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Sample records for mcnp user manual-version

  1. Sesame IO Library User Manual Version 8

    Energy Technology Data Exchange (ETDEWEB)

    Abhold, Hilary [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Young, Ginger Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-05-15

    This document is a user manual for SES_IO, a low-level library for reading and writing sesame files. The purpose of the SES_IO library is to provide a simple user interface for accessing and creating sesame files that does not change across sesame format type (such as binary, ascii, and xml).

  2. National Radiobiology Archives Distributed Access User`s Manual, Version 1.1. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Smith, S.K.; Prather, J.C.; Ligotke, E.K.; Watson, C.R.

    1992-06-01

    This supplement to the NRA Distributed Access User`s manual (PNL-7877), November 1991, describes installation and use of Version 1.1 of the software package; this is not a replacement of the previous manual. Version 1.1 of the NRA Distributed Access Package is a maintenance release. It eliminates several bugs, and includes a few new features which are described in this manual. Although the appearance of some menu screens has changed, we are confident that the Version 1.0 User`s Manual will provide an adequate introduction to the system. Users who are unfamiliar with Version 1.0 may wish to experiment with that version before moving on to Version 1.1.

  3. ANOPP2 User's Manual: Version 1.2

    Science.gov (United States)

    Lopes, L. V.; Burley, C. L.

    2016-01-01

    This manual documents the Aircraft NOise Prediction Program 2 (ANOPP2). ANOPP2 is a toolkit that includes a framework, noise prediction methods, and peripheral software to aid a user in predicting and understanding aircraft noise. This manual includes an explanation of the overall design and structure of ANOPP2, including a brief introduction to aircraft noise prediction and the ANOPP2 background, philosophy, and architecture. The concept of nested acoustic data surfaces and its application to a mixed-fidelity noise prediction are presented. The structure and usage of ANOPP2, which includes the communication between the user, the ANOPP2 framework, and noise prediction methods, are presented for two scenarios: wind-tunnel and flight. These scenarios serve to provide the user with guidance and documentation references for performing a noise prediction using ANOPP2.

  4. MULTIPLE PROJECTIONS SYSTEM (MPS) - USER'S MANUAL VERSION 1.0

    Science.gov (United States)

    The report is a user's manual for version 1.0 of the Multiple Projections Systems (MPS), a computer system that can perform "what if" scenario analysis and report the final results (i.e., Rate of Further Progress - ROP - inventories) to EPA (i.e., the Aerometric Information Retri...

  5. MULTIPLE PROJECTIONS SYSTEM (MPS): USER'S MANUAL VERSION 2.0

    Science.gov (United States)

    The document is a user's manual for Multiple Projections System (MPS) Version 2.0, based on the 3% reasonable further progress (RFP) tracking system that was developed in FY92/FY93. The 3% RFP tracking system is a Windows application, and enhancements to convert the 3% RFP track...

  6. Activity Catalog Tool (ACT) user manual, version 2.0

    Science.gov (United States)

    Segal, Leon D.; Andre, Anthony D.

    1994-01-01

    This report comprises the user manual for version 2.0 of the Activity Catalog Tool (ACT) software program, developed by Leon D. Segal and Anthony D. Andre in cooperation with NASA Ames Aerospace Human Factors Research Division, FLR branch. ACT is a software tool for recording and analyzing sequences of activity over time that runs on the Macintosh platform. It was designed as an aid for professionals who are interested in observing and understanding human behavior in field settings, or from video or audio recordings of the same. Specifically, the program is aimed at two primary areas of interest: human-machine interactions and interactions between humans. The program provides a means by which an observer can record an observed sequence of events, logging such parameters as frequency and duration of particular events. The program goes further by providing the user with a quantified description of the observed sequence, through application of a basic set of statistical routines, and enables merging and appending of several files and more extensive analysis of the resultant data.

  7. Manufactured Home Energy Audit (MHEA)Users Manual (Version 7)

    Energy Technology Data Exchange (ETDEWEB)

    Gettings, M.B.

    2003-01-27

    The Manufactured Home Energy Audit (MHEA) is a software tool that predicts manufactured home energy consumption and recommends weatherization retrofit measures. It was developed to assist local weatherization agencies working with the U.S. Department of Energy (DOE) Weatherization Assistance Program. Whether new or experienced, employed within or outside the Weatherization Assistance Program, all users can benefit from incorporating MHEA into their manufactured home weatherization programs. DOE anticipates that the state weatherization assistance programs that incorporate MHEA into their programs will find significant growth in the energy and cost savings achieved from manufactured home weatherization. The easy-to-use MHEA uses a relatively standard Windows graphical interface for entering simple inputs and provides understandable, usable results. The user enters information about the manufactured home construction, heating equipment, cooling equipment appliances, and weather site. MHEA then calculates annual energy consumption using a simplified building energy analysis technique. Weatherization retrofit measures are evaluated based on the predicted energy savings after installation of the measure, the measure cost, and the measure life. Finally, MHEA recommends retrofit measures that are energy and cost effective for the particular home being evaluated. MHEA evaluates each manufactured home individually and takes into account local weather conditions, retrofit measure costs, and fuel costs. The recommended package of weatherization retrofit measures is tailored to the home being evaluated. More traditional techniques apply the same package of retrofit measures to all manufactured homes, often the same set of measures that are installed into site-built homes. Effective manufactured home weatherization can be achieved only by installing measures developed specifically for manufactured homes. The unique manufactured home construction characteristics require that

  8. Water Security Toolkit User Manual Version 1.2.

    Energy Technology Data Exchange (ETDEWEB)

    Klise, Katherine A.; Siirola, John Daniel; Hart, David; Hart, William Eugene; Phillips, Cynthia Ann; Haxton, Terranna; Murray, Regan; Janke, Robert; Taxon, Thomas; Laird, Carl; Seth, Arpan; Hackebeil, Gabriel; McGee, Shawn; Mann, Angelica

    2014-08-01

    The Water Security Toolkit (WST) is a suite of open source software tools that can be used by water utilities to create response strategies to reduce the impact of contamination in a water distribution network . WST includes hydraulic and water quality modeling software , optimizati on methodologies , and visualization tools to identify: (1) sensor locations to detect contamination, (2) locations in the network in which the contamination was introduced, (3) hydrants to remove contaminated water from the distribution system, (4) locations in the network to inject decontamination agents to inactivate, remove, or destroy contaminants, (5) locations in the network to take grab sample s to help identify the source of contamination and (6) valves to close in order to isolate contaminate d areas of the network. This user manual describes the different components of WST , along w ith examples and case studies. License Notice The Water Security Toolkit (WST) v.1.2 Copyright c 2012 Sandia Corporation. Under the terms of Contract DE-AC04-94AL85000, there is a non-exclusive license for use of this work by or on behalf of the U.S. government. This software is distributed under the Revised BSD License (see below). In addition, WST leverages a variety of third-party software packages, which have separate licensing policies: Acro Revised BSD License argparse Python Software Foundation License Boost Boost Software License Coopr Revised BSD License Coverage BSD License Distribute Python Software Foundation License / Zope Public License EPANET Public Domain EPANET-ERD Revised BSD License EPANET-MSX GNU Lesser General Public License (LGPL) v.3 gcovr Revised BSD License GRASP AT&T Commercial License for noncommercial use; includes randomsample and sideconstraints executable files LZMA SDK Public Domain nose GNU Lesser General Public License (LGPL) v.2.1 ordereddict MIT License pip MIT License PLY BSD License PyEPANET Revised BSD License Pyro MIT License PyUtilib Revised BSD License Py

  9. Water Security Toolkit User Manual Version 1.2.

    Energy Technology Data Exchange (ETDEWEB)

    Klise, Katherine A.; Siirola, John Daniel; Hart, David; Hart, William Eugene; Phillips, Cynthia Ann; Haxton, Terranna; Murray, Regan; Janke, Robert; Taxon, Thomas; Laird, Carl; Seth, Arpan; Hackebeil, Gabriel; McGee, Shawn; Mann, Angelica

    2014-08-01

    The Water Security Toolkit (WST) is a suite of open source software tools that can be used by water utilities to create response strategies to reduce the impact of contamination in a water distribution network . WST includes hydraulic and water quality modeling software , optimizati on methodologies , and visualization tools to identify: (1) sensor locations to detect contamination, (2) locations in the network in which the contamination was introduced, (3) hydrants to remove contaminated water from the distribution system, (4) locations in the network to inject decontamination agents to inactivate, remove, or destroy contaminants, (5) locations in the network to take grab sample s to help identify the source of contamination and (6) valves to close in order to isolate contaminate d areas of the network. This user manual describes the different components of WST , along w ith examples and case studies. License Notice The Water Security Toolkit (WST) v.1.2 Copyright c 2012 Sandia Corporation. Under the terms of Contract DE-AC04-94AL85000, there is a non-exclusive license for use of this work by or on behalf of the U.S. government. This software is distributed under the Revised BSD License (see below). In addition, WST leverages a variety of third-party software packages, which have separate licensing policies: Acro Revised BSD License argparse Python Software Foundation License Boost Boost Software License Coopr Revised BSD License Coverage BSD License Distribute Python Software Foundation License / Zope Public License EPANET Public Domain EPANET-ERD Revised BSD License EPANET-MSX GNU Lesser General Public License (LGPL) v.3 gcovr Revised BSD License GRASP AT&T Commercial License for noncommercial use; includes randomsample and sideconstraints executable files LZMA SDK Public Domain nose GNU Lesser General Public License (LGPL) v.2.1 ordereddict MIT License pip MIT License PLY BSD License PyEPANET Revised BSD License Pyro MIT License PyUtilib Revised BSD License Py

  10. National Radiobiology Archives Distributed Access User's Manual, Version 1. 1

    Energy Technology Data Exchange (ETDEWEB)

    Smith, S.K.; Prather, J.C.; Ligotke, E.K.; Watson, C.R.

    1992-06-01

    This supplement to the NRA Distributed Access User's manual (PNL-7877), November 1991, describes installation and use of Version 1.1 of the software package; this is not a replacement of the previous manual. Version 1.1 of the NRA Distributed Access Package is a maintenance release. It eliminates several bugs, and includes a few new features which are described in this manual. Although the appearance of some menu screens has changed, we are confident that the Version 1.0 User's Manual will provide an adequate introduction to the system. Users who are unfamiliar with Version 1.0 may wish to experiment with that version before moving on to Version 1.1.

  11. MCNP-DSP users manual

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.

    1997-01-01

    The Monte Carlo code MCNP-DSP was developed from the Los Alamos MCNP4a code to calculate the time and frequency response statistics obtained from the {sup 252}Cf-source-driven frequency analysis measurements. This code can be used to validate calculational methods and cross section data sets from subcritical experiments. This code provides a more general model for interpretation and planning of experiments for nuclear criticality safety, nuclear safeguards, and nuclear weapons identification and replaces the use of point kinetics models for interpreting the measurements. The use of MCNP-DSP extends the usefulness of this measurement method to systems with much lower neutron multiplication factors.

  12. MCNP-DSP USERS MANUAL

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.

    2001-01-19

    The Monte Carlo code MCNP-DSP was developed from the Los Alamos MCNP4a code to calculate the time and frequency response statistics obtained from subcritical measurements. The code can be used to simulate a variety of subcritical measurements including source-driven noise analysis, Rossi-{alpha}, pulsed source, passive frequency analysis, multiplicity, and Feynman variance measurements. This code can be used to validate Monte Carlo methods and cross section data sets with subcritical measurements and replaces the use of point kinetics models for interpreting subcritical measurements.

  13. System cost model user`s manual, version 1.2

    Energy Technology Data Exchange (ETDEWEB)

    Shropshire, D.

    1995-06-01

    The System Cost Model (SCM) was developed by Lockheed Martin Idaho Technologies in Idaho Falls, Idaho and MK-Environmental Services in San Francisco, California to support the Baseline Environmental Management Report sensitivity analysis for the U.S. Department of Energy (DOE). The SCM serves the needs of the entire DOE complex for treatment, storage, and disposal (TSD) of mixed low-level, low-level, and transuranic waste. The model can be used to evaluate total complex costs based on various configuration options or to evaluate site-specific options. The site-specific cost estimates are based on generic assumptions such as waste loads and densities, treatment processing schemes, existing facilities capacities and functions, storage and disposal requirements, schedules, and cost factors. The SCM allows customization of the data for detailed site-specific estimates. There are approximately forty TSD module designs that have been further customized to account for design differences for nonalpha, alpha, remote-handled, and transuranic wastes. The SCM generates cost profiles based on the model default parameters or customized user-defined input and also generates costs for transporting waste from generators to TSD sites.

  14. The Weatherization Assistant User's Manual (Version 8.9)

    Energy Technology Data Exchange (ETDEWEB)

    Gettings, Michael B. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Malhotra, Mini [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ternes, Mark P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    The Weatherization Assistant is a Windows-based energy audit software tool that was developed by Oak Ridge National Laboratory (ORNL) to help states and their local weatherization agencies implement the U.S. Department of Energy (DOE) Weatherization Assistance Program. The Weatherization Assistant is an umbrella program for two individual energy audits or measure selection programs: the National Energy Audit Tool (NEAT) for site-built single-family homes and the Manufactured Home Energy Audit (MHEA) for mobile homes. The Weatherization Assistant User's Manual documents the operation of the user interface for Version 8.9 of the software. This includes how to install and setup the software, navigate through the program, and initiate an energy audit. All of the user interface forms associated with the software and the data fields on these forms are described in detail. The manual is intended to be a training manual for new users of the Weatherization Assistant and as a reference manual for experienced users.

  15. SERA: Simulation Environment for Radiotherapy Applications - Users Manual Version 1CO

    Energy Technology Data Exchange (ETDEWEB)

    Venhuizen, James Robert; Wessol, Daniel Edward; Wemple, Charles Alan; Wheeler, Floyd J; Harkin, G. J.; Frandsen, M. W.; Albright, C. L.; Cohen, M.T.; Rossmeier, M.; Cogliati, J.J.

    2002-06-01

    This document is the user manual for the Simulation Environment for Radiotherapy Applications (SERA) software program developed for boron-neutron capture therapy (BNCT) patient treatment planning by researchers at the Idaho National Engineering and Environmental Laboratory (INEEL) and students and faculty at Montana State University (MSU) Computer Science Department. This manual corresponds to the final release of the program, Version 1C0, developed to run under the RedHat Linux Operating System (version 7.2 or newer) or the Solaris™ Operating System (version 2.6 or newer). SERA is a suite of command line or interactively launched software modules, including graphical, geometric reconstruction, and execution interface modules for developing BNCT treatment plans. The program allows the user to develop geometric models of the patient as derived from Computed Tomography (CT) and Magnetic Resonance Imaging (MRI) images, perform dose computation for these geometric models, and display the computed doses on overlays of the original images as three dimensional representations. This manual provides a guide to the practical use of SERA, but is not an exhaustive treatment of each feature of the code.

  16. Thermal Insulation System Analysis Tool (TISTool) User's Manual. Version 1.0.0

    Science.gov (United States)

    Johnson, Wesley; Fesmire, James; Leucht, Kurt; Demko, Jonathan

    2010-01-01

    The Thermal Insulation System Analysis Tool (TISTool) was developed starting in 2004 by Jonathan Demko and James Fesmire. The first edition was written in Excel and Visual BasIc as macros. It included the basic shapes such as a flat plate, cylinder, dished head, and sphere. The data was from several KSC tests that were already in the public literature realm as well as data from NIST and other highly respectable sources. More recently, the tool has been updated with more test data from the Cryogenics Test Laboratory and the tank shape was added. Additionally, the tool was converted to FORTRAN 95 to allow for easier distribution of the material and tool. This document reviews the user instructions for the operation of this system.

  17. Federal Renewable Energy Screening Assistant (FRESA) User's Manual: Version 2.5

    Energy Technology Data Exchange (ETDEWEB)

    Brown, T.; Tapia, D.; Mas, C.

    2000-04-05

    The FRESA computer program, Version 2.5, provides an easy way to collect and process building and facility data to indicate opportunities for renewable energy applications in federal facilities and buildings. The purpose of this analytic tool is to focus feasibility study efforts on those applications most likely to prove cost-effective. The program is a supplement to energy and water conservation audits, which must be completed for all federal buildings and will flag renewable energy opportunities by facilitating the evaluation and ranking process. FRESA results alone are generally not sufficient to establish project feasibility. The FRESA User's Manual provides instruction on getting started; an overview of the FRESA program structure; an explanation of the screening process; detailed information on using the functions of Facility/Building Info, Building/Facility Analysis, Input/Output, and Weather Data or Adding a Zip Code; troubleshooting; and archiving data. Appendices include Algorithms Used in FRESA Prescreening, Excel Spreadsheets for FRESA Inputs, Other Useful Information, and Acronyms and Abbreviations.

  18. Social values for ecosystem services (SolVES): Documentation and user manual, version 2.0

    Science.gov (United States)

    Sherrouse, Benson C.; Semmens, Darius J.

    2012-01-01

    In response to the need for incorporating quantified and spatially explicit measures of social values into ecosystem services assessments, the Rocky Mountain Geographic Science Center (RMGSC), in collaboration with Colorado State University, developed a geographic information system (GIS) application, Social Values for Ecosystem Services (SolVES). With version 2.0 (SolVES 2.0), RMGSC has improved and extended the functionality of SolVES, which was designed to assess, map, and quantify the perceived social values of ecosystem services. Social values such as aesthetics, biodiversity, and recreation can be evaluated for various stakeholder groups as distinguished by their attitudes and preferences regarding public uses, such as motorized recreation and logging. As with the previous version, SolVES 2.0 derives a quantitative, 10-point, social-values metric, the Value Index, from a combination of spatial and nonspatial responses to public attitude and preference surveys and calculates metrics characterizing the underlying environment, such as average distance to water and dominant landcover. Additionally, SolVES 2.0 integrates Maxent maximum entropy modeling software to generate more complete social value maps and to produce robust statistical models describing the relationship between the social values maps and explanatory environmental variables. The performance of these models can be evaluated for a primary study area, as well as for similar areas where primary survey data are not available but where social value mapping could potentially be completed using value-transfer methodology. SolVES 2.0 also introduces the flexibility for users to define their own social values and public uses, model any number and type of environmental variable, and modify the spatial resolution of analysis. With these enhancements, SolVES 2.0 provides an improved public domain tool for decisionmakers and researchers to evaluate the social values of ecosystem services and to facilitate

  19. M3 User's Manual. Version 3.0

    Energy Technology Data Exchange (ETDEWEB)

    Laaksoharju, Marcus (Geopoint AB, Sollentuna (Sweden)); Skaarman, Erik (Abscondo Utveckling, Bromma (Sweden)); Gomez, Javier B. (Univ. of Zaragoza (Spain). Geochemical modelling Group); Gurban, Ioana (3D Terra (Canada))

    2006-07-15

    This report describes the Multivariate Mixing and Mass balance calculations (M3). This new method and computer code is developed to trace the mixing and reaction processes in the groundwater. The aim of the M3 concept is to decode the often hidden and complex information gathered in the groundwater analytical data. The manual presents shortly the theory and practice behind the M3 method. The M3 computer code is also presented and emphasis is put on the reference manual. This includes detailed reference to the M3 program's abilities and limitations, installation procedures and all functions and operations that the program can perform. It also describes sample cases of how the program is used to analyse a test data set. This guide is part of the Help Files distributed together with M3. Two accompanying reports cover other aspects: - Concepts, Methods, and Mathematical Formulation, gives a complete description of the mathematical framework of M3 and introduces concepts and methods useful for the end user. - M3 version 3.0: Verification and Validation, gathers a collection of validation and verification exercises, designed to test each part of M3 code and to build confidence in its methodology. The M3 method has been tested and modified over several years. The development work has been supported by the Swedish Nuclear Fuel and Waste Management Company (SKB). The main test site for the model was the underground Aespoe Hard Rock Laboratory (HRL). The examples used in this manual are from a Aespoe international groundwater modelling co-operation project where one of the tools used was M3. The M3 concept has been applied on the data from SKB's site investigation programme and in data from Canada, Japan, Jordan, Gabon and Finland. The groundwater composition is a result of mixing processes and water-rock interaction. Standard groundwater models based on thermodynamic laws may not be applicable in a normal temperature groundwater system where equilibrium with many

  20. Hydra-TH User's Manual, Version: LA-CC-11120, Dated: December 1, 2011

    Energy Technology Data Exchange (ETDEWEB)

    Christon, Mark A. [Los Alamos National Laboratory; Bakosi, Jozsef [Los Alamos National Laboratory; Lowrie, Robert B. [Los Alamos National Laboratory

    2012-07-19

    Hydra-TH is a hybrid finite-element/finite-volume code built using the Hydra toolkit specifically to attack a broad class of incompressible, viscous fluid dynamics problems prevalent in the thermalhydraulics community. The purpose for this manual is provide sufficient information for an experience analyst to use Hydra-TH in an effective way. The Hydra-TH User's Manual present a brief overview of capabilities and visualization interfaces. The execution and restart models are described before turning to the detailed description of keyword input. Finally, a series of example problems are presented with sufficient data to permit the user to verify the local installation of Hydra-TH, and to permit a convenient starting point for more detailed and complex analyses.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  2. Rn3D: A finite element code for simulating gas flow and radon transport in variably saturated, nonisothermal porous media. User`s manual, Version 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Holford, D.J.

    1994-01-01

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

  3. CREST Cost of Renewable Energy Spreadsheet Tool: A Model for Developing Cost-based Incentives in the United States. User Manual Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Gifford, Jason S. [Sustainable Energy Advantage, LLC, Framingham, MA (United States); Grace, Robert C. [Sustainable Energy Advantage, LLC, Framingham, MA (United States)

    2011-03-01

    This user manual helps model users understands how to use the CREST model to support renewable energy incentives, FITs, and other renewable energy rate-setting processes. It reviews the spreadsheet tool, including its layout and conventions, offering context on how and why it was created. It also provides instructions on how to populate the model with inputs that are appropriate for a specific jurisdiction’s policymaking objectives and context. And, it describes the results and outlines how these results may inform decisions about long-term renewable energy support programs.

  4. Minerva User Manual Version 1.0

    Energy Technology Data Exchange (ETDEWEB)

    J.J. Cogliati; M. L. Milvich; D. E. Wessol; C. A. Wemple

    2007-03-01

    MINERVA (Modality-Inclusive Environment for Radiotherapeutic Variable Analysis) is a Java-based patient-centric radiation treatment planning system (RTPS) for computational dosimetry and treatment planning in emerging areas of radiotherapy for cancer and other diseases. MINERVA was primarily developed at the Idaho National Laboratory (INL) and Montana State University (MSU). MINERVA allows the radiotherapist to make side-by-side comparison of plans for multiple treatment modalities with a common anatomical basis for the computational geometry, calculate doses for combinations of different radiotherapy modalities, and perform dose analysis and reporting functions. This provides the therapist with a consistent basis for selecting the modality or combination of modalities to use for treatment of the patient. MINERVA employs an integrated, lightweight plug-in architecture to accommodate multi-modal treatment planning using standard interface components. The MINERVA design facilitates integration of improved or emerging treatment planning technologies. MINERVA consists of the basic radiation treatment planning software modules managed by a consistent patient interface for developing multi-modal radiotherapy patient treatment plans. One of MINERVA's main functions is to provide a graphical environment for constructing and displaying uniform volume-element-based solid models derived from medical images. These solid models form the geometric basis of the target areas for the radiation transport model.

  5. A Patch to MCNP5 for Multiplication Inference: Description and User Guide

    Energy Technology Data Exchange (ETDEWEB)

    Solomon, Jr., Clell J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-05-05

    A patch to MCNP5 has been written to allow generation of multiple neutrons from a spontaneous-fission event and generate list-mode output. This report documents the implementation and usage of this patch.

  6. MCNP Progress & Performance Improvements

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bull, Jeffrey S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-04-14

    Twenty-eight slides give information about the work of the US DOE/NNSA Nuclear Criticality Safety Program on MCNP6 under the following headings: MCNP6.1.1 Release, with ENDF/B-VII.1; Verification/Validation; User Support & Training; Performance Improvements; and Work in Progress. Whisper methodology will be incorporated into the code, and run speed should be increased.

  7. CREST Cost of Renewable Energy Spreadsheet Tool: A Model for Developing Cost-Based Incentives in the United States; User Manual Version 4, August 2009 - March 2011 (Updated July 2013)

    Energy Technology Data Exchange (ETDEWEB)

    Gifford, J. S.; Grace, R. C.

    2013-07-01

    The objective of this document is to help model users understand how to use the CREST model to support renewable energy incentives, FITs, and other renewable energy rate-setting processes. This user manual will walk the reader through the spreadsheet tool, including its layout and conventions, offering context on how and why it was created. This user manual will also provide instructions on how to populate the model with inputs that are appropriate for a specific jurisdiction's policymaking objectives and context. Finally, the user manual will describe the results and outline how these results may inform decisions about long-term renewable energy support programs.

  8. Salinas. Theory Manual Version 2.8

    Energy Technology Data Exchange (ETDEWEB)

    Reese, Garth M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Walsh, Timothy [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bhardwaj, Manoj K. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2009-02-01

    Salinas provides a massively parallel implementation of structural dynamics finite element analysis, required for high fidelity, validated models used in modal, vibration, static and shock analysis of structural systems. This manual describes the theory behind many of the constructs in Salinas. For a more detailed description of how to use Salinas , we refer the reader to Salinas, Users Notes. Many of the constructs in Salinas are pulled directly from published material. Where possible, these materials are referenced herein. However, certain functions in Salinas are specific to our implementation. We try to be far more complete in those areas. The theory manual was developed from several sources including general notes, a programmer notes manual, the user's notes and of course the material in the open literature.

  9. EOSPAC user's manual: version 6.2

    Energy Technology Data Exchange (ETDEWEB)

    Pimentel, David A [Los Alamos National Laboratory

    2011-01-05

    The EOSPAC utility package is a collection of interface routines, which can be used to access the SESAME data library and perform various data adjustments and interpolations on the SESAME data. The SESAME data library contains both thermodynamic (e.g., equation of state) and transport coefficients (e.g., opacity and conductivity), and it is described in reference 1. Note, for simplicity, the term EOS (equation of state) used herein includes both thermodynamic variables and transport coefficients. The EOSPAC utility package is designed to be used by physics codes (henceforth 'host codes') written in multiple languages and on multiple platforms. The remainder of this manual is organized into several sections. Section 2 provides a general overview of basic theory and models implemented within EOSPAC. Section 3 provides a general overview of how to use the EOSPAC interface library. Section 4 discusses conventions such as data organization and routine names. Sections 5 through 7 describe the public interfaces of EOSPAC in detail. Section 8 provides details related to some selected numerical features of EOSPAC. Section 9 gives examples for using the interface routines described in sections 5 through 7. Section 10 provides technical support contact information. Section 11 contains a brief set of acknowledgments. Finally, section 12 contains a list of referenced documents. Appendices list the Table Type Definitions, the Option Flag Definitions, the Information Flag Definitions, and the Error Code Definitions.

  10. MODGRO User’s Manual, Version 1.2

    Science.gov (United States)

    1988-02-01

    geometries. MODGRO is a tool that can analyze many coiznon, as well as some uncommon, crack geometries. MODGRO inputs are normalized to allow the use of any...state of stress in the crack growth direction. YLD is simply the yieid strE.gth of the given material. tress intensity Factor Solucions MODGRO has stress...modify the appropriate, non-holed, standard solution. A standard hole solution may be used only if the stress distribution is normalized to that of an

  11. SAVEWS Jr. User’s Manual, Version 1.0

    Science.gov (United States)

    2014-04-01

    is more of a cosmetic fix than a true signal processing fix, but it can be handy in some situations. The reprocessing results (Figure 24) exhibit...Figure A1) consists of a Lowrance LSS-2 structure scan module and a 12-V DC power source housed in a waterproof case (Figure A2), a quick disconnect...swivel mount for the HDS unit, and a waterproof through- deck fitting for transducer and power connections (Figure A3). The system includes

  12. SIERRA Multimechanics Module: Aria User Manual Version 4.44

    Energy Technology Data Exchange (ETDEWEB)

    Sierra Thermal/Fluid Team

    2017-04-01

    Aria is a Galerkin fnite element based program for solving coupled-physics problems described by systems of PDEs and is capable of solving nonlinear, implicit, transient and direct-to-steady state problems in two and three dimensions on parallel architectures. The suite of physics currently supported by Aria includes thermal energy transport, species transport, and electrostatics as well as generalized scalar, vector and tensor transport equations. Additionally, Aria includes support for manufacturing process fows via the incompressible Navier-Stokes equations specialized to a low Reynolds number ( %3C 1 ) regime. Enhanced modeling support of manufacturing processing is made possible through use of either arbitrary Lagrangian- Eulerian (ALE) and level set based free and moving boundary tracking in conjunction with quasi-static nonlinear elastic solid mechanics for mesh control. Coupled physics problems are solved in several ways including fully-coupled Newton's method with analytic or numerical sensitivities, fully-coupled Newton- Krylov methods and a loosely-coupled nonlinear iteration about subsets of the system that are solved using combinations of the aforementioned methods. Error estimation, uniform and dynamic h -adaptivity and dynamic load balancing are some of Aria's more advanced capabilities. Aria is based upon the Sierra Framework.

  13. National Energy Audit (NEAT) Users Manual Version 7

    Energy Technology Data Exchange (ETDEWEB)

    Gettings, M.

    2001-05-10

    Welcome to the U.S. Department of Energy's (DOE's) energy auditing tool, called ''NEAT.'' NEAT, an acronym for National Energy Audit Tool, a program for personal computers that was designed for use by local agencies in the Weatherization Assistance Program. It is an approved alternative audit that meets all auditing requirements set forth by the Program. NEAT is easy to use. It applies engineering and economic calculations to evaluate energy conservation measures for single-family, detached houses or small multifamily buildings. You can use it to rank measures for each individual house, or to establish a priority list of conservation measures for nearly identical housing types. NEAT was written for the Weatherization Assistance Program by Oak Ridge National Laboratory. Many building energy consumption algorithms are taken from Lawrence Berkeley Laboratory's Computerized Instrumented Residential Audit (CIRA), published in 1982 for the Department of Energy. Equipment retrofit conservation measures are based on published reports on various heating retrofits. Heating and cooling system replacement conservation measures are based on the energy ratings of new heating and cooling equipment. The Weatherization Program anticipates that this computer-based energy audit will offer substantial performance improvements to many states who choose to incorporate it into their programs. When conservation measures are evaluated locally according to climate, fuel cost, measure cost, and existing house conditions, the Program will be closer to its goal of assuring the maximum return for every federal dollar spent.

  14. Generic Optimization Program User Manual Version 3.0.0

    Energy Technology Data Exchange (ETDEWEB)

    Wetter, Michael

    2009-05-11

    GenOpt is an optimization program for the minimization of a cost function that is evaluated by an external simulation program. It has been developed for optimization problems where the cost function is computationally expensive and its derivatives are not available or may not even exist. GenOpt can be coupled to any simulation program that reads its input from text files and writes its output to text files. The independent variables can be continuous variables (possibly with lower and upper bounds), discrete variables, or both, continuous and discrete variables. Constraints on dependent variables can be implemented using penalty or barrier functions. GenOpt uses parallel computing to evaluate the simulations. GenOpt has a library with local and global multi-dimensional and one-dimensional optimization algorithms, and algorithms for doing parametric runs. An algorithm interface allows adding new minimization algorithms without knowing the details of the program structure. GenOpt is written in Java so that it is platform independent. The platform independence and the general interface make GenOpt applicable to a wide range of optimization problems. GenOpt has not been designed for linear programming problems, quadratic programming problems, and problems where the gradient of the cost function is available. For such problems, as well as for other problems, special tailored software exists that is more efficient.

  15. SIERRA Multimechanics Module: Aria User Manual Version 4.46.

    Energy Technology Data Exchange (ETDEWEB)

    Sierra Thermal/Fluid Team

    2017-09-01

    Aria is a Galerkin fnite element based program for solving coupled-physics problems described by systems of PDEs and is capable of solving nonlinear, implicit, transient and direct-to-steady state problems in two and three dimensions on parallel architectures. The suite of physics currently supported by Aria includes thermal energy transport, species transport, and electrostatics as well as generalized scalar, vector and tensor transport equations. Additionally, Aria includes support for manufacturing process fows via the incompressible Navier-Stokes equations specialized to a low Reynolds number ( %3C 1 ) regime. Enhanced modeling support of manufacturing processing is made possible through use of either arbitrary Lagrangian- Eulerian (ALE) and level set based free and moving boundary tracking in conjunction with quasi-static nonlinear elastic solid mechanics for mesh control. Coupled physics problems are solved in several ways including fully-coupled Newton's method with analytic or numerical sensitivities, fully-coupled Newton- Krylov methods and a loosely-coupled nonlinear iteration about subsets of the system that are solved using combinations of the aforementioned methods. Error estimation, uniform and dynamic h -adaptivity and dynamic load balancing are some of Aria's more advanced capabilities. Aria is based upon the Sierra Framework.

  16. SIERRA/Aero User Manual Version 4.44

    Energy Technology Data Exchange (ETDEWEB)

    Sierra Thermal/Fluid Team

    2017-04-01

    SIERRA/Aero is a compressible fluid dynamics program intended to solve a wide variety compressible fluid flows including transonic and hypersonic problems. This document describes the commands for assembling a fluid model for analysis with this module, henceforth referred to simply as Aero for brevity. Aero is an application developed using the SIERRA Toolkit (STK). The intent of STK is to provide a set of tools for handling common tasks that programmers encounter when developing a code for numerical simulation. For example, components of STK provide field allocation and management, and parallel input/output of field and mesh data. These services also allow the development of coupled mechanics analysis software for a massively parallel computing environment. In the definitions of the commands that follow, the term Real_Max denotes the largest floating point value that can be represented on a given computer. Int_Max is the largest such integer value.

  17. Fast Scattering Code (FSC) User's Manual: Version 2

    Science.gov (United States)

    Tinetti, Ana F.; Dun, M. H.; Pope, D. Stuart

    2006-01-01

    The Fast Scattering Code (version 2.0) is a computer program for predicting the three-dimensional scattered acoustic field produced by the interaction of known, time-harmonic, incident sound with aerostructures in the presence of potential background flow. The FSC has been developed for use as an aeroacoustic analysis tool for assessing global effects on noise radiation and scattering caused by changes in configuration (geometry, component placement) and operating conditions (background flow, excitation frequency).

  18. MCNP: Multigroup/adjoint capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, J.C.; Redmond, E.L. II; Palmtag, S.P.; Hendricks, J.S.

    1994-04-01

    This report discusses various aspects related to the use and validity of the general purpose Monte Carlo code MCNP for multigroup/adjoint calculations. The increased desire to perform comparisons between Monte Carlo and deterministic codes, along with the ever-present desire to increase the efficiency of large MCNP calculations has produced a greater user demand for the multigroup/adjoint capabilities. To more fully utilize these capabilities, we review the applications of the Monte Carlo multigroup/adjoint method, describe how to generate multigroup cross sections for MCNP with the auxiliary CRSRD code, describe how to use the multigroup/adjoint capability in MCNP, and provide examples and results indicating the effectiveness and validity of the MCNP multigroup/adjoint treatment. This information should assist users in taking advantage of the MCNP multigroup/adjoint capabilities.

  19. The REBUS-MCNP linkage.

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, J. G.; Nuclear Engineering Division

    2009-04-24

    The Reduced Enrichment Research and Test Reactor (RERTR) Program uses the REBUS-PC computer code to provide reactor physics and core design information such as neutron flux distributions in space, energy, and time, and to track isotopic changes in fuel and neutron absorbers with burnup. REBUS-PC models the complete fuel cycle including shuffling capability. REBUS-PC evolved using the neutronic capabilities of multi-group diffusion theory code DIF3D 9.0, but was extended to apply the continuous energy Monte Carlo code MCNP for one-group fluxes and cross-sections. The linkage between REBUS-PC and MCNP has recently been modernized and extended, as described in this manual. REBUS-PC now calls MCNP via a system call so that the user can apply any valid MCNP executable. The interface between REBUS-PC and MCNP requires minimal changes to an existing MCNP model, and little additional input. The REBUS-MCNP interface can also be used in conjunction with DIF3D neutronics to update an MCNP model with fuel compositions predicted using a DIF3D based depletion.

  20. MCNP{trademark} Monte Carlo: A precis of MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Adams, K.J.

    1996-06-01

    MCNP{trademark} is a general purpose three-dimensional time-dependent neutron, photon, and electron transport code. It is highly portable and user-oriented, and backed by stringent software quality assurance practices and extensive experimental benchmarks. The cross section database is based upon the best evaluations available. MCNP incorporates state-of-the-art analog and adaptive Monte Carlo techniques. The code is documented in a 600 page manual which is augmented by numerous Los Alamos technical reports which detail various aspects of the code. MCNP represents over a megahour of development and refinement over the past 50 years and an ongoing commitment to excellence.

  1. MCNP6 Fission Multiplicity with FMULT Card

    Energy Technology Data Exchange (ETDEWEB)

    Wilcox, Trevor [Los Alamos National Laboratory; Fensin, Michael Lorne [Los Alamos National Laboratory; Hendricks, John S. [Los Alamos National Laboratory; James, Michael R. [Los Alamos National Laboratory; McKinney, Gregg W. [Los Alamos National Laboratory

    2012-06-18

    With the merger of MCNPX and MCNP5 into MCNP6, MCNP6 now provides all the capabilities of both codes allowing the user to access all the fission multiplicity data sets. Detailed in this paper is: (1) the new FMULT card capabilities for accessing these different data sets; (2) benchmark calculations, as compared to experiment, detailing the results of selecting these separate data sets for thermal neutron induced fission on U-235.

  2. MCNP(TM) Version 5.

    Energy Technology Data Exchange (ETDEWEB)

    Cox, L. J. (Lawrence J.); Barrett, R. F. (Richard F.); Booth, Thomas Edward; Briesmeister, Judith F.; Brown, F. B. (Forrest B.); Bull, J. S. (Jeffrey S.); Giesler, G. C. (Gregg Carl); Goorley, J. T. (John T.); Mosteller, R. D. (Russell D.); Forster, R. A. (R. Arthur); Post, S. E. (Susan E.); Prael, R. E. (Richard E.); Selcow, Elizabeth Carol,; Sood, A. (Avneet)

    2002-01-01

    The Monte Carlo transport workhorse, MCNP, is undergoing a massive renovation at Los Alamos National Laboratory (LANL) in support of the Eolus Project of the Advanced Simulation and Computing (ASCI) Program. MCNP Version 5 (V5) (expected to be released to RSICC in Spring, 2002) will consist of a major restructuring from FORTRAN-77 (with extensions) to ANSI-standard FORTRAN-90 with support for all of the features available in the present release (MCNP-4C2/4C3). To most users, the look-and-feel of MCNP will not change much except for the improvements (improved graphics, easier installation, better online documentation). For example, even with the major format change, full support for incremental patching will still be provided. In addition to the language and style updates, MCNP V5 will have various new user features. These include improved photon physics, neutral particle radiography, enhancements and additions to variance reduction methods, new source options, and improved parallelism support (PVM, MPI, OpenMP).

  3. User`s manual for elegant Program Version 12.4, Manual Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Borland, M.

    1993-05-06

    Elegant stands for ``Electron Generation and Tracking,`` a somewhat out-of-date description of a fully 6D accelerator program that now does much more than generate particle distributions and track them elegant, written entirely in the C programming language, uses a variant of the MAD input format to describe accelerators, which may be either transport lines, circular machines, or a combination thereof. Program execution is driven by commands in a namelist format. This document describes the features available in elegant, listing the commands and their arguments. The differences between elegant and MAD formats for describing accelerators are listed. A series of examples of elegant input and output are given. Finally, appendices are included describing the post-processing programs.

  4. The National Energy Audit (NEAT) Engineering Manual (Version 6)

    Energy Technology Data Exchange (ETDEWEB)

    Gettings, M.B.

    2001-04-20

    Government-funded weatherization assistance programs resulted from increased oil prices caused by the 1973 oil embargo. These programs were instituted to reduce US consumption of oil and help low-income families afford the increasing cost of heating their homes. In the summer of 1988, Oak Ridge National Laboratory (ORNL) began providing technical support to the Department of Energy (DOE) Weatherization Assistance Program (WAP). A preliminary study found no suitable means of cost-effectively selecting energy efficiency improvements (measures) for single-family homes that incorporated all the factors seen as beneficial in improving cost-effectiveness and usability. In mid-1989, ORNL was authorized to begin development of a computer-based measure selection technique. In November of 1992 a draft version of the program was made available to all WAP state directors for testing. The first production release, Version 4.3, was made available in october of 1993. The Department of Energy's Weatherization Assistance Program has continued funding improvements to the program increasing its user-friendliness and applicability. initial publication of this engineering manual coincides with availability of Version 6.1, November 1997, though algorithms described generally apply to all prior versions. Periodic updates of specific sections in the manual will permit maintaining a relevant document. This Engineering Manual delineates the assumptions used by NEAT in arriving at the measure recommendations based on the user's input of the building characteristics. Details of the actual data entry are available in the NEAT User's Manual (ORNL/Sub/91-SK078/1) and will not be discussed in this manual.

  5. MCNP4B{sup {trademark}} verification and validation

    Energy Technology Data Exchange (ETDEWEB)

    Hendricks, J.S.; Court, J.D.

    1996-08-01

    Several new features and bug fixes have been incorporated into the new release of MCNP. As required by the MCNP Software Quality Assurance Plan, these changes to the code and the test set are documented here for user reference. This document summarizes the new MCNP4B features and corrections, separated into major and minor groupings. Also included are a code cleanup section and a section delineating problems identified in LA-12839 which have not been corrected. Finally, we document the MCNP4B test set modifications and explain how test set coverage has been improved.

  6. Verification of MCNP6.2 for Nuclear Criticality Safety Applications

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-05-10

    Several suites of verification/validation benchmark problems were run in early 2017 to verify that the new production release of MCNP6.2 performs correctly for nuclear criticality safety applications (NCS). MCNP6.2 results for several NCS validation suites were compared to the results from MCNP6.1 [1] and MCNP6.1.1 [2]. MCNP6.1 is the production version of MCNP® released in 2013, and MCNP6.1.1 is the update released in 2014. MCNP6.2 includes all of the standard features for NCS calculations that have been available for the past 15 years, along with new features for sensitivity-uncertainty based methods for NCS validation [3]. Results from the benchmark suites were compared with results from previous verification testing [4-8]. Criticality safety analysts should consider testing MCNP6.2 on their particular problems and validation suites. No further development of MCNP5 is planned. MCNP6.1 is now 4 years old, and MCNP6.1.1 is now 3 years old. In general, released versions of MCNP are supported only for about 5 years, due to resource limitations. All future MCNP improvements, bug fixes, user support, and new capabilities are targeted only to MCNP6.2 and beyond.

  7. SIERRA Low Mach Module: Fuego User Manual Version 4.44

    Energy Technology Data Exchange (ETDEWEB)

    Sierra Thermal/Fluid Team

    2017-04-01

    The SIERRA Low Mach Module: Fuego along with the SIERRA Participating Media Radiation Module: Syrinx, henceforth referred to as Fuego and Syrinx, respectively, are the key elements of the ASCI fire environment simulation project. The fire environment simulation project is directed at characterizing both open large-scale pool fires and building enclosure fires. Fuego represents the turbulent, buoyantly-driven incompressible flow, heat transfer, mass transfer, combustion, soot, and absorption coefficient model portion of the simulation software. Syrinx represents the participating-media thermal radiation mechanics. This project is an integral part of the SIERRA multi-mechanics software development project. Fuego depends heavily upon the core architecture developments provided by SIERRA for massively parallel computing, solution adaptivity, and mechanics coupling on unstructured grids.

  8. Apparel Research Network (ARN); Apparel Order Processing Module (AOPM): Field User Manual, Version 1

    Science.gov (United States)

    2013-01-31

    Max=51 in). Around neck at center (Min=12 in; Max=19 in). Base of neck to top of hand, arm straight down (Min=30 in; Max=39 in). Shoulder... circumference (Min=38 in; Max=56 in). Height, barefoot (Min=60 in; Max=81 in). Weight, no shoes (Min=90 lb; Max=310 lb). Around head at the center of the...center of seat (Min=35 in; Max=60 in). Around knee at knee cap (Min=10 in; Max=18 in). Desired circumference at bottom or hem of trousers (Min=15 in; Max

  9. SIERRA Code Coupling Module: Arpeggio User Manual Version 4.44

    Energy Technology Data Exchange (ETDEWEB)

    Sierra Thermal/Fluid Team

    2017-04-01

    The SNL Sierra Mechanics code suite is designed to enable simulation of complex multiphysics scenarios. The code suite is composed of several specialized applications which can operate either in standalone mode or coupled with each other. Arpeggio is a supported utility that enables loose coupling of the various Sierra Mechanics applications by providing access to Framework services that facilitate the coupling. More importantly Arpeggio orchestrates the execution of applications that participate in the coupling. This document describes the various components of Arpeggio and their operability. The intent of the document is to provide a fast path for analysts interested in coupled applications via simple examples of its usage.

