFPA 1584--2008 Standard on the...

  8. 77 FR 67628 - National Fire Codes: Request for Public Input for Revision of Codes and Standards

    Science.gov (United States)

    2012-11-13

    ... Standard on Fire and 7/8/2013 Life Safety in Animal Housing Facilities. NFPA 160--2011 Standard for the Use... five years in Revision Cycles that begin twice each year and take approximately two years to complete. Each Revision Cycle proceeds according to a published schedule that includes final dates for all major...

  9. Effects of the input polarization on JET polarimeter horizontal channels

    International Nuclear Information System (INIS)

    Gaudio, P.; Gelfusa, M.; Murari, A.; Orsitto, F.; Boboc, A.

    2013-01-01

    In the past, the analysis of JET polarimetry measurements were carried out only for the vertical channels using a polarimetry propagation code based on the Stokes vector formalism [1,2]. A new propagation code has been developed therefore for the horizontal chords to simulate and interpret the measurements of the Faraday rotation and Cotton–Mouton phase shift in JET. The code has been used to develop a theoretical study to the effect of the input polarization on the eventual quality of the measurements. The results allow choosing the best polarization to optimize the polarimetric measurements for the various experiments

  10. PHYSICS

    CERN Multimedia

    J. D'Hondt

    The Electroweak and Top Quark Workshop (16-17th of July) A Workshop on Electroweak and Top Quark Physics, dedicated on early measurements, took place on 16th-17th July. We had more than 40 presentations at the Workshop, which was an important milestone for 2007 physics analyses in the EWK and TOP areas. The Standard Model has been tested empirically by many previous experiments. Observables which are nowadays known with high precision will play a major role for data-based CMS calibrations. A typical example is the use of the Z to monitor electron and muon reconstruction in di-lepton inclusive samples. Another example is the use of the W mass as a constraint for di-jets in the kinematic fitting of top-quark events, providing information on the jet energy scale. The predictions of the Standard Model, for what concerns proton collisions at the LHC, are accurate to a level that the production of W/Z and top-quark events can be used as a powerful tool to commission our experiment. On the other hand the measure...

  11. PHYSICS

    CERN Multimedia

    Christopher Hill

    2013-01-01

    Since the last CMS Bulletin, the CMS Physics Analysis Groups have completed more than 70 new analyses, many of which are based on the complete Run 1 dataset. In parallel the Snowmass whitepaper on projected discovery potential of CMS for HL-LHC has been completed, while the ECFA HL-LHC future physics studies has been summarised in a report and nine published benchmark analyses. Run 1 summary studies on b-tag and jet identification, quark-gluon discrimination and boosted topologies have been documented in BTV-13-001 and JME-13-002/005/006, respectively. The new tracking alignment and performance papers are being prepared for submission as well. The Higgs analysis group produced several new results including the search for ttH with H decaying to ZZ, WW, ττ+bb (HIG-13-019/020) where an excess of ~2.5σ is observed in the like-sign di-muon channel, and new searches for high-mass Higgs bosons (HIG-13-022). Search for invisible Higgs decays have also been performed both using the associ...

  12. PHYSICS

    CERN Multimedia

    C. Hill

    2013-01-01

    In the period since the last CMS Bulletin, the LHC – and CMS – have entered LS1. During this time, CMS Physics Analysis Groups have performed more than 40 new analyses, many of which are based on the complete 8 TeV dataset delivered by the LHC in 2012 (and in some cases on the full Run 1 dataset). These results were shown at, and well received by, several high-profile conferences in the spring of 2013, including the inaugural meeting of the Large Hadron Collider    Physics Conference (LHCP) in Barcelona, and the 26th International Symposium on Lepton Photon Interactions at High Energies (LP) in San Francisco. In parallel, there have been significant developments in preparations for Run 2 of the LHC and on “future physics” studies for both Phase 1 and Phase 2 upgrades of the CMS detector. The Higgs analysis group produced five new results for LHCP including a new H-to-bb search in VBF production (HIG-13-011), ttH with H to γ&ga...