  10. MPACT Standard Input User s Manual, Version 2.2.0

    Energy Technology Data Exchange (ETDEWEB)

    Collins, Benjamin S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Fitzgerald, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Gehin, Jess C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Graham, Aaron [Univ. of Michigan, Ann Arbor, MI (United States); Jabaay, Daniel [Univ. of Michigan, Ann Arbor, MI (United States); Kelley, Blake W. [Univ. of Michigan, Ann Arbor, MI (United States); Kim, Kang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kochunas, Brendan [Univ. of Michigan, Ann Arbor, MI (United States); Kulesza, Joel A. [Univ. of Michigan, Ann Arbor, MI (United States); Larsen, Edward W. [Univ. of Michigan, Ann Arbor, MI (United States); Liu, Yuxuan [Univ. of Michigan, Ann Arbor, MI (United States); Liu, Zhouyu [Univ. of Michigan, Ann Arbor, MI (United States); Martin, William R. [Univ. of Michigan, Ann Arbor, MI (United States); Nelson, Adam G. [Univ. of Michigan, Ann Arbor, MI (United States); Palmtag, Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rose, Michael [Univ. of Michigan, Ann Arbor, MI (United States); Saller, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Stimpson, Shane G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Trahan, Travis [Univ. of Michigan, Ann Arbor, MI (United States); Wang, J. W. [Univ. of Michigan, Ann Arbor, MI (United States); Wieselquist, William A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Young, Mitchell [Univ. of Michigan, Ann Arbor, MI (United States); Zhu, Ang [Univ. of Michigan, Ann Arbor, MI (United States)

    2016-08-01

    The MPACT (Michigan PArallel Charactistics based Transport) code is designed to perform high-fidelity light water reactor (LWR) analysis using whole-core pin-resolved neutron transport calculations on modern parallel-computing hardware. The code consists of several libraries which provide the functionality necessary to solve steady-state eigenvalue problems. Several transport capabilities are available within MPACT including both 2-D and 3-D Method of Characteristics (MOC). A three-dimensional whole core solution based on the 2D-1D solution method provides the capability for full core depletion calculations.

  11. Matrix Diffusion Toolkit. User’s Manual. Version 1.0

    Science.gov (United States)

    2012-09-01

    0.37 Soft slightly organic clay : 0.66 Soft very organic clay : 0.75 Soft bentonite : 0.84 One fractured microcrystalline limestone in Virginia... clay : 2.07 Soft slightly organic clay : 1.58 Soft very organic clay : 1.43 Soft bentonite : 1.27 Source of Data Either from an analysis of soil...0.20 Soft glacial clay : 0.57 Stiff glacial clay : 0.37 Soft slightly organic clay : 0.66 Soft very organic clay : 0.75 Soft bentonite : 0.84

  12. MPACT VERA Input User s Manual, Version 2.2.0

    Energy Technology Data Exchange (ETDEWEB)

    Collins, Benjamin S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Fitzgerald, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Gehin, Jess C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Graham, Aaron [Univ. of Michigan, Ann Arbor, MI (United States); Jabaay, Daniel [Univ. of Michigan, Ann Arbor, MI (United States); Kelley, Blake W. [Univ. of Michigan, Ann Arbor, MI (United States); Kim, Kang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kochunas, Brendan [Univ. of Michigan, Ann Arbor, MI (United States); Kulesza, Joel A. [Univ. of Michigan, Ann Arbor, MI (United States); Larsen, Edward W. [Univ. of Michigan, Ann Arbor, MI (United States); Liu, Yuxuan [Univ. of Michigan, Ann Arbor, MI (United States); Liu, Zhouyu [Univ. of Michigan, Ann Arbor, MI (United States); Martin, William R. [Univ. of Michigan, Ann Arbor, MI (United States); Nelson, Adam G. [Univ. of Michigan, Ann Arbor, MI (United States); Palmtag, Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rose, Michael [Univ. of Michigan, Ann Arbor, MI (United States); Saller, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Stimpson, Shane G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Trahan, Travis [Univ. of Michigan, Ann Arbor, MI (United States); Wang, J. W. [Univ. of Michigan, Ann Arbor, MI (United States); Wieselquist, William A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Young, Mitchell [Univ. of Michigan, Ann Arbor, MI (United States); Zhu, Ang [Univ. of Michigan, Ann Arbor, MI (United States)

    2016-06-09

    The MPACT (Michigan PArallel Charactistics based Transport) code is designed to perform high-fidelity light water reactor (LWR) analysis using whole-core pin-resolved neutron transport calculations on modern parallel-computing hardware. The code consists of several libraries which provide the functionality necessary to solve steady-state eigenvalue problems. Several transport capabilities are available within MPACT including both 2-D and 3-D Method of Characteristics (MOC). A three-dimensional whole core solution based on the 2D-1D solution method provides the capability for full core depletion calculations.

  13. An assessment of the MCNP4C weight window

    Energy Technology Data Exchange (ETDEWEB)

    Christopher N. Culbertson; John S. Hendricks

    1999-12-01

    A new, enhanced weight window generator suite has been developed for MCNP version 4C. The new generator correctly estimates importances in either a user-specified, geometry-independent, orthogonal grid or in MCNP geometric cells. The geometry-independent option alleviates the need to subdivide the MCNP cell geometry for variance reduction purposes. In addition, the new suite corrects several pathologies in the existing MCNP weight window generator. The new generator is applied in a set of five variance reduction problems. The improved generator is compared with the weight window generator applied in MCNP4B. The benefits of the new methodology are highlighted, along with a description of its limitations. The authors also provide recommendations for utilization of the weight window generator.

  14. MOCUP: MCNP-ORIGEN2 coupled utility program

    Energy Technology Data Exchange (ETDEWEB)

    Moore, R.L.; Schnitzler, B.G.; Wemple, C.A. [and others

    1995-09-30

    MOCUP is a system of external processors that allow for a limited treatment of the temporal composition of the user-selected MCNP cells in a time-dependent flux environment. The ORIGEN2 code computes the time-dependent compositions of these individually selected MCNP cells. All data communication between the two codes is accomplished through the MCNP and ORIGEN2 input/output files, the MOCUP Processor Output files, and two user supplied tables. MOCUP is either command line or interactively driven. The interactive interface is based on the portable XII window environment and the Motif tool kit. MOCUP was constructed so that no modifications to either MCNP or ORIGEN2 were necessary. Section 4 of the writeup contains the input instructions needed to set up the MOCUP run. MOCUP is extremely useful for analysts who perform isotope production, material transformation, and depletion and isotope analyses on complex, non-lattice geometries, and uniform and non-uniform lattices.

  15. Validation suite for MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Mosteller, R. D. (Russell D.)

    2002-01-01

    Two validation suites, one for criticality and another for radiation shielding, have been defined and tested for the MCNP Monte Carlo code. All of the cases in the validation suites are based on experiments so that calculated and measured results can be compared in a meaningful way. The cases in the validation suites are described, and results from those cases are discussed. For several years, the distribution package for the MCNP Monte Carlo code1 has included an installation test suite to verify that MCNP has been installed correctly. However, the cases in that suite have been constructed primarily to test options within the code and to execute quickly. Consequently, they do not produce well-converged answers, and many of them are physically unrealistic. To remedy these deficiencies, sets of validation suites are being defined and tested for specific types of applications. All of the cases in the validation suites are based on benchmark experiments. Consequently, the results from the measurements are reliable and quantifiable, and calculated results can be compared with them in a meaningful way. Currently, validation suites exist for criticality and radiation-shielding applications.

  16. Low-level RF LabVIEW{reg_sign} control software user`s manual: Version 1.0

    Energy Technology Data Exchange (ETDEWEB)

    1992-06-01

    This document details information on the low-level radio frequency (LLRF) software control package. The chapters in this manual cover the following topics: Chapter one describes the general operating principles of the LabVIEW software package, and also discusses the high-level menu panels which allow access to the individual control panels. Chapter two covers the control panels used for conditioning the cavity, and for controlling the accelerator under normal operating conditions. Chapter three provides information on the resonance detection and reflectometer calibration function, including the setup and status panels for each. Chapter four contain instructions on the use of those panels dedicated to controlling the cavity RF field. Chapter five discusses the control panels that provide setup and status information on the diagnostic monitor subsystem. Chapter six outlines those panels used to control the timing functions provided by the LLRF system. Finally, chapter seven describes the control panels used to monitor and adjust the alarm and limit functions of the system. Throughout the document, it is assumed that the reader has a general working knowledge of accelerators, high-power amplifier equipment, and low-level RF (LLRF) control systems. References are listed as footnotes as they occur in the text.

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

    Energy Technology Data Exchange (ETDEWEB)

    Forster, R.A.; Godfrey, T.N.K.

    1985-01-01

    MCNP is a very general Monte Carlo neutron photon transport code system with approximately 250 person years of Group X-6 code development invested. It is extremely portable, user-oriented, and a true production code as it is used about 60 Cray hours per month by about 150 Los Alamos users. It has as its data base the best cross-section evaluations available. MCNP contains state-of-the-art traditional and adaptive Monte Carlo techniques to be applied to the solution of an ever-increasing number of problems. Excellent user-oriented documentation is available for all facets of the MCNP code system. Many useful and important variants of MCNP exist for special applications. The Radiation Shielding Information Center (RSIC) in Oak Ridge, Tennessee is the contact point for worldwide MCNP code and documentation distribution. A much improved MCNP Version 3A will be available in the fall of 1985, along with new and improved documentation. Future directions in MCNP development will change the meaning of MCNP to Monte Carlo N Particle where N particle varieties will be transported.

  18. A new MCNP{trademark} test set

    Energy Technology Data Exchange (ETDEWEB)

    Brockhoff, R.C.; Hendricks, J.S.

    1994-09-01

    The MCNP test set is used to test the MCNP code after installation on various computer platforms. For MCNP4 and MCNP4A this test set included 25 test problems designed to test as many features of the MCNP code as possible. A new and better test set has been devised to increase coverage of the code from 85% to 97% with 28 problems. The new test set is as fast as and shorter than the MCNP4A test set. The authors describe the methodology for devising the new test set, the features that were not covered in the MCNP4A test set, and the changes in the MCNP4A test set that have been made for MCNP4B and its developmental versions. Finally, new bugs uncovered by the new test set and a compilation of all known MCNP4A bugs are presented.

  19. MCNP and GADRAS Comparisons

    Energy Technology Data Exchange (ETDEWEB)

    Klasky, Marc Louis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, Steven Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); James, Michael R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mayo, Douglas R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-04-19

    To facilitate the timely execution of System Threat Reviews (STRs) for DNDO, and also to develop a methodology for performing STRs, LANL performed comparisons of several radiation transport codes (MCNP, GADRAS, and Gamma-Designer) that have been previously utilized to compute radiation signatures. While each of these codes has strengths, it is of paramount interest to determine the limitations of each of the respective codes and also to identify the most time efficient means by which to produce computational results, given the large number of parametric cases that are anticipated in performing STR's. These comparisons serve to identify regions of applicability for each code and provide estimates of uncertainty that may be anticipated. Furthermore, while performing these comparisons, examination of the sensitivity of the results to modeling assumptions was also examined. These investigations serve to enable the creation of the LANL methodology for performing STRs. Given the wide variety of radiation test sources, scenarios, and detectors, LANL calculated comparisons of the following parameters: decay data, multiplicity, device (n,γ) leakages, and radiation transport through representative scenes and shielding. This investigation was performed to understand potential limitations utilizing specific codes for different aspects of the STR challenges.

  20. The MCNP6 Analytic Criticality Benchmark Suite

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Codes Group

    2016-06-16

    Analytical benchmarks provide an invaluable tool for verifying computer codes used to simulate neutron transport. Several collections of analytical benchmark problems [1-4] are used routinely in the verification of production Monte Carlo codes such as MCNP® [5,6]. Verification of a computer code is a necessary prerequisite to the more complex validation process. The verification process confirms that a code performs its intended functions correctly. The validation process involves determining the absolute accuracy of code results vs. nature. In typical validations, results are computed for a set of benchmark experiments using a particular methodology (code, cross-section data with uncertainties, and modeling) and compared to the measured results from the set of benchmark experiments. The validation process determines bias, bias uncertainty, and possibly additional margins. Verification is generally performed by the code developers, while validation is generally performed by code users for a particular application space. The VERIFICATION_KEFF suite of criticality problems [1,2] was originally a set of 75 criticality problems found in the literature for which exact analytical solutions are available. Even though the spatial and energy detail is necessarily limited in analytical benchmarks, typically to a few regions or energy groups, the exact solutions obtained can be used to verify that the basic algorithms, mathematics, and methods used in complex production codes perform correctly. The present work has focused on revisiting this benchmark suite. A thorough review of the problems resulted in discarding some of them as not suitable for MCNP benchmarking. For the remaining problems, many of them were reformulated to permit execution in either multigroup mode or in the normal continuous-energy mode for MCNP. Execution of the benchmarks in continuous-energy mode provides a significant advance to MCNP verification methods.

  1. AF-Geospace User’s Manual Version 2.5.1 and Version 2.51P

    Science.gov (United States)

    2012-08-01

    sponsored by the Committee on Space Research (COSPAR) and the Union Radio Scientifique Internationale (URSI). Code provided by NASA’s Space Physics...RELEASE; DISTRIBUTION IS UNLIMITED. AIR FORCE RESEARCH LABORATORY Space Vehicles Directorate 3550 Aberdeen Ave SE AIR FORCE MATERIEL COMMAND...NUMBER PPM00004262 5f. WORK UNIT NUMBER EF004375 7. PERFORMING ORGANIZATION NAME(S) AND ADDRESS(ES) Air Force Research Laboratory Space

  2. Repair Level Analysis Software (LSA Subtask 303.2.7), User’s Manual (Version 1.0)

    Science.gov (United States)

    1991-05-01

    Support Activity ( MRSA ), and includes current acquisition milestones from Operational and Organizational (O&O) plans through post fielding assessments...SYSTEM STRT-UP 3-10 04/15/91 REPAIR LEVEL ANALYSIS ROD ROBOT 08:10:04 OPERATIONS REPORTS PALMAN ASSISTANCE UTILITIES QUIT DIRECTORIES FOR RLA PROGRAM...05/22/91 REPAIR LEVEL ANALYSIS EOD ROBOT 11:15:33 OPERATIONS REPORTS PALMAN ASSISTANCE UTILITIES QUIT IDENTIFY LEVEL OF REPAIR CANDIDATES EQUIPMENT

  3. ILS Element E12 Standardization and Interoperability. Distribution Program and User’s Manual Version 1.0

    Science.gov (United States)

    1991-04-01

    EXPLAIN CONSIDERATIONS FOR PETROLEUM, OIL AND LUBRICANTS (POL): E12.1A3-27 Do the requirements and/or contract identify the need for the development system...UIPMENVT DATA 5 .n. Th.i r-- rro;-ides, n: rmatizn - -. a REPORT -v--emi pmen zein ; itb= s’s-em, eaui:men- se-’’.c-edz, ".. - uhemen- 3_-:n-’:n Screen...QSTAGs)? - Indicate whether there are any from AR 34-1 top 5 priority list and if there are any in the Petroleum Oil Lubricants (POL) category. - Which

  4. Aircraft noise prediction program propeller analysis system IBM-PC version user's manual version 2.0

    Science.gov (United States)

    Nolan, Sandra K.

    1988-01-01

    The IBM-PC version of the Aircraft Noise Prediction Program (ANOPP) Propeller Analysis System (PAS) is a set of computational programs for predicting the aerodynamics, performance, and noise of propellers. The ANOPP-PAS is a subset of a larger version of ANOPP which can be executed on CDC or VAX computers. This manual provides a description of the IBM-PC version of the ANOPP-PAS and its prediction capabilities, and instructions on how to use the system on an IBM-XT or IBM-AT personal computer. Sections within the manual document installation, system design, ANOPP-PAS usage, data entry preprocessors, and ANOPP-PAS functional modules and procedures. Appendices to the manual include a glossary of ANOPP terms and information on error diagnostics and recovery techniques.

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

    Energy Technology Data Exchange (ETDEWEB)

    Randolph Schwarz; Leland L. Carter; Alysia Schwarz

    2005-08-23

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Randolph Schwarz; Leland L. Carter; Alysia Schwarz

    2005-08-23

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

  7. A comparison between the Monte Carlo radiation transport codes MCNP and MCBEND

    Energy Technology Data Exchange (ETDEWEB)

    Sawamura, Hidenori; Nishimura, Kazuya [Computer Software Development Co., Ltd., Tokyo (Japan)

    2001-01-01

    In Japan, almost of all radiation analysts are using the MCNP code and MVP code on there studies. But these codes have not had automatic variance reduction. MCBEND code made by UKAEA have automatic variance reduction. And, MCBEND code is user friendly more than other Monte Carlo Radiation Transport Codes. Our company was first introduced MCBEND code in Japan. Therefore, we compared with MCBEND code and MCNP code about functions and production capacity. (author)

  8. MCNP-REN a Monte Carlo tool for neutron detector design

    CERN Document Server

    Abhold, M E

    2002-01-01

    The development of neutron detectors makes extensive use of the predictions of detector response through the use of Monte Carlo techniques in conjunction with the point reactor model. Unfortunately, the point reactor model fails to accurately predict detector response in common applications. For this reason, the general Monte Carlo code developed at Los Alamos National Laboratory, Monte Carlo N-Particle (MCNP), was modified to simulate the pulse streams that would be generated by a neutron detector and normally analyzed by a shift register. This modified code, MCNP-Random Exponentially Distributed Neutron Source (MCNP-REN), along with the Time Analysis Program, predicts neutron detector response without using the point reactor model, making it unnecessary for the user to decide whether or not the assumptions of the point model are met for their application. MCNP-REN is capable of simulating standard neutron coincidence counting as well as neutron multiplicity counting. Measurements of mixed oxide fresh fuel w...

  9. The New MCNP6 Depletion Capability

    Energy Technology Data Exchange (ETDEWEB)

    Fensin, Michael Lorne [Los Alamos National Laboratory; James, Michael R. [Los Alamos National Laboratory; Hendricks, John S. [Los Alamos National Laboratory; Goorley, John T. [Los Alamos National Laboratory

    2012-06-19

    The first MCNP based inline Monte Carlo depletion capability was officially released from the Radiation Safety Information and Computational Center as MCNPX 2.6.0. Both the MCNP5 and MCNPX codes have historically provided a successful combinatorial geometry based, continuous energy, Monte Carlo radiation transport solution for advanced reactor modeling and simulation. However, due to separate development pathways, useful simulation capabilities were dispersed between both codes and not unified in a single technology. MCNP6, the next evolution in the MCNP suite of codes, now combines the capability of both simulation tools, as well as providing new advanced technology, in a single radiation transport code. We describe here the new capabilities of the MCNP6 depletion code dating from the official RSICC release MCNPX 2.6.0, reported previously, to the now current state of MCNP6. NEA/OECD benchmark results are also reported. The MCNP6 depletion capability enhancements beyond MCNPX 2.6.0 reported here include: (1) new performance enhancing parallel architecture that implements both shared and distributed memory constructs; (2) enhanced memory management that maximizes calculation fidelity; and (3) improved burnup physics for better nuclide prediction. MCNP6 depletion enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. The enhancements described here help provide a powerful capability as well as dictate a path forward for future development to improve the usefulness of the technology.

  10. MCNP{trademark} Software Quality Assurance plan

    Energy Technology Data Exchange (ETDEWEB)

    Abhold, H.M.; Hendricks, J.S.

    1996-04-01

    MCNP is a computer code that models the interaction of radiation with matter. MCNP is developed and maintained by the Transport Methods Group (XTM) of the Los Alamos National Laboratory (LANL). This plan describes the Software Quality Assurance (SQA) program applied to the code. The SQA program is consistent with the requirements of IEEE-730.1 and the guiding principles of ISO 900.

  11. CONDOR-CITVAP-MCNP calculation line description

    Energy Technology Data Exchange (ETDEWEB)

    Villarino, Eduardo Anibal [INVAP S.E., San Carlos de Bariloche (Argentina)

    2002-07-01

    A general description of the CONDOR-CITVAP-MCNP calculation line is given. This calculation line starts at cross section library and allows burnup dependent detailed calculation using MCNP. This calculation line is divided in two main methodologies: CONDOR-CITVAP that allows the 3-Dimensional core burnup calculation and MCNP that performs detailed transport calculations, both methodologies are coupled using the NDDUMP code. A short description of the used codes are given: CONDOR code performs the cell calculation, generating burnup dependent macroscopic cross section and burnup dependent numerical densities per material. CITVAP codes perform the burnup dependent core calculation, including the fuel management and calculates the burnup distribution per material. NDDUMP code generates materials burnup dependent numerical densities to be used by MCNP code. This paper presents a detailed description of the CONDOR-CITVAP-MCNP calculation line and a numerical comparison of the proposed methodology. (author)

  12. Validation and verification of MCNP6 as a new simulation tool useful for medical applications

    Energy Technology Data Exchange (ETDEWEB)

    Mashnik, Stepan G [Los Alamos National Laboratory

    2011-01-06

    MCNP6, the latest and most advanced LANL transport code, representing a merger of MCNP5 and MCNPX has been Validated and Verified (V&V) against different experimental data and results by other codes relevant to medical applications. In the present work, we V&V MCNP6 using mainly the latest modifications of the Cascade-Exciton Model (CEM) and of the Los Alamos version of the Quark-Gluon String Model (LAQGSM) event generators CEM03.02 and LAQGSM03.03. We found that MCNP6 describes well data of interest for medical applications measured on both thin and thick targets and agrees very well with similar results obtained with other codes; MCNP6 may be a very useful tool for medical applications We plan to make MCNP6 available to the public via RSICC at Oak Ridge in the middle of 2011 but we are allowed to provide it to friendly US Beta-users outside LANL already now.

  13. MatMCNP: A Code for Producing Material Cards for MCNP

    Energy Technology Data Exchange (ETDEWEB)

    DePriest, Kendall Russell [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Saavedra, Karen C. [American Structurepoint, Inc., Indianapolis, IN (United States)

    2014-09-01

    A code for generating MCNP material cards (MatMCNP) has been written and verified for naturally occurring, stable isotopes. The program allows for material specification as either atomic or weight percent (fractions). MatMCNP also permits the specification of enriched lithium, boron, and/or uranium. In addition to producing the material cards for MCNP, the code calculates the atomic (or number) density in atoms/barn-cm as well as the multiplier that should be used to convert neutron and gamma fluences into dose in the material specified.

  14. MCNP APPLICATIONS FOR THE 21ST CENTURY

    Energy Technology Data Exchange (ETDEWEB)

    G. MCKINNEY; T. BOOTH; ET AL

    2000-10-01

    The Los Alamos National Laboratory (LANL) Monte Carlo N-Particle radiation transport code, MCNP, has become an international standard for a wide spectrum of neutron, photon, and electron radiation transport applications. The latest version of the code, MCNP 4C, was released to the Radiation Safety Information Computational Center (RSICC) in February 2000. This paper describes the code development philosophy, new features and capabilities, applicability to various problems, and future directions.

  15. MCNP application for the 21 century

    Energy Technology Data Exchange (ETDEWEB)

    McKinney, M.C. [and others

    2000-08-01

    The Los Alamos National Laboratory (LANL) Monte Carlo N-Particle radiation transport code, MCNP, has become an international standard for a wide spectrum of neutron, photon, and electron radiation transport applications. The latest version of the code, MCNP 4C, was released to the Radiation Safety Information Computational Center (RSICC) in February 2000. This paper describes the code development philosophy, new features and capabilities, applicability to various problems, and future directions.

  16. The prevalence and clinical characteristics associated with Diagnostic and Statistical Manual Version-5-defined anxious distress specifier in adults with major depressive disorder

    DEFF Research Database (Denmark)

    McIntyre, Roger S.; Woldeyohannes, Hanna O; Soczynska, Joanna K

    2016-01-01

    OBJECTIVES: The aim of the study was to evaluate the prevalence of and illness characteristics in adults with major depressive disorder (MDD) with anxious distress specifier (ADS) enrolled in the International Mood Disorders Collaborative Project, which is a collaborative research platform...... Manual Version-5-defined ADS was operationalized as the presence of at least two out of three proxy items instead of two out of five specifiers. RESULTS: A total of 464 individuals (i.e. 56%) met criteria for ADS. There were no between-group differences in sociodemographic variables (e.g. gender...

  17. Monte Carlo importance sampling for the MCNP{trademark} general source

    Energy Technology Data Exchange (ETDEWEB)

    Lichtenstein, H.

    1996-01-09

    Research was performed to develop an importance sampling procedure for a radiation source. The procedure was developed for the MCNP radiation transport code, but the approach itself is general and can be adapted to other Monte Carlo codes. The procedure, as adapted to MCNP, relies entirely on existing MCNP capabilities. It has been tested for very complex descriptions of a general source, in the context of the design of spent-reactor-fuel storage casks. Dramatic improvements in calculation efficiency have been observed in some test cases. In addition, the procedure has been found to provide an acceleration to acceptable convergence, as well as the benefit of quickly identifying user specified variance-reduction in the transport that effects unstable convergence.

  18. CTEx Beowulf cluster for MCNP performance

    Energy Technology Data Exchange (ETDEWEB)

    Gonzaga, Roberto N.; Amorim, Aneuri S. de; Balthar, Mario Cesar V. [Centro Tecnologico do Exercito (CTEx), Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil)

    2011-07-01

    This work is an introduction to the CTEx Nuclear Defense Department's Beowulf Cluster. Building a Beowulf Cluster is a complex learning process that greatly depends upon your hardware and software requirements. The feasibility and efficiency of performing MCNP5 calculations with a small, heterogeneous computing cluster built in Red Hat's Fedora Linux operating system personal computers (PC) are explored. The performance increases that may be expected with such clusters are estimated for cases that typify general radiation transport calculations. Our results show that the speed increase from additional slave PCs is nearly linear up to 10 processors. The pre compiled parallel binary version of MCNP uses the Message-Passing Interface (MPI) protocol. The use of this pre compiled parallel version of MCNP5 with the MPI protocol on a small, heterogeneous computing cluster built from Red Hat's Fedora Linux operating system PCs is the subject of this work. (author)

  19. Adjoint-Based Uncertainty Quantification with MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Seifried, Jeffrey E. [Univ. of California, Berkeley, CA (United States)

    2011-09-01

    This work serves to quantify the instantaneous uncertainties in neutron transport simulations born from nuclear data and statistical counting uncertainties. Perturbation and adjoint theories are used to derive implicit sensitivity expressions. These expressions are transformed into forms that are convenient for construction with MCNP6, creating the ability to perform adjoint-based uncertainty quantification with MCNP6. These new tools are exercised on the depleted-uranium hybrid LIFE blanket, quantifying its sensitivities and uncertainties to important figures of merit. Overall, these uncertainty estimates are small (< 2%). Having quantified the sensitivities and uncertainties, physical understanding of the system is gained and some confidence in the simulation is acquired.

  20. New methods for neutron response calculations with MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Hendricks, J.S. [Los Alamos National Lab., NM (United States). Applied Theoretical and Computational Physics Div.

    1997-05-01

    MCNP4B was released for international distribution in February, 1997. The author summarized the new MCNP4B features since the release of MCNP4A over three years earlier and compare some results. Then he describes new methods being developed for future code releases. The focus is methods and applications of ex-core neutron response calculations.

  1. MICRO-VERS. Micro-computer Software for the Vocational Education Reporting System. User's Guide and Reference Manual. Version 3.1. Apple II.

    Science.gov (United States)

    Illinois State Board of Education, Springfield. Dept. of Adult, Vocational and Technical Education.

    This manual is intended to accompany a software system for the Apple II microcomputer that is designed to aid local districts in completing vocational education enrollment claims and Vocational Education Data System (VEDS) reports. Part I, Introduction, gives a brief overview of the Microcomputer Vocational Education Reporting System (MICRO-VERS),…

  2. Gridded Surface Subsurface Hydrologic Analysis (GSSHA) User’s Manual; Version 1.43 for Watershed Modeling System 6.1

    Science.gov (United States)

    2006-09-01

    arithme- tic weighting of the values or a geometric average. The intercell hydraulic conductivity weighting method is selected with the RICHARDS_K_OPTION...ERDC/CHL SR-06-1 141 where: Cs is the total concentration of sand-size sediment particles in motion, USo is unit stream power (L T-1), U* is the

  3. CATDAT : A Program for Parametric and Nonparametric Categorical Data Analysis : User's Manual Version 1.0, 1998-1999 Progress Report.

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, James T.

    1999-12-01

    Natural resource professionals are increasingly required to develop rigorous statistical models that relate environmental data to categorical responses data. Recent advances in the statistical and computing sciences have led to the development of sophisticated methods for parametric and nonparametric analysis of data with categorical responses. The statistical software package CATDAT was designed to make some of these relatively new and powerful techniques available to scientists. The CATDAT statistical package includes 4 analytical techniques: generalized logit modeling; binary classification tree; extended K-nearest neighbor classification; and modular neural network.

  4. Evaluation of Geometric Progression (GP Buildup Factors using MCNP Codes (MCNP6.1 and MCNP5-1.60

    Directory of Open Access Journals (Sweden)

    Kim Kyung-O

    2016-01-01

    Full Text Available The gamma-ray buildup factors of three-dimensional point kernel code (QAD-CGGP are re-evaluated by using MCNP codes (MCNP6.1 and MCNPX5-1.60 and ENDF/B-VI.8 photoatomic data, which cover an energy range of 0.015–15 MeV and an iron thickness of 0.5–40 Mean Free Path (MFP. These new data are fitted to the Geometric Progression (GP fitting function and are then compared with ANS standard data equipped with QAD-CGGP. In addition, a simple benchmark calculation was performed to compare the QAD-CGGP results applied with new and existing buildup factors based on the MCNP codes. In the case of the buildup factors of low-energy gamma-rays, new data are evaluated to be about 5% higher than the existing data. In other cases, these new data present a similar trend based on the specific penetration depth, while existing data continuously increase beyond that depth. In a simple benchmark, the calculations using the existing data were slightly underestimated compared to the reference data at a deep penetration depth. On the other hand, the calculations with new data were stabilized with an increasing penetration depth, despite a slight overestimation at a shallow penetration depth.

  5. Criticality Calculations with MCNP6 - Practical Lectures

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Methods, Codes, and Applications (XCP-3)

    2016-11-29

    These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.

  6. CGMF & FREYA Verification in MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-05

    At the present time, the new and updated fission event generators included in MCNP6.2 have been verified to be functioning properly through a variety of detailed tests. This work describes the detailed verification steps taken to ensure these complicated fission event generators, FREYA and CGMF, are integrated into MCNP6 properly. Ultimately, with the knowledge that MCNP6 is making use of these models appropriately, we can now begin to validate the models against benchmarked experiments. Some benchmarks, including criticality and subcritical experiments interested in multiplication and bulk counting rates, are easy to model and understand but are likely insensitive to the detailed nature of these models. It will take some new measurements with coincidence detection capabilities to be able to stress the physics within each of these fission event generator models. Once the models are validated and it is understood where the models can truly be predictive, then we can study what SNM observables can be characterized for nonproliferation applications.

  7. Development of Monteburns: A Code That Links MCNP and ORIGEN2 in an Automated Fashion for Burnup Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Holly R. Trellue

    1998-12-01

    Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code 0RIGEN2. Monteburns produces many criticality and burnup computational parameters based on material feed/removal specifications, power(s), and time intervals. This code processes input from the user indicating the system geometry, initial material compositions, feed/removal, and other code-specific parameters. Results from MCNP, 0RIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to first transfer one-group cross sections and fluxes from MCNP to 0RIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from 0RIGEN2 back to MCNP in a repeated, cyclic fashion. The main requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with 0RIGEN2 and other calculations are performed by monteburns. This report presents the results obtained from the benchmarking of monteburns to measured and previously obtained data from traditional Light Water Reactor systems. The majority of the differences seen between the two were less than five percent. These were primarily a result of variances in cross sections between MCNP, cross section libraries used by other codes, and observed values. With this understanding, this code can now be used with confidence for burnup calculations in three-dimensional systems. It was designed for use in the Accelerator Transmutation of Waste project at Los Alamos National Laboratory but is also being applied to the analysis of isotopic production/destruction of transuranic actinides in a reactor system. The code has now been shown to sufficiently support these calculations.

  8. MCNP simulations of material exposure experiments (u)

    Energy Technology Data Exchange (ETDEWEB)

    Temple, Brian A [Los Alamos National Laboratory

    2010-12-08

    Simulations of proposed material exposure experiments were performed using MCNP6. The experiments will expose ampules containing different materials of interest with radiation to observe the chemical breakdown of the materials. Simulations were performed to map out dose in materials as a function of distance from the source, dose variation between materials, dose variation due to ampule orientation, and dose variation due to different source energy. This write up is an overview of the simulations and will provide guidance on how to use the data in the spreadsheet.

  9. GNU Octave Manual Version 3

    DEFF Research Database (Denmark)

    W. Eaton, John; Bateman, David; Hauberg, Søren

    This manual is the definitive guide to GNU Octave, an interactive environment for numerical computation. The manual covers the new version 3 of GNU Octave.......This manual is the definitive guide to GNU Octave, an interactive environment for numerical computation. The manual covers the new version 3 of GNU Octave....

  10. MCNP6 Cosmic & Terrestrial Background Particle Fluxes -- Release 4

    Energy Technology Data Exchange (ETDEWEB)

    McMath, Garrett E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Div.; McKinney, Gregg W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Div.; Wilcox, Trevor [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Nuclear Engineering and Nonproliferation Div.

    2015-01-23

    Essentially a set of slides, the presentation begins with the MCNP6 cosmic-source option, then continues with the MCNP6 transport model (atmospheric, terrestrial) and elevation scaling. It concludes with a few slides on results, conclusions, and suggestions for future work.

  11. Semi-Analytical Benchmarks for MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Grechanuk, Pavel Aleksandrovi [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-11-07

    Code verification is an extremely important process that involves proving or disproving the validity of code algorithms by comparing them against analytical results of the underlying physics or mathematical theory on which the code is based. Monte Carlo codes such as MCNP6 must undergo verification and testing upon every release to ensure that the codes are properly simulating nature. Specifically, MCNP6 has multiple sets of problems with known analytic solutions that are used for code verification. Monte Carlo codes primarily specify either current boundary sources or a volumetric fixed source, either of which can be very complicated functions of space, energy, direction and time. Thus, most of the challenges with modeling analytic benchmark problems in Monte Carlo codes come from identifying the correct source definition to properly simulate the correct boundary conditions. The problems included in this suite all deal with mono-energetic neutron transport without energy loss, in a homogeneous material. The variables that differ between the problems are source type (isotropic/beam), medium dimensionality (infinite/semi-infinite), etc.

  12. Benchmarking MCNP and TRIPOLI with PGNAA measurements

    Science.gov (United States)

    Carasco, C.; Perot, B.; Sikora, A.; Mauerhofer, E.; Havenith, A.; Payan, E.; Kettler, J.; Kring, T.; Ma, J. L.

    2014-06-01

    The French Alternative Energies and Atomic Energy Commission (CEA Cadarache), the Forschungszentrum Jülich GmbH (FZJ), and the RWTH Aachen University (RWTH) are involved in a cooperation aiming at characterizing toxic and reactive elements in radioactive waste packages by means of Prompt Gamma Neutron Activation Analysis (PGNAA). The design of an optimized measurement system and the assessment of its performances for realistic scenarios can be conveniently studied by numerical Monte Carlo simulation, provided the model and nuclear data offer a sufficient precision. Previous studies performed with MCNP have shown that when the nuclear data libraries lack of precision, relevant results can still be obtained by performing calculations in multiple steps (by first determining the radiative capture rate, and transporting the induced gamma toward the detector) and by injecting valid gamma-ray production data in-between [1]. In such cases, it is interesting to compare the results obtained with different codes. In the present paper, we propose to compare the MCNP and TRIPOLI codes with measurements obtained in MEDINA (Multi Element Detection based on Instrumental Neutron Activation), which is the new FZJ PGNAA facility [2]. The aim of the measurement campaign is to assess capture gamma rays of toxic elements that can be found in 200 L waste drums which are expected for geological repository.

  13. Voxel2MCNP: software for handling voxel models for Monte Carlo radiation transport calculations.

    Science.gov (United States)

    Hegenbart, Lars; Pölz, Stefan; Benzler, Andreas; Urban, Manfred

    2012-02-01

    Voxel2MCNP is a program that sets up radiation protection scenarios with voxel models and generates corresponding input files for the Monte Carlo code MCNPX. Its technology is based on object-oriented programming, and the development is platform-independent. It has a user-friendly graphical interface including a two- and three-dimensional viewer. A row of equipment models is implemented in the program. Various voxel model file formats are supported. Applications include calculation of counting efficiency of in vivo measurement scenarios and calculation of dose coefficients for internal and external radiation scenarios. Moreover, anthropometric parameters of voxel models, for instance chest wall thickness, can be determined. Voxel2MCNP offers several methods for voxel model manipulations including image registration techniques. The authors demonstrate the validity of the program results and provide references for previous successful implementations. The authors illustrate the reliability of calculated dose conversion factors and specific absorbed fractions. Voxel2MCNP is used on a regular basis to generate virtual radiation protection scenarios at Karlsruhe Institute of Technology while further improvements and developments are ongoing.

  14. Criticality calculations with MCNP{trademark}: A primer

    Energy Technology Data Exchange (ETDEWEB)

    Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A. [New Mexico Univ., Albuquerque, NM (United States)

    1994-06-06

    With the closure of many experimental facilities, the nuclear criticality safety analyst increasingly is required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his/her facility. This primer will help you, the analyst, understand and use the MCNP Monte Carlo code for nuclear criticality safety analyses. It assumes that you have a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with MCNP in particular. Appendix A gives an introduction to Monte Carlo techniques. The primer is designed to teach by example, with each example illustrating two or three features of MCNP that are useful in criticality analyses. Beginning with a Quickstart chapter, the primer gives an overview of the basic requirements for MCNP input and allows you to run a simple criticality problem with MCNP. This chapter is not designed to explain either the input or the MCNP options in detail; but rather it introduces basic concepts that are further explained in following chapters. Each chapter begins with a list of basic objectives that identify the goal of the chapter, and a list of the individual MCNP features that are covered in detail in the unique chapter example problems. It is expected that on completion of the primer you will be comfortable using MCNP in criticality calculations and will be capable of handling 80 to 90 percent of the situations that normally arise in a facility. The primer provides a set of basic input files that you can selectively modify to fit the particular problem at hand.

  15. Justine user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.R.

    1995-10-01

    Justine is the graphical user interface to the Los Alamos Radiation Modeling Interactive Environment (LARAMIE). It provides LARAMIE customers with a powerful, robust, easy-to-use, WYSIWYG interface that facilitates geometry construction and problem specification. It is assumed that the reader is familiar with LARAMIE, and the transport codes available, i.e., MCNPTM and DANTSYSTM. No attempt is made in this manual to describe these codes in detail. Information about LARAMIE, DANTSYS, and MCNP are available elsewhere. It i also assumed that the reader is familiar with the Unix operating system and with Motif widgets and their look and feel. However, a brief description of Motif and how one interacts with it can be found in Appendix A.

  16. Criticality calculations with MCNP{sup TM}: A primer

    Energy Technology Data Exchange (ETDEWEB)

    Mendius, P.W. [ed.; Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A.

    1994-08-01

    The purpose of this Primer is to assist the nuclear criticality safety analyst to perform computer calculations using the Monte Carlo code MCNP. Because of the closure of many experimental facilities, reliance on computer simulation is increasing. Often the analyst has little experience with specific codes available at his/her facility. This Primer helps the analyst understand and use the MCNP Monte Carlo code for nuclear criticality analyses. It assumes no knowledge of or particular experience with Monte Carlo codes in general or with MCNP in particular. The document begins with a Quickstart chapter that introduces the basic concepts of using MCNP. The following chapters expand on those ideas, presenting a range of problems from simple cylinders to 3-dimensional lattices for calculating keff confidence intervals. Input files and results for all problems are included. The Primer can be used alone, but its best use is in conjunction with the MCNP4A manual. After completing the Primer, a criticality analyst should be capable of performing and understanding a majority of the calculations that will arise in the field of nuclear criticality safety.

  17. Lecture Notes on Criticality Safety Validation Using MCNP & Whisper

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-11

    Training classes for nuclear criticality safety, MCNP documentation. The need for, and problems surrounding, validation of computer codes and data area considered first. Then some background for MCNP & Whisper is given--best practices for Monte Carlo criticality calculations, neutron spectra, S(α,β) thermal neutron scattering data, nuclear data sensitivities, covariance data, and correlation coefficients. Whisper is computational software designed to assist the nuclear criticality safety analyst with validation studies with the Monte Carlo radiation transport package MCNP. Whisper's methodology (benchmark selection – Ck's, weights; extreme value theory – bias, bias uncertainty; MOS for nuclear data uncertainty – GLLS) and usage are discussed.

  18. Recent Developments in the MCNP-POLIMI Postprocessing Code

    Energy Technology Data Exchange (ETDEWEB)

    Pozzi, S.A.

    2004-12-17

    The design and analysis of measurements performed with organic scintillators rely on the use of Monte Carlo codes to simulate the interaction of neutrons and photons, originating from fission and other reactions, with the materials present in the system and the radiation detectors. MCNP-PoliMi is a modification of the MCNP-4c code that models the physics of secondary particle emission from fission and other processes realistically. This characteristic allows for the simulation of the higher moments of the distribution of the number of neutrons and photons in a multiplying system. The present report describes the recent additions to the MCNP-PoliMi post-processing code. These include the simulation of detector dead time, multiplicity, and third order statistics.

  19. Features of MCNP6 Relevant to Medical Radiation Physics

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, H. Grady III [Los Alamos National Laboratory; Goorley, John T. [Los Alamos National Laboratory

    2012-08-29

    MCNP (Monte Carlo N-Particle) is a general-purpose Monte Carlo code for simulating the transport of neutrons, photons, electrons, positrons, and more recently other fundamental particles and heavy ions. Over many years MCNP has found a wide range of applications in many different fields, including medical radiation physics. In this presentation we will describe and illustrate a number of significant recently-developed features in the current version of the code, MCNP6, having particular utility for medical physics. Among these are major extensions of the ability to simulate large, complex geometries, improvement in memory requirements and speed for large lattices, introduction of mesh-based isotopic reaction tallies, advances in radiography simulation, expanded variance-reduction capabilities, especially for pulse-height tallies, and a large number of enhancements in photon/electron transport.

  20. Geometry creation for MCNP by Sabrina and XSM

    Energy Technology Data Exchange (ETDEWEB)

    Van Riper, K.A.

    1994-02-01

    The Monte Carlo N-Particle transport code MCNP is based on a surface description of 3-dimensional geometry. Cells are defined in terms of boolean operations on signed quadratic surfaces. MCNP geometry is entered as a card image file containing coefficients of the surface equations and a list of surfaces and operators describing cells. Several programs are available to assist in creation of the geometry specification, among them Sabrina and the new ``Smart Editor`` code XSM. We briefly describe geometry creation in Sabrina and then discuss XSM in detail. XSM is under development; our discussion is based on the state of XSM as of January 1, 1994.

  1. An Electron/Photon/Relaxation Data Library for MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Hughes, III, H. Grady [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-08-07

    The capabilities of the MCNP6 Monte Carlo code in simulation of electron transport, photon transport, and atomic relaxation have recently been significantly expanded. The enhancements include not only the extension of existing data and methods to lower energies, but also the introduction of new categories of data and methods. Support of these new capabilities has required major additions to and redesign of the associated data tables. In this paper we present the first complete documentation of the contents and format of the new electron-photon-relaxation data library now available with the initial production release of MCNP6.

  2. Determining the effects of microsphere and surrounding material composition on {sup 90}Y dose kernels using egsnrc and mcnp5

    Energy Technology Data Exchange (ETDEWEB)

    Paxton, Adam B.; Davis, Stephen D.; DeWerd, Larry A. [Department of Medical Physics, University of Wisconsin-Madison, Madison, Wisconsin 53705 (United States); Department of Medical Physics, University of Wisconsin-Madison, Madison, Wisconsin 53705 and McGill University Health Centre, Department of Medical Physics, Montreal, Quebec H3G 1A4 (Canada); Department of Medical Physics, University of Wisconsin-Madison, Madison, Wisconsin 53705 (United States)

    2012-03-15

    Purpose: Recent advances in the imaging of {sup 90}Y using positron emission tomography (PET) and improved uncertainty in the branching ratio for the internal pair production component of {sup 90}Y decay allow for a more accurate determination of the activity distribution of {sup 90}Y microspheres within a patient. This improved activity distribution can be convolved with the dose kernel of {sup 90}Y to calculate the dose distribution within a patient. This work investigates the effects of microsphere and surrounding material composition on {sup 90}Y dose kernels using egsnrc and mcnp5 and compares the results of these two transport codes. Methods: Monte Carlo simulations were performed with egsnrc and mcnp5 to calculate the dose rate at multiple radial distances around various {sup 90}Y sources. Point source simulations were completed with mcnp5 to determine the optimal electron transport settings for this work. After determining the optimal settings, point source simulations were completed using egsnrc (user code edknrc) and mcnp5 in water and liver [as defined by the International Commission on Radiation Units and Measurements (ICRU) Report 44]. The results were compared to ICRU Report 72 reference data. Point source simulations were also completed in water with a density of 1.06 g{center_dot}cm{sup -3} to evaluate the effect of the density of the surrounding material. Glass and resin microsphere simulations were performed with average and maximum diameter and density values (based on values given in the literature) in water and in liver. The results were compared to point source simulation results using the same transport code and in the same surrounding material. All simulations had statistical uncertainties less than 1%. Results: The optimal transport settings in mcnp5 for this work included using the energy-and step-specific algorithm (DBCN 17J 2) and ESTEP set to 10. These settings were used for all subsequent simulations with mcnp5. The point source

  3. MCNP6. Simulating Correlated Data in Fission Events

    Energy Technology Data Exchange (ETDEWEB)

    Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sood, Avneet [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-12-03

    This report is a series of slides discussing the MCNP6 code and its status in simulating fission. Applications of interest include global security and nuclear nonproliferation, detection of special nuclear material (SNM), passive and active interrogation techniques, and coincident neutron and photon leakage.