  13. PHYSICS

    CERN Multimedia

    C. Hill

    2013-01-01

    The period since the last CMS bulletin has seen the end of proton collisions at a centre-of-mass energy 8 TeV, a successful proton-lead collision run at 5 TeV/nucleon, as well as a “reference” proton run at 2.76 TeV. With these final LHC Run 1 datasets in hand, CMS Physics Analysis Groups have been busy analysing these data in preparation for the winter conferences. Moreover, despite the fact that the pp run only concluded in mid-December (and there was consequently less time to complete data analyses), CMS again made a strong showing at the Rencontres de Moriond in La Thuile (EW and QCD) where nearly 40 new results were presented. The highlight of these preliminary results was the eagerly anticipated updated studies of the properties of the Higgs boson discovered in July of last year. Meanwhile, preparations for Run 2 and physics performance studies for Phase 1 and Phase 2 upgrade scenarios are ongoing. The Higgs analysis group produced updated analyses on the full Run 1 dataset (~25 f...

  14. Generomak: Fusion physics, engineering and costing model

    International Nuclear Information System (INIS)

    Delene, J.G.; Krakowski, R.A.; Sheffield, J.; Dory, R.A.

    1988-06-01

    A generic fusion physics, engineering and economics model (Generomak) was developed as a means of performing consistent analysis of the economic viability of alternative magnetic fusion reactors. The original Generomak model developed at Oak Ridge by Sheffield was expanded for the analyses of the Senior Committee on Environmental Safety and Economics of Magnetic Fusion Energy (ESECOM). This report describes the Generomak code as used by ESECOM. The input data used for each of the ten ESECOM fusion plants and the Generomak code output for each case is given. 14 refs., 3 figs., 17 tabs

  15. Verification of the thermal module in the ELESIM code and the associated uncertainty analysis

    International Nuclear Information System (INIS)

    Arimescu, V.I.; Williams, A.F.; Klein, M.E.; Richmond, W.R.; Couture, M.

    1997-09-01

    Temperature is a critical parameter in fuel modelling because most of the physical processes that occur in fuel elements during irradiation are thermally activated. The focus of this paper is the temperature distribution calculation used in the computer code ELESIM, developed at AECL to model the steady-state behaviour of CANDU fuel. A validation procedure for fuel codes is described and applied to ELESIM's thermal calculation.The effects of uncertainties in model parameters, like Uranium Dioxide thermal conductivity, and input variables, such as fuel element linear power, are accounted for through an uncertainty analysis using Response Surface and Monte Carlo techniques

  16. EGS code system: computer programs for the Monte Carlo simulation of electromagnetic cascade showers. Version 3

    International Nuclear Information System (INIS)

    Ford, R.L.; Nelson, W.R.

    1978-06-01

    A code to simulate almost any electron--photon transport problem conceivable is described. The report begins with a lengthy historical introduction and a description of the shower generation process. Then the detailed physics of the shower processes and the methods used to simulate them are presented. Ideas of sampling theory, transport techniques, particle interactions in general, and programing details are discussed. Next, EGS calculations and various experiments and other Monte Carlo results are compared. The remainder of the report consists of user manuals for EGS, PEGS, and TESTSR codes; options, input specifications, and typical output are included. 38 figures, 12 tables

  17. Speech coding

    Energy Technology Data Exchange (ETDEWEB)

    Ravishankar, C., Hughes Network Systems, Germantown, MD

    1998-05-08

    Speech is the predominant means of communication between human beings and since the invention of the telephone by Alexander Graham Bell in 1876, speech services have remained to be the core service in almost all telecommunication systems. Original analog methods of telephony had the disadvantage of speech signal getting corrupted by noise, cross-talk and distortion Long haul transmissions which use repeaters to compensate for the loss in signal strength on transmission links also increase the associated noise and distortion. On the other hand digital transmission is relatively immune to noise, cross-talk and distortion primarily because of the capability to faithfully regenerate digital signal at each repeater purely based on a binary decision. Hence end-to-end performance of the digital link essentially becomes independent of the length and operating frequency bands of the link Hence from a transmission point of view digital transmission has been the preferred approach due to its higher immunity to noise. The need to carry digital speech became extremely important from a service provision point of view as well. Modem requirements have introduced the need for robust, flexible and secure services that can carry a multitude of signal types (such as voice, data and video) without a fundamental change in infrastructure. Such a requirement could not have been easily met without the advent of digital transmission systems, thereby requiring speech to be coded digitally. The term Speech Coding is often referred to techniques that represent or code speech signals either directly as a waveform or as a set of parameters by analyzing the speech signal. In either case, the codes are transmitted to the distant end where speech is reconstructed or synthesized using the received set of codes. A more generic term that is applicable to these techniques that is often interchangeably used with speech coding is the term voice coding. This term is more generic in the sense that the

  18. Introduction of SCIENCE code package

    International Nuclear Information System (INIS)

    Lu Haoliang; Li Jinggang; Zhu Ya'nan; Bai Ning

    2012-01-01

    The SCIENCE code package is a set of neutronics tools based on 2D assembly calculations and 3D core calculations. It is made up of APOLLO2F, SMART and SQUALE and used to perform the nuclear design and loading pattern analysis for the reactors on operation or under construction of China Guangdong Nuclear Power Group. The purpose of paper is to briefly present the physical and numerical models used in each computation codes of the SCIENCE code pack age, including the description of the general structure of the code package, the coupling relationship of APOLLO2-F transport lattice code and SMART core nodal code, and the SQUALE code used for processing the core maps. (authors)

  19. Channel coding techniques for wireless communications

    CERN Document Server

    Deergha Rao, K

    2015-01-01

    The book discusses modern channel coding techniques for wireless communications such as turbo codes, low-density parity check (LDPC) codes, space–time (ST) coding, RS (or Reed–Solomon) codes and convolutional codes. Many illustrative examples are included in each chapter for easy understanding of the coding techniques. The text is integrated with MATLAB-based programs to enhance the understanding of the subject’s underlying theories. It includes current topics of increasing importance such as turbo codes, LDPC codes, Luby transform (LT) codes, Raptor codes, and ST coding in detail, in addition to the traditional codes such as cyclic codes, BCH (or Bose–Chaudhuri–Hocquenghem) and RS codes and convolutional codes. Multiple-input and multiple-output (MIMO) communications is a multiple antenna technology, which is an effective method for high-speed or high-reliability wireless communications. PC-based MATLAB m-files for the illustrative examples are provided on the book page on Springer.com for free dow...

  20. Optimal codes as Tanner codes with cyclic component codes

    DEFF Research Database (Denmark)

    Høholdt, Tom; Pinero, Fernando; Zeng, Peng

    2014-01-01

    In this article we study a class of graph codes with cyclic code component codes as affine variety codes. Within this class of Tanner codes we find some optimal binary codes. We use a particular subgraph of the point-line incidence plane of A(2,q) as the Tanner graph, and we are able to describe ...

  1. Aztheca Code

    International Nuclear Information System (INIS)

    Quezada G, S.; Espinosa P, G.; Centeno P, J.; Sanchez M, H.

    2017-09-01

    This paper presents the Aztheca code, which is formed by the mathematical models of neutron kinetics, power generation, heat transfer, core thermo-hydraulics, recirculation systems, dynamic pressure and level models and control system. The Aztheca code is validated with plant data, as well as with predictions from the manufacturer when the reactor operates in a stationary state. On the other hand, to demonstrate that the model is applicable during a transient, an event occurred in a nuclear power plant with a BWR reactor is selected. The plant data are compared with the results obtained with RELAP-5 and the Aztheca model. The results show that both RELAP-5 and the Aztheca code have the ability to adequately predict the behavior of the reactor. (Author)

  2. Data entry system for INIS input using a personal computer

    International Nuclear Information System (INIS)

    Ishikawa, Masashi

    1990-01-01

    Input preparation for the INIS (International Nuclear Information System) has been performed by Japan Atomic Energy Research Institute since 1970. Instead of the input data preparation done by worksheets make out with the typewriters, new method with which data can be directly inputted into a diskette using personal computers is introduced. According to the popularization of personal computers and word processors, this system is easily applied to other system, so the outline and the future development on it are described. A shortcoming of this system is that spell-checking and data entry using authority files are hardly performed because of the limitation of hardware resources, and that data code conversion is needed because applied code systems between personal computer and main frame computer are quite different from each other. On the other hand, improving the timelyness of data entry is expected without duplication of keying. (author)

  3. Vocable Code

    DEFF Research Database (Denmark)

    Soon, Winnie; Cox, Geoff

    2018-01-01

    a computational and poetic composition for two screens: on one of these, texts and voices are repeated and disrupted by mathematical chaos, together exploring the performativity of code and language; on the other, is a mix of a computer programming syntax and human language. In this sense queer code can...... be understood as both an object and subject of study that intervenes in the world’s ‘becoming' and how material bodies are produced via human and nonhuman practices. Through mixing the natural and computer language, this article presents a script in six parts from a performative lecture for two persons...