  4. Duplicating MC-15 Output with Python and MCNP

    Energy Technology Data Exchange (ETDEWEB)

    McSpaden, Alexander Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-08-23

    Two Python scripts have been written that process the output files of MCNP6 into a format that mimics the list-mode output of Los Alamos National Laboratory’s MC-15 and NPOD neutron detection systems. This report details the methods implemented in these scripts and instructions on their use.

  5. Possible Improvements to MCNP6 and its CEM/LAQGSM Event-Generators

    Energy Technology Data Exchange (ETDEWEB)

    Mashnik, Stepan Georgievich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-08-04

    This report is intended to the MCNP6 developers and sponsors of MCNP6. It presents a set of suggested possible future improvements to MCNP6 and to its CEM03.03 and LAQGSM03.03 event-generators. A few suggested modifications of MCNP6 are quite simple, aimed at avoiding possible problems with running MCNP6 on various computers, i.e., these changes are not expected to change or improve any results, but should make the use of MCNP6 easier; such changes are expected to require limited man-power resources. On the other hand, several other suggested improvements require a serious further development of nuclear reaction models, are expected to improve significantly the predictive power of MCNP6 for a number of nuclear reactions; but, such developments require several years of work by real experts on nuclear reactions.

  6. Fission Matrix Capability for MCNP Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Carney, Sean E. [Los Alamos National Laboratory; Brown, Forrest B. [Los Alamos National Laboratory; Kiedrowski, Brian C. [Los Alamos National Laboratory; Martin, William R. [Los Alamos National Laboratory

    2012-09-05

    In a Monte Carlo criticality calculation, before the tallying of quantities can begin, a converged fission source (the fundamental eigenvector of the fission kernel) is required. Tallies of interest may include powers, absorption rates, leakage rates, or the multiplication factor (the fundamental eigenvalue of the fission kernel, k{sub eff}). Just as in the power iteration method of linear algebra, if the dominance ratio (the ratio of the first and zeroth eigenvalues) is high, many iterations of neutron history simulations are required to isolate the fundamental mode of the problem. Optically large systems have large dominance ratios, and systems containing poor neutron communication between regions are also slow to converge. The fission matrix method, implemented into MCNP[1], addresses these problems. When Monte Carlo random walk from a source is executed, the fission kernel is stochastically applied to the source. Random numbers are used for: distances to collision, reaction types, scattering physics, fission reactions, etc. This method is used because the fission kernel is a complex, 7-dimensional operator that is not explicitly known. Deterministic methods use approximations/discretization in energy, space, and direction to the kernel. Consequently, they are faster. Monte Carlo directly simulates the physics, which necessitates the use of random sampling. Because of this statistical noise, common convergence acceleration methods used in deterministic methods do not work. In the fission matrix method, we are using the random walk information not only to build the next-iteration fission source, but also a spatially-averaged fission kernel. Just like in deterministic methods, this involves approximation and discretization. The approximation is the tallying of the spatially-discretized fission kernel with an incorrect fission source. We address this by making the spatial mesh fine enough that this error is negligible. As a consequence of discretization we get a

  7. User Manual for Whisper-1.1

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Alwin, Jennifer Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-26

    Whisper is a statistical analysis package developed in 2014 to support nuclear criticality safety (NCS) validation [1-3]. It uses the sensitivity profile data for an application as computed by MCNP6 [4-6] along with covariance files [7,8] for the nuclear data to determine a baseline upper-subcritical-limit (USL) for the application. Whisper version 1.0 was first developed and used at LANL in 2014 [3]. During 2015- 2016, Whisper was updated to version 1.1 and is to be included with the upcoming release of MCNP6.2. This document describes the user input and options for running whisper-1.1, including 2 perl utility scripts that simplify ordinary NCS work, whisper_mcnp.pl and whisper_usl.pl. For many detailed references on the theory, applications, nuclear data & covariances, SQA, verification-validation, adjointbased methods for sensitivity-uncertainty analysis, and more – see the Whisper – NCS Validation section of the MCNP Reference Collection at mcnp.lanl.gov. There are currently over 50 Whisper reference documents available.

  8. A Verification of MCNP6 FMESH Tally Capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Swift, Alicia L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McKigney, Edward A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Schirato, Richard C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Robinson, Alex Philip [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Temple, Brian Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-02-10

    This work serves to verify the MCNP6 FMESH capability through comparison to two types of data. FMESH tallies, binned in time, were generated on an ideal detector face for neutrons undergoing a single scatter in a graphite target. For verification, FMESH results were compared to analytic calculations of the nonrelativistic TOF for elastic and inelastic single neutron scatters (TOF for the purposes of this paper is the time for a neutron to travel from its scatter location in the graphite target to the detector face). FMESH tally results were also compared to F4 tally results, an MNCP tally that calculates fluence in the same way as the FMESH tally. The FMESH tally results agree well with the analytic results and the F4 tally; hence, it is believed that, for simple geometries, MCNP6 FMESH tallies represent the physics of neutron scattering very well.

  9. Energetic Light Fragment Production Capability in MCNP6

    CERN Document Server

    Kerby, Leslie M; Gudima, Konstantin K; Sierk, Arnold J; Bull, Jeffrey S; James, Michael R

    2016-01-01

    The goal of this research is to enable MCNP6 to produce high-energy light fragments. These energetic light fragments may be emitted by our models through three processes: Fermi breakup, preequilibrium, and coalescence. We explore the emission of light fragments through each of these mechanisms and demonstrate an improved agreement with experimental data achieved by extending precompound models to include emission of fragments heavier than $^4$He.

  10. General purpose photoneutron production in MCNP4A

    Energy Technology Data Exchange (ETDEWEB)

    Gallmeier, F.X.

    1995-08-01

    A photoneutron production option was implemented in the MCNP4A code, mainly to supply a tool for reactor shielding calculations in beryllium and heavy water environments of complicated three-dimensional geometries. Photoneutron production cross sections for deuterium and beryllium were created. Subroutines were developed to calculate the probability of photoneutron production at photon collision sites and the energy and flight direction of the created photoneutrons. These subroutines were implemented into MCNP4A. Some small program changes were necessary for processing the input to read the photoneutron production cross sections and to install a photoneutron switch. Some arrays were installed or extended to sample photoneutron creation and loss information, and output routines were changed to give the appropriate summary tables. To verify and validate the photoneutron production data and the MCNP4A implementations, the yields of photoneutron sources were calculated and compared with experiments. In the case of deuterium-based photoneutron sources, the calculations agreed well with the experiments; the beryuium-based photoneutron source calculations were up to 30% higher compared with the measurements. More accurate beryllium photoneutron cross sections would be desirable. To apply the developed method to a real shielding problem, the fast neutron fluxes in the heavy-water-filled reflector vessel of the Advanced Neutron Source reactor were investigated and compared with published DORT calculations. Considering the complete independence between the calculations, the merely 10 to 20% lower fluxes obtained with MCNP4A, compared against the DORT results, were more than satisfactory, as the discrepancy is based primarily on differences in the calculated thermal neutron fluxes.

  11. Implementation of On-the-Fly Doppler Broadening in MCNP5 for Multiphysics Simulation of Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    William Martin

    2012-11-16

    A new method to obtain Doppler broadened cross sections has been implemented into MCNP, removing the need to generate cross sections for isotopes at problem temperatures. Previous work had established the scientific feasibility of obtaining Doppler-broadened cross sections "on-the-fly" (OTF) during the random walk of the neutron. Thus, when a neutron of energy E enters a material region that is at some temperature T, the cross sections for that material at the exact temperature T are immediately obtained by interpolation using a high order functional expansion for the temperature dependence of the Doppler-broadened cross section for that isotope at the neutron energy E. A standalone Fortran code has been developed that generates the OTF library for any isotope that can be processed by NJOY. The OTF cross sections agree with the NJOY-based cross sections for all neutron energies and all temperatures in the range specified by the user, e.g., 250K - 3200K. The OTF methodology has been successfully implemented into the MCNP Monte Carlo code and has been tested on several test problems by comparing MCNP with conventional ACE cross sections versus MCNP with OTF cross sections. The test problems include the Doppler defect reactivity benchmark suite and two full-core VHTR configurations, including one with multiphysics coupling using RELAP5-3D/ATHENA for the thermal-hydraulic analysis. The comparison has been excellent, verifying that the OTF libraries can be used in place of the conventional ACE libraries generated at problem temperatures. In addition, it has been found that using OTF cross sections greatly reduces the complexity of the input for MCNP, especially for full-core temperature feedback calculations with many temperature regions. This results in an order of magnitude decrease in the number of input lines for full-core configurations, thus simplifying input preparation and reducing the potential for input errors. Finally, for full-core problems with multiphysics

  12. Comparison of CAP88 and MCNP for Overhead Gamma-emitting Plumes

    Energy Technology Data Exchange (ETDEWEB)

    Mcnaughton, Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gillis, Jessica Mcdonnel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McClory, Aysha Reede [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Whicker, Jeffrey Jay [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fuehne, David Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-01-08

    The purpose of this paper is to use the Monte Carlo N-Particle Code (MCNP) to investigate the dose from gamma-emitting radionuclides such as Carbon-11 when a plume passes overhead. MCNP results are compared with results from the EPA program, CAP88. In some cases, typically near the source during stable conditions, the CAP88 results are less than the MCNP results. However, in the case of a receptor 800 m from a source at the Los Alamos Neutron Science Center (LANSCE), the CAP88 result is greater than the MCNP result.

  13. Social values for ecosystem services (SolVES): A GIS application for assessing, mapping, and quantifying the social values of ecosystem services-Documentation and user manual, version 1.0

    Science.gov (United States)

    Sherrouse, Benson C.; Riegle, Jodi L.; Semmens, Darius J.

    2010-01-01

    In response to the need for incorporating quantified and spatially explicit measures of social values into ecosystem services assessments, the Rocky Mountain Geographic Science Center, in collaboration with Colorado State University, has developed a geographic information system application, Social Values for Ecosystem Services (SolVES). SolVES can be used to assess, map, and quantify the perceived social values of ecosystem services. SolVES derives a quantitative social values metric, the Value Index, from a combination of spatial and nonspatial responses to public attitude and preference surveys. SolVES also generates landscape metrics, such as average elevation and distance to water, calculated from spatial data layers describing the underlying physical environment. Using kernel density calculations and zonal statistics, SolVES derives and maps the 10-point Value Index and reports landscape metrics associated with each index value for social value types such as aesthetics, biodiversity, and recreation. This can be repeated for various survey subgroups as distinguished by their attitudes and preferences regarding public uses of the forests such as motorized recreation and logging for fuels reduction. The Value Index provides a basis of comparison within and among survey subgroups to consider the effect of social contexts on the valuation of ecosystem services. SolVES includes regression coefficients linking the predicted value (the Value Index) to landscape metrics. These coefficients are used to generate predicted social value maps using value transfer techniques for areas where primary survey data are not available. SolVES was developed, and will continue to be enhanced through future versions, as a public domain tool to enable decision makers and researchers to map the social values of ecosystem services and to facilitate discussions among diverse stakeholders regarding tradeoffs between different ecosystem services in a variety of physical and social contexts.

  14. New gamma and neutron measurements and MCNP simulations

    Energy Technology Data Exchange (ETDEWEB)

    Crovisier, Ph.; Camus, L.; Marty, P. [CEA Centre de Valduc, Is sur Tille (France). Service de Protection contre les Rayonnements; Groetz, J.E [Univ. de Franche Comte, Besancon (France). Laboratoire de Microanalyses Nucleaires

    2003-05-01

    To take into account the criticality risk, the Radiation Protection Service of the CEA Valduc center has developed a new method allowing quickly fixed fissile material mass determination in complex configurations where the other classical techniques, such as gamma spectrometry, cannot be easily used (contaminated areas, large thickness shield protection). Then, the Radiation Protection Service in collaboration with the Nuclear Microanalyses Laboratory carried out ambient dose equivalent rate measurements coupled with a MCNP simulations in order to estimate 'holdup' nuclear materials. The methodology used is described below: Choice of measurement devices (gamma or neutron) according to the detection limits. Use of calibrated dose rate meters and new neutron spectrometer ROSPEC (measurement references and uncertainties). Ambient dose equivalent rate measurements should be performed at different locations in the vicinity of the system studied. Complete geometry system, shields and sources locations (if it's possible) should be modeled accurately in MCNP simulations. Ambient dose equivalent rate calculations at each measurement locations and for each source described are performed by using the MCNP code. All these measurements and calculations allow to set up a linear equations system with activities sources (mass) as unknowns. Due to the measurement uncertainties, this system cannot be exactly solved but by an iterative approach. The fissile material characteristics (i.e isotopic abundance, chemical form, nuclides) located in the system are very important to enable us the nuclear material mass estimations. Previously, these features can be determined by smears radiological analyses or by knowing the elaborated nuclear materials in the concerned plant. For the first time, this new method was successfully used to study a vessel containing metal plutonium located on the walls. The second estimation concerned the 'holdup' fissile material in a

  15. MCNP5 CALCULATIONS REPLICATING ARH-600 NITRATE DATA

    Energy Technology Data Exchange (ETDEWEB)

    FINFROCK SH

    2011-10-25

    This report serves to extend the previous document: 'MCNP Calculations Replicating ARH-600 Data' by replicating the nitrate curves found in ARH-600. This report includes the MCNP models used, the calculated critical dimension for each analyzed parameter set, and the resulting data libraries for use with the CritView code. As with the ARH-600 data, this report is not meant to replace the analysis of the fissile systems by qualified criticality personnel. The M CNP data is presented without accounting for the statistical uncertainty (although this is typically less than 0.001) or bias and, as such, the application of a reasonable safety margin is required. The data that follows pertains to the uranyl nitrate and plutonium nitrate spheres, infinite cylinders, and infinite slabs of varying isotopic composition, reflector thickness, and molarity. Each of the cases was modeled in MCNP (version 5.1.40), using the ENDF/B-VI cross section set. Given a molarity, isotopic composition, and reflector thickness, the fissile concentration and diameter (or thicknesses in the case of the slab geometries) were varied. The diameter for which k-effective equals 1.00 for a given concentration could then be calculated and graphed. These graphs are included in this report. The pages that follow describe the regions modeled, formulas for calculating the various parameters, a list of cross-sections used in the calculations, a description of the automation routine and data, and finally the data output. The data of most interest are the critical dimensions of the various systems analyzed. This is presented graphically, and in table format, in Appendix B. Appendix C provides a text listing of the same data in a format that is compatible with the CritView code. Appendices D and E provide listing of example Template files and MCNP input files (these are discussed further in Section 4). Appendix F is a complete listing of all of the output data (i.e., all of the analyzed dimensions and

  16. Validation of the Monte Carlo code MCNP-DSP

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.; Mihalczo, J.T. [Oak Ridge National Lab., TN (United States)

    1996-09-12

    Several calculations were performed to validate MCNP-DSP, which is a Monte Carlo code that calculates all the time and frequency analysis parameters associated with the {sup 252}Cf-source-driven time and frequency analysis method. The frequency analysis parameters are obtained in two ways: directly by Fourier transforming the detector responses and indirectly by taking the Fourier transform of the autocorrelation and cross-correlation functions. The direct and indirect Fourier processing methods were shown to produce the same frequency spectra and convergence, thus verifying the way to obtain the frequency analysis parameters from the time sequences of detector pulses. (Author).

  17. Uncertainty analysis in MCNP5 calculations for brachytherapy treatment

    Energy Technology Data Exchange (ETDEWEB)

    Gerardy, I., E-mail: gerardy@isib.be [Institut Superieur Industriel de Bruxelles, 150, Rue Royale, B-1000 Brussels (Belgium); Rodenas, J.; Gallardo, S. [Departamento de Ingenieria Quimica y Nuclear, Universidad Politecnica de Valencia (Spain)

    2011-08-15

    The Monte Carlo (MC) method can be applied to simulate brachytherapy treatment planning. The MCNP5 code gives, together with results, a statistical uncertainty associated with them. However, the latter is not the only existing uncertainty related to the simulation and other uncertainties must be taken into account. A complete analysis of all sources of uncertainty having some influence on results of the simulation of brachytherapy treatment is presented in this paper. This analysis has been based on the recommendations of the American Association for Physicist in Medicine (AAPM) and of the International Standard Organisation (ISO).

  18. Bigfoot Field Manual, Version 2.1

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, J.L.; Burrows, S.; Gower, S.T.; Cohen, W.B.

    1999-09-01

    The BigFoot Project is funded by the Earth Science Enterprise to collect and organize data to be used in the National Aeronautics and Space Administration's Earth Observing System (EOS) Validation Program. The data collected by the BigFoot Project are unique in being ground-based observations coincident with satellite overpasses. In addition to collecting data, the BigFoot project will develop and test new algorithms for scaling point measurements to the same spatial scales as the EOS satellite products. This BigFoot Field Manual will be used to achieve completeness and consistency of data collected at four initial BigFoot sites and at future sites that may collect similar validation data. Therefore, validation datasets submitted to the Oak Ridge National Laboratory Distributed Active Archive Center that have been compiled in a manner consistent with the field manual will be especially valuable in the validation program.

  19. MCNP6 and DRiFT modeling efforts for the NEUANCE/DANCE detector array

    Energy Technology Data Exchange (ETDEWEB)

    Pinilla, Maria Isabel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-01-30

    This report seeks to study and benchmark code predictions against experimental data; determine parameters to match MCNP-simulated detector response functions to experimental stilbene measurements; add stilbene processing capabilities to DRiFT; and improve NEUANCE detector array modeling and analysis using new MCNP6 and DRiFT features.

  20. MCNP modelling of a combined neutron/gamma counter

    CERN Document Server

    Bourva, L C A; Ottmar, H; Weaver, D R

    1999-01-01

    A series of Monte Carlo neutron calculations for a combined gamma/passive neutron coincidence counter has been performed. This type of device, part of a suite of non-destructive assay instruments utilised for the enforcement of the Euratom nuclear safeguards within the European Union, is to be used for high accuracy measurements of the plutonium content of small samples of nuclear materials. The multi-purpose Monte Carlo N-particle (MCNP) code version 4B has been used to model in detail the neutron coincidence detector and to investigate the leakage self-multiplication of PuO sub 2 and mixed U-Pu oxide (MOX) reference samples used to calibrate the instrument. The MCNP calculations have been used together with a neutron coincidence counting interpretative model to determine characteristic parameters of the detector. A comparative study to both experimental and previous numerical results has been performed. Sensitivity curves of the variation of the detector's efficiency, epsilon, to, alpha, the ratio of (alpha...

  1. MCNP6 fragmentation of light nuclei at intermediate energies

    CERN Document Server

    Mashnik, Stepan G

    2014-01-01

    Fragmentation reactions induced on light target nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the latest Los Alamos Monte Carlo transport code MCNP6 and with its cascade-exciton model (CEM) and Los Alamos version of the quark-gluon string model (LAQGSM) event generators, version 03.03, used as stand-alone codes. Such reactions are involved in different applications, like cosmic-ray-induced single event upsets (SEU's), radiation protection, and cancer therapy with proton and ion beams, among others; therefore, it is important that MCNP6 simulates them as well as possible. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC), followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after INC. Both CEM and LAQGSM account also for coalescence of light fragments (complex particles) up to He4 from energetic nucleons ...

  2. Estimation and interpretation of k{sub eff} confidence intervals in MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Urbatsch, T.J. [Michigan Univ., Ann Arbor, MI (United States). Dept. of Nuclear Engineering; Forster, R.A.; Prael, R.E.; Beckman, R.J. [Los Alamos National Lab., NM (United States)

    1995-07-01

    MCNP has three different, but correlated, estimators for Calculating k{sub eff} in nuclear criticality calculations: collision, absorption, and track length estimators. The combination of these three estimators, the three-combined k{sub eff} estimator, is shown to be the best k{sub eff} estimator available in MCNP for estimating k{sub eff} confidence intervals. Theoretically, the Gauss-Markov Theorem provides a solid foundation for MCNP`s three-combined estimator. Analytically, a statistical study, where the estimates are drawn using a known covariance matrix, shows that the three-combined estimator is superior to the individual estimator with the smallest variance. The importance of MCNP`s batch statistics is demonstrated by an investigation of the effects of individual estimator variance bias on the combination of estimators, both heuristically with the analytical study and emprically with MCNP.

  3. Testing the Delayed Gamma Capability in MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Weldon, Robert A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fensin, Michael L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); McKinney, Gregg W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-10-28

    . We examine five different decay chains (two-stage decay to stable) and show the predictability of the MCNP6 delayed gamma feature. Results do show that while the default delayed gamma calculations available in the MCNP6 1.0 release can give accurate results for some isotopes (e.g., 137Ba), the percent differences between the closed form analytic solutions and the MCNP6 calculations were often >40% (28Mg, 28Al, 42K, 47Ca, 47Sc, 60Co). With the MCNP6 1.1 Beta release, the tenth entry on the DBCN card allows improved calculation within <5% as compared to the closed form analytic solutions for immediate parent emissions and transient equilibrium systems. While the tenth entry on the DBCN card for MCNP6 1.1 gives much better results for transient equilibrium systems and parent emissions in general, it does little to improve daughter emissions of secular equilibrium systems. Finally, hypotheses were presented as to why daughter emissions of secular equilibrium systems might be mispredicted in some cases and not in others.

  4. Voxel2MCNP: a framework for modeling, simulation and evaluation of radiation transport scenarios for Monte Carlo codes.

    Science.gov (United States)

    Pölz, Stefan; Laubersheimer, Sven; Eberhardt, Jakob S; Harrendorf, Marco A; Keck, Thomas; Benzler, Andreas; Breustedt, Bastian

    2013-08-21

    The basic idea of Voxel2MCNP is to provide a framework supporting users in modeling radiation transport scenarios using voxel phantoms and other geometric models, generating corresponding input for the Monte Carlo code MCNPX, and evaluating simulation output. Applications at Karlsruhe Institute of Technology are primarily whole and partial body counter calibration and calculation of dose conversion coefficients. A new generic data model describing data related to radiation transport, including phantom and detector geometries and their properties, sources, tallies and materials, has been developed. It is modular and generally independent of the targeted Monte Carlo code. The data model has been implemented as an XML-based file format to facilitate data exchange, and integrated with Voxel2MCNP to provide a common interface for modeling, visualization, and evaluation of data. Also, extensions to allow compatibility with several file formats, such as ENSDF for nuclear structure properties and radioactive decay data, SimpleGeo for solid geometry modeling, ImageJ for voxel lattices, and MCNPX's MCTAL for simulation results have been added. The framework is presented and discussed in this paper and example workflows for body counter calibration and calculation of dose conversion coefficients is given to illustrate its application.

  5. A Validated MCNP(X) Cross Section Library based on JEFF 3.1

    Energy Technology Data Exchange (ETDEWEB)

    Haeck, W.; Verboomen, B.

    2006-10-15

    ALEPH-LIB is a multi-temperature neutron transport library for standard use by MCNP(X) and ALEPH generated with ALEPH-DLG. This is an auxiliary computer code to ALEPH, the Monte Carlo burn-up code under development at SCK-CEN in collaboration with Ghent university. ALEPH-DLG automates the entire process of generating library files with NJOY and takes care of the first requirement of a validated application library: verify the processing. It produces tailor made NJOY input files using data from the original ENDF file (initial temperature, the fact if the nuclide is fissile or if it has unresolved resonances, etc.) When the library files have been generated, ALEPH-DLG will also process the output from NJOY by extracting all messages and warnings. If ALEPH-DLG finds anything out of the ordinary, it will either warn the user or perform corrective actions. The temperatures included in the ALEPH-LIB library are 300, 600, 900, 1200, 1500 and 1800 K. Library files were produced for the JEF 2.2, JEFF 3.0, JEFF 3.1, JENDL 3.3 and ENDF/B-VI.8 nuclear data libraries. This will be extended with ENDF/B-VII when it becomes available. This report deals with the JEFF 3.1 files included in ALEPH-LIB that are now released by the NEA-OECD.

  6. MCNP(TM) Release 6.1.1 beta: Creating and Testing the Code Distribution

    Energy Technology Data Exchange (ETDEWEB)

    Cox, Lawrence J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Casswell, Laura [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-06-12

    This report documents the preparations for and testing of the production release of MCNP6™1.1 beta through RSICC at ORNL. It addresses tests on supported operating systems (Linux, MacOSX, Windows) with the supported compilers (Intel, Portland Group and gfortran). Verification and Validation test results are documented elsewhere. This report does not address in detail the overall packaging of the distribution. Specifically, it does not address the nuclear and atomic data collection, the other included software packages (MCNP5, MCNPX and MCNP6) and the collection of reference documents.

  7. Developing an interface between MCNP and McStas for simulation of neutron moderators

    DEFF Research Database (Denmark)

    Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik;

    2012-01-01

    Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using MCNP/X whereas simulations of neutron transport and instrument performance are carried out by neutron ray tracing codes such as McStas. The coupling between the two simulations suites...... typically consists of providing analytical fits from MCNP/X neutron spectra to McStas. This method is generally successful, but as will be discussed in the this paper, there are limitations and a more direct coupling between MCNP/X andMcStas could allow for more accurate simulations of e.g. complex...... moderator geometries, interference between beamlines as well as shielding requirements along the neutron guides. In this paper different possible interfaces between McStas and MCNP/X are discussed and first preliminary performance results are shown....

  8. Current status of MCNP6 as a simulation tool useful for space and accelerator applications

    CERN Document Server

    Mashnik, S G; Hughes, H G; Prael, R E; Sierk, A J

    2012-01-01

    For the past several years, a major effort has been undertaken at Los Alamos National Laboratory (LANL) to develop the transport code MCNP6, the latest LANL Monte-Carlo transport code representing a merger and improvement of MCNP5 and MCNPX. We emphasize a description of the latest developments of MCNP6 at higher energies to improve its reliability in calculating rare-isotope production, high-energy cumulative particle production, and a gamut of reactions important for space-radiation shielding, cosmic-ray propagation, and accelerator applications. We present several examples of validation and verification of MCNP6 compared to a wide variety of intermediate- and high-energy experimental data on reactions induced by photons, mesons, nucleons, and nuclei at energies from tens of MeV to about 1 TeV/nucleon, and compare to results from other modern simulation tools.

  9. Dosimetric characterization of a brachytherapy applicator using MCNP5 modelisation and in-phantom measurements.

    Science.gov (United States)

    Gerardy, I; Ródenas, J; van Dycke, M; Gallardo, S; Ceccolini, Elisa

    2010-01-01

    A gynaecological applicator consisting of a metallic intra-uterine tube with a plastic vaginal applicator and an HDR Ir-192 source have been simulated with MCNP5 (Monte Carlo code). A solid phantom has been designed to perform measurements around the applicator with radiochromic films. The isodose curves obtained are compared with curves calculated with the F4MESH tally of MCNP5 with a good agreement. A pinpoint ionization chamber has been used to evaluate dose at some reference points.

  10. Developing an interface between MCNP and McStas for simulation of neutron moderators

    OpenAIRE

    Klinkby, Esben Bryndt; Lauritzen, Bent; Nonbøl, Erik; Willendrup, Peter Kjær

    2012-01-01

    Simulations of target-moderator-reflector system at spallation sources are conventionally carried out using MCNP/X whereas simulations of neutron transport and instrument performance are carried out by neutron ray tracing codes such as McStas. The coupling between the two simulations suites typically consists of providing analytical fits from MCNP/X neutron spectra to McStas. This method is generally successful, but as will be discussed in the this paper, there are limitations and a more dire...

  11. Report on Thermal Neutron Diffusion Length Measurement in Reactor Grade Graphite Using MCNP and COMSOL Multiphysics

    OpenAIRE

    2013-01-01

    Neutron diffusion length in reactor grade graphite is measured both experimentally and theoretically. The experimental work includes Monte Carlo (MC) coding using 'MCNP' and Finite Element Analysis (FEA) coding suing 'COMSOL Multiphysics' and Matlab. The MCNP code is adopted to simulate the thermal neutron diffusion length in a reactor moderator of 2m x 2m with slightly enriched uranium ($^{235}U$), accompanied with a model designed for thermal hydraulic analysis using point kinetic equations...

  12. MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies

    Energy Technology Data Exchange (ETDEWEB)

    Mashnik, Stepan Georgievich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kerby, Leslie Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-05-22

    MCNP6, the latest and most advanced LANL Monte Carlo transport code, representing a merger of MCNP5 and MCNPX, is actually much more than the sum of those two computer codes; MCNP6 is available to the public via RSICC at Oak Ridge, TN, USA. In the present work, MCNP6 was validated and verified (V&V) against different experimental data on intermediate-energy fragmentation reactions, and results by several other codes, using mainly the latest modifications of the Cascade-Exciton Model (CEM) and of the Los Alamos version of the Quark-Gluon String Model (LAQGSM) event generators CEM03.03 and LAQGSM03.03. It was found that MCNP6 using CEM03.03 and LAQGSM03.03 describes well fragmentation reactions induced on light and medium target nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below, and can serve as a reliable simulation tool for different applications, like cosmic-ray-induced single event upsets (SEU’s), radiation protection, and cancer therapy with proton and ion beams, to name just a few. Future improvements of the predicting capabilities of MCNP6 for such reactions are possible, and are discussed in this work.

  13. IER-163 Post-Experiment MCNP Calculations (U)

    Energy Technology Data Exchange (ETDEWEB)

    Favorite, Jeffrey A. [Los Alamos National Laboratory

    2012-06-04

    IER-163 has been modeled with high fidelity in MCNP6. The model k{sub eff} was high, as in other similar calculations. The fission ratio {sup 238}U(n,f)/{sup 235}U(n,f) was 12.6% too small compared with measurements; the ratio {sup 239}Pu(n,f)/{sup 235}U(n,f) was 11.5% too small compared with measurements; the iridium ratio {sup 193}Ir(n,n{prime})/{sup 191}Ir(n,{gamma}) was 16.4% too large; and the gold ratios {sup 197}Au(n,2n)/{sup 197}Au(n,{gamma}), {sup 197}Au(n,2n)/{sup 235}U(n,f), and {sup 197}Au(n,{gamma})/{sup 235}U(n,f) were within one standard deviation of the measured values. It is suggested that the calculated {sup 235}U fission rate is too large and the calculated {sup 238}U fission rate is too small.

  14. MCNP5 study on kinetics parameters of coupled fast-thermal system HERBE

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2011-01-01

    Full Text Available New validation of the well-known Monte Carlo code MCNP5 against measured criticality and kinetics data for the coupled fast-thermal HERBE System at the Reactor B critical assembly is shown in this paper. Results of earlier calculations of these criticality and kinetics parameters, done by combination of transport and diffusion codes using two-dimension geometry model are compared to results of new calculations carried out by the MCNP5 code in three-dimension geometry. Satisfactory agreements in comparison of new results with experimental data, in spite complex heterogeneous composition of the HERBE core, are achieved confirming that MCNP5 code could apply successfully to study on HERBE kinetics parameters after uncertainties in impurities in material compositions and positions of fuel elements in fast zone were removed.

  15. MCNP6 Simulation of Reactions of Interest to FRIB, Medical, and Space Applications

    CERN Document Server

    Mashnik, Stepan G

    2014-01-01

    The latest, production, version of the Los Alamos Monte Carlo N-Particle transport code MCNP6 has been used to simulate a variety of particle-nucleus and nucleus-nucleus reactions of academic and applied interest to the Facility for Rare Isotope Beams (FRIB), medical isotope production, space-radiation shielding, cosmic-ray propagation, and accelerator applications, including several reactions induced by radioactive isotopes, analyzing production of both stable and radioactive residual nuclei. Here, we discuss examples of validation and verification of MCNP6 compared to recent neutron spectra measured at the Heavy Ion Medical Accelerator in Chiba, Japan; to spectra of light fragments from several reactions measured recently at GANIL, France; INFN Laboratori Nazionali del Sud, Catania, Italy; COSY of the Julich Research Center, Germany; and to cross sections of products from several reactions measured lately at GSI, Darmstadt, Germany; ITEP, Moscow, Russia; LANSCE, LANL, Los Alamos, USA. As a rule, MCNP6 provi...

  16. Report on Thermal Neutron Diffusion Length Measurement in Reactor Grade Graphite Using MCNP and COMSOL Multiphysics

    CERN Document Server

    Mirfayzi, S R

    2013-01-01

    Neutron diffusion length in reactor grade graphite is measured both experimentally and theoretically. The experimental work includes Monte Carlo (MC) coding using 'MCNP' and Finite Element Analysis (FEA) coding suing 'COMSOL Multiphysics' and Matlab. The MCNP code is adopted to simulate the thermal neutron diffusion length in a reactor moderator of 2m x 2m with slightly enriched uranium ($^{235}U$), accompanied with a model designed for thermal hydraulic analysis using point kinetic equations, based on partial and ordinary differential equation. The theoretical work includes numerical approximation methods including transcendental technique to illustrate the iteration process with the FEA method. Finally collision density of thermal neutron in graphite is measured, also specific heat relation dependability of collision density is also calculated theoretically, the thermal neutron diffusion length in graphite is evaluated at $50.85 \\pm 0.3cm$ using COMSOL Multiphysics and $50.95 \\pm 0.5cm$ using MCNP. Finally ...

  17. 3D neutronic calculations: CAD-MCNP methodology applied to vessel activation in KOYO-F

    Science.gov (United States)

    Herreras, Y.; Lafuente, A.; Sordo, F.; Cabellos, O.; Perlado, J. M.

    2008-05-01

    This paper presents a methodology for 3D neutronic calculations suitable for complex and extensive geometries. The geometry of the system design is first fully modelled with a CAD program, and subsequently processed through a MCNP-CAD interface in order to generate an MCNP geometry file. Neutronic irradiation results are finally achieved running the MCNPX program, where the geometry input card used is directly the MCNP-CAD interface output. This methodology enables accurate neutronic calculations for complex geometries characterised by high detail levels. This procedure will be applied to the Fast Ignition Fusion Reactor KOYO-F to determine first neutron fluxes calculations along the blanket as well as the material activation in the reduced martensitic 9Cr-1Mo steel vessel.

  18. Simulation of Photon energy Spectra Using MISC, SOURCES, MCNP and GADRAS

    Energy Technology Data Exchange (ETDEWEB)

    Tucker, Lucas P. [Los Alamos National Laboratory; Shores, Erik F. [Los Alamos National Laboratory; Myers, Steven C. [Los Alamos National Laboratory; Felsher, Paul D. [Los Alamos National Laboratory; Garner, Scott E. [Los Alamos National Laboratory; Solomon, Clell J. Jr. [Los Alamos National Laboratory

    2012-08-14

    The detector response functions included in the Gamma Detector Response and Analysis Software (GADRAS) are a valuable resource for simulating radioactive source emission spectra. Application of these response functions to the results of three-dimensional transport calculations is a useful modeling capability. Using a 26.2 kg shell of depleted uranium (DU) as a simple test problem, this work illustrates a method for manipulating current tally results from MCNP into the GAM file format necessary for a practical link to GADRAS detector response functions. MISC (MCNP Intrinsic Source Constructor) and SOURCES 4C were used to develop photon and neutron source terms for subsequent MCNP transport, and the resultant spectrum is shown to be in good agreement with that from GADRAS. A 1 kg DU sphere was also modeled with the method described here and showed similarly encouraging results.

  19. 3D neutronic calculations: CAD-MCNP methodology applied to vessel activation in KOYO-F

    Energy Technology Data Exchange (ETDEWEB)

    Herreras, Y; Cabellos, O; Perlado, J M [Instituto de Fusion Nuclear (DENIM)/ETSII/Universidad Politecnica, Madrid (Spain); Lafuente, A; Sordo, F [Universidad Politecnica de Madrid (UPM), Madrid (Spain)], E-mail: yuri@denim.upm.es

    2008-05-15

    This paper presents a methodology for 3D neutronic calculations suitable for complex and extensive geometries. The geometry of the system design is first fully modelled with a CAD program, and subsequently processed through a MCNP-CAD interface in order to generate an MCNP geometry file. Neutronic irradiation results are finally achieved running the MCNPX program, where the geometry input card used is directly the MCNP-CAD interface output. This methodology enables accurate neutronic calculations for complex geometries characterised by high detail levels. This procedure will be applied to the Fast Ignition Fusion Reactor KOYO-F to determine first neutron fluxes calculations along the blanket as well as the material activation in the reduced martensitic 9Cr-1Mo steel vessel.

  20. Implementation of a tree algorithm in MCNP code for nuclear well logging applications.

    Science.gov (United States)

    Li, Fusheng; Han, Xiaogang

    2012-07-01

    The goal of this paper is to develop some modeling capabilities that are missing in the current MCNP code. Those missing capabilities can greatly help for some certain nuclear tools designs, such as a nuclear lithology/mineralogy spectroscopy tool. The new capabilities to be developed in this paper include the following: zone tally, neutron interaction tally, gamma rays index tally and enhanced pulse-height tally. The patched MCNP code also can be used to compute neutron slowing-down length and thermal neutron diffusion length.

  1. Estimation and interpretation of k{sub eff} confidence intervals in MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Urbatsch, T.J. [Univ. of Michigan, Ann Arbor, MI (United States). Dept. of Nuclear Engineering; Forster, R.A.; Prael, R.E.; Beckman, R.J. [Los Alamos National Lab., NM (United States)

    1995-11-01

    MCNP`s criticality methodology and some basic statistics are reviewed. Confidence intervals are discussed, as well as how to build them and their importance in the presentation of a Monte Carlo result. The combination of MCNP`s three k{sub eff} estimators is shown, theoretically and empirically, by statistical studies and examples, to be the best k{sub eff} estimator. The method of combining estimators is based on a solid theoretical foundation, namely, the Gauss-Markov Theorem in regard to the least squares method. The confidence intervals of the combined estimator are also shown to have correct coverage rates for the examples considered.

  2. Comparison and validation of the results of the AZNHEX v.1.0 code with the MCNP code simulating the core of a fast reactor cooled with sodium; Comparacion y validacion de los resultados del codigo AZNHEX v.1.0 con el codigo MCNP simulando el nucleo de un reactor rapido refrigerado con sodio

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, J.; Francois L, J. L.; Bastida O, G. E. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Esquivel E, J., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The development of the AZTLAN platform for the analysis and design of nuclear reactors is led by Instituto Nacional de Investigaciones Nucleares (ININ) and divided into four working groups, which have well-defined activities to achieve significant progress in this project individually and jointly. Within these working groups is the users group, whose main task is to use the codes that make up the AZTLAN platform to provide feedback to the developers, and in this way to make the final versions of the codes are efficient and at the same time reliable and easy to understand. In this paper we present the results provided by the AZNHEX v.1.0 code when simulating the core of a fast reactor cooled with sodium at steady state. The validation of these results is a fundamental part of the platform development and responsibility of the users group, so in this research the results obtained with AZNHEX are compared and analyzed with those provided by the Monte Carlo code MCNP-5, software worldwide used and recognized. A description of the methodology used with MCNP-5 is also presented for the calculation of the interest variables and the difference that is obtained with respect to the calculated with AZNHEX. (Author)

  3. MCNP and other nuclear codes output graphical representation using python scripts; Representacion grafica de outputs de MCNP y codigos nucleares mediante el uso de scripts en python

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.

    2016-08-01

    Due to the lack of graphical representation capability of same nuclear codes like MCNP of GOTHIC, widely used in the industry, the following article describes the development of an interface to use a graphical representation open source (Paraview) with the outputs generated by the nuclear codes. Moreover, this article aims at describing the advantage of this type of visualization programs for the modeling and decision making in the calculation. (Author)

  4. Simulation of a nuclear densimeter using the Monte Carlo MCNP-4C code; Simulacao de um densimetro nuclear utilizando o codigo Monte Carlo MCNP-4C

    Energy Technology Data Exchange (ETDEWEB)

    Penna, Rodrigo [UNI-BH, Belo Horizonte, MG (Brazil). Dept. de Ciencias Biologicas, Ambientais e da Saude (DCBAS/DCET); Silva, Clemente Jose Gusmao Carneiro da [Universidade Estadual de Santa Cruz, UESC, Ilheus, BA (Brazil); Gomes, Paulo Mauricio Costa [Universidade FUMEC, Belo Horizonte, MG (Brazil)

    2008-07-01

    Viability of building a nuclear wood densimeter based on low energy photons Compton scattering was done using Monte Carlo code (MCNP- 4C). It is simulated a collimated 60 keV beam of gamma rays emitted by {sup 241}Am source reaching wood blocks. Backscattered radiation by these blocks was calculated. Photons scattered were correlated with blocks of different wood densities. Results showed a linear relationship on wood density and scattered photons, therefore the viability of this wood densimeter. (author)

  5. MCNP modelling of scintillation-detector gamma-ray spectra from natural radionuclides

    NARCIS (Netherlands)

    Hendriks, Peter; Maucec, M; de Meijer, RJ

    2002-01-01

    gamma-ray spectra of natural radionuclides are simulated for a BGO detector in a borehole geometry using the Monte Carlo code MCNP. All gamma-ray emissions of the decay of K-40 and the series of Th-232 and U-238 are used to describe the source. A procedure is proposed which excludes the time-consumi

  6. Using NJOY to Create MCNP ACE Files and Visualize Nuclear Data

    Energy Technology Data Exchange (ETDEWEB)

    Kahler, Albert Comstock [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-10-14

    We provide lecture materials that describe the input requirements to create various MCNP ACE files (Fast, Thermal, Dosimetry, Photo-nuclear and Photo-atomic) with the NJOY Nuclear Data Processing code system. Input instructions to visualize nuclear data with NJOY are also provided.

  7. Application of MCNP for neutronic calculations at VR-1 training reactor

    Science.gov (United States)

    Huml, Ondřej; Rataj, Jan; Bílý, Tomáš

    2014-06-01

    The paper presents utilization of Monte Carlo MCNP transport code for neutronic calculations of training reactor VR-1. Results of calculations are compared with results of measurements realized during last few critical experiments with various reactor core configurations. Very good agreement between calculations and measurements is observed.

  8. Simulasi MCNP5 dalam Eksperimen Kritikalitas Larutan Plutonium Uranium Nitrat Dengan Reflektor Air dan Polyethelene

    Directory of Open Access Journals (Sweden)

    Dinan A.

    2011-12-01

    Full Text Available Banyak perangkat kritik dibangun untuk memenuhi kebutuhan studi fenomena kecelakaan kritikalitas pada larutan fisil di fasilitas daur bahan bakar nuklir. Salah satu diantaranya adalah perangkat kritik SCAMP. Di perangkat ini dikerjakan eksperimen kritikalitas menggunakan bejana silindris stainless steel berisi larutan plutonium uranium nitrat (Pu ditambah U nitrat. Sebanyak 7 eksperimen didemonstrasikan dengan reflektor air di semua sisi permukaan bejana larutan kecuali di bagian atas bejana. Makalah ini membahas simulasi transport Monte Carlo MCNP5 dalam eksperimen kritikalitas larutan Pu ditambah U nitrat dengan reflektor air dan polyethylene. Simulasi MCNP5 dengan pustaka ENDF/BVI memberikan hasil yang paling dekat dengan data eksperimen terutama pada kasus A untuk varian geometri 4. Dibandingkan pustaka ENDF/BV, perhitungan kritikalitas dengan pustaka ENDF/B-VI memberikan hasil lebih dekat dengan perhitungan MONK dimana bias perhitungannya kurang dari 0,44%, khususnya pada kasus A namun pada kasus B dan C simulasi MCNP5dengan pustaka ENDF/BV memberikan hasil dengan kecenderungan lebih baik dibandingkan pustaka ENDFB/VI dengan bias perhitungan kurang dari 2,67% dan kurang dari 1,13%. Secara keseluruhan dapat disimpulkan bahwa MCNP5 telah menunjukkan reliabilitasnya dalam simulasi kritikalitas larutan Pu ditambah U nitrat.

  9. Certification of MCNP Version 4A for WHC computer platforms. Revision 7

    Energy Technology Data Exchange (ETDEWEB)

    Carter, L.L.

    1995-05-03

    MCNP is a general-purpose Monte Carlo code that can be used for neutron, photon, or coupled neutron/photon transport, including the capability to calculate eigenvalues for critical systems. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces, and some special fourth-degree surfaces (elliptical tori).

  10. MCNP5 CRITICALITY VALIDATION AND BIAS FOR INTERMEDIATE ENRICHED URANIUM SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    FINFROCK SH

    2009-12-10

    The purpose of this analysis is to validate the Monte Carlo N-Particle 5 (MCNP5) code Version 1.40 (LA-UR-03-1987, 2005) and its cross-section database for k-code calculations of intermediate enriched uranium systems on INTEL{reg_sign} processor based PC's running any version of the WINDOWS operating system. Configurations with intermediate enriched uranium were modeled with the moderator range of 39 {le} H/Fissile {le} 1438. See Table 2-1 for brief descriptions of selected cases and Table 3-1 for the range of applicability for this validation. A total of 167 input cases were evaluated including bare and reflected systems in a single body or arrays. The 167 cases were taken directly from the previous (Version 4C [Lan 2005]) validation database. Section 2.0 list data used to calculate k-effective (k{sub eff}) for the 167 experimental criticality benchmark cases using the MCNP5 code v1.40 and its cross section database. Appendix B lists the MCNP cross-section database entries validated for use in evaluating the intermediate enriched uranium systems for criticality safety. The dimensions and atom densities for the intermediate enriched uranium experiments were taken from NEA/NSC/DOC(95)03, September 2005, which will be referred to as the benchmark handbook throughout the report. For these input values, the experimental benchmark k{sub eff} is approximately 1.0. The MCNP validation computer runs ran to an accuracy of approximately {+-} 0.001. For the cases where the reported benchmark k{sub eff} was not equal to 1.0000 the MCNP calculational results were normalized. The difference between the MCNP validation computer runs and the experimentally measured k{sub eff} is the MCNP5 v1.40 bias. The USLSTATS code (ORNL 1998) was utilized to perform the statistical analysis and generate an acceptable maximum k{sub eff} limit for calculations of the intermediate enriched uranium type systems.

  11. An update to the computation of the Goudsmit-Saunderson distribution in MCNP ® version 6.2

    Energy Technology Data Exchange (ETDEWEB)

    Dixon, David A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-10-17

    To demonstrate how the number of terms used to compute the Goudsmit- Saunderson distribution impacts results, we study a thin foil problem where the outgoing angular distribution is observed for 20-MeV electrons. The foil thickness is roughly 1.5 times the default substep size for a 20-MeV electron in gold ( 2.8e-3 cm). Therefore, for the default substep size, electrons are guaranteed to undergo at least one collision before encountering a boundary (at which point an approximation is applied). In theory, one can improve the accuracy of the MCNP6.2 electron transport method by reducing the substep size. However, one must assume that the underlying data is valid. We show that when a user reduces the substep size, the angular distribution observed is different than expected. The primary reason being that the underlying data was not computed using a sufficient number of terms. We also show that the distribution is recovered when the number of terms is increased from hundreds to thousands. Here, we showed that stabilizing the underlying angular deflection distributions used in transport improves simulation results, particularly, when default parameters are adjusted such that the substep size is reduced. Stabilization is achieved by adding more terms when computing the Goudsmit-Sanderson distribution.

  12. Obtaining of primary rays of spectrum X codes Penelope and MCNP5; Obtencion del espectro primario de Rayos X con los codigos Penelope y MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Pozuelo, F.; Querol, A.; Gallardo, S.; Rodenas, J.; Verdu, G.

    2012-07-01

    In this case, used codes PENELOPE MCNP5, based on the Monte Carlo method for x-ray spectrum taking into account the characteristics of the x-ray tube. In order to achieve a greater fit of simulated by the theoretical spectrum. It carried out a sensitivity analysis of the parameters available in both codes. The obtaining of the simulated spectrum could lead to an improvement in quality control of the x-ray tube to incorporate it as a method complementary to techniques.

  13. Graphical User Interface for Simplified Neutron Transport Calculations

    Energy Technology Data Exchange (ETDEWEB)

    Schwarz, Randolph; Carter, Leland L

    2011-07-18

    A number of codes perform simple photon physics calculations. The nuclear industry is lacking in similar tools to perform simplified neutron physics shielding calculations. With the increased importance of performing neutron calculations for homeland security applications and defense nuclear nonproliferation tasks, having an efficient method for performing simple neutron transport calculations becomes increasingly important. Codes such as Monte Carlo N-particle (MCNP) can perform the transport calculations; however, the technical details in setting up, running, and interpreting the required simulations are quite complex and typically go beyond the abilities of most users who need a simple answer to a neutron transport calculation. The work documented in this report resulted in the development of the NucWiz program, which can create an MCNP input file for a set of simple geometries, source, and detector configurations. The user selects source, shield, and tally configurations from a set of pre-defined lists, and the software creates a complete MCNP input file that can be optionally run and the results viewed inside NucWiz.

  14. Coupling MCNP-DSP and LAHET Monte Carlo codes for designing subcriticality monitors for accelerator-driven systems

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.; Perez, R. [Oak Ridge National Lab., TN (United States); Rugama, Y.; Munoz-Cobo, J.L. [Poly. Tech. Univ. of Valencia (Spain). Chemical and Nuclear Engineering Dept.

    2001-07-01

    The design of reactivity monitoring systems for accelerator-driven systems must be investigated to ensure that such systems remain subcritical during operation. The Monte Carlo codes LAHET and MCNP-DSP were combined together to facilitate the design of reactivity monitoring systems. The coupling of LAHET and MCNP-DSP provides a tool that can be used to simulate a variety of subcritical measurements such as the pulsed neutron, Rossi-{alpha}, or noise analysis measurements. (orig.)

  15. Coupling MCNP-DSP and LAHET Monte Carlo Codes for Designing Subcriticality Monitors for Accelerator-Driven Systems

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E.; Rugama, Y. Munoz-Cobos, J.; Perez, R.

    2000-10-23

    The design of reactivity monitoring systems for accelerator-driven systems must be investigated to ensure that such systems remain subcritical during operation. The Monte Carlo codes LAHET and MCNP-DSP were combined together to facilitate the design of reactivity monitoring systems. The coupling of LAHET and MCNP-DSP provides a tool that can be used to simulate a variety of subcritical measurements such as the pulsed neutron, Rossi-{alpha}, or noise analysis measurements.

  16. Implementation of a tree algorithm in MCNP code for nuclear well logging applications

    Energy Technology Data Exchange (ETDEWEB)

    Li Fusheng, E-mail: fusheng.li@bakerhughes.com [Baker Hughes Incorporated, 2001 Rankin Rd. Houston, TX 77073-5101 (United States); Han Xiaogang [Baker Hughes Incorporated, 2001 Rankin Rd. Houston, TX 77073-5101 (United States)

    2012-07-15

    The goal of this paper is to develop some modeling capabilities that are missing in the current MCNP code. Those missing capabilities can greatly help for some certain nuclear tools designs, such as a nuclear lithology/mineralogy spectroscopy tool. The new capabilities to be developed in this paper include the following: zone tally, neutron interaction tally, gamma rays index tally and enhanced pulse-height tally. The patched MCNP code also can be used to compute neutron slowing-down length and thermal neutron diffusion length. - Highlights: Black-Right-Pointing-Pointer Tree structure programming is suitable for Monte-Carlo based particle tracking. Black-Right-Pointing-Pointer Enhanced pulse height tally is developed for oilwell logging tool simulation. Black-Right-Pointing-Pointer Neutron interaction tally and gamma ray index tally for geochemical logging.

  17. MCNP6 Simulation of Light and Medium Nuclei Fragmentation at Intermediate Energies

    Energy Technology Data Exchange (ETDEWEB)

    Mashnik, Stepan Georgievich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kerby, Leslie Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Univ. of Idaho, Moscow, ID (United States)

    2015-08-24

    Fragmentation reactions induced on light and medium nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the Los Alamos transport code MCNP6 and with its CEM03.03 and LAQGSM03.03 event generators. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC), followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after the INC. CEM and LAQGSM account also for coalescence of light fragments (complex particles) up to sup>4He from energetic nucleons emitted during INC. We investigate the validity and performance of MCNP6, CEM, and LAQGSM in simulating fragmentation reactions at intermediate energies and discuss possible ways of further improving these codes.

  18. MCNP6 simulation of light and medium nuclei fragmentation at intermediate energies

    Directory of Open Access Journals (Sweden)

    Mashnik Stepan G.

    2016-01-01

    Full Text Available Fragmentation reactions induced on light and medium nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the Los Alamos transport code MCNP6 and with its CEM03.03 and LAQGSM03.03 event generators. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC, followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after the INC. CEM and LAQGSM account also for coalescence of light fragments (complex particles up to 4He from energetic nucleons emitted during INC. We investigate the validity and performance of MCNP6, CEM, and LAQGSM in simulating fragmentation reactions at intermediate energies and discuss possible ways of further improving these codes.

  19. MCNP6 simulation of light and medium nuclei fragmentation at intermediate energies

    CERN Document Server

    Mashnik, Stepan G

    2015-01-01

    Fragmentation reactions induced on light and medium nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the Los Alamos transport code MCNP6 and with its CEM03.03 and LAQGSM03.03 event generators. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC), followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after the INC. CEM and LAQGSM account also for coalescence of light fragments (complex particles) up to 4He from energetic nucleons emitted during INC. We investigate the validity and performance of MCNP6, CEM, and LAQGSM in simulating fragmentation reactions at intermediate energies and discuss possible ways of further improving these codes

  20. Use experiences of MCNP in nuclear energy study. 2. Review of variance reduction techniques

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Kiyoshi; Yamamoto, Toshihiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [eds.

    1998-03-01

    `MCNP Use Experience` Working Group was established in 1996 under the Special Committee on Nuclear Code Evaluation. This year`s main activity of the working group has been focused on the review of variance reduction techniques of Monte Carlo calculations. This working group dealt with the variance reduction techniques of (1) neutron and gamma ray transport calculation of fusion reactor system, (2) concept design of nuclear transmutation system using accelerator, (3) JMTR core calculation, (4) calculation of prompt neutron decay constant, (5) neutron and gamma ray transport calculation for exposure evaluation, (6) neutron and gamma ray transport calculation of shielding system, etc. Furthermore, this working group started an activity to compile `Guideline of Monte Carlo Calculation` which will be a standard in the future. The appendices of this report include this `Guideline`, the use experience of MCNP 4B and examples of Monte Carlo calculations of high energy charged particles. The 11 papers are indexed individually. (J.P.N.)

  1. MCNP6 simulation of light and medium nuclei fragmentation at intermediate energies

    Science.gov (United States)

    Mashnik, Stepan G.; Kerby, Leslie M.

    2016-05-01

    Fragmentation reactions induced on light and medium nuclei by protons and light nuclei of energies around 1 GeV/nucleon and below are studied with the Los Alamos transport code MCNP6 and with its CEM03.03 and LAQGSM03.03 event generators. CEM and LAQGSM assume that intermediate-energy fragmentation reactions on light nuclei occur generally in two stages. The first stage is the intranuclear cascade (INC), followed by the second, Fermi breakup disintegration of light excited residual nuclei produced after the INC. CEM and LAQGSM account also for coalescence of light fragments (complex particles) up to 4He from energetic nucleons emitted during INC. We investigate the validity and performance of MCNP6, CEM, and LAQGSM in simulating fragmentation reactions at intermediate energies and discuss possible ways of further improving these codes.

  2. Establishment and Verification of MCNP Neutron Transport Model About Tianwan Nuclear Power Plant

    Institute of Scientific and Technical Information of China (English)

    ZHOU; Qi

    2012-01-01

    <正>In order to calculating the neutron flux in the surveillance box and reactor pressure vessel of the Tianwan NPP, we need to build up the neutron transport model by using the Monte Carlo code MCNP. The core of the NPP is very complicated for modeling so we put forward some assumptions that can simplify the neutron transport model. A lot of calculation works have been done to prove that the assumptions are right and suitable.

  3. VVER-440 Ex-Core Neutron Transport Calculations by MCNP-5 Code and Comparison with Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Borodkin, Pavel; Khrennikov, Nikolay [Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS) Malaya Krasnoselskaya ul., 2/8, bld. 5, 107140 Moscow (Russian Federation)

    2008-07-01

    Ex-core neutron transport calculations are needed to evaluate radiation loading parameters (neutron fluence, fluence rate and spectra) on the in-vessel equipment, reactor pressure vessel (RPV) and support constructions of VVER type reactors. Due to these parameters are used for reactor equipment life-time assessment, neutron transport calculations should be carried out by precise and reliable calculation methods. In case of RPVs, especially, of first generation VVER-440s, the neutron fluence plays a key role in the prediction of RPV lifetime. Main part of VVER ex-core neutron transport calculations are performed by deterministic and Monte-Carlo methods. This paper deals with precise calculations of the Russian first generation VVER-440 by MCNP-5 code. The purpose of this work was an application of this code for expert calculations, verification of results by comparison with deterministic calculations and validation by neutron activation measured data. Deterministic discrete ordinates DORT code, widely used for RPV neutron dosimetry and many times tested by experiments, was used for comparison analyses. Ex-vessel neutron activation measurements at the VVER-440 NPP have provided space (in azimuth and height directions) and neutron energy (different activation reactions) distributions data for experimental (E) validation of calculated results. Calculational intercomparison (DORT vs. MCNP-5) and comparison with measured values (MCNP-5 and DORT vs. E) have shown agreement within 10-15% for different space points and reaction rates. The paper submits a discussion of results and makes conclusions about practice use of MCNP-5 code for ex-core neutron transport calculations in expert analysis. (authors)

  4. UW MCNP source patch for the EPFL Haefely source. EPFL (Swiss) fusion-fission hybrid experiment

    Energy Technology Data Exchange (ETDEWEB)

    McKinney, G; Woodruff, G L

    1986-06-01

    The development of a source patch which describes the Haefely neutron source for use in the MCNP Monte Carlo code has been described in progress reports of the EPFL (Swiss) Fusion Blanket Project at the University of Washington. The most recent of these reports dealing with the source patch was Progress Report No. 14. This report reviews some of the physical description included in the report, and also includes additional details of the patch as well as a listing of the patch itself.

  5. Development of Multi-physics (Multiphase CFD + MCNP) simulation for generic solution vessel power calculation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Jun [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Buechler, Cynthia Eileen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-07-17

    The current study aims to predict the steady state power of a generic solution vessel and to develop a corresponding heat transfer coefficient correlation for a Moly99 production facility by conducting a fully coupled multi-physics simulation. A prediction of steady state power for the current application is inherently interconnected between thermal hydraulic characteristics (i.e. Multiphase computational fluid dynamics solved by ANSYS-Fluent 17.2) and the corresponding neutronic behavior (i.e. particle transport solved by MCNP6.2) in the solution vessel. Thus, the development of a coupling methodology is vital to understand the system behavior at a variety of system design and postulated operating scenarios. In this study, we report on the k-effective (keff) calculation for the baseline solution vessel configuration with a selected solution concentration using MCNP K-code modeling. The associated correlation of thermal properties (e.g. density, viscosity, thermal conductivity, specific heat) at the selected solution concentration are developed based on existing experimental measurements in the open literature. The numerical coupling methodology between multiphase CFD and MCNP is successfully demonstrated, and the detailed coupling procedure is documented. In addition, improved coupling methods capturing realistic physics in the solution vessel thermal-neutronic dynamics are proposed and tested further (i.e. dynamic height adjustment, mull-cell approach). As a key outcome of the current study, a multi-physics coupling methodology between MCFD and MCNP is demonstrated and tested for four different operating conditions. Those different operating conditions are determined based on the neutron source strength at a fixed geometry condition. The steady state powers for the generic solution vessel at various operating conditions are reported, and a generalized correlation of the heat transfer coefficient for the current application is discussed. The assessment of multi

  6. Assessment of CANDU reactor physics effects using a simplified whole-core MCNP model

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    2002-07-01

    A whole-core Monte Carlo n-particle (MCNP) model of a simplified CANDU reactor was developed and used to study core configurations and reactor physics phenomena of interest in CANDU safety analysis. The resulting reactivity data were compared with values derived from corresponding WIMS-AECL/RFSP, two-neutron-energy-group diffusion theory core simulations, thereby extending the range of CANDU-related code-to-code benchmark comparisons to include whole-core representations. These comparisons show a systematic discrepancy of about 6 mk between the respective absolute k{sub eff} values, but very good agreement to within about -0.15 {+-} 0.06 mk for the reactivity perturbation induced by G-core checkerboard coolant voiding. These findings are generally consistent with the results of much simpler uniform-lattice comparisons involving only WIMS-AECL and MCNP. In addition, MCNP fission-energy tallies were used to evaluate other core-wide properties, such as fuel bundle and total-channel power distributions, as well as intra-bundle details, such as outer-fuel-ring relative power densities and outer-ring fuel element azimuthal power variations, which cannot be determined directly from WIMS-AECL/RFSP core calculations. The average MCNP values for the ratio of outer fuel element to average fuel element power density agreed well with corresponding values derived from WIMS-AECL lattice-cell cases, showing a small systematic discrepancy of about 0.5 %, independent of fuel bum-up. For fuel bundles containing the highest-power fuel elements, the maximum peak-to-average outer-element azimuthal power variation was about 2.5% for cases where a statistically significant trend was observed, while much larger peak-to-average outer-element azimuthal power variations of up to around 42% were observed in low-power fuel bundles at the core/radial-neutron-reflector interface. (author)

  7. Development of an MCNP-tally based burnup code and validation through PWR benchmark exercises

    Energy Technology Data Exchange (ETDEWEB)

    El Bakkari, B. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco)], E-mail: bakkari@gmail.com; El Bardouni, T.; Merroun, O.; El Younoussi, Ch.; Boulaich, Y. [ERSN-LMR, Department of physics, Faculty of Sciences P.O.Box 2121, Tetuan (Morocco); Chakir, E. [EPTN-LPMR, Faculty of Sciences Kenitra (Morocco)

    2009-05-15

    The aim of this study is to evaluate the capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5, through the use of the MCNP tally information. BUCAL1 uses the fourth order Runge Kutta method with the predictor-corrector approach as the integration method to determine the fuel composition at a desired burnup step. Validation of BUCAL1 was done by code vs. code comparison. Results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multi-group two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. Eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300 K) and hot (900 K) conditions, respectively. The eigenvalue comparison between BUCAL1 and the aforementioned two kinds of codes shows a good prediction of the systems'k-inf values during the entire burnup history, and the maximum difference is within 2%. The differences between the BUCAL1 isotope concentrations and the predictions of CASMO-4, MCODE and MOCUP are generally better, and only for a few sets of isotopes these differences exceed 10%.

  8. Current status of ACE format libraries for MCNP at nuclear date center of KAERI

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Gil, Choong Sup; Lee, Young Ouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-09-15

    The current status of ACE format MCNP/MCNPX libraries by NDC of KAERI is presented with a short description of each library. Validation calculations with recent nuclear data evaluations ENDF/B-VII.0, ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0 have been carried out by the MCNP5 code for 119 criticality benchmark problems taken from the expanded criticality validation suite supplied by LANL. The overall performances of the ACE format KN-libraries have been analyzed in comparison with the results calculated with the ENDF/B-VII.0-based ENDF70 library of LANL. It was confirmed that the ENDF/B-VII.1-based KNE71 library showed better performances than the others by comparing the RMS errors and χ2 values for five benchmark categories as well as whole benchmark problems. ENDF/B-VII.1 and JEFF-3.2 have a tendency to yield more reliable MCNP calculation results within certain confidence intervals regarding the total uncertainties for the keff values. It is found that the adoption of the latest evaluated nuclear data might ensure better outcomes in various research and development areas.

  9. A simulation of a pebble bed reactor core by the MCNP-4C computer code

    Directory of Open Access Journals (Sweden)

    Bakhshayesh Moshkbar Khalil

    2009-01-01

    Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

  10. V&V of MCNP 6.1.1 Beta Against Intermediate and High-Energy Experimental Data

    Energy Technology Data Exchange (ETDEWEB)

    Mashnik, Stepan G [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-09-08

    This report presents a set of validation and verification (V&V) MCNP 6.1.1 beta results calculated in parallel, with MPI, obtained using its event generators at intermediate and high-energies compared against various experimental data. It also contains several examples of results using the models at energies below 150 MeV, down to 10 MeV, where data libraries are normally used. This report can be considered as the forth part of a set of MCNP6 Testing Primers, after its first, LA-UR-11-05129, and second, LA-UR-11-05627, and third, LA-UR-26944, publications, but is devoted to V&V with the latest, 1.1 beta version of MCNP6. The MCNP6 test-problems discussed here are presented in the /VALIDATION_CEM/and/VALIDATION_LAQGSM/subdirectories in the MCNP6/Testing/directory. README files that contain short descriptions of every input file, the experiment, the quantity of interest that the experiment measures and its description in the MCNP6 output files, and the publication reference of that experiment are presented for every test problem. Templates for plotting the corresponding results with xmgrace as well as pdf files with figures representing the final results of our V&V efforts are presented. Several technical “bugs” in MCNP 6.1.1 beta were discovered during our current V&V of MCNP6 while running it in parallel with MPI using its event generators. These “bugs” are to be fixed in the following version of MCNP6. Our results show that MCNP 6.1.1 beta using its CEM03.03, LAQGSM03.03, Bertini, and INCL+ABLA, event generators describes, as a rule, reasonably well different intermediate- and high-energy measured data. This primer isn’t meant to be read from cover to cover. Readers may skip some sections and go directly to any test problem in which they are interested.

  11. MCNP5 and GEANT4 comparisons for preliminary Fast Neutron Pencil Beam design at the University of Utah TRIGA system

    Science.gov (United States)

    Adjei, Christian Amevi

    The main objective of this thesis is twofold. The starting objective was to develop a model for meaningful benchmarking of different versions of GEANT4 against an experimental set-up and MCNP5 pertaining to photon transport and interactions. The following objective was to develop a preliminary design of a Fast Neutron Pencil Beam (FNPB) Facility to be applicable for the University of Utah research reactor (UUTR) using MCNP5 and GEANT4. The three various GEANT4 code versions, GEANT4.9.4, GEANT4.9.3, and GEANT4.9.2, were compared to MCNP5 and the experimental measurements of gamma attenuation in air. The average gamma dose rate was measured in the laboratory experiment at various distances from a shielded cesium source using a Ludlum model 19 portable NaI detector. As it was expected, the gamma dose rate decreased with distance. All three GEANT4 code versions agreed well with both the experimental data and the MCNP5 simulation. Additionally, a simple GEANT4 and MCNP5 model was developed to compare the code agreements for neutron interactions in various materials. Preliminary FNPB design was developed using MCNP5; a semi-accurate model was developed using GEANT4 (because GEANT4 does not support the reactor physics modeling, the reactor was represented as a surface neutron source, thus a semi-accurate model). Based on the MCNP5 model, the fast neutron flux in a sample holder of the FNPB is obtained to be 6.52×107 n/cm2s, which is one order of magnitude lower than gigantic fast neutron pencil beam facilities existing elsewhere. The MCNP5 model-based neutron spectrum indicates that the maximum expected fast neutron flux is at a neutron energy of ~1 MeV. In addition, the MCNP5 model provided information on gamma flux to be expected in this preliminary FNPB design; specifically, in the sample holder, the gamma flux is to be expected to be around 108 γ/cm 2s, delivering a gamma dose of 4.54×103 rem/hr. This value is one to two orders of magnitudes below the gamma

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

    Science.gov (United States)

    2014-03-27

    want to express my sincere love, respect, and admiration for my wife, who motivated and supported me throughout this long endeavor; this document ...widely utilized radiation transport code is MCNP. First created at Los Alamos National Laboratory ( LANL ) in 1957, the code simulated neutral...explanation of the current capabilities of MCNP will occur within the next chapter of this document ; however, it is important to note that MCNP

  13. Accuracy of the electron transport in mcnp5 and its suitability for ionization chamber response simulations: A comparison with the egsnrc and penelope codes

    Energy Technology Data Exchange (ETDEWEB)

    Koivunoro, Hanna; Siiskonen, Teemu; Kotiluoto, Petri; Auterinen, Iiro; Hippelaeinen, Eero; Savolainen, Sauli [Department of Physics, University of Helsinki, P.O. Box 64, FI-00014 Helsinki University (Finland) and Department of Oncology, Helsinki University Central Hospital, FI-00029 HUS (Finland); STUK-Radiation and Nuclear Safety Authority, P.O. Box 14, FI-00881 Helsinki (Finland); VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland); Department of Physics, University of Helsinki, P.O. Box 64, FI-00014 Helsinki University (Finland); HUS Medical Imaging Centre, Helsinki University Central Hospital, FI-00029 HUS (Finland)

    2012-03-15

    Purpose: In this work, accuracy of the mcnp5 code in the electron transport calculations and its suitability for ionization chamber (IC) response simulations in photon beams are studied in comparison to egsnrc and penelope codes. Methods: The electron transport is studied by comparing the depth dose distributions in a water phantom subdivided into thin layers using incident energies (0.05, 0.1, 1, and 10 MeV) for the broad parallel electron beams. The IC response simulations are studied in water phantom in three dosimetric gas materials (air, argon, and methane based tissue equivalent gas) for photon beams ({sup 60}Co source, 6 MV linear medical accelerator, and mono-energetic 2 MeV photon source). Two optional electron transport models of mcnp5 are evaluated: the ITS-based electron energy indexing (mcnp5{sub ITS}) and the new detailed electron energy-loss straggling logic (mcnp5{sub new}). The electron substep length (ESTEP parameter) dependency in mcnp5 is investigated as well. Results: For the electron beam studies, large discrepancies (>3%) are observed between the mcnp5 dose distributions and the reference codes at 1 MeV and lower energies. The discrepancy is especially notable for 0.1 and 0.05 MeV electron beams. The boundary crossing artifacts, which are well known for the mcnp5{sub ITS}, are observed for the mcnp5{sub new} only at 0.1 and 0.05 MeV beam energies. If the excessive boundary crossing is eliminated by using single scoring cells, the mcnp5{sub ITS} provides dose distributions that agree better with the reference codes than mcnp5{sub new}. The mcnp5 dose estimates for the gas cavity agree within 1% with the reference codes, if the mcnp5{sub ITS} is applied or electron substep length is set adequately for the gas in the cavity using the mcnp5{sub new}. The mcnp5{sub new} results are found highly dependent on the chosen electron substep length and might lead up to 15% underestimation of the absorbed dose. Conclusions: Since the mcnp5 electron

  14. Simulation of radiation transport using MCNP for a teletherapy machine; Simulacion del transporte de radiacion usando MCNP para una maquina de teleterapia

    Energy Technology Data Exchange (ETDEWEB)

    Flores O, F.E.; Mireles G, F.; Davila R, J.I.; Pinedo V, J.L.; Risorios M, C.; Lopez del Rio, H. [UAZ, Unidad Academica de Estudios Nucleares, 98068 Zacatecas (Mexico)

    2008-07-01

    The MCNP code is used to simulate the radiation transport taking as tools the transport physics of each particle, either photon, neutron or electron, and the generation of random numbers. Developed in the Los Alamos National Laboratory, this code has been used thoroughly with great success, because the results of the simulations are broadly validated with representative experiments. In the one present work the room of radiotherapy of the Institute Zacatecano of the Tumor it is simulated, located in the city of Zacatecas where one is Theratron 780C machine manufactured by MSD Nordion, with the purpose of estimating the contribution to the dose that would be received in different points of the structure, included three directly under the source. Three results of analytical calculations for points located at different distances from the source are presented, and they are compared against those obtained by the simulation. Its are also presented results for the simulation of 10 points more distributed around the source. (Author)

  15. User 2020

    DEFF Research Database (Denmark)

    Porras, Jari; Heikkinen, Kari; Kinnula, Marianne

    2014-01-01

    and environment, and each has had its effect on the development of technology. The closer we come to the current generation, the bigger is the effect of technology on the characteristics of that generation. User needs guide the technology and the technology shapes the users. This WWRF Outlook analyses......The User 2020 vision is of the changing needs and habits of a user in the future digital world. In order to understand the needs of the future users, we need to look at how users and technology have changed during recent years. The different generations of users are products of their own time...... determined by the era in which they were born. This is due to the fact that digital natives, born in an already “fully” digitalized world with a plethora of ICT services, have a much closer relationship to these solutions than generations before them. This has also shaped the users perspectives and had...

  16. Understanding users

    DEFF Research Database (Denmark)

    Johannsen, Carl Gustav Viggo

    2014-01-01

    Segmentation of users can help libraries in the process of understanding user similarities and differences. Segmentation can also form the basis for selecting segments of target users and for developing tailored services for specific target segments. Several approaches and techniques have been...... segmentation project using computer-generated clusters. Compared to traditional marketing texts, this article also tries to identify user segments or images or metaphors by the library profession itself....

  17. Understanding users

    DEFF Research Database (Denmark)

    Johannsen, Carl Gustav Viggo

    2014-01-01

    Segmentation of users can help libraries in the process of understanding user similarities and differences. Segmentation can also form the basis for selecting segments of target users and for developing tailored services for specific target segments. Several approaches and techniques have been te...... segmentation project using computer-generated clusters. Compared to traditional marketing texts, this article also tries to identify user segments or images or metaphors by the library profession itself....

  18. Simulations of neutron multiplicity measurements of a weapons-grade plutonium sphere with MCNP-PoliMi.

    Energy Technology Data Exchange (ETDEWEB)

    Mattingly, John K.; Pozzi, Sara A. (University of Michigan, Ann Arbor, MI); Clarke, Shaun D. (University of Michigan, Ann Arbor, MI); Dennis, Ben D. (University of Michigan, Ann Arbor, MI); Miller, Eric C. (University of Michigan, Ann Arbor, MI); Padovani, E. (Polytechnic of Milan, Italy)

    2010-06-01

    With increasing concern over the ability to detect and characterize special nuclear materials, the need for computer codes that can successfully predict the response of detector systems to various measurement scenarios is extremely important. These computer algorithms need to be benchmarked against a variety of experimental configurations to ensure their accuracy and understand their limitations. The Monte Carlo code MCNP-PoliMi is a modified version of the MCNP-4c code. Recently these modifications have been ported into the new MCNPX 2.6.0 code, which gives the new MCNPX-PoliMi a wider variety of options and abilities, taking advantage of the improvements made to MCNPX. To verify the ability of the MCNPX-PoliMi code to simulate the response of a neutron multiplicity detector simulated results were compared to experimental data. The experiment consisted of a 4.5-kg sphere of alpha-phase plutonium that was moderated with various thicknesses of polyethylene. The results showed that our code system can simulate the multiplicity distributions with relatively good agreement with measured data. The enhancements made to MCNP since the release of MCNP-4c have had little to no effect on the ability of the MCNP-PoliMi to resolve the discrepancies observed in the simulated neutron multiplicity distributions when compared experimental data.

  19. Validation and Verification of MCNP6 Against Intermediate and High-Energy Experimental Data and Results by Other Codes

    CERN Document Server

    Mashnik, Stepan G

    2010-01-01

    MCNP6, the latest and most advanced LANL transport code representing a recent merger of MCNP5 and MCNPX, has been Validated and Verified (V&V) against a variety of intermediate and high-energy experimental data and against results by different versions of MCNPX and other codes. In the present work, we V&V MCNP6 using mainly the latest modifications of the Cascade-Exciton Model (CEM) and of the Los Alamos version of the Quark-Gluon String Model (LAQGSM) event generators CEM03.02 and LAQGSM03.03. We found that MCNP6 describes reasonably well various reactions induced by particles and nuclei at incident energies from 18 MeV to about 1 TeV per nucleon measured on thin and thick targets and agrees very well with similar results obtained with MCNPX and calculations by CEM03.02, LAQGSM03.01 (03.03), INCL4 + ABLA, and Bertini INC + Dresner evaporation, EPAX, ABRABLA, HIPSE, and AMD, used as stand alone codes. Most of several computational bugs and more serious physics problems observed in MCNP6/X during our V...

  20. MCNP/MCNPX几何栅元划分方法对精确放疗剂量计算的影响研究%Effect of Different Voxel-uniting Methods on the Dose Calculation of MCNP/MCNPX

    Institute of Scientific and Technical Information of China (English)

    赵攀; 陈义学; 林辉; 郑善良; 吴宜灿

    2006-01-01

    复杂几何模型的建立是Monte Carlo粒子输运程序MCNP/MCNPX在放疗领域广泛应用的关键与难点, 发展了基于医学CT影像的MCNP/MCNPX自动建模软件, 提出并实现了3种几何柵元划分的方法.根据临床实例数据, 分别建立了3种MCNP几何模型. 在此基础上, 研究分析了3种几何柵元划分方法及重复结构描述方法对计算结果的影响, 为MCNP/MCNPX在放疗中的应用提供基础.

  1. New Tools to Prepare ACE Cross-section Files for MCNP Analytic Test Problems

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Monte Carlo Codes Group

    2016-06-17

    Monte Carlo calculations using one-group cross sections, multigroup cross sections, or simple continuous energy cross sections are often used to: (1) verify production codes against known analytical solutions, (2) verify new methods and algorithms that do not involve detailed collision physics, (3) compare Monte Carlo calculation methods with deterministic methods, and (4) teach fundamentals to students. In this work we describe 2 new tools for preparing the ACE cross-section files to be used by MCNP® for these analytic test problems, simple_ace.pl and simple_ace_mg.pl.

  2. ANALISIS LAJU DOSIS NEUTRON REAKTOR PLTN PWR 1000 MWe MENGGUNAKAN PROGRAM MCNP

    Directory of Open Access Journals (Sweden)

    Amir Hamzah

    2015-03-01

    Full Text Available Dalam rangka menyongsong PLTN pertama di Indonesia, dilakukan kajian dan analisis berbagai aspek teknologi reaktor tersebut. Tujuan dari penelitian ini adalah menentukan laju dosis neutron di luar perisai biologik reaktor PLTN PWR 1000 MWe yang merupakan bagian dari kegiatan besar di atas. Data hasil analisis laju dosis radiasi pada posisi tertentu sangat dibutuhkan untuk menunjukkan tingkat paparan radiasi di posisi tersebut. Analisis laju dosis neutron ditentukan berdasarkan hasil analisis fluks dan spektrum neutron. Analisis fluks dan spektrum neutron di teras reaktor daya PWR 1000 Mwe dilakukan menggunakan program MCNP. Model perhitungan yang dilakukan meliputi 9 zona material yaitu, teras, air, selimut, air, tong, air, bejana tekan, beton dan lapisan udara luar. Penentuan distribusi fluks dan spektrum neutron dilakukan ke arah radial hingga di luar perisai beton dengan akurasi antara 10% hingga 30% dalam tiap kelompok energi yang jumlahnya 1 dan 50 kelompok. Hasil analisis laju dosis neutron di permukaan perisai biologik reaktor PLTN PWR 1000 MWe pada kondisi reaktor beroperasi daya penuh sudah di bawah nilai batas keselamatan. Maka dapat disimpulkan bahwa dari segi paparan radiasi neutron, penggunaan perisai radiasi beton setebal dua meter sudah memenuhi persyaratan keselamatan. Kata kunci: PLTN PWR, fluks neutron, perisai, laju dosis neutron, MCNP.   In order to meet the first nuclear power plant in Indonesia, it has been conducted a study and analysis of various aspects of reactor technology. The purpose of this study was to determine the neutron dose rates at the outside of biological shield of NPP PWR 1000 MWe reactor that is a part of the activities described above. The analysis data of radiation dose rate at a specific position is needed to show the level of radiation exposure in those positions. Analysis neutron dose rate is determined based on the results of the analysis of neutron flux. Analysis of flux and neutron spectrum in

  3. Simulations of X-ray spectrum and HVL for mammographic equipment using MCNP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Rafael Toledo F. de; Alvarez, Matheus; Velo, Alexandre F.; Oliveira, Marcela de; Miranda, Jose Ricardo A. [Universidade Estadual Paulista Julio de mesquita Filho (UNESP), Botucatu, SP (Brazil). Inst. de Biociencias de Botucatu. Dept. de Fisica e Biofisica; Pina, Diana R. [Universidade Estadual Paulista Julio de mesquita Filho (UNESP), Botucatu, SP (Brazil). Fac. de Medicina. Dept. de Doencas Tropicais e Diagnostico por Imagem

    2012-07-01

    Full text: The main goal of mammography is early detection of breast cancer. Thus, the mammograph should be designed so that the X-ray photons are emitted within an appropriate energy range, to distinguish the normal breast tissue and cancerous tissue. The distribution of the photons amount of X-ray beam, with their respective energies, is called the spectrum. From the spectrum it is possible to estimate the quality of the X-ray beam from the Half Value Layer (HVL). Objectives: This study aims to simulate the Senographe 600T mammography unit, manufactured by General Electric (GE), using the MCNP5 Monte Carlo code, to obtain its spectrum and HVL, and compare the HVL of the simulated model with experimental data. Method: the mammography unit was simulated using a simplified model which a beam of 2x10{sup 8} electrons focuses on a Mo target angled 12 degrees, within a capsule filled with vacuum. The incident electrons were converted into photons. The capsule has a beryllium window, allowing the passage of the X-ray beam. The beam is detected by an air cylinder with 1 cm thickness placed 60 cm from the target. On the path of X-ray beam, is inserted a 0.03 mm Mo filter located 1.6 cm after the beryllium window. The space between the capsule and the detector cylinder was filled with air. The quality of X-ray beam was verified from the HVL using the MCNP5 code and the experimental method for the voltage range typically used in clinical routine (26-31 kVp). Results and discussion: the X-ray spectrum of the mammography device is satisfactorily simulated by MCNP5, showing the characteristic radiation peaks of molybdenum at 17.479 keV and 19.602 keV, the filtered spectrum generated by Bremsstrahlung, and reducing the total number of photons with the decrease in applied tension (kVp). The HVL obtained by MCNP5 and experimental measurements show a maximum difference of 5.31% (for 31 kVp). The result of both methods are within acceptable limits established by national

  4. A system of materials composition and geometry arrangement for fast neutron beam thermalization: An MCNP study

    Science.gov (United States)

    Uhlář, Radim; Alexa, Petr; Pištora, Jaromír

    2013-03-01

    Compact deuterium-tritium neutron generators emit fast neutrons (14.2 MeV) that have to be thermalized for neutron activation analysis experiments. To maximize thermal neutron flux and minimize epithermal and fast neutron fluxes across the output surface of the neutron generator facility, Monte Carlo calculations (MCNP5; Los Alamos National Laboratory) for different moderator types and widths and collimator and reflector designs have been performed. A thin lead layer close to the neutron generator as neutron multiplier followed by polyethylene moderator and surrounded by a massive lead and nickel collimator and reflector was obtained as the optimum setup.

  5. Isodose distributions and dose uniformity in the Portuguese gamma irradiation facility calculated using the MCNP code

    CERN Document Server

    Oliveira, C

    2001-01-01

    A systematic study of isodose distributions and dose uniformity in sample carriers of the Portuguese Gamma Irradiation Facility was carried out using the MCNP code. The absorbed dose rate, gamma flux per energy interval and average gamma energy were calculated. For comparison purposes, boxes filled with air and 'dummy' boxes loaded with layers of folded and crumpled newspapers to achieve a given value of density were used. The magnitude of various contributions to the total photon spectra, including source-dependent factors, irradiator structures, sample material and other origins were also calculated.

  6. Using MCNP in the design of neutron sources and neutron beams

    Energy Technology Data Exchange (ETDEWEB)

    Hergenreder, Daniel F.; Lecot, Carlos A.; Lovotti, Osvaldo P. [INVAP S.A., San Carlos de Bariloche (Argentina). Nuclear Projects Department. Nuclear Engineering Division

    2002-07-01

    The calculation methodology used to design cold, thermal and hot neutron sources and their associated neutron beam transport systems is presented. The design goal is to evaluate the performance of the neutron sources, their beam tubes and neutron guides at specific experimental locations in the reactor hall as well as in the neutron hall. The Monte Carlo method is a unique and powerful tool to transport neutrons. Its use in a bootstrap scheme appears to be an appropriate solution for this type of system. The proper use of MCNP as the main tool leads to a fast and reliable method to perform calculations in a relatively short time with low statistical errors. (author)

  7. Retrieval of gamma cell 220 irradiator isodose curves with MCNP simulations and experimental measurements

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, R.R.; Grynberg, S.E.; Ferreira, A.V.; Belo, L.C.M.; Squair, P.L.; Ribeiro, M.A. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Sousa, R.V.; Sebastiao, R.C.O. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Quimica

    2010-03-15

    Gamma irradiator facilities can be used in a wide range of applications such as biological and chemical researches, sterilization of medical devices and products. Dose mapping must be performed in these equipment in order to establish plant operational parameters, as dose uniformity, source utilization efficiency and maximum and minimum dose positions. The isodoses curves are measured using dosimeters or computer simulations. This work evaluates the absorbed dose in the CDTN/CNEN Gamma Cell Irradiation Facility, using the Monte Carlo N-Particles (MCNP) code. (author)

  8. Improvements in the simulation of the efficiency of a HPGe detector with Monte Carlo code MCNP5; Mejoras en la simulacion de la eficiencia de un detector HPGe con el codigo Monte Carlo MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, S.; Querol, A.; Rodenas, J.; Verdu, G.

    2014-07-01

    in this paper we propose to perform a simulation model using the MCNP5 code and a registration form meshing to improve the simulation efficiency of the detector in the range of energies ranging from 50 to 2000 keV. This meshing is built by FMESH MCNP5 registration code that allows a mesh with cells of few microns. The photon and electron flow is calculated in the different cells of the mesh which is superimposed on detector geometry. It analyzes the variation of efficiency (related to the variation of energy deposited in the active volume). (Author)

  9. Simulation of reactor noise analysis measurement for light-water critical assembly TCA using MCNP-DSP

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, Toshihiro; Sakurai, Kiyoshi; Tonoike, Kotaro; Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-01-01

    Reactor noise analysis methods using Monte Carlo technique have been proposed and developed in the field of nuclear criticality safety. The Monte Carlo simulation for noise analysis can be made by simulating physical phenomena in the course of neutron transport in a nuclear fuel as practically as possible. MCNP-DSP was developed by T. Valentine of ORNL for this purpose and it is a modified version of MCNP-4A. The authors applied this code to frequency analysis measurements performed in light-water critical assembly TCA. Prompt neutron generation times for critical and subcritical cores were measured by doing the frequency analysis of detector signals. The Monte Carlo simulations for these experiments were carried out using MCNP-DSP, and prompt neutron generation times were calculated. (author)

  10. Comparison of TG-43 dosimetric parameters of brachytherapy sources obtained by three different versions of MCNP codes.

    Science.gov (United States)

    Zaker, Neda; Zehtabian, Mehdi; Sina, Sedigheh; Koontz, Craig; Meigooni, Ali S

    2016-03-01

    Monte Carlo simulations are widely used for calculation of the dosimetric parameters of brachytherapy sources. MCNP4C2, MCNP5, MCNPX, EGS4, EGSnrc, PTRAN, and GEANT4 are among the most commonly used codes in this field. Each of these codes utilizes a cross-sectional library for the purpose of simulating different elements and materials with complex chemical compositions. The accuracies of the final outcomes of these simulations are very sensitive to the accuracies of the cross-sectional libraries. Several investigators have shown that inaccuracies of some of the cross section files have led to errors in  125I and  103Pd parameters. The purpose of this study is to compare the dosimetric parameters of sample brachytherapy sources, calculated with three different versions of the MCNP code - MCNP4C, MCNP5, and MCNPX. In these simulations for each source type, the source and phantom geometries, as well as the number of the photons, were kept identical, thus eliminating the possible uncertainties. The results of these investigations indicate that for low-energy sources such as  125I and  103Pd there are discrepancies in gL(r) values. Discrepancies up to 21.7% and 28% are observed between MCNP4C and other codes at a distance of 6 cm for  103Pd and 10 cm for  125I from the source, respectively. However, for higher energy sources, the discrepancies in gL(r) values are less than 1.1% for  192Ir and less than 1.2% for  137Cs between the three codes. PACS number(s): 87.56.bg.

  11. RoadPlan Manual version 3.1

    NARCIS (Netherlands)

    Schoute, Albert L.

    2002-01-01

    RoadPlan is an interactive planning and preprocessing tool to analyse and optimize multi-agv traffic. Currently it provides facilities for: * Traffic road map configuration and visualization; * Interactive drawing and modification of traffic layouts; * Collision and deadlock analysis of multiple agv

  12. MPACT Theory Manual, Version 2.2.0

    Energy Technology Data Exchange (ETDEWEB)

    Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Collins, Benjamin S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gehin, Jess C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jabaay, Daniel [Univ. of Michigan, Ann Arbor, MI (United States); Kelley, Blake W. [Univ. of Michigan, Ann Arbor, MI (United States); Clarno, Kevin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kim, Kang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kochunas, Brendan [Univ. of Michigan, Ann Arbor, MI (United States); Larsen, Edward W. [Univ. of Michigan, Ann Arbor, MI (United States); Liu, Yuxuan [Univ. of Michigan, Ann Arbor, MI (United States); Liu, Zhouyu [Univ. of Michigan, Ann Arbor, MI (United States); Martin, William R. [Univ. of Michigan, Ann Arbor, MI (United States); Palmtag, Scott [Core Physics, Inc., Cary, NC (United States); Rose, Michael [Univ. of Michigan, Ann Arbor, MI (United States); Saller, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Stimpson, Shane [Univ. of Michigan, Ann Arbor, MI (United States); Trahan, Travis [Univ. of Michigan, Ann Arbor, MI (United States); Wang, J. W. [Univ. of Michigan, Ann Arbor, MI (United States); Wieselquist, William A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Young, Mitchell [Univ. of Michigan, Ann Arbor, MI (United States); Zhu, Ang [Univ. of Michigan, Ann Arbor, MI (United States)

    2016-06-09

    This theory manual describes the three-dimensional (3-D) whole-core, pin-resolved transport calculation methodology employed in the MPACT code. To provide sub-pin level power distributions with sufficient accuracy, MPACT employs the method of characteristics (MOC) solutions in the framework of a 3-D coarse mesh finite difference (CMFD) formulation. MPACT provides a 3D MOC solution, but also a 2D/1D solution in which the 2D planar solution is provided by MOC and the axial coupling is resolved by one-dimensional (1-D) lower order (diffusion or P3) solutions. In Chapter 2 of the manual, the MOC methodology is described for calculating the regional angular and scalar fluxes from the Boltzmann transport equation. In Chapter 3, the 2D/1D methodology is described, together with the description of the CMFD iteration process involving dynamic homogenization and solution of the multigroup CMFD linear system. A description of the MPACT depletion algorithm is given in Chapter 4, followed by a discussion of the subgroup and ESSM resonance processing methods in Chapter 5. The final Chapter 6 describes a simplified thermal hydraulics model in MPACT.

  13. SIERRA/Aero Theory Manual Version 4.44

    Energy Technology Data Exchange (ETDEWEB)

    Sierra Thermal/Fluid Team

    2017-04-01

    SIERRA/Aero is a two and three dimensional, node-centered, edge-based finite volume code that approximates the compressible Navier-Stokes equations on unstructured meshes. It is applicable to inviscid and high Reynolds number laminar and turbulent flows. Currently, two classes of turbulence models are provided: Reynolds Averaged Navier-Stokes (RANS) and hybrid methods such as Detached Eddy Simulation (DES). Large Eddy Simulation (LES) models are currently under development. The gas may be modeled either as ideal, or as a non-equilibrium, chemically reacting mixture of ideal gases. This document describes the mathematical models contained in the code, as well as certain implementation details. First, the governing equations are presented, followed by a description of the spatial discretization. Next, the time discretization is described, and finally the boundary conditions. Throughout the document, SIERRA/ Aero is referred to simply as Aero for brevity.