  4. GESDATA: A failure-data management code

    International Nuclear Information System (INIS)

    Garcia Gay, J.; Francia Gonzalez, L.; Ortega Prieto, P.; Mira McWilliams, J.; Aguinaga Zapata, M.

    1987-01-01

    GESDATA is a failure data management code for both qualitative and quantitative fault-tree evaluation. Data management using the code should provide the analyst, in the quickest and easiest way, with the reliability data which constitute the input values for fault-tree evaluation programs. (orig./HSCH)

  5. MCNP: a general Monte Carlo code for neutron and photon transport. Version 3A. Revision 2

    International Nuclear Information System (INIS)

    Briesmeister, J.F.

    1986-09-01

    This manual is a practical guide for the use of our general-purpose Monte Carlo code MCNP. The first chapter is a primer for the novice user. The second chapter describes the mathematics, data, physics, and Monte Carlo simulation found in MCNP. This discussion is not meant to be exhaustive - details of the particular techniques and of the Monte Carlo method itself will have to be found elsewhere. The third chapter shows the user how to prepare input for the code. The fourth chapter contains several examples, and the fifth chapter explains the output. The appendices show how to use MCNP on particular computer systems at the Los Alamos National Laboratory and also give details about some of the code internals that those who wish to modify the code may find useful. 57 refs

  6. Material input of nuclear fuel

    International Nuclear Information System (INIS)

    Rissanen, S.; Tarjanne, R.

    2001-01-01

    The Material Input (MI) of nuclear fuel, expressed in terms of the total amount of natural material needed for manufacturing a product, is examined. The suitability of the MI method for assessing the environmental impacts of fuels is also discussed. Material input is expressed as a Material Input Coefficient (MIC), equalling to the total mass of natural material divided by the mass of the completed product. The material input coefficient is, however, only an intermediate result, which should not be used as such for the comparison of different fuels, because the energy contents of nuclear fuel is about 100 000-fold compared to the energy contents of fossil fuels. As a final result, the material input is expressed in proportion to the amount of generated electricity, which is called MIPS (Material Input Per Service unit). Material input is a simplified and commensurable indicator for the use of natural material, but because it does not take into account the harmfulness of materials or the way how the residual material is processed, it does not alone express the amount of environmental impacts. The examination of the mere amount does not differentiate between for example coal, natural gas or waste rock containing usually just sand. Natural gas is, however, substantially more harmful for the ecosystem than sand. Therefore, other methods should also be used to consider the environmental load of a product. The material input coefficient of nuclear fuel is calculated using data from different types of mines. The calculations are made among other things by using the data of an open pit mine (Key Lake, Canada), an underground mine (McArthur River, Canada) and a by-product mine (Olympic Dam, Australia). Furthermore, the coefficient is calculated for nuclear fuel corresponding to the nuclear fuel supply of Teollisuuden Voima (TVO) company in 2001. Because there is some uncertainty in the initial data, the inaccuracy of the final results can be even 20-50 per cent. The value

  7. 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.

  8. PHYSICS

    CERN Multimedia

    V.Ciulli

    2011-01-01

    The main programme of the Physics Week held between 16th and 20th May was a series of topology-oriented workshops on di-leptons, di-photons, inclusive W, and all-hadronic final states. The goal of these workshops was to reach a common understanding for the set of objects (ID, cleaning...), the handling of pile-up, calibration, efficiency and purity determination, as well as to revisit critical common issues such as the trigger. Di-lepton workshop Most analysis groups use a di-lepton trigger or a combination of single and di-lepton triggers in 2011. Some groups need to collect leptons with as low PT as possible with strong isolation and identification requirements as for Higgs into WW at low mass, others with intermediate PT values as in Drell-Yan studies, or high PT as in the Exotica group. Electron and muon reconstruction, identification and isolation, was extensively described in the workshop. For electrons, VBTF selection cuts for low PT and HEEP cuts for high PT were discussed, as well as more complex d...