  14. Total reaction cross sections in CEM and MCNP6 at intermediate energies

    Science.gov (United States)

    Kerby, Leslie M.; Mashnik, Stepan G.

    2015-08-01

    Accurate total reaction cross section models are important to achieving reliable predictions from spallation and transport codes. The latest version of the Cascade Exciton Model (CEM) as incorporated in the code CEM03.03, and the Monte Carlo N-Particle transport code (MCNP6), both developed at Los Alamos National Laboratory (LANL), each use such cross sections. Having accurate total reaction cross section models in the intermediate energy region (∼ 50 MeV to ∼ 5 GeV) is very important for different applications, including analysis of space environments, use in medical physics, and accelerator design, to name just a few. The current inverse cross sections used in the preequilibrium and evaporation stages of CEM are based on the Dostrovsky et al. model, published in 1959. Better cross section models are available now. Implementing better cross section models in CEM and MCNP6 should yield improved predictions for particle spectra and total production cross sections, among other results. Our current results indicate this is, in fact, the case.

  15. Transport calculation of thermal and cold neutrons using NMTC/JAERI-MCNP4A code system

    Energy Technology Data Exchange (ETDEWEB)

    Iga, Kiminori [Kyushu Univ., Fukuoka (Japan); Takada, Hiroshi; Nagao, Tadashi

    1998-01-01

    In order to investigate the applicability of the NMTC/JAERI-MCNP4A code system to the neutronics design study in the neutron science research project of JAERI, transport calculations of thermal and cold neutrons are performed with the code system on a spallation neutron source composed of light water cooled tantalum target with a moderator and a reflector system. The following neutronic characteristics are studied in the calculation : the variation of the intensity of neutrons emitted from a light water moderator or a liquid hydrogen with/without the B{sub 4}C decoupler, which are installed to produce sharp pulse, and that dependent on the position of external source neutrons in the tantalum target. The calculated neutron energy spectra are reproduced well by the semi-empirical formula with the parameter values reliable in physical meanings. It is found to be necessary to employ proper importance sampling technique in the statistics. It is confirmed from this work that the NMTC/JAERI-MCNP4A code system is applicable to the neutronics design study of spallation neutron sources proposed for the neutron science research project. (author)

  16. Total Reaction Cross Sections in CEM and MCNP6 at Intermediate Energies

    CERN Document Server

    Kerby, Leslie M

    2015-01-01

    Accurate total reaction cross section models are important to achieving reliable predictions from spallation and transport codes. The latest version of the Cascade Exciton Model (CEM) as incorporated in the code CEM03.03, and the Monte Carlo N-Particle transport code (MCNP6), both developed at Los Alamos National Laboratory (LANL), each use such cross sections. Having accurate total reaction cross section models in the intermediate energy region ($\\sim$50 MeV to $\\sim$5 GeV) is very important for different applications, including analysis of space environments, use in medical physics, and accelerator design, to name just a few. The current inverse cross sections used in the preequilibrium and evaporation stages of CEM are based on the Dostrovsky {\\it et al.} model, published in 1959. Better cross section models are available now. Implementing better cross section models in CEM and MCNP6 should yield improved predictions for particle spectra and total production cross sections, among other results. Our current...

  17. Physics and Algorithm Enhancements for a Validated MCNP/X Monte Carlo Simulation Tool, Phase VII

    Energy Technology Data Exchange (ETDEWEB)

    McKinney, Gregg W [Los Alamos National Laboratory

    2012-07-17

    Currently the US lacks an end-to-end (i.e., source-to-detector) radiation transport simulation code with predictive capability for the broad range of DHS nuclear material detection applications. For example, gaps in the physics, along with inadequate analysis algorithms, make it difficult for Monte Carlo simulations to provide a comprehensive evaluation, design, and optimization of proposed interrogation systems. With the development and implementation of several key physics and algorithm enhancements, along with needed improvements in evaluated data and benchmark measurements, the MCNP/X Monte Carlo codes will provide designers, operators, and systems analysts with a validated tool for developing state-of-the-art active and passive detection systems. This project is currently in its seventh year (Phase VII). This presentation will review thirty enhancements that have been implemented in MCNPX over the last 3 years and were included in the 2011 release of version 2.7.0. These improvements include 12 physics enhancements, 4 source enhancements, 8 tally enhancements, and 6 other enhancements. Examples and results will be provided for each of these features. The presentation will also discuss the eight enhancements that will be migrated into MCNP6 over the upcoming year.

  18. Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Love, E.F.; Pauley, K.A.; Reid, B.D.

    1995-09-01

    This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy`s Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste.

  19. User design

    CERN Document Server

    Carr-Chellman, Alison A

    2012-01-01

    User Design offers a fresh perspective on how front-line learners (users) can participate in the design of learning environments. The author challenges the universal assumption that front-line users must be relegated to the role of offering input, and that the actual design activity of learning systems must still be conducted only by experts. The book presents a new set of methods and strategies that show how the tools of professional designers can be effectively shared with broad groups of users and other participants in the process of creating their own learning. Drawing

  20. MCNP - transport calculations in ducts using multigroup albedo coefficients; Calculos de transporte em dutos utilizando coeficientes de albedo multigrupo no codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Ono, Shizuca; Vieira, Wilson J.; Garcia, Roberto D.M. [Centro Tecnico Aeroespacial (CTA-IEAv), Sao Jose dos Campos, SP (Brazil). Inst. de Estudos Avancados

    2000-07-01

    In this work, the use of multigroup albedo coefficients in Monte Carlo calculations of particle reflection and transmission by ducts is investigated. The procedure consists in modifying the MCNP code so that an albedo matrix computed previously by deterministic methods or Monte Carlo is introduced into the program to describe particle reflection by a surface. This way it becomes possible to avoid the need of considering particle transport in the duct wall explicitly, changing the problem to a problem of transport in the duct interior only and reducing significantly the difficulty of the real problem. The probability of particle reflection at the duct wall is given, for each group, as the sum of the albedo coefficients over the final groups. The calculation is started by sampling a source particle and simulating its reflection on the duct wall by sampling a group for the emerging particle. The particle weight is then reduced by the reflection probability. Next, a new direction and trajectory for the particle is selected. Numerical results obtained for the model are compared with results from a discrete ordinates code and results from Monte Carlo simulations that take particle transport in the wall into account. (author)

  1. MCNP-DSP calculations of measurements with uranyl nitrate solution system

    Energy Technology Data Exchange (ETDEWEB)

    Valentine, T.E. [Oak Ridge National Lab., TN (United States)

    1998-09-01

    The {sup 252}Cf-source-driven noise analysis method has been used to determine the subcriticality of various configurations of fissile materials. In the past, the application of this method was limited because point-kinetics models had to be used to interpret the data; however, with the development of the Monte Carlo code MCNP-DSP, the measurements can be analyzed using the more general Monte Carlo models. The results of the Monte carlo calculations will be dependent on the ability to model the experiment accurately and on the nuclear data used to perform the calculations. This paper presents a comparison of the measured and calculated ratio of spectral densities for a subset of measurements performed with a uranyl nitrate solution tank filled to various heights. The results presented are for calculations that were performed with both ENDF/B-IV and ENDF/B-V cross-section data sets.

  2. Mcnp calculation of neutron scatter in the Main Bay of the Chadwick Building, NPL

    Energy Technology Data Exchange (ETDEWEB)

    Naismith, O.F.; Thomas, D.J.

    1996-02-01

    The Monte Carlo neutron transport code MCNP has been used to calculate the room and air scattered neutron component at 75 cm from a radionuclide source located at the center of the low-scatter area in the Chadwick Building, Bldg. 47, at National Physical Laboratory (NPL). This is the standard distance used for calibrating personal dosemeters, and the calculation provides information for correcting the response of dosemeters to the scattered radiation. Calculations were performed for both an Am-Be and a (252)Cf source. These measurements revealed that the model used for features within the low-scatter area needs to be refined for calculating scatter at distances further from the source than 75 cm.

  3. Comparison of a laboratory spectrum of Eu-152 with results of simulation using the MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Rodenas, J. [Departamento de Ingenieria Quimica y Nuclear, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain); Gallardo, S. [Departamento de Ingenieria Quimica y Nuclear, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain)], E-mail: sergalbe@iqn.upv.es; Ortiz, J. [Laboratorio de Radiactividad Ambiental, Universidad Politecnica de Valencia, Apartado 22012, E-46071 Valencia (Spain)

    2007-09-21

    Detectors used for gamma spectrometry must be calibrated for each geometry considered in environmental radioactivity laboratories. This calibration is performed using a standard solution containing gamma emitter sources. Nevertheless, the efficiency curves obtained are periodically checked using a source such as {sup 152}Eu emitting many gamma rays that cover a wide energy range (20-1500 keV). {sup 152}Eu presents a problem because it has a lot of peaks affected by True Coincidence Summing (TCS). Two experimental measures have been performed placing the source (a Marinelli beaker) at 0 and 10 cm from the detector. Both spectra are simulated by the MCNP 4C code, where the TCS is not reproduced. Therefore, the comparison between experimental and simulated peak net areas permits one to choose the most convenient peaks to check the efficiency curves of the detector.

  4. MCNP Simulation to Hard X-Ray Emission of KSU Dense Plasma Focus Machine

    CERN Document Server

    Mohamed, Amgad E

    2015-01-01

    The MCNP program used to simulate the hard x-ray emission from KSU dense plasma focus device, an electron beam spectrum of maximum energy 100 keV was used to hit anode target. The bremsstrahlung radiation was measured using the F2 tally functions on the chamber walls and on a virtual sphere surrounding the machine, the radiation spectrum was recorded for various anode materials like tungsten, stainless steel and molybdenum. It was found that tungsten gives the best and the most intense radiation for the same electron beam. An aluminum filter of thickness 2mm and 4mm was used to cutoff the lower energy band from the x-ray spectrum. It was found that the filters achieved the mission and there is no distinct difference in between.

  5. Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model

    Science.gov (United States)

    Kazeminezhad, F.; Anghaie, S.

    2008-01-01

    Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.

  6. Image enhancement using MCNP5 code and MATLAB in neutron radiography.

    Science.gov (United States)

    Tharwat, Montaser; Mohamed, Nader; Mongy, T

    2014-07-01

    This work presents a method that can be used to enhance the neutron radiography (NR) image for objects with high scattering materials like hydrogen, carbon and other light materials. This method used Monte Carlo code, MCNP5, to simulate the NR process and get the flux distribution for each pixel of the image and determines the scattered neutron distribution that caused image blur, and then uses MATLAB to subtract this scattered neutron distribution from the initial image to improve its quality. This work was performed before the commissioning of digital NR system in Jan. 2013. The MATLAB enhancement method is quite a good technique in the case of static based film neutron radiography, while in neutron imaging (NI) technique, image enhancement and quantitative measurement were efficient by using ImageJ software. The enhanced image quality and quantitative measurements were presented in this work.

  7. Pebble bed reactors simulation using MCNP: The Chinese HTR-10 reactor

    Directory of Open Access Journals (Sweden)

    SA Hosseini

    2013-09-01

    Full Text Available   Given the role of Gas-Graphite reactors as the fourth generation reactors and their recently renewed importance, in 2002 the IAEA proposed a set of Benchmarking problems. In this work, we propose a model both efficient in time and resources and exact to simulate the HTR-10 reactor using MCNP-4C code. During the present work, all of the pressing factors in PBM reactor design such as the inter-pebble leakage, fuel particle distribution and fuel pebble packing fraction effects have been taken into account to obtain an exact and easy to run model. Finally, the comparison between the results of the present work and other calculations made at INEEL proves the exactness of the proposed model.

  8. Using MCNP to estimate nuclear energy deposition in a cold neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Lecot, Carlos A.; Hergenreder, Daniel F.; Lovotti, Osvaldo P. [INVAP S.A., San Carlos de Bariloche (Argentina). Nuclear Projects Department. Nuclear Engineering Division

    2002-07-01

    The location of a Cold Neutron Source (CNS) implies a careful cost/benefit balance between neutron performance and heat removal capacity of the required cryogenic equipment. To justify this balance, the calculation of the total heat deposited in the device is a critical parameter. It depends on many different contributions, i.e. neutron and gamma radiation, beta decay, fission product decay gammas, among others. With minor modifications to some standard cross section sets, the Monte Carlo code MCNP offers the possibility to calculate the total heat load in a single calculation, without the utilization of intermediate calculations and/or auxiliary codes. This paper describes the methodology used to modify the cross section sets, to calculate the energy deposited in the CNS and to evaluate the cold neutron flux which is the variable used to compare performance at different locations. (author)

  9. Assessment of doses caused by electrons in thin layers of tissue-equivalent materials, using MCNP.

    Science.gov (United States)

    Heide, Bernd

    2013-10-01

    Absorbed doses caused by electron irradiation were calculated with Monte Carlo N-Particle transport code (MCNP) for thin layers of tissue-equivalent materials. The layers were so thin that the calculation of energy deposition was on the border of the scope of MCNP. Therefore, in this article application of three different methods of calculation of energy deposition is discussed. This was done by means of two scenarios: in the first one, electrons were emitted from the centre of a sphere of water and also recorded in that sphere; and in the second, an irradiation with the PTB Secondary Standard BSS2 was modelled, where electrons were emitted from an (90)Sr/(90)Y area source and recorded inside a cuboid phantom made of tissue-equivalent material. The speed and accuracy of the different methods were of interest. While a significant difference in accuracy was visible for one method in the first scenario, the difference in accuracy of the three methods was insignificant for the second one. Considerable differences in speed were found for both scenarios. In order to demonstrate the need for calculating the dose in thin small zones, a third scenario was constructed and simulated as well. The third scenario was nearly equal to the second one, but a pike of lead was assumed to be inside the phantom in addition. A dose enhancement (caused by the pike of lead) of ∼113 % was recorded for a thin hollow cylinder at a depth of 0.007 cm, which the basal-skin layer is referred to in particular. Dose enhancements between 68 and 88 % were found for a slab with a radius of 0.09 cm for all depths. All dose enhancements were hardly noticeable for a slab with a cross-sectional area of 1 cm(2), which is usually applied to operational radiation protection.

  10. Thorium-based mixed oxide fuel in a pressurized water reactor: A feasibility analysis with MCNP

    Science.gov (United States)

    Tucker, Lucas Powelson

    This dissertation investigates techniques for spent fuel monitoring, and assesses the feasibility of using a thorium-based mixed oxide fuel in a conventional pressurized water reactor for plutonium disposition. Both non-paralyzing and paralyzing dead-time calculations were performed for the Portable Spectroscopic Fast Neutron Probe (N-Probe), which can be used for spent fuel interrogation. Also, a Canberra 3He neutron detector's dead-time was estimated using a combination of subcritical assembly measurements and MCNP simulations. Next, a multitude of fission products were identified as candidates for burnup and spent fuel analysis of irradiated mixed oxide fuel. The best isotopes for these applications were identified by investigating half-life, photon energy, fission yield, branching ratios, production modes, thermal neutron absorption cross section and fuel matrix diffusivity. 132I and 97Nb were identified as good candidates for MOX fuel on-line burnup analysis. In the second, and most important, part of this work, the feasibility of utilizing ThMOX fuel in a pressurized water reactor (PWR) was first examined under steady-state, beginning of life conditions. Using a three-dimensional MCNP model of a Westinghouse-type 17x17 PWR, several fuel compositions and configurations of a one-third ThMOX core were compared to a 100% UO2 core. A blanket-type arrangement of 5.5 wt% PuO2 was determined to be the best candidate for further analysis. Next, the safety of the ThMOX configuration was evaluated through three cycles of burnup at several using the following metrics: axial and radial nuclear hot channel factors, moderator and fuel temperature coefficients, delayed neutron fraction, and shutdown margin. Additionally, the performance of the ThMOX configuration was assessed by tracking cycle length, plutonium destroyed, and fission product poison concentration.

  11. ACTI - An MCNP Data Library for Prompt Gamma-Ray Spectroscopy.

    Energy Technology Data Exchange (ETDEWEB)

    Frankle, S. C. (Stephanie C.); Reedy, R. C. (Robert C.); Young, P. G. (Phillip Gaffney),

    2002-01-01

    Prompt gamma-ray spectroscopy is used in a wide variety of applications for determining material compositions. High-quality photon-production data from thermal-neutron capture reactions are essential for these applications. Radiation transport codes, such as MCNP{trademark}, are often used to design detector systems, determine minimum detection thresholds, etc. These transport codes rely on evaluated nuclear databases such as ENDF (Evaluated Nuclear Data File) to provide the fundamental data used in the transport calculations. Often the photon-production data from incident neutron reactions in the evaluations are of relatively poor quality. We have compiled the best experimental data for thermal-neutron capture for the naturally occurring isotopes for elements from H through Zn as well as for {sup 70,72,73,74,76}Ge, {sup 149}Sm, {sup 155,157}Gd, {sup 181}Ta and {sup 182,183,184,186}W. This compilation has been used to update the ENDF evaluations for {sup 1}H, {sup 4}He, {sup 9}Be, {sup 14}N, {sup 16}O, {sup 19}F, Na, Mg, {sup 27}Al, {sup 32}S, S, {sup 35,37}Cl, K, Ca, {sup 45}Sc, Ti, {sup 51}V, {sup 50,52,53,54}Cr, {sup 55}Mn, {sup 54,56,57,58}Fe, {sup 58,60,61,62,64}Ni, {sup 63,65}Cu and {sup 182,183,184,186}W. In addition, the inelastic cross sections and corresponding secondary-photon distributions were updated for {sup 16}O. Complete new evaluations were submitted to ENDF for {sup 35,37}Cl. This paper will discuss the evaluation effort and the production of the MCNP data library, ACTI, based on the new evaluations. Data from the ENDF evaluations for {sup 28-30}Si were also included in the ACTI library for completeness. The silicon evaluations were updated in 1997 and include the latest experimental data for radiative capture.

  12. Energy and spatial dependence of MCNP simulations for ZED-2 critical experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)], E-mail: kozierk@aecl.ca

    2008-07-01

    MCNP simulations of ZED-2 critical experiments provide a good test of the reliability of the nuclear data involved in the simulation of reactor physics phenomena of importance to CANDU reactors, particularly the coolant void reactivity. Recent work has therefore focused on the impact of the new ENDF/B-VII.0 nuclear data library. One feature of this library is the provision of thermal scattering law data for UO{sub 2}. Initial MCNP results using preliminary ACE-format data files for UO{sub 2} thermal scattering suggested that a consistent reduction was obtained in the coolant void reactivity simulation bias, especially for ZED-2 critical experiments involving slightly enriched uranium (0.95 wt% {sup 235}U) and H{sub 2}O/air coolant. However, subsequent work using UO{sub 2} thermal scattering data files that correctly include the coherent elastic scattering component indicated that the net reactivity impact is quite small. The present work extends this investigation to examine in detail the energy dependence of the impact of the UO{sub 2} thermal scattering data and, more generally, the energy and spatial dependence of the coolant void reactivity simulation bias for some of these experiments. In addition, results are presented using MCNPX with an improved treatment for thermal scattering. It is found that the net reactivity impact results from the cancellation of larger positive and negative effects at different energies and in different fuel regions, and which generally highlight the reactor physics changes that occur when the coolant is removed. (author)

  13. Determination of dosimetric parameters for 125I seed source using MCNP5 and EGSnrc MC codes%MCNP5与EGSnrc比较计算125I种子源剂量参数

    Institute of Scientific and Technical Information of China (English)

    曹振; 阮锡超; 孟贝蒂; 石翠燕

    2014-01-01

    根据AAPM TG43U1的推荐,使用MCNP5与EGSnrc两种蒙特卡罗程序计算6711型125I种子源剂量计算参数,并将两者计算结果和AAPM推荐值比较,得到相对偏差结果如下:剂量率常数,MCNP5为0.62%,EGSnrc为2.07%;径向剂量函数,MNCP5为0.15%-5.12%,EGSnrc为0%-2.18%.两者计算结果均与推荐值符合得很好,而EGSnrc的计算结果更具优势.

  14. Development of an interface between MCNP and ORIGEN codes for calculations of fuel evolution in nuclear systems. Initial project; Desenvolvimento de uma interface entre os codigos MCNP e ORIGEN para calculos de evolucao de combustiveis em sistemas nucleares. Projeto inicial

    Energy Technology Data Exchange (ETDEWEB)

    Campolina, Daniel de Almeida Magalhaes

    2009-07-01

    In Many situations of nuclear system study, it is necessary to know the detailed particle flux in a geometry. Deterministic 1-D and 2-D methods aren't suitable to represent some strong 3-D behavior configurations, for example in cores where the neutron flux varies considerably in the space and Monte Carlo analysis are necessary. The majority of Monte Carlo transport calculation codes, performs time static simulations, in terms of fuel isotopic composition. This work is a initial project to incorporate depletion capability to the MCNP code, by means of a connection with ORIGEN2.1 burnup code. The method to develop the program proposed followed the methodology of other programs used to the same purpose. Essentially, MCNP data library are used to generate one group microscopic cross sections that override default ORIGEN libraries. To verify the actual implemented part, comparisons which MCNPX (version 2.6.0) results were made. The neutron flux and criticality value of core agree. The neutron flux and criticality value of the core agree, especially in beginning of burnup when the influence of fission products are not very considerable. The small difference encountered was probably caused by the difference in the number of isotopes considered in the transport models (89 MCNPX x 25 GB). Next step of this work is to adapt MCNP version 4C to work with a memory higher than its standard value (4MB), in order to allow a greater number of isotopes in the transport model. (author)

  15. Determination of {beta}{sub eff} using MCNP-4C2 and application to the CROCUS and PROTEUS reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vollaire, J. [European Organization for Nuclear Research CERN, CH-1211 Geneve 23 (Switzerland); Plaschy, M.; Jatuff, F. [Paul Scherrer Institut PSI, CH-5232 Villigen PSI (Switzerland); Chawla, R. [Paul Scherrer Institut PSI, CH-5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, CH-1015 Lausanne (Switzerland)

    2006-07-01

    A new Monte Carlo method for the determination of {beta}{sub eff} has been recently developed and tested using appropriate models of the experimental reactors CROCUS and PROTEUS. The current paper describes the applied methodology and highlights the resulting improvements compared to the simplest MCNP approach, i.e. the 'prompt method' technique. In addition, the flexibility advantages of the developed method are presented. Specifically, the possibility to obtain the effective delayed neutron fraction {beta}{sub eff} per delayed neutron group, per fissioning nuclide and per reactor region is illustrated. Finally, the MCNP predictions of {beta}{sub eff} are compared to the results of deterministic calculations. (authors)

  16. Uncertainty analysis in the simulation of X-ray spectra in the diagnostic range using the MCNP5 code.

    Science.gov (United States)

    Gallardo, S; Querol, A; Ródenas, J; Verdú, G

    2011-01-01

    An accurate knowledge of the photonic spectra emitted by X-ray tubes in radiodiagnostics is essential to better estimate the imparted dose to patients and to improve the image quality obtained with these devices. In this work, several X-ray spectra have been simulated using the MCNP5 code to simulate X-ray production in a commercial device. To validate the Monte Carlo results, simulated spectra have been compared to those extracted from the IPEM 78 database. The uncertainty associated to some geometrical features of the tube and its effect on the simulated spectra has been analyzed using the Noether-Wilks formula. This analysis has been focused on the thickness of collimators, filters, shielding and barrel shutter. Furthermore, results show that the uncertainty due to geometrical parameters (0.98% in terms of Root Mean Squared) is higher than the statistical uncertainty associated to the MCNP5 calculations.

  17. Using MCNP6 to Estimate Fission Neutron Properties of a Reflected Plutonium Sphere

    Energy Technology Data Exchange (ETDEWEB)

    Clark, Alexander Rich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Nelson, Mark Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hutchinson, Jesson D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-08-08

    The purpose of this project was to determine the fission multiplicity distribution, p(v), for the Beryllium Reflected Plutonium (BeRP) ball and to determine whether or not it changed appreciably for various High Density Polyethylene (HDPE) reflected configurations. The motivation for this project was to determine whether or not the average number of neutrons emitted per fission, v, changed significantly enough to reduce the discrepancy between MCNP6 and Robba, Dowdy, Atwater (RDA) point kinetic model estimates of multiplication. The energy spectrum of neutrons that induced fissions in the BeRP ball, NIF (E), was also computed in order to determine the average energy of neutrons inducing fissions, NIF . p(v) was computed using the FMULT card, NIF (E) and NIF were computed using an F4 tally with an FM tally modifier (F4/FM) card, and the multiplication factor, keff, was computed using the KCODE card. Although NIF (E) changed significantly between bare and HDPE reflected configurations of the BeRP ball, the change in p(v), and thus the change in v, was insignificant. This is likely due to a difference between the way that NIF is computed using the FMULT and F4/FM cards. The F4/FM card indicated that NIF (E) was essentially Watt-fission distributed for a bare configuration and highly thermalized for all HDPE reflected configurations, while the FMULT card returned an average energy between 1 and 2 MeV for all configurations, which would indicate that the spectrum is Watt-fission distributed, regardless of the amount of HDPE reflector. The spectrum computed with the F4/FM cards is more physically meaningful and so the discrepancy between it and the FMULT card result is being investigated. It is hoped that resolving the discrepancy between the FMULT and F4/FM card estimates of NIF(E) will provide better v estimates that will lead to RDA multiplication estimates that are in better agreement with MCNP6 simulations.

  18. MCNP6 model of the University of Washington clinical neutron therapy system (CNTS)

    Science.gov (United States)

    Moffitt, Gregory B.; Stewart, Robert D.; Sandison, George A.; Goorley, John T.; Argento, David C.; Jevremovic, Tatjana

    2016-01-01

    A MCNP6 dosimetry model is presented for the Clinical Neutron Therapy System (CNTS) at the University of Washington. In the CNTS, fast neutrons are generated by a 50.5 MeV proton beam incident on a 10.5 mm thick Be target. The production, scattering and absorption of neutrons, photons, and other particles are explicitly tracked throughout the key components of the CNTS, including the target, primary collimator, flattening filter, monitor unit ionization chamber, and multi-leaf collimator. Simulations of the open field tissue maximum ratio (TMR), percentage depth dose profiles, and lateral dose profiles in a 40 cm  ×  40 cm  ×  40 cm water phantom are in good agreement with ionization chamber measurements. For a nominal 10  ×  10 field, the measured and calculated TMR values for depths of 1.5 cm, 5 cm, 10 cm, and 20 cm (compared to the dose at 1.7 cm) are within 0.22%, 2.23%, 4.30%, and 6.27%, respectively. For the three field sizes studied, 2.8 cm  ×  2.8 cm, 10.4 cm  ×  10.3 cm, and 28.8 cm  ×  28.8 cm, a gamma test comparing the measured and simulated percent depth dose curves have pass rates of 96.4%, 100.0%, and 78.6% (depth from 1.5 to 15 cm), respectively, using a 3% or 3 mm agreement criterion. At a representative depth of 10 cm, simulated lateral dose profiles have in-field (⩾10% of central axis dose) pass rates of 89.7% (2.8 cm  ×  2.8 cm), 89.6% (10.4 cm  ×  10.3 cm), and 100.0% (28.8 cm  ×  28.8 cm) using a 3% and 3 mm criterion. The MCNP6 model of the CNTS meets the minimum requirements for use as a quality assurance tool for treatment planning and provides useful insights and information to aid in the advancement of fast neutron therapy.

  19. The application of the Monte-Carlo neutron transport code MCNP to a small "nuclear battery" system

    OpenAIRE

    Puigdellívol Sadurní, Roger

    2009-01-01

    The project consist in calculate the keff to a small nuclear battery. The code Monte- Carlo neutron transport code MCNP is used to calculate the keff. The calculations are done at the beginning of life to know the capacity of the core becomes critical in different conditions. These conditions are the study parameters that determine the criticality of the core. These parameters are the uranium enrichment, the coated particles (TRISO) packing factor and the size of the core. More...

  20. DESAIN TERAS PLTN JENIS PEBBLE BED MODULAR REACTOR (PBMR) MENGGUNAKAN PAKET PROGRAM MCNP-5 PADA KONDISI BEGINNING OF LIFE

    OpenAIRE

    Ralind Re Marla; Yohannes Sardjono; Supardi Supardi

    2015-01-01

    Telah dilakukan desain teras Pembangkit Listrik Tenaga Nuklir (PLTN) untuk jenis Pebble Bed Modular Reactor (PBMR) dengan daya 70 MWe untuk keperluan proses smelter pada keadaan beginning of life (BOL). Analisis ini bertujuan untuk mengetahui persen pengkayaan, distribusi suhu dan nilai keselamatan dengan koefisien reaktivitas teras yang negatif pada reaktor jenis PBMR apabila daya reaktor 70 MWe. Analisis menggunakan program Monte Carlo N-Particle-5 (MCNP5) dan dari hasil analisis ini dihara...

  1. Availability of MCNP & MATLAB for reconstructing the water-vapor two-phase flow pattern in neutron radiography

    Institute of Scientific and Technical Information of China (English)

    FENG Qixi; FENG Quanke; TAKESHI Kawai

    2008-01-01

    The China Advanced Research Reactor (CARR) is scheduled to be operated in the autumn of 2008. In this paper, we report preparations for installing the neutron radiography instrument (NRI) and for utilizing it efficiently. The 2-D relative neutron intensity profiles for the water-vapor two-phase flow inside the tube were obtained using the MCNP code without influence of γ-ray and electronic-noise. The MCNP simulation of the 2-D neutron intensity profile for the water-vapor two-phase flow was demonstrated. The simulated 2-D neutron intensity profiles could be used as the benchmark data base by calibrating part of the data measured by the CARR-NRI. The 3-D objective images allow us to understand the flow pattern more clearly and it is reconstructed using the MATLAB through the threshold transformation techniques. And thus it is concluded that the MCNP code and the MATLAB are very useful for constructing the benchmark data base for the investigation of the water-vapor two-phase flow using the CARR-NRI.

  2. Designing an epithermal neutron beam for boron neutron capture therapy for a DIDO type reactor using MCNP

    Science.gov (United States)

    Ross, D.; Constantine, G.; Weaver, D. R.; Beynon, T. D.

    1993-10-01

    This paper describes work undertaken to design an epithermal neutron beam for a DIDO type reactor for use in boron neutron capture therapy, a form of cancer treatment. It involved extensive use of MCNP, a Monte Carlo computer code. Initially, calculations were made with MCNP to simulate earlier experiments with an epithermal beam on the DIDO reactor. This comparison made it possible both to validate the Monte Carlo modelling of the reactor and to gain an insight into the important features of the simulation. Following this, MCNP was used to design a filtered epithermal neutron beam facility for DIDO's largest beam tube, a 13.7 cm radius horizontal tube which extends radially away from the core. First a selection was made of the optimum filter components for the beam. Then the research concentrated on combining these filter elements to construct a practical epithermal beam design. The results suggest that the optimum method of generating the epithermal neutron source is to employ a filter combination consisting principally of liquid argon with the addition of cadmium, aluminium, titanium and possibly tin. The calculations also show that the resultant neutron beam would have a flux greater than 1.0 × 10 9 n cm -2 s -1 and have sufficiently low fast-neutron and gamma-ray contamination.

  3. MCNP Super Lattice Method for VHTR ORIGEN2.2 Nuclear Library Improvement Based on ENDF/B-VII

    Energy Technology Data Exchange (ETDEWEB)

    G. S. Chang; J. R. Parry

    2010-10-01

    The advanced Very High Temperature gas-cooled Reactor (VHTR) achieves simplification of safety through reliance on innovative features and passive systems. One of the VHTRs innovative features is the reliance on ceramic-coated fuel particles to retain the fission products under extreme accident conditions. The effect of the random fuel kernel distribution in the fuel prismatic block creates a double-heterogeneous lattice, which needs to be addressed through the use of the newly developed prismatic super Kernel-by-Kernel Fuel (KbKF) lattice model method. Based on the new ENDF/B-VII nuclear cross section evaluated data, the developed KbKF super lattice model was then used with MCNP to calculate the material isotopes neutron reaction rates, such as, (n,?); (n,n’); (n,2n’); (n,f); (n,p); (n,?). Then, the MCNP-calculated results are rearranged to generate a set of new libraries “VHTRXS.lib,” for the ORIGEN2.2 isotopes depletion and build-up analysis code. The libraries contain one group cross section data for the structural light elements, actinides, and fission products that can be applied in the VHTR related fuel burnup and material transmutation analysis codes. The efficiency and ease of use of the MCNP method to generate and update the ORIGEN2.2 one-group spectrum weighed cross section library for VHTR was demonstrated.

  4. MCNP{trademark} simulations for identifying environmental contaminants using prompt gamma-rays from thermal neutron capture reactions

    Energy Technology Data Exchange (ETDEWEB)

    Frankle, S.C.; Conaway, J.G.

    1996-12-31

    The primary purposes of the Multispectral Neutron Logging Project, (MSN Project, funded by the U.S. Department of Energy), were to assess the effectiveness of existing neutron- induced spectral gamma-ray logging techniques for identifying environmental contaminants along boreholes, to further improve the technology, and to transfer that technology to industry. Using a pulsed neutron source with a high-resolution gamma-ray detector, spectra from thermal neutron capture reactions may be used to identify contaminants in the borehole environment. Direct borehole measurements such as this complement physical sampling and are useful in environmental restoration projects where characterization of contaminated sites is required and long-term monitoring may be needed for many years following cleanup or stabilization. In the MSN Project, a prototype logging instrument was designed which incorporated a pulsed 14-MeV neutron source and HPGe detector. Experimental measurements to determine minimum detection thresholds with the prototype instrument were conducted in the variable-contaminant test model for Cl, Cd, Sm, Gd, and Hg. We benchmarked an enhanced version of the Monte Carlo N-Particle computer code MCNP{trademark} using experimental data for Cl provide by our collaborators and experimental data from the variable-contaminant test model. MCNP was then used to estimate detection thresholds for the other contaminants used in the variable-contaminant model with the goal of validating the use of MCNP to estimate detection thresholds for many other contaminants that were not measured.

  5. Use of a Boron Doped Spherical Phantom for the Investigation of Neutron Directional Properties: Comparison Between Experiment and MCNP Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Drake, P.; Kierkegaard, J

    1999-07-01

    A boron doped 19 cm diameter spherical phantom was constructed to give information on the direction of neutrons inside the Ringhals 4 containment. The phantom was made of 40% paraffin and 60% boric acid. 10B contributes 2% of the total phantom weight. The phantom was tested for its angular sensitivity to neutrons. The response was tested with a {sup 252}Cf source and with a Monte Carlo calculation (MCNP) simulating a {sup 252}Cf source. In these investigations the phantom showed a strong directional response. However, there was only a fair correspondence between the experiment and the simulation. The discrepancies are, at least in part, due to the difference in energy and angular response of the dosemeters as compared with the idealised response characteristics in the MCNP calculation. In the MCNP calculation the experimental conditions were not fully simulated. The investigations also showed that the addition of boron to the phantom reduces the leakage of thermalised neutrons from the phantom, and the production of neutron induced photons in the phantom to insignificant levels. (author)

  6. Performance of the MTR core with MOX fuel using the MCNP4C2 code.

    Science.gov (United States)

    Shaaban, Ismail; Albarhoum, Mohamad

    2016-08-01

    The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U3O8&PuO2) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U3O8-Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U3O8-Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with (235)U and the amount of loaded (235)U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively.

  7. Human eye analytical and mesh-geometry models for ophthalmic dosimetry using MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Angelocci, Lucas V.; Fonseca, Gabriel P.; Yoriyaz, Helio, E-mail: hyoriyaz@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Eye tumors can be treated with brachytherapy using Co-60 plaques, I-125 seeds, among others materials. The human eye has regions particularly vulnerable to ionizing radiation (e.g. crystalline) and dosimetry for this region must be taken carefully. A mathematical model was proposed in the past [1] for the eye anatomy to be used in Monte Carlo simulations to account for dose distribution in ophthalmic brachytherapy. The model includes the description for internal structures of the eye that were not treated in previous works. The aim of this present work was to develop a new eye model based on the Mesh geometries of the MCNP6 code. The methodology utilized the ABAQUS/CAE (Simulia 3DS) software to build the Mesh geometry. For this work, an ophthalmic applicator containing up to 24 model Amersham 6711 I-125 seeds (Oncoseed) was used, positioned in contact with a generic tumor defined analytically inside the eye. The absorbed dose in eye structures like cornea, sclera, choroid, retina, vitreous body, lens, optical nerve and optical nerve wall were calculated using both models: analytical and MESH. (author)

  8. MCNP modeling of a neutron generator and its shielding at Missouri University of Science and Technology

    Science.gov (United States)

    Sharma, Manish K.; Alajo, Ayodeji Babatunde; Liu, Xin

    2014-12-01

    The shielding of a neutron generator producing fast neutrons should be sufficient to limit the dose rates to the prescribed values. A deuterium-deuterium neutron generator has been installed in the Nuclear Engineering Department at Missouri University of Science and Technology (Missouri S&T). The generator produces fast neutrons with an approximate energy of 2.5 MeV. The generator is currently shielded with different materials like lead, high-density polyethylene, and borated polyethylene. An MCNP transport simulation has been performed to estimate the dose rates at various places in and around the facility. The simulations incorporated the geometric and composition information of these shielding materials to determine neutron and photon dose rates at three central planes passing through the neutron source. Neutron and photon dose rate contour plots at these planes were provided using a MATLAB program. Furthermore, the maximum dose rates in the vicinity of the facility were used to estimate the annual limit for the generator's hours of operation. A successful operation of this generator will provide a convenient neutron source for basic and applied research at the Nuclear Engineering Department of Missouri S&T.

  9. Simulation of the BNCT of Brain Tumors Using MCNP Code: Beam Designing and Dose Evaluation

    Directory of Open Access Journals (Sweden)

    Fatemeh Sadat Rasouli

    2012-09-01

    Full Text Available Introduction BNCT is an effective method to destroy brain tumoral cells while sparing the healthy tissues. The recommended flux for epithermal neutrons is 109 n/cm2s, which has the most effectiveness on deep-seated tumors. In this paper, it is indicated that using D-T neutron source and optimizing of Beam Shaping Assembly (BSA leads to treating brain tumors in a reasonable time where all IAEA recommended criteria are met. Materials and Methods The proposed BSA based on a D-T neutron generator consists of a neutron multiplier system, moderators, reflector, and collimator. The simulated Snyder head phantom is used to evaluate dose profiles in tissues due to the irradiation of designed beam. Monte Carlo Code, MCNP-4C, was used in order to perform these calculations.   Results The neutron beam associated with the designed and optimized BSA has an adequate epithermal flux at the beam port and neutron and gamma contaminations are removed as much as possible. Moreover, it was showed that increasing J/Φ, as a measure of beam directionality, leads to improvement of beam performance and survival of healthy tissues surrounding the tumor. Conclusion According to the simulation results, the proposed system based on D-T neutron source, which is suitable for in-hospital installations, satisfies all in-air parameters. Moreover, depth-dose curves investigate proper performance of designed beam in tissues. The results are comparable with the performances of other facilities.

  10. Simulation of dental intensifying screen for intraoral radiographic using MCNP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, Vanessa M.; Oliveira, Renato C.M., E-mail: vanessamachado@ufmg.br [Curso Superior de Tecnologia em Radiologia. Faculdade de Medicina da Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil); Barros, Graiciany P.; Oliveira, Arno H.; Veloso, M. Auxiliadora F. [Departamento de Engenharia Nuclear. Escola de Engenharia. Universidade Federal de Minas Gerais, Belo Horizonte, MG (Brazil)

    2011-07-01

    One of basic principles for radiological protection is the optimization of techniques for obtain radiographic images, in way that the dose in the patient is kept as low as reasonably achievable (ALARA). Intensifying screens are used in medical radiology, which reduce considerably the dose rates in the production of radiographic images, maintaining the quality of these, while in dental radiology, there is no a intensifying screen available for intraoral examinations. From this technological requirement, this paper evaluates a computational modeling of an intensifying screen for use in intraoral radiography. For this, it was used the Monte Carlo code MCNP5 that allows the radiography simulation through the transport of electrons and photons in the different materials present in this examination. The goal of an intensifying screen is the conversion of X-ray photons to photons in the visible spectrum, knowing that radiographic films are more sensitive to light photons than to X-ray photons. So the screen should be composed of an efficient material for converting x-rays photons in light photons, therefore was made simulations using different materials, thicknesses and positions possible for placing screen in radiographic film in order to find the way more technically feasible. (author)

  11. Comparative dosimetry of prostate brachytherapy with I-125 and Pd-103 seeds via SISCODES/MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Trindade, Bruno Machado; Falcao, Patricia Lima, E-mail: bmtrindade@yahoo.com [Nucleo de Radiacoes Ionizantes - Universidade Federal de Minas Gerais (NRI/UFMG), Belo Horizonte, MG (Brazil); Christovao, Marilia Tavares [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Trindade, Daniela de Fatima Maia [Centro Universitario Una, Belo Horizonte, MG (Brazil); Campos, Tarcisio Passos Ribeiro de [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2012-09-15

    Objective: The present paper is aimed at presenting a comparative dosimetric study of prostate brachytherapy with I-125 and Pd-103 seeds. Materials and Methods: A protocol for both implants with 148 seeds was simulated on a heterogeneous three-dimensional pelvic phantom by means of the SISCODES/MCNP5 codes. Dose-volume histograms on prostate, rectum and bladder, dose indexes D10, D30, D90, D0.5cc, D2cc and D7cc, and representations of the spatial dose distribution were evaluated. Results: For a D90 index equivalent to the prescription dose, the initial activity of each I-125 seed was calculated as 0.42 mCi and of Pd-103 as 0.94 mCi. The maximum dose on the urethra was 90% and 108% of the prescription dose for I-125 and Pd-103, respectively. The D2cc for I-125 was 30 Gy on the rectum and 127 Gy on the bladder; for Pd-103 was 29 Gy on the rectum and 189 Gy on the bladder. The D10 on the pubic bone was 144 Gy for I-125 and 66 Gy for Pd-103. Conclusion: The results indicate that Pd-103 and I-125 implants could deposit the prescribed dose on the target volume. Among the findings of the present study, there is an excessive radiation exposure of the pelvic bones, particularly with the I-125 protocol. (author)

  12. Production of Energetic Light Fragments in CEM, LAQGSM, and MCNP6

    CERN Document Server

    Mashnik, Stepan G; Gudima, Konstantin K; Sierk, Arnold J; Bull, Jeffrey S; James, Michael R

    2016-01-01

    We extend the cascade-exciton model (CEM), and the Los Alamos version of the quark-gluon string model (LAQGSM), event generators of the Monte-Carlo N-particle transport code version 6 (MCNP6), to describe production of energetic light fragments (LF) heavier than 4He from various nuclear reactions induced by particles and nuclei at energies up to about 1 TeV/nucleon. In these models, energetic LF can be produced via Fermi break-up, preequilibrium emission, and coalescence of cascade particles. Initially, we study several variations of the Fermi break-up model and choose the best option for these models. Then, we extend the modified exciton model (MEM) used by these codes to account for a possibility of multiple emission of up to 66 types of particles and LF (up to 28Mg) at the preequilibrium stage of reactions. Then, we expand the coalescence model to allow coalescence of LF from nucleons emitted at the intranuclear cascade stage of reactions and from lighter clusters, up to fragments with mass numbers A < ...

  13. Correction Method of Considering Resonance Elastic Scattering in MCNP%在 MCNP 中考虑共振弹性散射的修正方法

    Institute of Scientific and Technical Information of China (English)

    贺清明; 曹良志; 吴宏春; 郑友琦

    2014-01-01

    The free gas model is adopted to consider the thermal scattering effect of the elastic collision between neutron and target .The conventional model assumes that the elastic scattering cross sections are constant at 0 K , w hich neglects the influence of resonance effect .In order to consider the resonance elastic scattering in the free gas model ,the Doppler broadening rejection correction (DBRC ) method was applied to correct the free gas model of MCNP .The Mosteller’s Doppler defect benchmark for LWR pin cell was analyzed .The numerical results show that neglect of resonance elastic scattering effect contributes to overestimation of the infinite multiplicative factor to the extent of 40‐100 pcm and 140‐200 pcm for hot zero power and hot full power cases , respectively . T he fuel temperature coefficients are also overestimated 7%‐15% . T he computational time of the newly developed sampling technique was studied and the influence of the resonance elastic scattering effect on the emergent energy distribution w as analyzed .%蒙特卡罗方法采用自由气体模型来考虑中子与靶核的弹性碰撞中的热效应。传统的模型假设绝对零度下的弹性散射截面是常数,忽略了截面的共振效应所带来的影响。为在自由气体模型中考虑共振弹性散射效应,采用多普勒展宽舍弃修正方法,修正了连续能量蒙特卡罗程序MCNP的自由气体模型,并对Mosteller轻水堆多普勒基准题进行了分析。数值结果表明:对于轻水堆,在热态零功率的情况下,忽略共振弹性散射会高估燃料棒的无限介质增殖因数(k∞)40~100 pcm ,热态满功率下高估140~200 pcm ;忽略共振弹性散射给燃料温度系数带来7%~15%正的偏差。同时分析了新的抽样方法对计算时间的影响,以及共振弹性散射效应对中子出射能量分布的影响。

  14. Validation of MCNP6 Version 1.0 with the ENDF/B-VII.1 Cross Section Library for Uranium Metal, Oxide, and Solution Systems on the High Performance Computing Platform Moonlight

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, Bryan Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); MacQuigg, Michael Robert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wysong, Andrew Russell [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-04-21

    In this document, the code MCNP is validated with ENDF/B-VII.1 cross section data under the purview of ANSI/ANS-8.24-2007, for use with uranium systems. MCNP is a computer code based on Monte Carlo transport methods. While MCNP has wide reading capability in nuclear transport simulation, this validation is limited to the functionality related to neutron transport and calculation of criticality parameters such as keff.