  9. The Aster code; Code Aster

    Energy Technology Data Exchange (ETDEWEB)

    Delbecq, J.M

    1999-07-01

    The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)

  10. Hypoxia in the St. Lawrence Estuary: How a Coding Error Led to the Belief that “Physics Controls Spatial Patterns”

    NARCIS (Netherlands)

    Bourgault, D.; Cyr, F.

    2015-01-01

    Two fundamental sign errors were found in a computer code used for studying the oxygenminimum zone (OMZ) and hypoxia in the Estuary and Gulf of St. Lawrence. These errorsinvalidate the conclusions drawn from the model, and call into question a proposed mechanismfor generating OMZ that challenges

  11. Phasing Out a Polluting Input

    OpenAIRE

    Eriksson, Clas

    2015-01-01

    This paper explores economic policies related to the potential conflict between economic growth and the environment. It applies a model with directed technological change and focuses on the case with low elasticity of substitution between clean and dirty inputs in production. New technology is substituted for the polluting input, which results in a gradual decline in pollution along the optimal long-run growth path. In contrast to some recent work, the era of pollution and environmental polic...

  12. Objective Oriented Design of Architecture for TH System Safety Analysis Code and Verification

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong

    2008-03-15

    In this work, objective oriented design of generic system analysis code has been tried based on the previous works in KAERI for two phase three field Pilot code. It has been performed to implement of input and output design, TH solver, component model, special TH models, heat structure solver, general table, trip and control, and on-line graphics. All essential features for system analysis has been designed and implemented in the final product SYSTF code. The computer language C was used for implementation in the Visual studio 2008 IDE (Integrated Development Environment) since it has easier and lighter than C++ feature. The code has simple and essential features of models and correlation, special component, special TH model and heat structure model. However the input features is able to simulate the various scenarios, such as steady state, non LOCA transient and LOCA accident. The structure validity has been tested through the various verification tests and it has been shown that the developed code can treat the non LOCA and LOCA simulation. However more detailed design and implementation of models are required to get the physical validity of SYSTF code simulation.

  13. Objective Oriented Design of Architecture for TH System Safety Analysis Code and Verification

    International Nuclear Information System (INIS)

    Chung, Bub Dong

    2008-03-01

    In this work, objective oriented design of generic system analysis code has been tried based on the previous works in KAERI for two phase three field Pilot code. It has been performed to implement of input and output design, TH solver, component model, special TH models, heat structure solver, general table, trip and control, and on-line graphics. All essential features for system analysis has been designed and implemented in the final product SYSTF code. The computer language C was used for implementation in the Visual studio 2008 IDE (Integrated Development Environment) since it has easier and lighter than C++ feature. The code has simple and essential features of models and correlation, special component, special TH model and heat structure model. However the input features is able to simulate the various scenarios, such as steady state, non LOCA transient and LOCA accident. The structure validity has been tested through the various verification tests and it has been shown that the developed code can treat the non LOCA and LOCA simulation. However more detailed design and implementation of models are required to get the physical validity of SYSTF code simulation

  14. Coding Class

    DEFF Research Database (Denmark)

    Ejsing-Duun, Stine; Hansbøl, Mikala

    Denne rapport rummer evaluering og dokumentation af Coding Class projektet1. Coding Class projektet blev igangsat i skoleåret 2016/2017 af IT-Branchen i samarbejde med en række medlemsvirksomheder, Københavns kommune, Vejle Kommune, Styrelsen for IT- og Læring (STIL) og den frivillige forening...... Coding Pirates2. Rapporten er forfattet af Docent i digitale læringsressourcer og forskningskoordinator for forsknings- og udviklingsmiljøet Digitalisering i Skolen (DiS), Mikala Hansbøl, fra Institut for Skole og Læring ved Professionshøjskolen Metropol; og Lektor i læringsteknologi, interaktionsdesign......, design tænkning og design-pædagogik, Stine Ejsing-Duun fra Forskningslab: It og Læringsdesign (ILD-LAB) ved Institut for kommunikation og psykologi, Aalborg Universitet i København. Vi har fulgt og gennemført evaluering og dokumentation af Coding Class projektet i perioden november 2016 til maj 2017...