  15. SU-E-T-212: Comparison of TG-43 Dosimetric Parameters of Low and High Energy Brachytherapy Sources Obtained by MCNP Code Versions of 4C, X and 5

    Energy Technology Data Exchange (ETDEWEB)

    Zehtabian, M; Zaker, N; Sina, S [Shiraz University, Shiraz, Fars (Iran, Islamic Republic of); Meigooni, A Soleimani [Comprehensive Cancer Center of Nevada, Las Vegas, Nevada (United States)

    2015-06-15

    Purpose: Different versions of MCNP code are widely used for dosimetry purposes. The purpose of this study is to compare different versions of the MCNP codes in dosimetric evaluation of different brachytherapy sources. Methods: The TG-43 parameters such as dose rate constant, radial dose function, and anisotropy function of different brachytherapy sources, i.e. Pd-103, I-125, Ir-192, and Cs-137 were calculated in water phantom. The results obtained by three versions of Monte Carlo codes (MCNP4C, MCNPX, MCNP5) were compared for low and high energy brachytherapy sources. Then the cross section library of MCNP4C code was changed to ENDF/B-VI release 8 which is used in MCNP5 and MCNPX codes. Finally, the TG-43 parameters obtained using the MCNP4C-revised code, were compared with other codes. Results: The results of these investigations indicate that for high energy sources, the differences in TG-43 parameters between the codes are less than 1% for Ir-192 and less than 0.5% for Cs-137. However for low energy sources like I-125 and Pd-103, large discrepancies are observed in the g(r) values obtained by MCNP4C and the two other codes. The differences between g(r) values calculated using MCNP4C and MCNP5 at the distance of 6cm were found to be about 17% and 28% for I-125 and Pd-103 respectively. The results obtained with MCNP4C-revised and MCNPX were similar. However, the maximum difference between the results obtained with the MCNP5 and MCNP4C-revised codes was 2% at 6cm. Conclusion: The results indicate that using MCNP4C code for dosimetry of low energy brachytherapy sources can cause large errors in the results. Therefore it is recommended not to use this code for low energy sources, unless its cross section library is changed. Since the results obtained with MCNP4C-revised and MCNPX were similar, it is concluded that the difference between MCNP4C and MCNPX is their cross section libraries.

  16. Uncertainty analysis in the simulation of an HPGe detector using the Monte Carlo Code MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, Sergio; Pozuelo, Fausto; Querol, Andrea; Verdu, Gumersindo; Rodenas, Jose, E-mail: sergalbe@upv.es [Universitat Politecnica de Valencia, Valencia, (Spain). Instituto de Seguridad Industrial, Radiofisica y Medioambiental (ISIRYM); Ortiz, J. [Universitat Politecnica de Valencia, Valencia, (Spain). Servicio de Radiaciones. Lab. de Radiactividad Ambiental; Pereira, Claubia [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2013-07-01

    A gamma spectrometer including an HPGe detector is commonly used for environmental radioactivity measurements. Many works have been focused on the simulation of the HPGe detector using Monte Carlo codes such as MCNP5. However, the simulation of this kind of detectors presents important difficulties due to the lack of information from manufacturers and due to loss of intrinsic properties in aging detectors. Some parameters such as the active volume or the Ge dead layer thickness are many times unknown and are estimated during simulations. In this work, a detailed model of an HPGe detector and a petri dish containing a certified gamma source has been done. The certified gamma source contains nuclides to cover the energy range between 50 and 1800 keV. As a result of the simulation, the Pulse Height Distribution (PHD) is obtained and the efficiency curve can be calculated from net peak areas and taking into account the certified activity of the source. In order to avoid errors due to the net area calculation, the simulated PHD is treated using the GammaVision software. On the other hand, it is proposed to use the Noether-Wilks formula to do an uncertainty analysis of model with the main goal of determining the efficiency curve of this detector and its associated uncertainty. The uncertainty analysis has been focused on dead layer thickness at different positions of the crystal. Results confirm the important role of the dead layer thickness in the low energy range of the efficiency curve. In the high energy range (from 300 to 1800 keV) the main contribution to the absolute uncertainty is due to variations in the active volume. (author)

  17. Study of radiation dose attenuation by skull bone in head during radiotherapy treatment using MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Menezes, Artur F.; Boia, Leonardo S.; Trombetta, Debora M.; Martins, Maximiano C.; Reis Junior, Juraci P.; Silva, Ademir X., E-mail: ademir@con.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Batista, Delano V.S., E-mail: delano@inca.gov.b [Instituto Nacional do Cancer (INCa), Rio de Janeiro, RJ (Brazil). Dept. de Fisica Medica

    2011-07-01

    In this study the MCNPX code was used to investigate possible influences of the attenuation beam by the surface bone during radiotherapy treatments of the skull. The computer simulation was performed on topographic image obtained from the National Cancer Institute, in Rio de Janeiro, database of patients treated with radiotherapy. The image segmentation process were performed using the SAPDI program developed to this purpose. The segmented image conversion for the input file recognized by MCNPX code was performed by SCAN2MCNP Software. The simulation was done using 10MeV Clinac 2300C spectrum considering two opposite parallel beams, with field size 2x2 and 4x4 cm{sup 2}, incident on a slice located above the eyes, containing two row of detectors positioned on the central region with a radius of 0.03 cm and arranged perpendicular to the radiation beams. After analyze the results, the relative error values in the range of 2 at 4% for the high dose region, and 26 at 37% for the low dose area were found, respectively. These differences were attributed to the radiation field attenuation on the bone surface at the entrance of the beam. It was observed that most situations on the high dose region the beam profile, from more realistic scenarios, became smaller than the one obtained when the tomography image was considered consisting of water. However for the low dose area the profile, obtained of the realistic situation, became higher than the one which was obtained when the tomography image was considered consisting of water. The results showed significant differences between both analyzed cases which show the need to use a correction factor by the treatment planning system used in radiotherapy services when the real chemical composition of patient head is unconsidered during the patient treatment planning. (author)

  18. Evaluation of the Biological Shields of the Secondary Standard Dosimetry Laboratory of Ghana Using MCNP5

    Directory of Open Access Journals (Sweden)

    P. Deatanyah

    2012-03-01

    Full Text Available The primary objective with radiation sources and facilities is the protection of both radiation workers and the general public. The biological shields of the Secondary Standard Dosimetry Laboratory of the Radiation Protection Institute (RPI Ghana had been evaluated for a collimated isotropic cesium-137 source for calibration purpose using MCNP5 code. The dose rate at supervised areas ranged from 0.57 to 8.35 :Sv/h and 0.26 to 10.22 :Sv/h at control areas when the source was panoramic. When the source was collimated, the dose rate ranged from 0.05 to 0.30 :Sv/h at supervised areas and 0.23 to 8.88 :Sv/h at control areas for 22.2 GBq of the cesium-137 source. The scatter contribution from the surfaces of the walls and roofs were also accounted for. The scatter radiation in the room decreased to 400 :Sv/h when the source was first collimated and to 3.5 :Sv/h when the source was further collimated. These results agreed quite well with experimental measurement. To effectively protect the staff, a narrow beam of 1.2 cm diameter which was defined at 1.0 m by the total surface of the ISO slab phantom was recommended to reduce the dose rate to less than 1.5 :Sv/h outside the calibration bunker even when the current activity is doubled. It was concluded that the 4.7 cm diameter of the existing narrow beam should be decreased to 1.2 cm by further collimation of the beam.

  19. Dose calculation for {sup 40}K ingestion in samples of beans using spectrometry and MCNP; Calculo de dose devido a ingestao de {sup 40}K em amostras de feijao utilizando espectrometria e MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Garcez, R.W.D.; Lopes, J.M.; Silva, A.X., E-mail: marqueslopez@yahoo.com.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/PEN/UFRJ), Rio de Janeiro, RJ (Brazil). Centro de Tecnologia; Domingues, A.M. [Universidade Federal Fluminense (UFF), Niteroi, RJ (Brazil). Instituto de Fisica; Lima, M.A.F. [Universidade Federal Fluminense (UFF), Niteroi, RJ (Brazil). Instituto de Biologia

    2014-07-01

    A method based on gamma spectroscopy and on the use of voxel phantoms to calculate dose due to ingestion of {sup 40}K contained in bean samples are presented in this work. To quantify the activity of radionuclide, HPGe detector was used and the data entered in the input file of MCNP code. The highest value of equivalent dose was 7.83 μSv.y{sup -1} in the stomach for white beans, whose activity 452.4 Bq.Kg{sup -1} was the highest of the five analyzed. The tool proved to be appropriate when you want to calculate the dose in organs due to ingestion of food. (author)

  20. Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP.

    Science.gov (United States)

    Abrefah, R G; Sogbadji, R B M; Ampomah-Amoako, E; Birikorang, S A; Odoi, H C; Nyarko, B J B

    2011-01-01

    The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1.

  1. Neutronic analysis for core conversion (HEU–LEU) of the low power research reactor using the MCNP4C code

    OpenAIRE

    Aldawahra Saadou; Khattab Kassem; Saba Gorge

    2015-01-01

    Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR) have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad) and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad) cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The propos...

  2. MCNP Calculations for the Shielding Design of a Beam Tube to Be Installed at the Portuguese Research Reactor

    Science.gov (United States)

    Gonçalves, I. F.; Ramalho, A. G.; Gonçalves, I. C.; Salgado, J.

    The work presented concerns the calculation of the external biological shielding for a neutron beam tube that will be installed at the Portuguese Research Reactor, RPI. This tube will have enough versatility to be used in fields so different as the analysis of the composition of samples or research work in Boron Neutron Capture Therapy, BNCT. The calculation was made by using the MCNP code. This code is a well validated and widely used code, and has therefore become an important tool in the design and optimisation work of experiences related to neutrons and gamma radiation.

  3. Continuous energy cross section library for MCNP/MCNPX based on JENDL high energy file 2007; FXJH7

    OpenAIRE

    佐々 敏信; 菅原 隆徳; 小迫 和明; 深堀 智生

    2008-01-01

    The latest JENDL High Energy File (JENDL/HE) was released in 2007 to respond the requirements of reaction data in high energy range up to several GeV to design accelerator facilities such as accelerator-driven systems and research complex like J-PARC. To apply the JENDL/HE-2007 file to the design study, the cross section library of FXJH7 series was constructed from the JENDL/HE file for the calculation using MCNP and MCNPX codes which are widely used in the field of nuclear reactors, fusion r...

  4. MCID: personalized dosimetric tool to simulate voxelized studies using MCNP5; MCID: herramienta dosimetrica personalizada para simular estudios voxelizados con MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Gil, Alex Vergara [Centro de Proteccion e Higiene de las Radiaciones (CPHR), La Habana (Cuba); Perez, Marco A. Coca; Aroche, Leonel A. Torres, E-mail: mcoca@infomed.sld.cu, E-mail: leonel@infomed.sld.cu [Centro de Investigaciones Clinicas (CIC), La Habana (Cuba); Pacilio, Massimiliano, E-mail: mpacilio@scamilloforlanini.rm.it [Hospital S. Camillo Forlanini (AOSCF), Roma (Italy). Departmento de Fisica Medica

    2013-07-01

    The purpose of this paper is to present the MCID software, a tool for calculating specific absorbed dose of patients in nuclear medicine, based on Monte Carlo simulation. This paper evaluates new clinical cases and new phantoms whose results validate the methodology implemented in MCID, which has followed a process of incorporating new materials, image processing in DICOM and Analyze format, a module of regions of interest and improvements in user interface. Now it has a tool to calculate the patient-specific absorbed doses in nuclear medicine that can be applied in clinical practice.

  5. MCNP6 Study of Fragmentation Products from 112Sn + 112Sn and 124Sn + 124Sn at 1 GeV/nucleon

    CERN Document Server

    Mashnik, Stepan G

    2013-01-01

    Isotope production cross sections from 112Sn + 112Sn and 124Sn + 124Sn reactions at 1 GeV/nucleon, which were measured recently at GSI using the heavy-ion accelerator SIS18 and the Fragment Separator (FRS), have been analyzed with the latest Los Alamos Monte-Carlo transport code MCNP6 using the LAQGSM03.03 event generator. MCNP6 reproduces reasonably well all the measured cross sections. Comparison of the MCNP6 results with the measured data and with calculations by a modification of the Los Alamos version of the Quark-Gluon String Model allowing for multifragmentation processes in the framework of the Statistical Multifragmentation Model (SMM) by Botvina and coauthors, as realized in the code LAQGSM03.S1, does not suggest unambiguous evidence of a multifragmentation signature.

  6. Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

    Energy Technology Data Exchange (ETDEWEB)

    Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

    2017-04-01

    Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

  7. Calculation of the X-Ray Spectrum of a Mammography System with Various Voltages and Different Anode-Filter Combinations Using MCNP Code

    OpenAIRE

    Lida Gholamkar; Mahdi Sadeghi; Ali Asghar Mowlavi; Mitra Athari

    2016-01-01

    Introduction One of the best methods in the diagnosis and control of breast cancer is mammography. The importance of mammography is directly related to its value in the detection of breast cancer in the early stages, which leads to a more effective treatment. The purpose of this article was to calculate the X-ray spectrum in a mammography system with Monte Carlo codes, including MCNPX and MCNP5. Materials and Methods The device, simulated using the MCNP code, was Planmed Nuance digital mammog...

  8. Gamma ray shielding study of barium-bismuth-borosilicate glasses as transparent shielding materials using MCNP-4C code, XCOM program, and available experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Bagheri, Reza; Yousefinia, Hassan [Nuclear Fuel Cycle Research School (NFCRS), Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of); Moghaddam, Alireza Khorrami [Radiology Department, Paramedical Faculty, Mazandaran University of Medical Sciences, Sari (Iran, Islamic Republic of)

    2017-02-15

    In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and 10th value layer values of barium-bismuth-borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium-bismuth-borosilicate glasses have good gamma ray shielding properties from the shielding point of view.

  9. Varian 2100C/D Clinac 18 MV photon phase space file characterization and modeling by using MCNP Code

    Science.gov (United States)

    Ezzati, Ahad Ollah

    2015-07-01

    Multiple points and a spatial mesh based surface source model (MPSMBSS) was generated for 18MV Varian 2100 C/D Clinac phase space file (PSF) and implemented in MCNP code. The generated source model (SM) was benchmarked against PSF and measurements. PDDs and profiles were calculated using the SM and original PSF for different field sizes from 5 × 5 to 20 × 20 cm2. Agreement was within 2% of the maximum dose at 100cm SSD for beam profiles at the depths of 4cm and 15cm with respect to the original PSF. Differences between measured and calculated points were less than 2% of the maximum dose or 2mm distance to agreement (DTA) at 100 cm SSD. Thus it can be concluded that the modified MCNP code can be used for radiotherapy calculations including multiple source model (MSM) and using the source biasing capability of MPSMBSS can increase the simulation speed up to 3600 for field sizes smaller than 5 × 5 cm2.

  10. Photopeak efficiency response function of an underwater gamma-ray NaI(Tl) detector using MCNP-X

    Energy Technology Data Exchange (ETDEWEB)

    Salgado, William L., E-mail: william.otero@hotmail.com [Instituto Federal do Rio de Janeiro (IFRJ), RJ (Brazil); Silva, Ademir X., E-mail: ademir@con.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (PEN/COPPE-DNC/UFRJ/EE/CT), Rio de Janeiro, RJ (Brazil); Salgado, Cesar M., E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    This work presents a study to calculate the response function of a 1.5″ x 1″ NaI(Tl) scintillation detector when it is used in the marine environment in the energy range from 20 keV to 662 keV. The method takes into account both the scattering of photons in the water and the detection mechanism of the detector. In addition, the calculation of the response function of the whole system is essential for suppressing the background of the measurement and for estimating the concentration of the involved radionuclides, especially given the greater probability of primary gamma photons undergoing multiple scattering events before they interact with the detector. The experimental photopeak efficiency measurements for point sources were compared with the simulated results under the same conditions of the experimental setup to validate the simulation of the detector. Monte Carlo simulations were performed using the MCNP-X code for the investigation of gamma-ray absorption in water in different brines. The energy resolution curve was used to improve the response of the mathematical simulation of the detector. The detector’s simulation was based on information obtained from the gammagraphy technique. Both dimensions and materials were used for the calculation with the MCNP-X code. The photopeak efficiency of a NaI(Tl) detector for different radionuclides in the aquatic environment with different salinities was calculated. (author)

  11. Simulation of irradiation exposure of electronic devices due to heavy ion therapy with Monte Carlo Code MCNP6

    Science.gov (United States)

    Lapins, Janis; Guilliard, Nicole; Bernnat, Wolfgang; Buck, Arnulf

    2017-09-01

    During heavy ion irradiation therapy the patient has to be located exactly at the right position to make sure that the Bragg peak occurs in the tumour. The patient has to be moved in the range of millimetres to scan the ill tissue. For that reason a special table was developed which allows exact positioning. The electronic control can be located outside the surgery. But that has some disadvantage for the construction. To keep the system compact it would be much more comfortable to put the electronic control inside the surgery. As a lot of high energetic secondary particles are produced during the therapy causing a high dose in the room it is important to find positions with low dose rates. Therefore, investigations are needed where the electronic devices should be located to obtain a minimum of radiation, help to prevent the failure of sensitive devices. The dose rate was calculated for carbon ions with different initial energy and protons over the entire therapy room with Monte Carlo particle tracking using MCNP6. The types of secondary particles were identified and the dose rate for a thin silicon layer and an electronic mixture material was determined. In addition, the shielding effect of several selected material layers was calculated using MCNP6.

  12. Determination of the detection efficiency of a HPGe detector by means of the MCNP 4A simulation code; Determinacion de la eficiencia de deteccion de un detector HPGe mediante el codigo de simulacion MCNP 4A

    Energy Technology Data Exchange (ETDEWEB)

    Leal, B. [Centro Regional de Estudios Nucleares, A.P. 579C, 98068 Zacatecas (Mexico)

    2004-07-01

    In the majority of the laboratories, the calibration in efficiency of the detector is carried out by means of the standard sources measurement of gamma photons that have a determined activity, or for matrices that contain a variety of radionuclides that can embrace the energy range of interest. Given the experimental importance that has the determination from the curves of efficiency to the effects of establishing the quantitative results, is appealed to the simulation of the response function of the detector used in the Regional Center of Nuclear Studies inside the energy range of 80 keV to 1400 keV varying the density of the matrix, by means of the application of the Monte Carlo code MCNP-4A. The adjustment obtained shows an acceptance grade in the range of 100 to 600 keV, with a smaller percentage discrepancy to 5%. (Author)

  13. Criticality calculations of a generic fuel container for fuel assemblies PWR, by means of the code MCNP; Calculos de criticidad de un contenedor de combustible generico para ensambles combustibles PWR, mediante el codigo MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Vargas E, S.; Esquivel E, J.; Ramirez S, J. R., E-mail: samuel.vargas@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    The purpose of the concept of burned consideration (Burn-up credit) is determining the capacity of the calculation codes, as well as of the nuclear data associates to predict the isotopic composition and the corresponding neutrons effective multiplication factor in a generic container of spent fuel during some time of relevant storage. The present work has as objective determining this capacity of the calculation code MCNP in the prediction of the neutrons effective multiplication factor for a fuel assemblies arrangement type PWR inside a container of generic storage. The calculations are divided in two parts, the first, in the decay calculations with specified nuclide concentrations by the reference for a pressure water reactor (PWR) with enriched fuel to 4.5% and a discharge burned of 50 GW d/Mtu. The second, in criticality calculations with isotopic compositions dependent of the time for actinides and important fission products, taking 30 time steps, for two actinide groups and fission products. (Author)

  14. Introduction to the simulation with MCNP Monte Carlo code and its applications in Medical Physics; Introduccion a la simulacion con el codigo de Monte Carlo MCNP y sus aplicaciones en Fisica Medica

    Energy Technology Data Exchange (ETDEWEB)

    Parreno Z, F.; Paucar J, R.; Picon C, C. [Instituto Peruano de Energia Nuclear, Av. Canada 1470, San Borja, Lima 41 (Peru)

    1998-12-31

    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. Comparative analysis of results between CASMO, MCNP and Serpent for a suite of Benchmark problems on BWR reactors; Analisis comparativo de resultados entre CASMO, MCNP y SERPENT para una suite de problemas Benchmark en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J. V.; Vargas E, S.; Gomez T, A. M. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Reyes F, M. del C.; Del Valle G, E., E-mail: vicente.xolocostli@inin.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, UP - Adolfo Lopez Mateos, Edif. 9, 07738 Mexico D. F. (Mexico)

    2014-10-15

    In this paper a comparison is made in analyzing the suite of Benchmark problems for reactors type BWR between CASMO-4, MCNP6 and Serpent code. The Benchmark problem consists of two different geometries: a fuel cell of a pin and assembly type BWR. To facilitate the study of reactors physics in the fuel pin their nuclear characteristics are provided to detail, such as burnt dependence, the reactivity of selected nuclide, etc. With respect to the fuel assembly, the presented results are regarding to infinite multiplication factor for burning different steps and different vacuum conditions. Making the analysis of this set of Benchmark problems provides comprehensive test problems for the next fuels generation of BWR reactors with high extended burned. It is important to note that when making this comparison the purpose is to validate the methodologies used in modeling for different operating conditions, if the case is of other BWR assembly. The results will be within a range with some uncertainty, considering that does not depend on code that is used. Escuela Superior de Fisica y Matematicas of Instituto Politecnico Nacional (IPN (Mexico) has accumulated some experience in using Serpent, due to the potential of this code over other commercial codes such as CASMO and MCNP. The obtained results for the infinite multiplication factor are encouraging and motivate the studies to continue with the generation of the X S of a core to a next step a respective nuclear data library is constructed and this can be used by codes developed as part of the development project of the Mexican Analysis Platform of Nuclear Reactors AZTLAN. (Author)

  16. Comparison of MCNP4B and WIMS-AECL calculations of coolant-void-reactivity effects for uniform lattices of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    1999-05-01

    This paper compares the results of coolant-void reactivity (CVR) reactor-physics calculations performed using the Monte Carlo N-particle transport code, MCNP version 4B, with those obtained using Atomic Energy of Canada Limited's (AECL's) latest version of the Winfrith improved multigroup scheme (WIMS) code, WIMS-AECL version 2-5c. Cross sections derived from the evaluated nuclear data file version B-VI (ENDF/B-VI) are used for both the WIMS-AECL and MCNP4B calculations. The comparison is made for uniform lattices at room temperature containing either fresh natural uranium or mixed oxide (MOX) 37-element CANDU fuel. The MOX fuel composition corresponds roughly to that of irradiated CANDU fuel at a burnup of about 4500 MWd/tU. The level of agreement between the CVR predictions of WIMS-AECL and MCNP4B is studied as a function of lattice buckling (a measure of the curvature of the neutron-flux distribution) over the range from 0.0 to 4.1 m{sup -2} . For the cases studied, it is found that the absolute k values calculated by WIMS-AECL are higher than those of MCNP4B by several mk (1 mk is a change of 0.001 in k), amounts that depend on the fuel type being modelled and the particular cross-section data used. However, the agreement between WIMS-AECL and MCNP4B is much better for the CVR (i.e., the {delta}k on coolant voiding), and is relatively insensitive to the fuel type. (author)

  17. Comparison of MCNP4B and WIMS-AECL calculations of coolant-void-reactivity effects for uniform lattices of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1999-07-01

    This paper compares the results of coolant-void reactivity (CVR) reactor-physics calculations performed using the Monte Carlo N-particle transport code, MCNP version 4B, with those obtained using Atomic Energy of Canada Limited's (AECL's) latest version of the Winfrith improved multigroup scheme (WIMS) code, WIMS-AECL version 2-5c. Cross sections derived from the evaluated nuclear data file version B-VT (ENDF/B-VI) are used for both the WIMS-AECL and MCNP4B calculations. The comparison is made for uniform lattices at room temperature containing either fresh natural uranium or mixed oxide (MOX) 37-element CANDU fuel. The MOX fuel composition corresponds roughly to that of irradiated CANDU fuel at a burnup of about 4500 MWd/tU. The level of agreement between the CVR predictions of WIMS-AECL and MCNP4B is studied as a function of lattice buckling (a measure of the curvature of the neutron-flux distribution) over the range from 0.0 to 4.1 m{sup -2}. For the cases studied, it is found that the absolute keff values calculated by WIMS-AECL are higher than those of MCNP4B by several mk (1 mk is a change of 0.001 in keff), amounts that depend on the fuel type being modelled and the particular cross-section data used. However, the agreement between WIMS-AECL and MCNP4B is much better for the CVR (i.e., the {delta}keff on coolant voiding), and is relatively insensitive to the fuel type. (author)

  18. EGS4 and MCNP4b MC Simulation of a Siemens KD2 Accelerator in 6 MV Photon Mode

    CERN Document Server

    Chaves, A; Fragoso, M; Lopes, C; Oliveira, C; Peralta, L; Rodrigues, P; Seco, J; Trindade, A

    2001-01-01

    The geometry of a Siemens Mevatron KD2 linear accelerator in 6 MV photon mode was modeled with EGS4 and MCNP4b. Energy spectra and other phase space distributions have been extensively compared in different plans along the beam line. The differences found have been evaluated both qualitative and quantitatively. The final aim was that both codes, running in different operating systems and with a common set of simulation conditions, met the requirement of fitting the experimental depth dose curves and dose profiles, measured in water for different field sizes. Whereas depth dose calculations are in a certain extent insensible to some simulation parameters like electron nominal energy, dose profiles have revealed to be a much better indicator to appreciate that feature. Fine energy tuning has been tried and the best fit was obtained for a nominal electron energy of 6.15 MeV.

  19. Monte Carlo Simulation of Electron Beams for Radiotherapy - EGS4, MCNP4b and GEANT3 Intercomparison

    CERN Document Server

    Trindade, A; Alves, C M; Chaves, A; Lopes, C; Oliveira, C; Peralta, L

    2000-01-01

    In medical radiation physics, an increasing number of Monte Carlo codes are being used, which requires intercomparison between them to evaluated the accuracy of the simulated results against benchmark experiments. The Monte Carlo code EGS4, commonly used to simulate electron beams from medical linear accelerators, was compared with GEANT3 and MCNP4b. Intercomparison of electron energy spectra, angular and spatial distribution were carried out for the Siemens KD2 linear accelerator, at beam energies of 10 and 15 MeV for a field size of 10x10 cm2. Indirect validation was performed against electron depth doses curves and beam profiles measured in a MP3-PTW water phantom using a Markus planar chamber. Monte Carlo isodose lines were reconstructed and compared to those from commercial treatment planning systems (TPS's) and with experimental data.

  20. Photon attenuation coefficients of Heavy-Metal Oxide glasses by MCNP code, XCOM program and experimental data: A comparison study

    Science.gov (United States)

    El-Khayatt, A. M.; Ali, A. M.; Singh, Vishwanath P.

    2014-01-01

    The mass attenuation coefficients, μ/ρ, total interaction cross-section, σt, and mean free path (MFP) of some Heavy Metal Oxides (HMO) glasses, with potential applications as gamma ray shielding materials, have been investigated using the MCNP-4C code. Appreciable variations are noted for all parameters by changing the photon energy and the chemical composition of HMO glasses. The numerical simulations parameters are compared with experimental data wherever possible. Comparisons are also made with predictions from the XCOM program in the energy region from 1 keV to 100 MeV. Good agreement noticed indicates that the chosen Monte Carlo method may be employed to make additional calculations on the photon attenuation characteristics of different glass systems, a capability particularly useful in cases where no analogous experimental data exist.

  1. Calibration with MCNP of NaI detector for the determination of natural radioactivity levels in the field.

    Science.gov (United States)

    Cinelli, Giorgia; Tositti, Laura; Mostacci, Domiziano; Baré, Jonathan

    2016-05-01

    In view of assessing natural radioactivity with on-site quantitative gamma spectrometry, efficiency calibration of NaI(Tl) detectors is investigated. A calibration based on Monte Carlo simulation of detector response is proposed, to render reliable quantitative analysis practicable in field campaigns. The method is developed with reference to contact geometry, in which measurements are taken placing the NaI(Tl) probe directly against the solid source to be analyzed. The Monte Carlo code used for the simulations was MCNP. Experimental verification of the calibration goodness is obtained by comparison with appropriate standards, as reported. On-site measurements yield a quick quantitative assessment of natural radioactivity levels present ((40)K, (238)U and (232)Th). On-site gamma spectrometry can prove particularly useful insofar as it provides information on materials from which samples cannot be taken.

  2. Evaluation of a 50-MV photon therapy beam from a racetrack microtron using MCNP4B Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Gudowska, I.; Svensson, R. [Karolinska Inst. (Sweden). Dept. of Medical Radiation Physics]|[Huddinge Univ. Hospital, Stockholm (Sweden). Dept. of Medical Physics; Sorcini, B. [Karolinska Inst. (Sweden). Dept. of Medical Radiation Physics]|[Stockholm Univ. (Sweden)

    2001-07-01

    High energy photon therapy beam from the 50 MV racetrack microtron has been evaluated using the Monte Carlo code MCNP4B. The spatial and energy distribution of photons, radial and depth dose distributions in the phantom are calculated for the stationary and scanned photon beams from different targets. The calculated dose distributions are compared to the experimental data using a silicon diode detector. Measured and calculated depth-dose distributions are in fairly good agreement, within 2-3% for the positions in the range 2-30 cm in the phantom, whereas the larger discrepancies up to 10% are observed in the dose build-up region. For the stationary beams the differences in the calculated and measured radial dose distributions are about 2-10%. (orig.)

  3. The performance test of anti-scattering x-ray grid with inclined shielding material by MCNP code simulation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Jun Woo; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2016-06-15

    The scattered photons cause reduction of the contrast of radiographic image and it results in the degradation of the quality of the image. In order to acquire better quality image, an anti-scattering x-ray gird should be equipped in radiography system. The X-ray anti-scattering grid of the inclined type based on the hybrid concept for that of parallel and focused type was tested by MCNP code. The MCNPX 2.7.0 was used for the simulation based test. The geometry for the test was based on the IEC 60627 which was an international standard for diagnostic X-ray imaging equipment-Characteristics of general purpose and mammographic anti-scatter grids. The performance of grids with four inclined shielding material types was compared with that of the parallel type. The grid with completely tapered type the best performance where there were little performance difference according to the degree of inclination.

  4. Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

    Energy Technology Data Exchange (ETDEWEB)

    Pecchia, M.; D' Auria, F. [San Piero A Grado Nuclear Research Group GRNSPG, Univ. of Pisa, via Diotisalvi, 2, 56122 - Pisa (Italy); Mazzantini, O. [Nucleo-electrica Argentina Societad Anonima NA-SA, Buenos Aires (Argentina)

    2012-07-01

    Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)

  5. Monte Carlo Simulation of Electron Beams for Radiotherapy - EGS4, MCNP4b and GEANT3 Intercomparison

    Science.gov (United States)

    Trindade, A.; Rodrigues, P.; Alves, C.; Chaves, A.; Lopes, M. C.; Oliveira, C.; Peralta, L.

    In medical radiation physics, an increasing number of Monte Carlo codes are being used, which requires intercomparison between them to evaluated the accuracy of the simulated results against benchmark experiments. The Monte Carlo code EGS4, commonly used to simulate electron beams from medical linear accelerators, was compared with GEANT3 and MCNP4b. Intercomparison of electron energy spectra, angular and spatial distribution were carried out for the Siemens KD2 linear accelerator, at beam energies of 10 and 15 MeV for a field size of 10x10 cm2. Indirect validation was performed against electron depth doses curves and beam profiles measured in a MP3-PTW water phantom using a Markus planar chamber. Monte Carlo isodose lines were reconstructed and compared to those from commercial treatment planning systems (TPS's) and with experimental data.

  6. High-fidelity MCNP modeling of a D-T neutron generator for active interrogation of special nuclear material

    Science.gov (United States)

    Katalenich, Jeff; Flaska, Marek; Pozzi, Sara A.; Hartman, Michael R.

    2011-10-01

    Fast and robust methods for interrogation of special nuclear material (SNM) are of interest to many agencies and institutions in the United States. It is well known that passive interrogation methods are typically sufficient for plutonium identification because of a relatively high neutron production rate from 240Pu [1]. On the other hand, identification of shielded uranium requires active methods using neutron or photon sources [2]. Deuterium-deuterium (2.45 MeV) and deuterium-tritium (14.1 MeV) neutron-generator sources have been previously tested and proven to be relatively reliable instruments for active interrogation of nuclear materials [3,4]. In addition, the newest generators of this type are small enough for applications requiring portable interrogation systems. Active interrogation techniques using high-energy neutrons are being investigated as a method to detect hidden SNM in shielded containers [4,5]. Due to the thickness of some containers, penetrating radiation such as high-energy neutrons can provide a potential means of probing shielded SNM. In an effort to develop the capability to assess the signal seen from various forms of shielded nuclear materials, the University of Michigan Neutron Science Laboratory's D-T neutron generator and its shielding were accurately modeled in MCNP. The generator, while operating at nominal power, produces approximately 1×10 10 neutrons/s, a source intensity which requires a large amount of shielding to minimize the dose rates around the generator. For this reason, the existing shielding completely encompasses the generator and does not include beam ports. Therefore, several MCNP simulations were performed to estimate the yield of uncollided 14.1-MeV neutrons from the generator for active interrogation experiments. Beam port diameters of 5, 10, 15, 20, and 25 cm were modeled to assess the resulting neutron fluxes. The neutron flux outside the beam ports was estimated to be approximately 2×10 4 n/cm 2 s.

  7. The User Experience

    Science.gov (United States)

    Schmidt, Aaron

    2010-01-01

    User experience (UX) is about arranging the elements of a product or service to optimize how people will interact with it. In this article, the author talks about the importance of user experience and discusses the design of user experiences in libraries. He first looks at what UX is. Then he describes three kinds of user experience design: (1)…

  8. User Innovation Management

    DEFF Research Database (Denmark)

    Kanstrup, Anne Marie; Bertelsen, Pernille

    User Innovation Management (UIM) is a method for fo-opereation with users in innovation projects. The UIM method emphasizes the practice of a participatorty attitude.......User Innovation Management (UIM) is a method for fo-opereation with users in innovation projects. The UIM method emphasizes the practice of a participatorty attitude....

  9. User Behavior Analytics

    Energy Technology Data Exchange (ETDEWEB)

    Turcotte, Melissa [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Moore, Juston Shane [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-28

    User Behaviour Analytics is the tracking, collecting and assessing of user data and activities. The goal is to detect misuse of user credentials by developing models for the normal behaviour of user credentials within a computer network and detect outliers with respect to their baseline.

  10. Franklin: User Experiences

    Energy Technology Data Exchange (ETDEWEB)

    National Energy Research Supercomputing Center; He, Yun (Helen); Kramer, William T.C.; Carter, Jonathan; Cardo, Nicholas

    2008-05-07

    The newest workhorse of the National Energy Research Scientific Computing Center is a Cray XT4 with 9,736 dual core nodes. This paper summarizes Franklin user experiences from friendly early user period to production period. Selected successful user stories along with top issues affecting user experiences are presented.

  11. The User Experience

    Science.gov (United States)

    Schmidt, Aaron

    2010-01-01

    User experience (UX) is about arranging the elements of a product or service to optimize how people will interact with it. In this article, the author talks about the importance of user experience and discusses the design of user experiences in libraries. He first looks at what UX is. Then he describes three kinds of user experience design: (1)…

  12. Development And Implementation Of Photonuclear Cross-section Data For Mutually Coupled Neutron-photon Transport Calculations In The Monte Carlo N-particle (mcnp) Radiation Transport Code

    CERN Document Server

    White, M C

    2000-01-01

    The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron tran...

  13. Implementation and qualification of MCNP 5 through the intercomparison with the benchmark for the calculation of critical systems Godiva and Jezebel; Implementacao e qualificacao do MCNP5 atraves da intercomparacao com o benchmark para o calculo dos sistemas criticos Godiva e Jezebel

    Energy Technology Data Exchange (ETDEWEB)

    Lara, Rafael G.; Maiorino, Jose R., E-mail: rafael.lara@aluno.ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Sociais Aplicadas

    2013-07-01

    This work aimed at the implementation and qualification of MCNP code in a supercomputer of the Universidade Federal do ABC, so that may be available a next-generation simulation tool for precise calculations of nuclear reactors and systems subject to radiation. The implementation of this tool will have multidisciplinary applications, covering various areas of engineering (nuclear, aerospace, biomedical), radiation physics and others.

  14. User evaluation in practice

    DEFF Research Database (Denmark)

    Krogstrup, Hanne Kathrine

    2004-01-01

    The BIKVA-model (brugerinddragelse i kvalitetsvurdering) or in english UPQA (User Participation in Quality Assessment) are presented......The BIKVA-model (brugerinddragelse i kvalitetsvurdering) or in english UPQA (User Participation in Quality Assessment) are presented...

  15. User Privacy and Empowerment:

    DEFF Research Database (Denmark)

    Dhotre, Prashant Shantaram; Olesen, Henning; Khajuria, Samant

    2017-01-01

    of personal information and its manage-ment. Thus, empowering users and enhancing awareness are essential to compre-hending the value of secrecy. This paper also introduced latest advances in the domain of privacy issues like User Managed Access (UMA) can state suitable requirements for user empowerment...... and will cater to redefine the trustworthy relationship between service providers and users. Subsequently, this paper con-cludes with suggestions for providing empowerment to the user and developing user-centric, transparent business models.......Today, the service providers are capable of assembling a huge measure of user information using Big data techniques. For service providers, user infor-mation has become a vital asset. The present business models are attentive to collect extensive users’ information to extract useful knowledge...

  16. Estimation of doses received by operators in the 1958 RB reactor accident using the MCNP5 computer code simulation

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2012-01-01

    Full Text Available A numerical simulation of the radiological consequences of the RB reactor reactivity excursion accident, which occurred on October 15, 1958, and an estimation of the total doses received by the operators were run by the MCNP5 computer code. The simulation was carried out under the same assumptions as those used in the 1960 IAEA-organized experimental simulation of the accident: total fission energy of 80 MJ released in the accident and the frozen positions of the operators. The time interval of exposure to high doses received by the operators has been estimated. Data on the RB1/1958 reactor core relevant to the accident are given. A short summary of the accident scenario has been updated. A 3-D model of the reactor room and the RB reactor tank, with all the details of the core, created. For dose determination, 3-D simplified, homogenised, sexless and faceless phantoms, placed inside the reactor room, have been developed. The code was run for a number of neutron histories which have given a dose rate uncertainty of less than 2%. For the determination of radiation spectra escaping the reactor core and radiation interaction in the tissue of the phantoms, the MCNP5 code was run (in the KCODE option and “mode n p e”, with a 55-group neutron spectra, 35-group gamma ray spectra and a 10-group electron spectra. The doses were determined by using the conversion of flux density (obtained by the F4 tally in the phantoms to doses using factors taken from ICRP-74 and from the deposited energy of neutrons and gamma rays (obtained by the F6 tally in the phantoms’ tissue. A rough estimation of the time moment when the odour of ozone was sensed by the operators is estimated for the first time and given in Appendix A.1. Calculated total absorbed and equivalent doses are compared to the previously reported ones and an attempt to understand and explain the reasons for the obtained differences has been made. A Root Cause Analysis of the accident was done and

  17. Calculation of the X-Ray Spectrum of a Mammography System with Various Voltages and Different Anode-Filter Combinations Using MCNP Code

    Directory of Open Access Journals (Sweden)

    Lida Gholamkar

    2016-09-01

    Full Text Available Introduction One of the best methods in the diagnosis and control of breast cancer is mammography. The importance of mammography is directly related to its value in the detection of breast cancer in the early stages, which leads to a more effective treatment. The purpose of this article was to calculate the X-ray spectrum in a mammography system with Monte Carlo codes, including MCNPX and MCNP5. Materials and Methods The device, simulated using the MCNP code, was Planmed Nuance digital mammography device (Planmed Oy, Finland, equipped with an amorphous selenium detector. Different anode/filter materials, such as molybdenum-rhodium (Mo-Rh, molybdenum-molybdenum (Mo-Mo, tungsten-tin (W-Sn, tungsten-silver (W-Ag, tungsten-palladium (W-Pd, tungsten-aluminum (W-Al, tungsten-molybdenum (W-Mo, molybdenum-aluminum (Mo-Al, tungsten-rhodium (W-Rh, rhodium-aluminum (Rh-Al, and rhodium-rhodium (Rh-Rh, were simulated in this study. The voltage range of the X-ray tube was between 24 and 34 kV with a 2 kV interval. Results The charts of changing photon flux versus energy were plotted for different types of anode-filter combinations. The comparison with the findings reported by others indicated acceptable consistency. Also, the X-ray spectra, obtained from MCNP5 and MCNPX codes for W-Ag and W-Rh combinations, were compared. We compared the present results with the reported data of MCNP4C and IPEM report No. 78 for Mo-Mo, Mo-Rh, and W-Al combinations. Conclusion The MCNPX calculation outcomes showed acceptable results in a low-energy X-ray beam range (10-35 keV. The obtained simulated spectra for different anode/filter combinations were in good conformity with the finding of previous research.

  18. Analysis of the variation of the attenuation curve in function of the radiation field size for k Vp X-ray beams using the MCNP-5C code

    Energy Technology Data Exchange (ETDEWEB)

    Fernandes, Marco A.R., E-mail: marco@cetea.com.b, E-mail: marfernandes@fmb.unesp.b [Universidade Estadual Paulista Julio de Mesquita Filho (FMB/UNESP), Botucatu, SP (Brazil). Fac. de Medicina; Ribeiro, Victor A.B. [Universidade Estadual Paulista Julio de Mesquita Filho (IBB/UNESP), Botucatu, SP (Brazil). Inst. de Biociencias; Viana, Rodrigo S.S.; Coelho, Talita S. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The paper illustrates the use of the Monte Carlo method, MCNP-5C code, to analyze the attenuation curve behavior of the 50 kVp radiation beam from superficial radiotherapy equipment as Dermopan2 model. The simulations seek to verify the MCNP-5C code performance to study the variation of the attenuation curve - percentage depth dose (PDD) curve - in function of the radiation field dimension used at radiotherapy of skin tumors with 50 kVp X-ray beams. The PDD curve was calculated for six different radiation field sizes with circular geometry of 1.0, 2.0, 3.0, 4.0, 5.0 and 6.0 cm in diameter. The radiation source was modeled considering a tungsten target with inclination 30 deg, focal point of 6.5 mm in diameter and energy beam of 50 kVp; the X-ray spectrum was calculated with the MCNP-5C code adopting total filtration (beryllium window of 1 mm and aluminum additional filter of 1 mm). The PDD showed decreasing behavior with the attenuation depth similar what is presented on the literature. There was not significant variation at the PDD values for the radiation field between 1.0 and 4.0 cm in diameter. The differences increased for fields of 5.0 and 6.0 cm and at attenuation depth higher than 1.0 cm. When it is compared the PDD values for fields of 3.0 and 6.0 cm in diameter, it verifies the greater difference (12.6 %) at depth of 5.7 cm, proving the scattered radiation effect. The MCNP-5C code showed as an appropriate procedure to analyze the attenuation curves of the superficial radiotherapy beams. (author)

  19. Analysis of uncertainty in the X-ray simulation using MCNPS; Analisis de incertidumbres en la simulacion de Rayos X utilizados MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Querol, A.; Gallardo, S.; Rodenas, J.; Verdu, G.

    2011-07-01

    The characterization of the primary X-ray spectrum is a very useful tool for quality control of radiology equipment, as it allows a better estimation of the dose received by the patient, and better image quality. In this work, we have developed a Monte Carlo model with MCNP5 program to simulate X-ray spectra from an electron source and the characteristics of a commercial X-ray tube.

  20. Production of energetic light fragments in extensions of the CEM and LAQGSM event generators of the Monte Carlo transport code MCNP6

    Science.gov (United States)

    Mashnik, Stepan G.; Kerby, Leslie M.; Gudima, Konstantin K.; Sierk, Arnold J.; Bull, Jeffrey S.; James, Michael R.

    2017-03-01

    We extend the cascade-exciton model (CEM), and the Los Alamos version of the quark-gluon string model (LAQGSM), event generators of the Monte Carlo N -particle transport code version 6 (MCNP6), to describe production of energetic light fragments (LF) heavier than 4He from various nuclear reactions induced by particles and nuclei at energies up to about 1 TeV/nucleon. In these models, energetic LF can be produced via Fermi breakup, preequilibrium emission, and coalescence of cascade particles. Initially, we study several variations of the Fermi breakup model and choose the best option for these models. Then, we extend the modified exciton model (MEM) used by these codes to account for a possibility of multiple emission of up to 66 types of particles and LF (up to 28Mg) at the preequilibrium stage of reactions. Then, we expand the coalescence model to allow coalescence of LF from nucleons emitted at the intranuclear cascade stage of reactions and from lighter clusters, up to fragments with mass numbers A ≤7 , in the case of CEM, and A ≤12 , in the case of LAQGSM. Next, we modify MCNP6 to allow calculating and outputting spectra of LF and heavier products with arbitrary mass and charge numbers. The improved version of CEM is implemented into MCNP6. Finally, we test the improved versions of CEM, LAQGSM, and MCNP6 on a variety of measured nuclear reactions. The modified codes give an improved description of energetic LF from particle- and nucleus-induced reactions; showing a good agreement with a variety of available experimental data. They have an improved predictive power compared to the previous versions and can be used as reliable tools in simulating applications involving such types of reactions.

  1. Comparison of Fuel Temperature Coefficients of PWR UO{sub 2} Fuel from Monte Carlo Codes (MCNP6.1 and KENO6)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung-O; Roh, Gyuhong; Lee, Byungchul [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    As a result, there was a difference within about 300-400 pcm between keff values at each enrichment due to the difference of codes and nuclear data used in the evaluations. The FTC was changed to be less negative with the increase of uranium enrichment, and it followed the form of asymptotic curve. However, it is necessary to perform additional study for investigating what factor causes the differences more than two standard deviation (2σ) among the FTCs at partial enrichment region. The interaction probability of incident neutron with nuclear fuel is depended on the relative velocity between the neutron and the target nuclei. The Fuel Temperature Coefficient (FTC) is defined as the change of Doppler effect with respect to the change in fuel temperature without any other change such as moderator temperature, moderator density, etc. In this study, the FTCs for UO{sub 2} fuel were evaluated by using MCNP6.1 and KENO6 codes based on a Monte Carlo method. In addition, the latest neutron cross-sections (ENDF/B-VI and VII) were applied to analyze the effect of these data on the evaluation of FTC, and nuclear data used in MCNP calculations were generated from the makxsf code. An evaluation of the Doppler effect and FTC for UO{sub 2} fuel widely used in PWR was conducted using MCNP6.1 and KENO6 codes. The ENDF/B-VI and VII were also applied to analyze what effect these data has on those evaluations. All cross-sections needed for MCNP calculation were produced using makxsf code. The calculation models used in the evaluations were based on the typical PWR UO{sub 2} lattice.

  2. Transparent User Authentication

    CERN Document Server

    Clarke, Nathan

    2011-01-01

    This groundbreaking text examines the problem of user authentication from a completely new viewpoint. Rather than describing the requirements, technologies and implementation issues of designing point-of-entry authentication, the book introduces and investigates the technological requirements of implementing transparent user authentication -- where authentication credentials are captured during a user's normal interaction with a system. This approach would transform user authentication from a binary point-of-entry decision to a continuous identity confidence measure. Topics and features: discu

  3. Development and validation of a model TRIGA Mark III reactor with code MCNP5; Desarrollo y validacion de un modelo del reactor Triga Mark III con el codigo MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K{sub eff} was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)

  4. Response function of an HPGe detector simulated through MCNP 4A varying the density and chemical composition of the matrix; Funcion respuesta de un detector HPGe simulada mediante MCNP 4A variando la densidad y composicion quimica de la matriz

    Energy Technology Data Exchange (ETDEWEB)

    Leal A, B.; Mireles G, F.; Quirino T, L.; Pinedo, J.L. [Universidad Autonoma de Zacatecas, Zacatecas (Mexico)]. e-mail: bleal79@yahoo.com.mx

    2005-07-01

    In the area of the Radiological Safety it is required of a calibrated detection system in energy and efficiency for the determination of the concentration in activity in samples that vary in chemical composition and by this in density. The area of Nuclear Engineering requires to find the grade of isotopic enrichment of the uranium of the Sub-critic Nuclear Chicago 9000 Mark. Given the experimental importance that has the determination from the curves of efficiency to the effects of establishing the quantitative results, is appealed to the simulation of the response function of the detector used in the Regional Center of Nuclear Studies inside the range of energy of 80 keV to 1400 keV varying the density of the matrix and the chemical composition by means of the application of the Monte Carlo code MCNP-4A. The obtained results in the simulation of the response function of the detector show a grade of acceptance in the range from 500 to 1400 keV energy, with a smaller percentage discrepancy to 10%, in the range of low energy that its go from 59 to 400 keV, the percentage discrepancy varies from 17% until 30%, which is manifested in the opposing isotopic relationship for 5 fuel rods of the Sub critic nuclear assemble. (Author)

  5. Specific absorbed fractions of photons calculated in voxel phantoms using the Monte Carlo EGS4 and MCNP4 codes; Fracoes absorvidas especificas de fotons calculadas em fantomas de voxels utilizando os codigos Monte Carlo EGS4 e MCNP4

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, R.; Khoury, H. J. [Pernambuco Univ. (UFPE), Recife, PE (Brazil). Dept. de Energia Nuclear; Yoriyaz, H. [Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), Sao Paulo, SP (Brazil); Lima, F.R.A. [Centro Regional de Ciencias Nucleares (CRCN-NE/CNEN), Recife, PE (Brazil); Faculdade Boa Viagem (FBV), Recife, PE (Brazil)]. E-mail: rkramer@uol.com.br

    2004-07-01

    Dose coefficients for intakes of radionuclides published by the International Commission on Radiological Protection (ICRP) are based on specific absorbed fractions (SAFs), which have been calculated in the mathematical MIRD phantoms. The replacement of the MIRD phantoms by voxel phantoms proposed by the ICRP raises the question about the changes to be expected for the SAFs, and consequently also for the dose coefficients. In order to investigate the dosimetric consequences of this replacement, SAFs have been calculated in the recently introduced MAX (Male Adult voXel) and FAXht (Female Adult voXel) head + trunk phantoms for photon energies between 10 keV and 4 MeV. For this purpose the phantoms have been connected to the EGS4 as well as to the MCNP4 code, which at present are probably the most used general-purpose Monte Carlo codes. Thereby it was possible to assess the impact on the SAFs, if different radiation transport methods are used. The mathematical MIRD phantoms have also been connected to the EGS4 code, and their elemental compositions of body tissues were replaced by those used in the voxel phantoms. In this manner it was possible to compare the SAFs of the MIRD phantoms on the one hand and the MAX and FAX phantom on the other hand as a function of the geometrical anatomy only, i.e. the volume, the shape and the position of organs at risk. (author)

  6. User Requirements for Wireless

    DEFF Research Database (Denmark)

    technologies or software has been developed. A variety of user requirements are provided illustrating the effect of changing the targeted user group with respect to age,; to the context and the different technologies or software as well as to the difference in viewpoint on ways of involving users...

  7. Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

    Energy Technology Data Exchange (ETDEWEB)

    Nasrabadi, M.N. [Department of Physics, Faculty of Science, University of Kashan, Km. 6, Ravand Road, Kashan (Iran, Islamic Republic of)], E-mail: mnnasri@kashanu.ac.ir; Jalali, M. [Isfahan Nuclear Science and Technology Research Institute, Atomic Energy organization of Iran (Iran, Islamic Republic of); Mohammadi, A. [Department of Physics, Faculty of Science, University of Kashan, Km. 6, Ravand Road, Kashan (Iran, Islamic Republic of)

    2007-10-15

    In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF{sub 3} detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required.

  8. Assessment of evaluated (n,d) energy-angle elastic scattering distributions using MCNP simulations of critical measurements and simplified calculation benchmarks

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario (Canada)

    2008-07-01

    Different evaluated (n,d) energy-angle elastic scattering distributions produce k-effective differences in MCNP5 simulations of critical experiments involving heavy water (D{sub 2}O) of sufficient magnitude to suggest a need for new (n,d) scattering measurements and/or distributions derived from modern theoretical nuclear models, especially at neutron energies below a few MeV. The present work focuses on the small reactivity change of < 1 mk that is observed in the MCNP5 D{sub 2}O coolant-void-reactivity calculation bias for simulations of two pairs of critical experiments performed in the ZED-2 reactor at the Chalk River Laboratories when different nuclear data libraries are used for deuterium. The deuterium data libraries tested include Endf/B-VII.0, Endf/B-VI.4, JENDL-3.3 and a new evaluation, labelled Bonn-B, which is based on recent theoretical nuclear-model calculations. Comparison calculations were also performed for a simplified, two-region, spherical model having an inner, 250-cm radius, homogeneous sphere of UO{sub 2}, without and with deuterium, and an outer 20-cm-thick deuterium reflector. A notable observation from this work is the reduction of about 0.4 mk in the MCNP5 ZED-2 CVR calculation bias that is obtained when the O-in-UO{sub 2} thermal scattering data comes from Endf-B-VII.0. (author)

  9. CaMath user`s guide

    Energy Technology Data Exchange (ETDEWEB)

    Cha, Ben-chin; Daly, B.

    1994-07-13

    CaMath is an external Mathematica package which can be loaded into Mathematica by a user. CaMath consists of a special set of channel access functions which provides the Mathematica users with easy and flexible access of channel information across the IOC networks. It also provides a complete set of process variable event monitoring functions. The available functions for CaMath, their functionality, and their syntax are described herein. This document also gives examples how a Mathematica user can interface to channel access devices. It is assumed that the user is already familiar with using Mathematica. Few examples of Mathematica module of using CaMath functions are also given in this document.

  10. DOSFAC2 user`s guide

    Energy Technology Data Exchange (ETDEWEB)

    Young, M.L.; Chanin, D.

    1997-12-01

    This document describes the DOSFAC2 code, which is used for generating dose-to-source conversion factors for the MACCS2 code. DOSFAC2 is a revised and updated version of the DOSFAC code that was distributed with version 1.5.11 of the MACCS code. included are (1) an overview and background of DOSFAC2, (2) a summary of two new functional capabilities, and (3) a user`s guide. 20 refs., 5 tabs.

  11. Neutronic analysis for core conversion (HEU–LEU of the low power research reactor using the MCNP4C code

    Directory of Open Access Journals (Sweden)

    Aldawahra Saadou

    2015-06-01

    Full Text Available Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff, excess reactivity (ρex, control rod worth (CRW, shutdown margin (SDM, safety reactivity factor (SRF, delayed neutron fraction (βeff and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

  12. Validation of MCNP NPP Activation Simulations for Decommissioning Studies by Analysis of NPP Neutron Activation Foil Measurement Campaigns

    Directory of Open Access Journals (Sweden)

    Volmert Ben

    2016-01-01

    Full Text Available In this paper, an overview of the Swiss Nuclear Power Plant (NPP activation methodology is presented and the work towards its validation by in-situ NPP foil irradiation campaigns is outlined. Nuclear Research and consultancy Group (NRG in The Netherlands has been given the task of performing the corresponding neutron metrology. For this purpose, small Aluminium boxes containing a set of circular-shaped neutron activation foils have been prepared. After being irradiated for one complete reactor cycle, the sets have been successfully retrieved, followed by gamma-spectrometric measurements of the individual foils at NRG. Along with the individual activities of the foils, the reaction rates and thermal, intermediate and fast neutron fluence rates at the foil locations have been determined. These determinations include appropriate corrections for gamma self-absorption and neutron self-shielding as well as corresponding measurement uncertainties. The comparison of the NPP Monte Carlo calculations with the results of the foil measurements is done by using an individual generic MCNP model functioning as an interface and allowing the simulation of individual foil activation by predetermined neutron spectra. To summarize, the comparison between calculation and measurement serve as a sound validation of the Swiss NPP activation methodology by demonstrating a satisfying agreement between measurement and calculation. Finally, the validation offers a chance for further improvements of the existing NPP models by ensuing calibration and/or modelling optimizations for key components and structures.

  13. Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

    Energy Technology Data Exchange (ETDEWEB)

    Zamani, M. [National Radiation Protection Department - NRPD, Atomic Energy Organization of Iran - AEOI, Tehran (Iran, Islamic Republic of); End of North Kargar st, Atomic Energy Organization of Iran, P.O. Box: 14155-1339, Tehran (Iran, Islamic Republic of); Kasesaz, Y.; Khalafi, H.; Shayesteh, M. [Radiation Application School, Nuclear Science and Technology Research Institute, AEOI, Tehran (Iran, Islamic Republic of)

    2015-07-01

    In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials with different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)

  14. Validation of MCNP6.1 for Criticality Safety of Pu-Metal, -Solution, and -Oxide Systems

    Energy Technology Data Exchange (ETDEWEB)

    Kiedrowski, Brian C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Conlin, Jeremy Lloyd [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Favorite, Jeffrey A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kahler, III, Albert C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kersting, Alyssa R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parsons, Donald K. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Walker, Jessie L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-05-13

    Guidance is offered to the Los Alamos National Laboratory Nuclear Criticality Safety division towards developing an Upper Subcritical Limit (USL) for MCNP6.1 calculations with ENDF/B-VII.1 nuclear data for three classes of problems: Pu-metal, -solution, and -oxide systems. A benchmark suite containing 1,086 benchmarks is prepared, and a sensitivity/uncertainty (S/U) method with a generalized linear least squares (GLLS) data adjustment is used to reject outliers, bringing the total to 959 usable benchmarks. For each class of problem, S/U methods are used to select relevant experimental benchmarks, and the calculational margin is computed using extreme value theory. A portion of the margin of sub criticality is defined considering both a detection limit for errors in codes and data and uncertainty/variability in the nuclear data library. The latter employs S/U methods with a GLLS data adjustment to find representative nuclear data covariances constrained by integral experiments, which are then used to compute uncertainties in keff from nuclear data. The USLs for the classes of problems are as follows: Pu metal, 0.980; Pu solutions, 0.973; dry Pu oxides, 0.978; dilute Pu oxide-water mixes, 0.970; and intermediate-spectrum Pu oxide-water mixes, 0.953.

  15. Shielding calculations for industrial 5/7.5MeV electron accelerators using the MCNP Monte Carlo Code

    Science.gov (United States)

    Peri, Eyal; Orion, Itzhak

    2017-09-01

    High energy X-rays from accelerators are used to irradiate food ingredients to prevent growth and development of unwanted biological organisms in food, and by that extend the shelf life of the products. The production of X-rays is done by accelerating 5 MeV electrons and bombarding them into a heavy target (high Z). Since 2004, the FDA has approved using 7.5 MeV energy, providing higher production rates with lower treatments costs. In this study we calculated all the essential data needed for a straightforward concrete shielding design of typical food accelerator rooms. The following evaluation is done using the MCNP Monte Carlo code system: (1) Angular dependence (0-180°) of photon dose rate for 5 MeV and 7.5 MeV electron beams bombarding iron, aluminum, gold, tantalum, and tungsten targets. (2) Angular dependence (0-180°) spectral distribution simulations of bremsstrahlung for gold, tantalum, and tungsten bombarded by 5 MeV and 7.5 MeV electron beams. (3) Concrete attenuation calculations in several photon emission angles for the 5 MeV and 7.5 MeV electron beams bombarding a tantalum target. Based on the simulation, we calculated the expected increase in dose rate for facilities intending to increase the energy from 5 MeV to 7.5 MeV, and the concrete width needed to be added in order to keep the existing dose rate unchanged.

  16. Computational model of Amersham I-125 source model 6711 and Prosper Pd-103 source model MED3633 using MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Menezes, Artur F.; Reis Junior, Juraci P.; Silva, Ademir X., E-mail: ademir@con.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Rosa, Luiz A.R. da, E-mail: lrosa@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Facure, Alessandro [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Cardoso, Simone C., E-mail: Simone@if.ufrj.b [Universidade Federal do Rio de Janeiro (IF/UFRJ), RJ (Brazil). Inst. de Fisica. Dept. de Fisica Nuclear

    2011-07-01

    Brachytherapy is used in cancer treatment at shorter distances through the use of small encapsulated source of ionizing radiation. In such treatment, a radiation source is positioned directly into or near the target volume to be treated. In this study the Monte Carlo based MCNP code was used to model and simulate the I-125 Amersham Health source model 6711 and the Pd-103 Prospera source model MED3633 in order to obtain the dosimetric parameter dose rate constant ({Lambda}) . The sources geometries were modeled and implemented in MCNPX code. The dose rate constant is an important parameter prostate LDR brachytherapy's treatments planning. This study was based on American Association of Physicists in Medicine (AAPM) recommendations which were produced by its Task Group 43. The results obtained were 0.941 and 0.65 for the dose rate constants of I-125 and Pd-103 sources, respectively. They present good agreement with the literature values based on different Monte Carlo codes. (author)

  17. NEPHTIS: 2D/3D validation elements using MCNP4c and TRIPOLI4 Monte-Carlo codes

    Energy Technology Data Exchange (ETDEWEB)

    Courau, T.; Girardi, E. [EDF R and D/SINETICS, 1av du General de Gaulle, F92141 Clamart CEDEX (France); Damian, F.; Moiron-Groizard, M. [DEN/DM2S/SERMA/LCA, CEA Saclay, F91191 Gif-sur-Yvette CEDEX (France)

    2006-07-01

    High Temperature Reactors (HTRs) appear as a promising concept for the next generation of nuclear power applications. The CEA, in collaboration with AREVA-NP and EDF, is developing a core modeling tool dedicated to the prismatic block-type reactor. NEPHTIS (Neutronics Process for HTR Innovating System) is a deterministic codes system based on a standard two-steps Transport-Diffusion approach (APOLLO2/CRONOS2). Validation of such deterministic schemes usually relies on Monte-Carlo (MC) codes used as a reference. However, when dealing with large HTR cores the fission source stabilization is rather poor with MC codes. In spite of this, it is shown in this paper that MC simulations may be used as a reference for a wide range of configurations. The first part of the paper is devoted to 2D and 3D MC calculations of a HTR core with control devices. Comparisons between MCNP4c and TRIPOLI4 MC codes are performed and show very consistent results. Finally, the last part of the paper is devoted to the code to code validation of the NEPHTIS deterministic scheme. (authors)

  18. Gamma Knife Simulation Using the MCNP4C Code and the Zubal Phantom and Comparison with Experimental Data

    Directory of Open Access Journals (Sweden)

    Somayeh Gholami

    2010-06-01

    Full Text Available Introduction: Gamma Knife is an instrument specially designed for treating brain disorders. In Gamma Knife, there are 201 narrow beams of cobalt-60 sources that intersect at an isocenter point to treat brain tumors. The tumor is placed at the isocenter and is treated by the emitted gamma rays. Therefore, there is a high dose at this point and a low dose is delivered to the normal tissue surrounding the tumor. Material and Method: In the current work, the MCNP simulation code was used to simulate the Gamma Knife. The calculated values were compared to the experimental ones and previous works. Dose distribution was compared for different collimators in a water phantom and the Zubal brain-equivalent phantom. The dose profiles were obtained along the x, y and z axes. Result: The evaluation of the developed code was performed using experimental data and we found a good agreement between our simulation and experimental data. Discussion: Our results showed that the skull bone has a high contribution to both scatter and absorbed dose. In other words, inserting the exact material of brain and other organs of the head in digital phantom improves the quality of treatment planning. This work is regarding the measurement of absorbed dose and improving the treatment planning procedure in Gamma-Knife radiosurgery in the brain.

  19. The dose distribution inside the irradiation chamber of the gamma cell 220 at KACST using MCNP4B

    Energy Technology Data Exchange (ETDEWEB)

    Hefne, Jameel [King Abdulaziz City for Science and Technology, Riyadh (Saudi Arabia)

    2000-03-01

    In the irradiation chamber of gamma cell-220 at KACST the dose distribution must be determined. The determination of this dose distribution gives an idea about the amount of the dose in any place inside the irradiation chamber which helps also to find the average dose given to any object that needs to be irradiated. The Monte Carlo N Particle code (MCNP4B) was used to estimate the dose distribution inside the irradiation chamber. Point detectors were used in this simulation. The code was run for sufficient numbers of history which shows a symmetrical distribution around the axis of the irradiation chamber, and the errors is less than 5%. The dose map shows that the dose increases as it is calculated from the center of the chamber to the chamber perimeter edge, and it decreases as moving to the top or the bottom of the chamber. The calculation was compared with a measurement, which was done by Dr. Abdelrehim. A good agreement between the calculation and the measurement was obtained. (author)

  20. User interface design considerations

    DEFF Research Database (Denmark)

    Andersen, Simon Engedal; Jakobsen, Arne; Rasmussen, Bjarne D.

    1999-01-01

    When designing a user interface for a simulation model there are several important issues to consider: Who is the target user group, and which a priori information can be expected. What questions do the users want answers to and what questions are answered using a specific model?When developing...... the user interface of EESCoolTools these issues led to a series of simulation tools each with a specific purpose and a carefully selected set of input and output variables. To allow a more wide range of questions to be answered by the same model, the user can change between different sets of input...... and output variables. This feature requires special attention when designing the user interface and a special approach for controlling the user selection of input and output variables are developed. To obtain a consistent system description the different input variables are grouped corresponding...

  1. International user studies

    DEFF Research Database (Denmark)

    Nielsen, Lene; Madsen, Sabine; Jensen, Iben;

    in Sydhavnen, and it is funded by InfinIT. Based on a qualitative interview study with 15 user researchers from 11 different companies, we have investigated how companies collect and present data about users on international markets. Key findings are: Companies do not collect data about end users in all....../region. The preferred data collection method is field studies. If possible, user researchers choose to go to the field themselves to gain rich insights and to control the data collection process. The main insights companies gain from international user studies are (1) that there are many similarities among end users...... across nationalities and (2) that it often is more important to focus on and take differences in market conditions into account than national culture per se. Companies are in the process of finding out how best to present the insights about international end users to their employees. However, so far...

  2. Monte Carlo determination of the conversion coefficients Hp(3)/Ka in a right cylinder phantom with 'PENELOPE' code. Comparison with 'MCNP' simulations.

    Science.gov (United States)

    Daures, J; Gouriou, J; Bordy, J M

    2011-03-01

    This work has been performed within the frame of the European Union ORAMED project (Optimisation of RAdiation protection for MEDical staff). The main goal of the project is to improve standards of protection for medical staff for procedures resulting in potentially high exposures and to develop methodologies for better assessing and for reducing, exposures to medical staff. The Work Package WP2 is involved in the development of practical eye-lens dosimetry in interventional radiology. This study is complementary of the part of the ENEA report concerning the calculations with the MCNP-4C code of the conversion factors related to the operational quantity H(p)(3). In this study, a set of energy- and angular-dependent conversion coefficients (H(p)(3)/K(a)), in the newly proposed square cylindrical phantom made of ICRU tissue, have been calculated with the Monte-Carlo code PENELOPE and MCNP5. The H(p)(3) values have been determined in terms of absorbed dose, according to the definition of this quantity, and also with the kerma approximation as formerly reported in ICRU reports. At a low-photon energy (up to 1 MeV), the two results obtained with the two methods are consistent. Nevertheless, large differences are showed at a higher energy. This is mainly due to the lack of electronic equilibrium, especially for small angle incidences. The values of the conversion coefficients obtained with the MCNP-4C code published by ENEA quite agree with the kerma approximation calculations obtained with PENELOPE. We also performed the same calculations with the code MCNP5 with two types of tallies: F6 for kerma approximation and *F8 for estimating the absorbed dose that is, as known, due to secondary electrons. PENELOPE and MCNP5 results agree for the kerma approximation and for the absorbed dose calculation of H(p)(3) and prove that, for photon energies larger than 1 MeV, the transport of the secondary electrons has to be taken into account.

  3. MuSim, a graphical user interface for multiple simulation programs

    CERN Document Server

    Roberts, Thomas J; Johnson, Rolland Paul; Neuffer, David Vincent

    2016-01-01

    MuSim is a new user-friendly program designed to interface to many different particle simulation codes, regardless of their data formats or geometry descriptions. It presents the user with a compelling graphical user interface that includes a flexible 3-D view of the simulated world plus powerful editing and drag-and-drop capabilities. All aspects of the design can be parameterized so that parameter scans and optimizations are easy. It is simple to create plots and display events in the 3-D viewer (with a slider to vary the transparency of solids), allowing for an effortless comparison of different simulation codes. Simulation codes: G4beamline 3.02 and MCNP 6.1; more are coming. Many accelerator design tools and beam optics codes were written long ago, with primitive user interfaces by today's standards. MuSim is specifically designed to make it easy to interface to such codes, providing a common user experience for all, and permitting the construction and exploration of models with very little overhead. For...

  4. MuSim, a Graphical User Interface for Multiple Simulation Programs

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, Thomas [MUONS Inc., Batavia; Cummings, Mary Anne [MUONS Inc., Batavia; Johnson, Rolland [MUONS Inc., Batavia; Neuffer, David [Fermilab

    2016-06-01

    MuSim is a new user-friendly program designed to interface to many different particle simulation codes, regardless of their data formats or geometry descriptions. It presents the user with a compelling graphical user interface that includes a flexible 3-D view of the simulated world plus powerful editing and drag-and-drop capabilities. All aspects of the design can be parametrized so that parameter scans and optimizations are easy. It is simple to create plots and display events in the 3-D viewer (with a slider to vary the transparency of solids), allowing for an effortless comparison of different simulation codes. Simulation codes: G4beamline, MAD-X, and MCNP; more coming. Many accelerator design tools and beam optics codes were written long ago, with primitive user interfaces by today's standards. MuSim is specifically designed to make it easy to interface to such codes, providing a common user experience for all, and permitting the construction and exploration of models with very little overhead. For today's technology-driven students, graphical interfaces meet their expectations far better than text-based tools, and education in accelerator physics is one of our primary goals.

  5. 基于 MELCOR 与 MCNP 程序的安全壳剂量率计算方法%Calculating Method of Containment Dose Rate Based on MELCOR and MCNP Codes

    Institute of Scientific and Technical Information of China (English)

    史晓磊; 许倩; 魏严凇; 季松涛

    2015-01-01

    严重事故条件下,评估安全壳内的放射性剂量率水平对核电厂严重事故管理、应急响应等环节具有重要指导意义。本工作利用M ELCOR程序模拟严重事故序列,计算不同核素组释放进入安全壳内的质量;利用ORIGEN2程序计算不同核素组的堆芯积存量及核素的γ源强;利用MCNP程序计算每组核素100%释放进入安全壳所产生的剂量率水平;最后根据拟合公式求解安全壳剂量率。中核核电运行管理有限公司30万千瓦机组安全壳剂量率的计算结果说明该方法切实可行。%It is important to evaluate the containment dose rate under severe accident conditions for some aspects of a nuclear power plant ,such as severe accident manage‐ment and emergency response .In this work ,the MELCOR code was used to simulate the sequence of severe accidents , calculate masses of radioactive fission products released to containment .The ORIGEN2 code was used to calculate the γsource intensity . The MCNP code was used to calculate the containment dose rate when each group of radioactive fission products was all released to containment .The containment dose rate was finally calculated by a fitting formula .This method was used in the 300 MW units of CNNP Nuclear Power Operations Management Co .Ltd and was proved to be available .

  6. Evaluation of the criticality of a concrete container for storage of spent fuel in dry with MCNP; Evaluacion de la criticidad de un contenedor de concreto para almacenamiento de combustible gastado en seco con MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Xolocostli M, J. V.; Ramirez S, J. R., E-mail: vicente.xolocostli@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    A main concern exists inside the nuclear power plants in operation around the world that is the with respect to the storage capacity of the spent fuel, due to the useful life of the plant and the storage capacity in the spent fuel pool. In diverse countries is believed that one of the best alternatives for the spent fuel is the reprocessing of the same one since exists a great quantity of fissile material that can be profitable as the Pu-239, but even so the costs for the reprocessing continue being high, what limits taking this process to great scale. Is for that reason the importance of the containers for storage of spent fuel in dry which has had a great apogee in the last years, since they represent an alternative to store the spent fuel before making a decision on the reprocessing of the same one or the final disposal. In this work an evaluation of the criticality of a concrete container for storage of spent fuel in dry commercially available is made, and which is useful for fuel assemblies type PWR like BWR, in our case only the type BWR is considered. For the analysis of the evaluation was used the code MCNP5, considering the characteristics of the concrete container according to the available data, although the type of fuel assembly is BWR one of the models of the ABB company was considered with which the comparative of the results is made. The made calculations were carried out considering the inundation of the gap that exist and the external cavity, being this the most extreme condition to arrive to the criticality or in the case of happening an accident to have the filtration of the water toward the space of the gap. (author)

  7. Measuring user engagement

    CERN Document Server

    Lalmas, Mounia; Yom-Tov, Elad

    2014-01-01

    User engagement refers to the quality of the user experience that emphasizes the positive aspects of interacting with an online application and, in particular, the desire to use that application longer and repeatedly. User engagement is a key concept in the design of online applications (whether for desktop, tablet or mobile), motivated by the observation that successful applications are not just used, but are engaged with. Users invest time, attention, and emotion in their use of technology, and seek to satisfy pragmatic and hedonic needs. Measurement is critical for evaluating whether online

  8. The User Reconfigured

    DEFF Research Database (Denmark)

    Bardzell, Jeffrey; Bardzell, Shaowen

    2015-01-01

    Foundational to HCI is the notion of “the user.” Whether a cognitive processor, social actor, consumer, or even a non- user, the user in HCI has always been as much a technical construct as actual people using systems. We explore an emerging formulation of the user—the subjectivity of in- formation......, and activism. We argue that subjectivi- ties of information clarifies the relationships between de- sign choices and embodied experiences, ways that designers design users and not just products, and ways to cultivate and transform, rather than merely support, human agency....

  9. Demonstrator 1: User Interface and User Functions

    DEFF Research Database (Denmark)

    Gram, Christian

    1999-01-01

    Describes the user interface and its functionality in a prototype system used for a virtual seminar session. The functionality is restricted to what is needed for a distributed seminar discussion among not too many people. The system is designed to work with the participants distributed at several...

  10. Comparison of KENO-VI and MCNP5 Criticality Analyses for a Lunar Regolith Clustered-Reactor System

    Science.gov (United States)

    Bess, John Darrell

    2008-01-01

    The Lunar Regolith Clustered-Reactor System design has been presented as an alternative method for providing surface power to a lunar facility using a fast-fission, heatpipe-cooled nuclear reactor. The reactor system is divided into subcritical units that can be safely launched into orbit without risk of inadvertent criticality in the event of a launch accident. The reactor subunits are emplaced into the lunar surface to form a clustered-reactor system, utilizing the regolith as both radiation shielding and neutron-reflector material. Coordinated placement of multiple subunits can provision a critical reactor system proportional to localized lunar surface power demand. Reactor units assembled using proven and tested materials in radiation environments such as UO2 fuel, stainless-steel cladding and support, and compatible liquid-metal heatpipes promote safety and reliability, with ease of manufacture and testing. Reactor power levels of approximately 100 kWth per subunit significantly reduces the negative effects of elevated temperature and radiation environments associated with single nuclear power reactors operated at higher power levels. The analysis of subunit criticality in various accident scenarios differs by up to 4% (~$6 in reactivity) between results generated using conventional criticality analysis codes, MCNP5 and KENO-VI. A demonstrated trend exists between results of the two criticality codes as accident conditions approach a multiplication factor of one. Code comparison of a tri-cluster system on the lunar surface provides comparable results with calculated system reactivity within 0.5%. Iron concentration is confirmed as the dominant element in the lunar regolith influencing system reactivity.

  11. Investigation of Anisotropy Caused by Cylinder Applicator on Dose Distribution around Cs-137 Brachytherapy Source using MCNP4C Code

    Directory of Open Access Journals (Sweden)

    Sedigheh Sina

    2011-06-01

    Full Text Available Introduction: Brachytherapy is a type of radiotherapy in which radioactive sources are used in proximity of tumors normally for treatment of malignancies in the head, prostate and cervix. Materials and Methods: The Cs-137 Selectron source is a low-dose-rate (LDR brachytherapy source used in a remote afterloading system for treatment of different cancers. This system uses active and inactive spherical sources of 2.5 mm diameter, which can be used in different configurations inside the applicator to obtain different dose distributions. In this study, first the dose distribution at different distances from the source was obtained around a single pellet inside the applicator in a water phantom using the MCNP4C Monte Carlo code. The simulations were then repeated for six active pellets in the applicator and for six point sources.  Results: The anisotropy of dose distribution due to the presence of the applicator was obtained by division of dose at each distance and angle to the dose at the same distance and angle of 90 degrees. According to the results, the doses decreased towards the applicator tips. For example, for points at the distances of 5 and 7 cm from the source and angle of 165 degrees, such discrepancies reached 5.8% and 5.1%, respectively.  By increasing the number of pellets to six, these values reached 30% for the angle of 5 degrees. Discussion and Conclusion: The results indicate that the presence of the applicator causes a significant dose decrease at the tip of the applicator compared with the dose in the transverse plane. However, the treatment planning systems consider an isotropic dose distribution around the source and this causes significant errors in treatment planning, which are not negligible, especially for a large number of sources inside the applicator.

  12. Applicability of the MCNP-ACAB system to inventory prediction in high-burnup fuels: sensitivity/uncertainty estimates

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Herranz, N.; Cabellos, O. [Madrid Polytechnic Univ., Dept. of Nuclear Engineering (Spain); Cabellos, O.; Sanz, J. [Madrid Polytechnic Univ., 2 Instituto de Fusion Nuclear (Spain); Sanz, J. [Univ. Nacional Educacion a Distancia, Dept. of Power Engineering, Madrid (Spain)

    2005-07-01

    We present a new code system which combines the Monte Carlo neutron transport code MCNP-4C and the inventory code ACAB as a suitable tool for high burnup calculations. Our main goal is to show that the system, by means of ACAB capabilities, enables us to assess the impact of neutron cross section uncertainties on the inventory and other inventory-related responses in high burnup applications. The potential impact of nuclear data uncertainties on some response parameters may be large, but only very few codes exist which can treat this effect. In fact, some of the most reported effective code systems in dealing with high burnup problems, such as CASMO-4, MCODE and MONTEBURNS, lack this capability. As first step, the potential of our system, ruling out the uncertainty capability, has been compared with that of those code systems, using a well referenced high burnup pin-cell benchmark exercise. It is proved that the inclusion of ACAB in the system allows to obtain results at least as reliable as those obtained using other inventory codes, such as ORIGEN2. Later on, the uncertainty analysis methodology implemented in ACAB, including both the sensitivity-uncertainty method and the uncertainty analysis by the Monte Carlo technique, is applied to this benchmark problem. We estimate the errors due to activation cross section uncertainties in the prediction of the isotopic content up to the high-burnup spent fuel regime. The most relevant uncertainties are remarked, and some of the most contributing cross sections to those uncertainties are identified. For instance, the most critical reaction for Am{sup 242m} is Am{sup 241}(n,{gamma}-m). At 100 MWd/kg, the cross-section uncertainty of this reaction induces an error of 6.63% on the Am{sup 242m} concentration.The uncertainties in the inventory of fission products reach up to 30%.

  13. Quantitative comparison between PGNAA measurements and MCNP calculations in view of the characterization of radioactive wastes in Germany and France

    Energy Technology Data Exchange (ETDEWEB)

    Mauerhofer, E. [FZJ, Institute for Energy and Climate Research - Nuclear Waste Management and Reactor Safety, Wilhelm-Johnen-Strasse, D-52428 Juelich (Germany); Havenith, A.; Kettler, J. [RWTH Aachen University, Institute of Nuclear Fuel Cycle, Elisabethstrasse 16, D-52062 Aachen (Germany); Carasco, C.; Payan, E.; Ma, J. L.; Perot, B. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 St Paul-lez-Durance (France)

    2013-04-19

    The Forschungszentrum Juelich GmbH (FZJ), together with the Aachen University Rheinisch-Westfaelische Technische Hochschule (RWTH) and the French Alternative Energies and Atomic Energy Commission (CEA Cadarache) are involved in a cooperation aiming at characterizing toxic and reactive elements in radioactive waste packages by means of Prompt Gamma Neutron Activation Analysis (PGNAA). The French and German waste management agencies have indeed defined acceptability limits concerning these elements in view of their projected geological repositories. A first measurement campaign was performed in the new Prompt Gamma Neutron Activation Analysis (PGNAA) facility called MEDINA, at FZJ, to assess the capture gamma-ray signatures of some elements of interest in large samples up to waste drums with a volume of 200 liter. MEDINA is the acronym for Multi Element Detection based on Instrumental Neutron Activation. This paper presents MCNP calculations of the MEDINA facility and quantitative comparison between measurement and simulation. Passive gamma-ray spectra acquired with a high purity germanium detector and calibration sources are used to qualify the numerical model of the crystal. Active PGNAA spectra of a sodium chloride sample measured with MEDINA then allow for qualifying the global numerical model of the measurement cell. Chlorine indeed constitutes a usual reference with reliable capture gamma-ray production data. The goal is to characterize the entire simulation protocol (geometrical model, nuclear data, and postprocessing tools) which will be used for current measurement interpretation, extrapolation of the performances to other types of waste packages or other applications, as well as for the study of future PGNAA facilities.

  14. Quantitative comparison between PGNAA measurements and MCNP calculations in view of the characterization of radioactive wastes in Germany and France

    Science.gov (United States)

    Mauerhofer, E.; Havenith, A.; Carasco, C.; Payan, E.; Kettler, J.; Ma, J. L.; Perot, B.

    2013-04-01

    The Forschungszentrum Jülich GmbH (FZJ), together with the Aachen University Rheinisch-Westfaelische Technische Hochschule (RWTH) and the French Alternative Energies and Atomic Energy Commission (CEA Cadarache) are involved in a cooperation aiming at characterizing toxic and reactive elements in radioactive waste packages by means of Prompt Gamma Neutron Activation Analysis (PGNAA) [1]. The French and German waste management agencies have indeed defined acceptability limits concerning these elements in view of their projected geological repositories. A first measurement campaign was performed in the new Prompt Gamma Neutron Activation Analysis (PGNAA) facility called MEDINA, at FZJ, to assess the capture gamma-ray signatures of some elements of interest in large samples up to waste drums with a volume of 200 liter. MEDINA is the acronym for Multi Element Detection based on Instrumental Neutron Activation. This paper presents MCNP calculations of the MEDINA facility and quantitative comparison between measurement and simulation. Passive gamma-ray spectra acquired with a high purity germanium detector and calibration sources are used to qualify the numerical model of the crystal. Active PGNAA spectra of a sodium chloride sample measured with MEDINA then allow for qualifying the global numerical model of the measurement cell. Chlorine indeed constitutes a usual reference with reliable capture gamma-ray production data. The goal is to characterize the entire simulation protocol (geometrical model, nuclear data, and postprocessing tools) which will be used for current measurement interpretation, extrapolation of the performances to other types of waste packages or other applications, as well as for the study of future PGNAA facilities.

  15. A New On-the-Fly Sampling Method for Incoherent Inelastic Thermal Neutron Scattering Data in MCNP6

    Energy Technology Data Exchange (ETDEWEB)

    Pavlou, Andrew Theodore [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Brown, Forrest B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Ji, Wei [Rensselaer Polytechnic Inst., Troy, NY (United States)

    2014-09-02

    At thermal energies, the scattering of neutrons in a system is complicated by the comparable velocities of the neutron and target, resulting in competing upscattering and downscattering events. The neutron wavelength is also similar in size to the target's interatomic spacing making the scattering process a quantum mechanical problem. Because of the complicated nature of scattering at low energies, the thermal data files in ACE format used in continuous-energy Monte Carlo codes are quite large { on the order of megabytes for a single temperature and material. In this paper, a new storage and sampling method is introduced that is orders of magnitude less in size and is used to sample scattering parameters at any temperature on-the-fly. In addition to the reduction in storage, the need to pre-generate thermal scattering data tables at fine temperatures has been eliminated. This is advantageous for multiphysics simulations which may involve temperatures not known in advance. A new module was written for MCNP6 that bypasses the current S(α,β) table lookup in favor of the new format. The new on-the-fly sampling method was tested for graphite for two benchmark problems at ten temperatures: 1) an eigenvalue test with a fuel compact of uranium oxycarbide fuel homogenized into a graphite matrix, 2) a surface current test with a \\broomstick" problem with a monoenergetic point source. The largest eigenvalue difference was 152pcm for T= 1200K. For the temperatures and incident energies chosen for the broomstick problem, the secondary neutron spectrum showed good agreement with the traditional S(α,β) sampling method. These preliminary results show that sampling thermal scattering data on-the-fly is a viable option to eliminate both the storage burden of keeping thermal data at discrete temperatures and the need to know temperatures before simulation runtime.

  16. Sensitivity of MCNP5 calculations for a spherical numerical benchmark problem to the angular scattering distributions for deuterium

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K. S. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ont. K0J 1J0 (Canada)

    2006-07-01

    This paper examines the sensitivity of MCNP5 k{sub eff} results to various deuterium data files for a simple benchmark problem consisting of an 8.4-cm radius sphere of uranium surrounded by an annulus of deuterium at the nuclide number density corresponding to heavy water. This study was performed to help clarify why {Delta}k{sub eff} values of about 10 mk are obtained when different ENDF/B deuterium data files are used in simulations of critical experiments involving solutions of high-enrichment uranyl fluoride in heavy water, while simulations of low-leakage, heterogeneous critical lattices of natural-uranium fuel rods in heavy water show differences of <1 mk. The benchmark calculations were performed as a function of deuterium reflector thickness for several uranium compositions using deuterium ACE files derived from ENDF/B-VII.b1 (release beta 1), ENDF/B-VI.4 and JENDL-3.3, which differ primarily in the energy/angle distributions for elastic scattering <3.2 MeV. Calculations were also performed using modified ACE files having equiprobable cosine bin values in the centre-of-mass reference frame in a progressive manner with increasing energy. It was found that the {Delta}k{sub eff} values increased with deuterium reflector thickness and uranium enrichment. The studies using modified ACE files indicate that most of the reactivity differences arise at energies <1 MeV; hence, this energy range should be given priority if new scattering distribution measurements are undertaken. (authors)

  17. DESAIN TERAS PLTN JENIS PEBBLE BED MODULAR REACTOR (PBMR MENGGUNAKAN PAKET PROGRAM MCNP-5 PADA KONDISI BEGINNING OF LIFE

    Directory of Open Access Journals (Sweden)

    Ralind Re Marla

    2015-03-01

    Full Text Available Telah dilakukan desain teras Pembangkit Listrik Tenaga Nuklir (PLTN untuk jenis Pebble Bed Modular Reactor (PBMR dengan daya 70 MWe untuk keperluan proses smelter pada keadaan beginning of life (BOL. Analisis ini bertujuan untuk mengetahui persen pengkayaan, distribusi suhu dan nilai keselamatan dengan koefisien reaktivitas teras yang negatif pada reaktor jenis PBMR apabila daya reaktor 70 MWe. Analisis menggunakan program Monte Carlo N-Particle-5 (MCNP5 dan dari hasil analisis ini diharapkan dapat memenuhi syarat dalam mendukung program percepatan pembangunan kelistrikan batubara 10.000 MWe khususnya untuk proses smelter, yang tersebar merata di wilayah Indonesia. Hasil penelitian menunjukkan bahwa, faktor perlipatan efektif (k-eff Reaktor jenis PBMR daya 70 MWe mengalami kondisi kritis pada pengkayaan 5,626 % dengan nilai faktor perlipatan efektif 1,00031±0,00087 dan nilai koefisien reaktivitas suhu pada -10,0006 pcm/K. Dari hasil analisis daat disimpulkan bahwa reaktor jenis PBMR daya 70 MWe adalah aman.   ABSTRACT The core design of Nuclear Power Plant for Pebble Bed Modular Reactor (PBMR type with 70 MWe capacity power in Beginning of Life (BOL has been performed. The aim of this analysis, to know percent enrichment, temperature distribution and safety value by negative temperature coefficient at type PBMR if reactor power become lower equal to 70 MWe. This analysis was expected become one part of overview project development the power plant with 10.000 MWe of total capacity, spread evenly in territory of Indonesia especially to support of smelter industries. The results showed that, effective multiplication factor (keff with power 70 MWe critical condition at enrichment 5,626 %is 1,00031±0,00087, based on enrichment result, a value of the temperature coefficient reactivity is - 10,0006 pcm/K. Based on the results of these studies, it can beconcluded that the PBMR 70 MWe design is theoritically safe.

  18. User programmable virtualized networks

    NARCIS (Netherlands)

    Meijer, R.J.; Strijkers, R.J.; Gommans, L.; Laat, C.de

    2006-01-01

    This paper introduces the concept of a User Programmable Virtualized Network, which allows networks to deliver application specific services using network element components that developers can program as part of a users application. The use of special tokens in data or control packets is the basis

  19. Additional user needs

    Energy Technology Data Exchange (ETDEWEB)

    Rorschach, H.E.; Hayter, J.B.

    1986-08-15

    This paper summarizes the conclusions of a discussion group on users' needs held at the Workshop on an Advanced Steady-State Neutron Facility. The discussion was devoted to reactor characteristics, special facilities and siting considerations suggested by user needs.

  20. EMI New User Communities

    CERN Document Server

    Riedel, M

    2013-01-01

    This document provides pieces of information about new user communities that directly or indirectly take advantage of EMI Products. Each user community is described via one specific EMI product use case to understand and communicate the current usage of EMI Products in practice.