  15. Uplink Coding

    Science.gov (United States)

    Andrews, Ken; Divsalar, Dariush; Dolinar, Sam; Moision, Bruce; Hamkins, Jon; Pollara, Fabrizio

    2007-01-01

    This slide presentation reviews the objectives, meeting goals and overall NASA goals for the NASA Data Standards Working Group. The presentation includes information on the technical progress surrounding the objective, short LDPC codes, and the general results on the Pu-Pw tradeoff.

  16. Network Coding

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 15; Issue 7. Network Coding. K V Rashmi Nihar B Shah P Vijay Kumar. General Article Volume 15 Issue 7 July 2010 pp 604-621. Fulltext. Click here to view fulltext PDF. Permanent link: https://www.ias.ac.in/article/fulltext/reso/015/07/0604-0621 ...

  17. MCNP code

    International Nuclear Information System (INIS)

    Cramer, S.N.

    1984-01-01

    The MCNP code is the major Monte Carlo coupled neutron-photon transport research tool at the Los Alamos National Laboratory, and it represents the most extensive Monte Carlo development program in the United States which is available in the public domain. The present code is the direct descendent of the original Monte Carlo work of Fermi, von Neumaum, and Ulam at Los Alamos in the 1940s. Development has continued uninterrupted since that time, and the current version of MCNP (or its predecessors) has always included state-of-the-art methods in the Monte Carlo simulation of radiation transport, basic cross section data, geometry capability, variance reduction, and estimation procedures. The authors of the present code have oriented its development toward general user application. The documentation, though extensive, is presented in a clear and simple manner with many examples, illustrations, and sample problems. In addition to providing the desired results, the output listings give a a wealth of detailed information (some optional) concerning each state of the calculation. The code system is continually updated to take advantage of advances in computer hardware and software, including interactive modes of operation, diagnostic interrupts and restarts, and a variety of graphical and video aids

  18. Expander Codes

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 10; Issue 1. Expander Codes - The Sipser–Spielman Construction. Priti Shankar. General Article Volume 10 ... Author Affiliations. Priti Shankar1. Department of Computer Science and Automation, Indian Institute of Science Bangalore 560 012, India.

  19. 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)

  20. Conceptual Design of GRIG (GUI Based RETRAN Input Generator)

    International Nuclear Information System (INIS)

    Lee, Gyung Jin; Hwang, Su Hyun; Hong, Soon Joon; Lee, Byung Chul; Jang, Chan Su; Um, Kil Sup

    2007-01-01

    For the development of high performance methodology using advanced transient analysis code, it is essential to generate the basic input of transient analysis code by rigorous QA procedures. There are various types of operating NPPs (Nuclear Power Plants) in Korea such as Westinghouse plants, KSNP(Korea Standard Nuclear Power Plant), APR1400 (Advance Power Reactor), etc. So there are some difficulties to generate and manage systematically the input of transient analysis code reflecting the inherent characteristics of various types of NPPs. To minimize the user faults and investment man power and to generate effectively and accurately the basic inputs of transient analysis code for all domestic NPPs, it is needed to develop the program that can automatically generate the basic input, which can be directly applied to the transient analysis, from the NPP design material. ViRRE (Visual RETRAN Running Environment) developed by KEPCO (Korea Electric Power Corporation) and KAERI (Korea Atomic Energy Research Institute) provides convenient working environment for Kori Unit 1/2. ViRRE shows the calculated results through on-line display but its capability is limited on the convenient execution of RETRAN. So it can not be used as input generator. ViSA (Visual System Analyzer) developed by KAERI is a NPA (Nuclear Plant Analyzer) using RETRAN and MARS code as thermal-hydraulic engine. ViSA contains both pre-processing and post-processing functions. In the pre-processing, only the trip data cards and boundary conditions can be changed through GUI mode based on pre-prepared text-input, so the capability of input generation is very limited. SNAP (Symbolic Nuclear Analysis Package) developed by Applied Programming Technology, Inc. and NRC (Nuclear Regulatory Commission) provides efficient working environment for the use of nuclear safety analysis codes such as RELAP5 and TRAC-M codes. SNAP covers wide aspects of thermal-hydraulic analysis from model creation through data analysis