  1. Dosimetry analysis of distribution radial dose profiles of {sup 90}Sr + {sup 90}Y beta therapy applicators using the MCNP-4C code and radio chromium films; Analise dosimetrica de perfis de distribuicoes radiais de doses relativas de um aplicador de betaterapia de {sup 90}Sr + {sup 90}Y utilizando o codigo MCNP-4C e filmes radiocromicos

    Energy Technology Data Exchange (ETDEWEB)

    Coelho, T.S.; Yoriyaz, H. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Fernandes, M.A.R. [Universidade Estadual Paulista Julio de Mesquita Filho (UNESP), Botucatu, SP (Brazil). Fac. de Medicina. Servico de Radioterapia; Louzada, M.J.Q. [Universidade Estadual Paulista Julio de Mesquita Filho (UNESP), Aracatuba, SP (Brazil). Curso de Medicina Veterinaria

    2010-07-01

    Although they are no longer manufactured, the applicators of {sup 90}Sr +{sup 90}Y acquired in the decades of 1990 are still in use, by having half-life of 28.5 years. These applicators have calibration certificate given by their manufacturers, where few have been recalibrated. Thus it becomes necessary to accomplish thorough dosimetry of these applicators. This paper presents a dosimetric analysis distribution radial dose profiles for emitted by an {sup 90}Sr+{sup 90}Y beta therapy applicator, using the MCNP-4C code to simulate the distribution radial dose profiles and radiochromium films to get them experimentally . The results with the simulated values were compared with the results of experimental measurements, where both curves show similar behavior, which may validate the use of MCNP-4C and radiochromium films for this type of dosimetry. (author)

  2. Dosimetry analysis of distributions radials dose profiles of {sup 90}Sr + {sup 90}Y beta therapy applicators using the MCNP-4C code and radio chromium films; Analise dosimetrica de perfis de distribuicoes radias de doses relativas de um aplicador de betaterapia de {sup 90}Sr + {sup 90}Y utilizando o codigo MCNP-4C e filmes radiocromicos

    Energy Technology Data Exchange (ETDEWEB)

    Coelho, Talita S.; Yoriyaz, Helio [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Fernandes, Marco A.R., E-mail: tasallesc@gmail.co [UNESP, Botucatu, SP (Brazil). Faculdade de Medicina. Servico de Radioterapia; Louzada, Mario J.Q. [UNESP, Aracatuba, SP (Brazil). Curso de Medicina Veterinaria

    2011-07-01

    Although they are no longer manufactured, the applicators of {sup 90}Sr + {sup 90}Y acquired in the decades of 1990 are still in use, by having half-life of 28.5 years. These applicators have calibration certificate given by their manufacturers, where few have been re calibrated. Thus it becomes necessary to accomplish thorough dosimetry of these applicators. This paper presents a dosimetric analysis distribution radial dose profiles for emitted by an {sup 90}Sr + {sup 90}Y beta therapy applicator, using the MCNP-4C code to simulate the distribution radial dose profiles and radio chromium films to get them experimentally . The results with the simulated values were compared with the results of experimental measurements, where both curves show similar behavior, which may validate the use of MCNP-4C and radio chromium films for this type of dosimetry. (author)

  3. User Frustrations as Opportunities

    Directory of Open Access Journals (Sweden)

    Michael Weiss

    2012-04-01

    Full Text Available User frustrations are an excellent source of new product ideas. Starting with this observation, this article describes an approach that entrepreneurs can use to discover business opportunities. Opportunity discovery starts with a problem that the user has, but may not be able to articulate. User-centered design techniques can help elicit those latent needs. The entrepreneur should then try to understand how users are solving their problem today, before proposing a solution that draws on the unique skills and technical capabilities available to the entrepreneur. Finally, an in-depth understanding of the user allows the entrepreneur to hone in on the points of difference and resonance that are the foundation of a strong customer value proposition.

  4. Lead User Innovation

    DEFF Research Database (Denmark)

    Brem, Alexander; Larsen, Henry

    2015-01-01

    , deliver and capture the value of an innovatively new device together. From the perspective of the lead user, we show antecedents and effects of social interaction between organizational actors and the lead user on the development of social capital, especially trust and shared imagination. The second case......User innovation and especially the integration of lead users is a key topic in the innovation management literature of recent years. This paper contributes by providing a rare perspective into what easily could be seen as innovation failure, shown from two perspectives. We show how a lack of shared...... imagination hampers participation and kills innovation between interdependent stakeholders at the threshold between invention and innovation in practice. We present a first case in the fun-sport industry where an external lead user and diverse firm representatives in different functions fail to create...

  5. International user studies

    DEFF Research Database (Denmark)

    Nielsen, Lene; Madsen, Sabine; Jensen, Iben

    across nationalities and (2) that it often is more important to focus on and take differences in market conditions into account than national culture per se. Companies are in the process of finding out how best to present the insights about international end users to their employees. However, so far...... a company’s general attitude and approach to (1) international markets and (2) user studies. Lastly, we present the theoretical ideas and concepts about culture that has informed the research....... in Sydhavnen, and it is funded by InfinIT. Based on a qualitative interview study with 15 user researchers from 11 different companies, we have investigated how companies collect and present data about users on international markets. Key findings are: Companies do not collect data about end users in all...

  6. User participation in implementation

    DEFF Research Database (Denmark)

    Fleron, Benedicte; Rasmussen, Rasmus; Simonsen, Jesper

    2012-01-01

    Systems development has been claimed to benefit from user participation, yet user participation in implementation activities may be more common and is a growing focus of participatory-design work. We investigate the effect of the extensive user participation in the implementation of a clinical...... system by empirically analyzing how management, participating staff, and non-participating staff view the implementation process with respect to areas that have previously been linked to user participation such as system quality, emergent interactions, and psychological buy-in. The participating staff...... experienced more uncertainty and frustration than management and non-participating staff, especially concerning how to run an implementation process and how to understand and utilize the configuration possibilities of the system. This suggests that user participation in implementation introduces a need...

  7. The PANTHER User Experience

    Energy Technology Data Exchange (ETDEWEB)

    Coram, Jamie L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Morrow, James D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Perkins, David Nikolaus [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    This document describes the PANTHER R&D Application, a proof-of-concept user interface application developed under the PANTHER Grand Challenge LDRD. The purpose of the application is to explore interaction models for graph analytics, drive algorithmic improvements from an end-user point of view, and support demonstration of PANTHER technologies to potential customers. The R&D Application implements a graph-centric interaction model that exposes analysts to the algorithms contained within the GeoGraphy graph analytics library. Users define geospatial-temporal semantic graph queries by constructing search templates based on nodes, edges, and the constraints among them. Users then analyze the results of the queries using both geo-spatial and temporal visualizations. Development of this application has made user experience an explicit driver for project and algorithmic level decisions that will affect how analysts one day make use of PANTHER technologies.

  8. Extension and Test of Limits on MCNP Geometry Description%MCNP程序几何描述能力扩展及应用测试

    Institute of Scientific and Technical Information of China (English)

    刘镇洲; 李刚; 邓力; 柴晓明

    2013-01-01

    为使MCNP程序能模拟数百万规模的反应堆“pin-by-pin”问题和医学体素模型,本文对MCNP程序进行了改进,使几何块、几何面数量可扩展.改进后的程序对硼中子俘获治疗(BNCT)的人体大脑进行几何建模,栅元数量达百万量级;计算了大脑的中子、光子吸收剂量率随深度的变化,为大脑BNCT提供理论支持.此外,对百万规模的“Like n But”重复结构模型进行了串、并行测试,验证了几何规模扩展后程序计算的正确性.%To set up models for 'pin-by-pin' facilities (reactors) and human phantoms with millions of cells by using the MCNP code, the MCNP code was modified and limits on the number of cell, surface, material, etc. were extended. The Snyder head phantom for boron neutron capture therapy was modeled. Neutron and induced gamma-ray absorbed dose in human brain was calculated. Besides, a critical calculation model of repeated structures described by the 'Like n But' cards was calculated employing both series and parallel computing. As a conclusion, modifications of the MCNP code were proved to be correct.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  10. Evaluation of computational models and cross sections used by MCNP6 for simulation of characteristic X-ray emission from thick targets bombarded by kiloelectronvolt electrons

    Science.gov (United States)

    Poškus, A.

    2016-09-01

    This paper evaluates the accuracy of the single-event (SE) and condensed-history (CH) models of electron transport in MCNP6.1 when simulating characteristic Kα, total K (=Kα + Kβ) and Lα X-ray emission from thick targets bombarded by electrons with energies from 5 keV to 30 keV. It is shown that the MCNP6.1 implementation of the CH model for the K-shell impact ionization leads to underestimation of the K yield by 40% or more for the elements with atomic numbers Z 25. The Lα yields are underestimated by more than an order of magnitude in CH mode, because MCNP6.1 neglects X-ray emission caused by electron-impact ionization of L, M and higher shells in CH mode (the Lα yields calculated in CH mode reflect only X-ray fluorescence, which is mainly caused by photoelectric absorption of bremsstrahlung photons). The X-ray yields calculated by MCNP6.1 in SE mode (using ENDF/B-VII.1 library data) are more accurate: the differences of the calculated and experimental K yields are within the experimental uncertainties for the elements C, Al and Si, and the calculated Kα yields are typically underestimated by (20-30)% for the elements with Z > 25, whereas the Lα yields are underestimated by (60-70)% for the elements with Z > 49. It is also shown that agreement of the experimental X-ray yields with those calculated in SE mode is additionally improved by replacing the ENDF/B inner-shell electron-impact ionization cross sections with the set of cross sections obtained from the distorted-wave Born approximation (DWBA), which are also used in the PENELOPE code system. The latter replacement causes a decrease of the average relative difference of the experimental X-ray yields and the simulation results obtained in SE mode to approximately 10%, which is similar to accuracy achieved with PENELOPE. This confirms that the DWBA inner-shell impact ionization cross sections are significantly more accurate than the corresponding ENDF/B cross sections when energy of incident electrons

  11. Simulação de um densímetro nuclear utilizando o código Monte Carlo MCNP-4C

    OpenAIRE

    Penna, Rodrigo; Comitê Científico; da Silva, Clemente José Gusmão Carneiro; Professor; Gomes, Paulo Maurício Costa; Professor

    2008-01-01

    Foi Utilizado o código Monte Carlo (MCNP-4C) para simular um densímetro nuclear capaz de medir a densidade da madeira superficialmente. Utilizou-se uma fonte de Amerício-241, de baixa energia (E= 60 Kev) o que permite uma maior segurança na operação. Os resultados mostraram que a densidade da madeira pode ser medida partir da radiação espalhada devido ao Efeito Compton. A técnica representa um avanço em relação à metodologia atual.

  12. User`s guide to MIDAS

    Energy Technology Data Exchange (ETDEWEB)

    Tisue, S.A.; Williams, N.B.; Huber, C.C. [Argonne National Lab., IL (United States). Decision and Information Sciences Div.; Chun, K.C. [Argonne National Lab., IL (United States). Environmental Assessment Div.

    1995-12-01

    Welcome to the MIDAS User`s Guide. This document describes the goals of the Munitions Items Disposition Action System (MIDAS) program and documents the MIDAS software. The main text first describes the equipment and software you need to run MIDAS and tells how to install and start it. It lists the contents of the database and explains how it is organized. Finally, it tells how to perform various functions, such as locating, entering, viewing, deleting, changing, transferring, and printing both textual and graphical data. Images of the actual computer screens accompany these explanations and guidelines. Appendix A contains a glossary of names for the various abbreviations, codes, and chemicals; Appendix B is a list of modem names; Appendix C provides a database dictionary and rules for entering data; and Appendix D describes procedures for troubleshooting problems associated with connecting to the MIDAS server and using MIDAS.

  13. Aztec user`s guide. Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Hutchinson, S.A.; Shadid, J.N.; Tuminaro, R.S.

    1995-10-01

    Aztec is an iterative library that greatly simplifies the parallelization process when solving the linear systems of equations Ax = b where A is a user supplied n x n sparse matrix, b is a user supplied vector of length n and x is a vector of length n to be computed. Aztec is intended as a software tool for users who want to avoid cumbersome parallel programming details but who have large sparse linear systems which require an efficiently utilized parallel processing system. A collection of data transformation tools are provided that allow for easy creation of distributed sparse unstructured matrices for parallel solution. Once the distributed matrix is created, computation can be performed on any of the parallel machines running Aztec: nCUBE 2, IBM SP2 and Intel Paragon, MPI platforms as well as standard serial and vector platforms. Aztec includes a number of Krylov iterative methods such as conjugate gradient (CG), generalized minimum residual (GMRES) and stabilized biconjugate gradient (BICGSTAB) to solve systems of equations. These Krylov methods are used in conjunction with various preconditioners such as polynomial or domain decomposition methods using LU or incomplete LU factorizations within subdomains. Although the matrix A can be general, the package has been designed for matrices arising from the approximation of partial differential equations (PDEs). In particular, the Aztec package is oriented toward systems arising from PDE applications.

  14. Game user experience evaluation

    CERN Document Server

    Bernhaupt, Regina

    2015-01-01

    Evaluating interactive systems for their user experience (UX) is a standard approach in industry and research today. This book explores the areas of game design and development and Human Computer Interaction (HCI) as ways to understand the various contributing aspects of the overall gaming experience. Fully updated, extended and revised this book is based upon the original publication Evaluating User Experience in Games, and provides updated methods and approaches ranging from user- orientated methods to game specific approaches. New and emerging methods and areas explored include physiologi

  15. Designing for user engagement

    CERN Document Server

    Geisler, Cheryl

    2013-01-01

    Designing for User Engagement on the Web: 10 Basic Principles is concerned with making user experience engaging. The cascade of social web applications we are now familiar with - blogs, consumer reviews, wikis, and social networking - are all engaging experiences. But engagement is an increasingly common goal in business and productivity environments as well. This book provides a foundation for all those seeking to design engaging user experiences rich in communication and interaction. Combining a handbook on basic principles with case studies, it provides readers with a ric

  16. Safety for Users

    CERN Document Server

    HR Department

    2008-01-01

    CERN welcomes more than 8000 Users every year. The PH Department as host to these scientific associates requires the highest safety standards. The PH Safety Office has published a Safety Flyer for Users. Important safety topics and procedures are presented. Although the Flyer is intended primarily to provide safety information for Users, the PH Safety Office invites all those on the CERN sites to keep a copy of the flyer as it gives guidance in matters of safety and explains what to do in the event of an emergency. Link: http://ph-dep.web.cern.ch/ph-dep/Safety/SafetyOffice.html PH-Safety Office PH Department

  17. Safety for Users

    CERN Document Server

    HR Department

    2008-01-01

    CERN welcomes more than 8000 Users every year. The PH Department as host to these scientific associates requires the highest safety standards. The PH Safety Office has published a safety flyer for Users. Important safety topics and procedures are presented. Although the flyer is intended primarily to provide safety information for Users, the PH Safety Office invites all those on the CERN sites to keep a copy of the flyer as it gives guidance in matters of safety and explains what to do in the event of an emergency. The flyer is available at: http://ph-dep.web.cern.ch/ph-dep/Safety/SafetyOffice.html PH-Safety Office PH Department

  18. Distributed User Interfaces

    CERN Document Server

    Gallud, Jose A; Penichet, Victor M R

    2011-01-01

    The recent advances in display technologies and mobile devices is having an important effect on the way users interact with all kinds of devices (computers, mobile devices, laptops, tablets, and so on). These are opening up new possibilities for interaction, including the distribution of the UI (User Interface) amongst different devices, and implies that the UI can be split and composed, moved, copied or cloned among devices running the same or different operating systems. These new ways of manipulating the UI are considered under the emerging topic of Distributed User Interfaces (DUIs). DUIs

  19. Rapid MCNP simulation of DNA double strand break (DSB) relative biological effectiveness (RBE) for photons, neutrons, and light ions.

    Science.gov (United States)

    Stewart, Robert D; Streitmatter, Seth W; Argento, David C; Kirkby, Charles; Goorley, John T; Moffitt, Greg; Jevremovic, Tatjana; Sandison, George A

    2015-11-07

    To account for particle interactions in the extracellular (physical) environment, information from the cell-level Monte Carlo damage simulation (MCDS) for DNA double strand break (DSB) induction has been integrated into the general purpose Monte Carlo N-particle (MCNP) radiation transport code system. The effort to integrate these models is motivated by the need for a computationally efficient model to accurately predict particle relative biological effectiveness (RBE) in cell cultures and in vivo. To illustrate the approach and highlight the impact of the larger scale physical environment (e.g. establishing charged particle equilibrium), we examined the RBE for DSB induction (RBEDSB) of x-rays, (137)Cs γ-rays, neutrons and light ions relative to γ-rays from (60)Co in monolayer cell cultures at various depths in water. Under normoxic conditions, we found that (137)Cs γ-rays are about 1.7% more effective at creating DSB than γ-rays from (60)Co (RBEDSB  =  1.017) whereas 60-250 kV x-rays are 1.1 to 1.25 times more efficient at creating DSB than (60)Co. Under anoxic conditions, kV x-rays may have an RBEDSB up to 1.51 times as large as (60)Co γ-rays. Fission neutrons passing through monolayer cell cultures have an RBEDSB that ranges from 2.6 to 3.0 in normoxic cells, but may be as large as 9.93 for anoxic cells. For proton pencil beams, Monte Carlo simulations suggest an RBEDSB of about 1.2 at the tip of the Bragg peak and up to 1.6 a few mm beyond the Bragg peak. Bragg peak RBEDSB increases with decreasing oxygen concentration, which may create opportunities to apply proton dose painting to help address tumor hypoxia. Modeling of the particle RBE for DSB induction across multiple physical and biological scales has the potential to aid in the interpretation of laboratory experiments and provide useful information to advance the safety and effectiveness of hadron therapy in the treatment of cancer.

  20. Rapid MCNP simulation of DNA double strand break (DSB) relative biological effectiveness (RBE) for photons, neutrons, and light ions

    Science.gov (United States)

    Stewart, Robert D.; Streitmatter, Seth W.; Argento, David C.; Kirkby, Charles; Goorley, John T.; Moffitt, Greg; Jevremovic, Tatjana; Sandison, George A.

    2015-11-01

    To account for particle interactions in the extracellular (physical) environment, information from the cell-level Monte Carlo damage simulation (MCDS) for DNA double strand break (DSB) induction has been integrated into the general purpose Monte Carlo N-particle (MCNP) radiation transport code system. The effort to integrate these models is motivated by the need for a computationally efficient model to accurately predict particle relative biological effectiveness (RBE) in cell cultures and in vivo. To illustrate the approach and highlight the impact of the larger scale physical environment (e.g. establishing charged particle equilibrium), we examined the RBE for DSB induction (RBEDSB) of x-rays, 137Cs γ-rays, neutrons and light ions relative to γ-rays from 60Co in monolayer cell cultures at various depths in water. Under normoxic conditions, we found that 137Cs γ-rays are about 1.7% more effective at creating DSB than γ-rays from 60Co (RBEDSB  =  1.017) whereas 60-250 kV x-rays are 1.1 to 1.25 times more efficient at creating DSB than 60Co. Under anoxic conditions, kV x-rays may have an RBEDSB up to 1.51 times as large as 60Co γ-rays. Fission neutrons passing through monolayer cell cultures have an RBEDSB that ranges from 2.6 to 3.0 in normoxic cells, but may be as large as 9.93 for anoxic cells. For proton pencil beams, Monte Carlo simulations suggest an RBEDSB of about 1.2 at the tip of the Bragg peak and up to 1.6 a few mm beyond the Bragg peak. Bragg peak RBEDSB increases with decreasing oxygen concentration, which may create opportunities to apply proton dose painting to help address tumor hypoxia. Modeling of the particle RBE for DSB induction across multiple physical and biological scales has the potential to aid in the interpretation of laboratory experiments and provide useful information to advance the safety and effectiveness of hadron therapy in the treatment of cancer.

  1. ARM User Survey Report

    Energy Technology Data Exchange (ETDEWEB)

    Roeder, LR

    2010-06-22

    The objective of this survey was to obtain user feedback to, among other things, determine how to organize the exponentially growing data within the Atmospheric Radiation Measurement (ARM) Climate Research Facility, and identify users’ preferred data analysis system. The survey findings appear to have met this objective, having received approximately 300 responses that give insight into the type of work users perform, usage of the data, percentage of data analysis users might perform on an ARM-hosted computing resource, downloading volume level where users begin having reservations, opinion about usage if given more powerful computing resources (including ability to manipulate data), types of tools that would be most beneficial to them, preferred programming language and data analysis system, level of importance for certain types of capabilities, and finally, level of interest in participating in a code-sharing community.

  2. The User Reconfigured

    DEFF Research Database (Denmark)

    Bardzell, Jeffrey; Bardzell, Shaowen

    2015-01-01

    , and activism. We argue that subjectivi- ties of information clarifies the relationships between de- sign choices and embodied experiences, ways that designers design users and not just products, and ways to cultivate and transform, rather than merely support, human agency.......—by laying out what it means and why research- ers are being drawn to it. We then use it to guide a case study of a relatively marginal use of computing—digitally mediated sexuality—to holistically explore design in rela- tion to embodiment, tactual experience, sociability, power, ideology, selfhood......Foundational to HCI is the notion of “the user.” Whether a cognitive processor, social actor, consumer, or even a non- user, the user in HCI has always been as much a technical construct as actual people using systems. We explore an emerging formulation of the user—the subjectivity of in- formation...

  3. Interactive Office user's manual

    Science.gov (United States)

    Montgomery, Edward E.; Lowers, Benjamin; Nabors, Terri L.

    1990-01-01

    Given here is a user's manual for Interactive Office (IO), an executive office tool for organization and planning, written specifically for Macintosh. IO is a paperless management tool to automate a related group of individuals into one productive system.

  4. User Interface History

    DEFF Research Database (Denmark)

    Jørgensen, Anker Helms; Myers, Brad A

    2008-01-01

    User Interfaces have been around as long as computers have existed, even well before the field of Human-Computer Interaction was established. Over the years, some papers on the history of Human-Computer Interaction and User Interfaces have appeared, primarily focusing on the graphical interface era...... and early visionaries such as Bush, Engelbart and Kay. With the User Interface being a decisive factor in the proliferation of computers in society and since it has become a cultural phenomenon, it is time to paint a more comprehensive picture of its history. This SIG will investigate the possibilities...... of  launching a concerted effort towards creating a History of User Interfaces. ...

  5. "Playing" with our users

    DEFF Research Database (Denmark)

    Brooks, Anthony Lewis

    2014-01-01

    was from the amazing Dr Anthony Lewis Brooks (aka Tony) who has conceived the concepts GameAbilitation, ArtAbilitation, and Ludic Engagement Designs for All. While presenting some of his work on GameAbilitation and ArtAbilitation he brought up the subject of conducting research with users with disabilities......, about what happens to our users when research is over, funds are gone and the curtain of experiments has fallen. Dr Brooks presented the case of a young user who while unable to move and communicate had to part with the test device that provided him with interactive playful experience. We’ve all been...... confined in a house. For researchers that work with people with disabilities and in my case with playful interactions and positive immersive experience, we might have to think harder when we write project proposals or sketch our methodology. Devices, software and experience should be available to the users...

  6. Water, Energy, and Biogeochemical Model (WEBMOD), user’s manual, version 1

    Science.gov (United States)

    Webb, Richard M.T.; Parkhurst, David L.

    2017-02-08

    The Water, Energy, and Biogeochemical Model (WEBMOD) uses the framework of the U.S. Geological Survey (USGS) Modular Modeling System to simulate fluxes of water and solutes through watersheds. WEBMOD divides watersheds into model response units (MRU) where fluxes and reactions are simulated for the following eight hillslope reservoir types: canopy; snowpack; ponding on impervious surfaces; O-horizon; two reservoirs in the unsaturated zone, which represent preferential flow and matrix flow; and two reservoirs in the saturated zone, which also represent preferential flow and matrix flow. The reservoir representing ponding on impervious surfaces, currently not functional (2016), will be implemented once the model is applied to urban areas. MRUs discharge to one or more stream reservoirs that flow to the outlet of the watershed. Hydrologic fluxes in the watershed are simulated by modules derived from the USGS Precipitation Runoff Modeling System; the National Weather Service Hydro-17 snow model; and a topography-driven hydrologic model (TOPMODEL). Modifications to the standard TOPMODEL include the addition of heterogeneous vertical infiltration rates; irrigation; lateral and vertical preferential flows through the unsaturated zone; pipe flow draining the saturated zone; gains and losses to regional aquifer systems; and the option to simulate baseflow discharge by using an exponential, parabolic, or linear decrease in transmissivity. PHREEQC, an aqueous geochemical model, is incorporated to simulate chemical reactions as waters evaporate, mix, and react within the various reservoirs of the model. The reactions that can be specified for a reservoir include equilibrium reactions among water; minerals; surfaces; exchangers; and kinetic reactions such as kinetic mineral dissolution or precipitation, biologically mediated reactions, and radioactive decay. WEBMOD also simulates variations in the concentrations of the stable isotopes deuterium and oxygen-18 as a result of varying inputs, mixing, and evaporation. This manual describes the WEBMOD input and output files, along with the algorithms and procedures used to simulate the hydrology and water quality in a watershed. Examples are presented that demonstrate hydrologic processes, weathering reactions, and isotopic evolution in an alpine watershed and the effect of irrigation on water flows and salinity in an intensively farmed agricultural area.

  7. Model-Based Spectrum Management. Part 1: Modeling and Computation Manual, Version 2.0

    Science.gov (United States)

    2013-12-01

    systems such as Wireless Interoperability for Microwave Access ( WiMAX ) and Long Term Evolution (LTE) technologies. The fundamental concept underlying...Geodetic System WiMAX Wireless Interoperability for Microwave Access WRC World Radiocommunication Conference XML Extensible Markup Language F-3 This page intentionally left blank.

  8. Durra: A Task-Level Description Language Reference Manual. Version 2

    Science.gov (United States)

    1989-09-01

    Introduction Durra, also called "Indian millet" and "Guinea corn," Is a type of grain sorghum with slender stalks, widely grown in warm dry regions. Durra...GlobalProcessName ProcessNameListperiod ProcessQueueName GlobalProcessName I Globa lQueueName 6 CMU/SEI-89-TR-34 1.6. Literal Values Each of the non

  9. E&V (Evaluation and Validation) Manual. Version 1.0.

    Science.gov (United States)

    1987-12-29

    THE MANUAL Chapter 2 discusses the structure of the E&V Reference System (Refer- ence Manual plus Guidebook) and the Classification Schema upon which...should also be specified. This is because if the scores for reliability are high, but the software is incorrect or difficult to verify, the true...7.1.6.9] Experiments, Monitored (3.4.4) t- xpert System, APSE Viewed as a Knowledge-Based (3.2.5) Export (see: Import/Export] Fault Tolerance [Anomaly

  10. MOPAC Manual (Version 3.10) a General Molecular Orbital Package -- IBM-PC Version.

    Science.gov (United States)

    1986-08-20

    DCART DEBUG - DEBUG OPTION TURNED ON DEBUGPULAY PRINT DETAILS OF WORKING IN PULAY DENOUT - DENSITY MATRIX OUTPUT (CHANNEL 10) DENSITY - FINAL DENSITY...This is useful in debugging ITER. DEBUG can also increase the amount of output produced when a key-word is used, e.g. COMPFG. DENOUT The density...occurs during the current run then DENOUT is invoked automatically. (see RESTART) DENSITY At the end of a job, when the results are being printed, the

  11. MOPAC Manual (Version 3.10) A General Molecular Orbital Package -- Cray- XMP Version

    Science.gov (United States)

    1986-06-01

    PRINT DETAILS OF WORKING IN PULAY DENOUT - DENSITY MATRIX OUTPUT (CHANNEL 10) DENSITY - FINAL DENSITY MATRIX TO BE PRINTED DEP - GENERATE...the amount of output produced when a key-word Is used, e.g. COMPFG. - 15 - KEY-MORDS Page 2-8 DENOUT The density matrix at the end of the...then DENOUT Is Invoked automatically, (see RESTART) DENSITY At the end of a job. when the results are being printed, the density matrix Is also

  12. SIERRA Low Mach Module: Fuego Theory Manual Version 4.44

    Energy Technology Data Exchange (ETDEWEB)

    Sierra Thermal/Fluid Team

    2017-04-01

    The SIERRA Low Mach Module: Fuego along with the SIERRA Participating Media Radiation Module: Syrinx, henceforth referred to as Fuego and Syrinx, respectively, are the key elements of the ASCI fire environment simulation project. The fire environment simulation project is directed at characterizing both open large-scale pool fires and building enclosure fires. Fuego represents the turbulent, buoyantly-driven incompressible flow, heat transfer, mass transfer, combustion, soot, and absorption coefficient model portion of the simulation software. Syrinx represents the participating-media thermal radiation mechanics. This project is an integral part of the SIERRA multi-mechanics software development project. Fuego depends heavily upon the core architecture developments provided by SIERRA for massively parallel computing, solution adaptivity, and mechanics coupling on unstructured grids.

  13. FAST User Guide

    Science.gov (United States)

    Walatka, Pamela P.; Clucas, Jean; McCabe, R. Kevin; Plessel, Todd; Potter, R.; Cooper, D. M. (Technical Monitor)

    1994-01-01

    The Flow Analysis Software Toolkit, FAST, is a software environment for visualizing data. FAST is a collection of separate programs (modules) that run simultaneously and allow the user to examine the results of numerical and experimental simulations. The user can load data files, perform calculations on the data, visualize the results of these calculations, construct scenes of 3D graphical objects, and plot, animate and record the scenes. Computational Fluid Dynamics (CFD) visualization is the primary intended use of FAST, but FAST can also assist in the analysis of other types of data. FAST combines the capabilities of such programs as PLOT3D, RIP, SURF, and GAS into one environment with modules that share data. Sharing data between modules eliminates the drudgery of transferring data between programs. All the modules in the FAST environment have a consistent, highly interactive graphical user interface. Most commands are entered by pointing and'clicking. The modular construction of FAST makes it flexible and extensible. The environment can be custom configured and new modules can be developed and added as needed. The following modules have been developed for FAST: VIEWER, FILE IO, CALCULATOR, SURFER, TOPOLOGY, PLOTTER, TITLER, TRACER, ARCGRAPH, GQ, SURFERU, SHOTET, and ISOLEVU. A utility is also included to make the inclusion of user defined modules in the FAST environment easy. The VIEWER module is the central control for the FAST environment. From VIEWER, the user can-change object attributes, interactively position objects in three-dimensional space, define and save scenes, create animations, spawn new FAST modules, add additional view windows, and save and execute command scripts. The FAST User Guide uses text and FAST MAPS (graphical representations of the entire user interface) to guide the user through the use of FAST. Chapters include: Maps, Overview, Tips, Getting Started Tutorial, a separate chapter for each module, file formats, and system

  14. VOLTTRON: User Guide

    Energy Technology Data Exchange (ETDEWEB)

    Lutes, Robert G.; Katipamula, Srinivas; Akyol, Bora A.; Tenney, Nathan D.; Haack, Jereme N.; Monson, Kyle E.; Carpenter, Brandon J.

    2014-04-24

    This document is a user guide for the deployment of the Transactional Network platform and agent/application development within the VOLTTRON. The intent of this user guide is to provide a description of the functionality of the Transactional Network Platform. This document describes how to deploy the platform, including installation, use, guidance, and limitations. It also describes how additional features can be added to enhance its current functionality.

  15. SILMUSCEN and CLIGEN User`s Guide

    Energy Technology Data Exchange (ETDEWEB)

    Carter, T.; Tuomenvirta, H. [Finnish Meteorological Inst., Helsinki (Finland); Posch, M. [Water and Environment Research Inst., Helsinki (Finland)

    1995-12-31

    This User`s Guide has been prepared to provide recommendations for the selection and application of climatic scenarios in the Finnish Research Programme on Climate Change (SILMU). These scenarios are required for conducting impact studies in SILMU. They should reflect the current range of estimates of future climate in the Finnish region. In addition, they should be consistent with other projections of importance in impact studies, such as future atmospheric composition and sea level. Section 2 provides some background information about the types of scenarios required in SILMU and Section 3 offers a general description of the scenarios. In Section 4 there is some advice on applying sensitivity studies to complement the use of scenarios. Section 5 explains the installation of the SILMUSCEN program and Section 6 guides the user through some examples to illustrate how SILMUSCEN can be used. Section 7 offers some recommendations on which scenarios to adopt for different impact assessments. In order to ensure some compatibility between impact studies in SILMU, it is very important that the recommendations in this section are followed as far as possible. Section 8 addresses important omissions from the computer program and suggests procedures to adopt in their absence. Section 9 explores alternative methods of specifying the baseline climate, and shows how scenario adjustments to the baseline can be made. in Section 10, the stochastic weather generator, CLIGEN, is described and its use illustrated by means of examples. Finally, possible refinements of the programs are outlined in Section 11, along with contact names and addresses for obtaining further information. (36 refs.)

  16. Metadata: A user`s view

    Energy Technology Data Exchange (ETDEWEB)

    Bretherton, F.P. [Univ. of Wisconsin, Madison, WI (United States); Singley, P.T. [Oak Ridge National Lab., TN (United States)

    1994-12-31

    An analysis is presented of the uses of metadata from four aspects of database operations: (1) search, query, retrieval, (2) ingest, quality control, processing, (3) application to application transfer; (4) storage, archive. Typical degrees of database functionality ranging from simple file retrieval to interdisciplinary global query with metadatabase-user dialog and involving many distributed autonomous databases, are ranked in approximate order of increasing sophistication of the required knowledge representation. An architecture is outlined for implementing such functionality in many different disciplinary domains utilizing a variety of off the shelf database management subsystems and processor software, each specialized to a different abstract data model.

  17. Hanford inventory program user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Hinkelman, K.C.

    1994-09-12

    Provides users with instructions and information about accessing and operating the Hanford Inventory Program (HIP) system. The Hanford Inventory Program is an integrated control system that provides a single source for the management and control of equipment, parts, and material warehoused by Westinghouse Hanford Company in various site-wide locations. The inventory is comprised of spare parts and equipment, shop stock, special tools, essential materials, and convenience storage items. The HIP replaced the following systems; ACA, ASP, PICS, FSP, WSR, STP, and RBO. In addition, HIP manages the catalog maintenance function for the General Supplies inventory stocked in the 1164 building and managed by WIMS.

  18. Neutron flux distribution inside the cylindrical core of minor excess of reactivity in the IPEN/MB-01 reactor and comparison with citation code and MCNP- 5 code

    Energy Technology Data Exchange (ETDEWEB)

    Aredes, Vitor Ottoni; Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto C.; Santos, Diogo Feliciano dos; Lima, Ana Cecilia de Souza, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    This study aims to determine the distribution of thermal neutron flux in the IPEN/MB-01 nuclear reactor core assembled with cylindrical core configuration of minor excess of reactivity with 568 fuel rods (28 fuel rods in diameter). The thermal neutron flux at the positions of irradiation derive from the method of reaction rate using gold foils. The experiment consists in inserting gold activations foils with and without cadmium coverage (cadmium boxes with 0.0502 cm thickness) in several positions throughout the active core. After irradiation, activity induced by nuclear reaction rates over gold foils is assessed by gamma ray spectrometry using a high-purity germanium (HPGe) detector. Experimental results are compared to those derived from calculations performed using a three dimensional CITATION diffusion code and MCNP-5 code and a proper nuclear data library. While calculated neutron flux data shows good agreement with experimental values in regions with little disturbance in the neutron flux, also showing that in the region of the reflectors of neutrons and near the control rods, the diffusion theory is not very precise. The average value of thermal neutron flux obtained experimentally compared to the calculated value by CITATION code and MCNP-5 code respectively show a difference of 1.18% and 0.84% at a nuclear power level of 74.65 ± 3.28 % watts. The average measured value of thermal neutron flux is 4.10 10{sup 8} ± 5.25% n/cm{sup 2}s. (author)

  19. Investigation of reactivity variations of the Isfahan MNSR reactor due to variations in the thickness of the core top beryllium layer using WIMSD and MCNP codes

    Directory of Open Access Journals (Sweden)

    A Shirani

    2010-12-01

    Full Text Available In this work, the Isfahan Miniature Neutron Source Reactor (MNSR is first simulated using the WIMSD code, and its fuel burn-up after 7 years of operation ( when the reactor was revived by adding a 1.5 mm thick beryllium shim plate to the top of its core and also after 14 years of operation (total operation time of the reactor is calculated. The reactor is then simulated using the MCNP code, and its reactivity variation due to adding a 1.5 mm thick beryllium shim plate to the top of the reactor core, after 7 years of operation, is calculated. The results show good agreement with the available data collected at the revival time. Exess reactivity of the reactor at present time (after 14 years of operation and after 7 years of the the reactor revival time is also determined both experimentally and by calculation, which show good agreement, and indicate that at the present time there is no need to add any further beryllium shim plate to the top of the reactor core. Furthermore, by adding more beryllium layers with various thicknesses to the top of the reactor core, in the input program of the MCNP program, reactivity value of these layers is calculated. From these results, one can predict the necessary beryllium thickness needed to reach a desired reactivity in the MNSR reactor.

  20. Different Materials for γ-ray Attenuation Coefficient of MCNP Simulation%不同材料对γ射线衰减系数的 MCNP 模拟

    Institute of Scientific and Technical Information of China (English)

    刘自霞; 陆春海; 陈敏; 张志程; 黄硕

    2013-01-01

    线衰减系数是研究材料屏蔽性能的重要参数。为了研究防辐射材料的线衰减系数,基于蒙特卡罗方法,运用MCNP程序模拟不同厚度、不同材料、不同组分材料对γ射线的线衰减系数,进而对材料的线衰减系数做了精确的计算,分析了几种材料的能谱图。同时建立了屏蔽材料的衰减系数-能量拟合方程,并分析了误差。模拟计算结果表明:相同能量下,不同厚度的同种材料线衰减系数相同;线衰减系数跟原子序数Z、材料的密度、组成成分有关;在0.02~1 MeV能量区间,几种材料的线衰减系数都是随着能量的增加而减小;PbO-B2 O3玻璃系统中,随着PbO含量的增加,其线衰减系数也增大。%Linear attenuation coefficient is important parameters in the study of the materials shielding perform -ance.In order to study the linear attenuation coefficient of the radiation protective material shielding , based on the Monte Carlo method , the MCNP program is used to simulate the linear attenuation coefficient of gamma rays of different thickness、different materials、different components of material .And precise calculations of material linear attenuation coefficient are carried out , the energy spectrum of several materials is analyzed .Also the line-ar attenuation coefficient -energy fitting equation of shielding materials is established , with the error analysed . Simulation results show that:under the same energy , same material has the same linear attenuation coefficient with different thickness;linear attenuation coefficient is associated with atomic number Z , the density of materi-al and composition of materials;In the energy range of 0.02~1 keV, with the increase of source energy E , lin-ear attenuation coefficient of several materials shows the tendency of decreases .In the PbO-B2 O3 glass system , with the increase of the content of PbO , the linear attenuation coefficient is also increased .

  1. Evaluating User Participation and User Influence in an Enterprise System

    Science.gov (United States)

    Gibbs, Martin D.

    2010-01-01

    Does user influence have an impact on the data quality of an information systems development project? What decision making should users have? How can users effectively be engaged in the process? What is success? User participation is considered to be a critical success factor for Enterprise Resource Planning (ERP) projects, yet there is little…

  2. Evaluating User Participation and User Influence in an Enterprise System

    Science.gov (United States)

    Gibbs, Martin D.

    2010-01-01

    Does user influence have an impact on the data quality of an information systems development project? What decision making should users have? How can users effectively be engaged in the process? What is success? User participation is considered to be a critical success factor for Enterprise Resource Planning (ERP) projects, yet there is little…

  3. GRSAC Users Manual

    Energy Technology Data Exchange (ETDEWEB)

    Ball, S.J.; Nypaver, D.J.

    1999-02-01

    An interactive workstation-based simulation code (GRSAC) for studying postulated severe accidents in gas-cooled reactors has been developed to accommodate user-generated input with ''smart front-end'' checking. Code features includes on- and off-line plotting, on-line help and documentation, and an automated sensitivity study option. The code and its predecessors have been validated using comparisons with a variety of experimental data and similar codes. GRSAC model features include a three-dimensional representation of the core thermal hydraulics, and optional ATWS (anticipated transients without scram) capabilities. The user manual includes a detailed description of the code features, and includes four case studies which guide the user through four different examples of the major uses of GRSAC: an accident case; an initial conditions setup and run; a sensitivity study; and the setup of a new reactor model.

  4. Engaging with users

    DEFF Research Database (Denmark)

    Riisberg, Vibeke; Bang, Anne Louise

    to change the education of future designers. This is an emerging field at a number of design schools across the world, among these Design School Kolding in Denmark. In this paper we discuss ways in which we as design educators can teach fashion and textile students ways to engage with users during...... the creative process. To a large degree it is not common to engage direct with users in fashion and textile design. However, we see an increasing interest in this subject among the design students and also in recent research within fashion and textiles. We therefore argue that there is a need for participatory...... with the biggest sense organ – our skin. Thus, the aim of our research is to develop new dialogue tools for teaching fashion and textile students in order to stimulate new ways of thinking and engaging with users. By developing and employing participatory design methods in the field of fashion and textiles, we...

  5. End User Evaluations

    Science.gov (United States)

    Jay, Caroline; Lunn, Darren; Michailidou, Eleni

    As new technologies emerge, and Web sites become increasingly sophisticated, ensuring they remain accessible to disabled and small-screen users is a major challenge. While guidelines and automated evaluation tools are useful for informing some aspects of Web site design, numerous studies have demonstrated that they provide no guarantee that the site is genuinely accessible. The only reliable way to evaluate the accessibility of a site is to study the intended users interacting with it. This chapter outlines the processes that can be used throughout the design life cycle to ensure Web accessibility, describing their strengths and weaknesses, and discussing the practical and ethical considerations that they entail. The chapter also considers an important emerging trend in user evaluations: combining data from studies of “standard” Web use with data describing existing accessibility issues, to drive accessibility solutions forward.

  6. GLAST User Support

    Science.gov (United States)

    Band, David L.; Science Support Center, GLAST

    2006-12-01

    The Gamma-ray Large Area Space Telescope (GLAST) mission will provide the user community with many scientific opportunities. The mission's interface with the user community is the GLAST Science Support Center (GSSC). Yearly guest investigator (GI) cycles will support research related to GLAST. After the first year GIs may propose pointed observations; however, as a consequence of the large field-of-view of GLAST's instruments, pointed observations will rarely have an advantage over the default survey mode. Data, analysis software and documentation will be provided through the GSSC website (http://glast.gsfc.nasa.gov/ssc/); the website also includes a library of scientific results, and a helpdesk.

  7. User Centered Design

    DEFF Research Database (Denmark)

    Egbert, Maria; Matthews, Ben

    2012-01-01

    The interdisciplinary approach of User Centered Design is presented here with a focus on innovation in the design and use of hearing technologies as well as on the potential of innovation in interaction. This approach is geared towards developing new products, systems, technologies and practices...... based on an understanding of why so few persons with hearing loss use the highly advanced hearing technologies. In integrating Conversation Analysis (“CA”), audiology and User Centered Design, three disciplines which are collaborating together for the first time, we are addressing the following...

  8. EPRINT ARCHIVE USER SURVEY

    CERN Multimedia

    2001-01-01

    University of Southampton invites the CERN community to participate in a survey Professor Stevan Harnad is conducting on current users and non-users of Eprint Archives. http://www.eprints.org/survey/ The findings will be used to suggest potential enhancements of the services as well as to get a deeper understanding of the very rapid developments in the on-line dissemination and use of scientific and scholarly research. (The survey is anonymous. Revealing your identity is optional and it will be kept confidential.)

  9. RADTRAN 5 user guide.

    Energy Technology Data Exchange (ETDEWEB)

    Kanipe, Frances L.; Neuhauser, Karen Sieglinde

    2003-07-01

    This User Guide for the RADTRAN 5 computer code for transportation risk analysis describes basic risk concepts and provides the user with step-by-step directions for creating input files by means of either the RADDOG input file generator software or a text editor. It also contains information on how to interpret RADTRAN 5 output, how to obtain and use several types of important input data, and how to select appropriate analysis methods. Appendices include a glossary of terms, a listing of error messages, data-plotting information, images of RADDOG screens, and a table of all data in the internal radionuclide library.

  10. TRANS-USERS

    DEFF Research Database (Denmark)

    to redesign production and business processes to accommodate for users' requirements (Maisons MACCHI), and the client as driver of innovation on the construction and renovation of the low budget hotel brand Formule 1 of ACCOR Hotels. In the third part, the discussion and conclusion addresses three interlinked...... in Denmark, Sweden and France. The five case studies are: The industrialised home building concept BoKlok, a web based product configurator for kitchens by HTH, the innovative potential of the dual role of employees as both user and employee in Rockwool, the application of quality management systems...

  11. The OSIRIS user guide

    CERN Document Server

    Telling, M T F

    2003-01-01

    This user guide contains all the information necessary to perform a successful neutron scattering experiment on the OSIRIS spectrometer at ISIS, RAL, UK. Since OSIRIS is a continually evolving and improving instrument some information contained within this manual may become redundant. However, the basic instrument operating procedures should remain essentially unchanged. While updated manuals will be produced when appropriate, the most comprehensive source of information concerning OSIRIS is the Instrument Scientist/Local Contact. It would be appreciated, however, if this user guide were the first point of call should problems arise

  12. TIA Software User's Manual

    Science.gov (United States)

    Cramer, K. Elliott; Syed, Hazari I.

    1995-01-01

    This user's manual describes the installation and operation of TIA, the Thermal-Imaging acquisition and processing Application, developed by the Nondestructive Evaluation Sciences Branch at NASA Langley Research Center, Hampton, Virginia. TIA is a user friendly graphical interface application for the Macintosh 2 and higher series computers. The software has been developed to interface with the Perceptics/Westinghouse Pixelpipe(TM) and PixelStore(TM) NuBus cards and the GW Instruments MacADIOS(TM) input-output (I/O) card for the Macintosh for imaging thermal data. The software is also capable of performing generic image-processing functions.

  13. "Playing" with our users

    DEFF Research Database (Denmark)

    Brooks, Anthony Lewis

    2014-01-01

    . Unfortunately if donated in the school they are rarely being used by the students. In the case of virtual reality or artistic installations it is extremely difficult to provide such equipment to users. Last but not least we are not sure how the software will be used and if the experience will continue...... after the conduct of the research. If not due to restrictions, user should at least continue to be part of the research’s debrief and next steps. While I was in Nottingham I realised that sometimes our research, our playful educational experience, our DIY VR helmet, our beta, glitchy, research-only game...

  14. Perspectives on User Satisfaction Surveys.

    Science.gov (United States)

    Cullen, Rowena

    2001-01-01

    Discusses academic libraries, digital environments, increasing competition, the relationship between service quality and user satisfaction, and user surveys. Describes the SERVQUAL model that measures service quality and user satisfaction in academic libraries; considers gaps between user expectations and managers' perceptions of user…

  15. Personal lifelong user model clouds

    DEFF Research Database (Denmark)

    Dolog, Peter; Kay, Judy; Kummerfeld, Bob

    This paper explores an architecture for very long term user modelling, based upon personal user model clouds. These ensure that the individual's applications can access their model whenever it is needed. At the same time, the user can control the use of their user model. So, they can ensure...

  16. MCNP Simulations of Measurement of Insulation Compaction in the Cryogenic Rocket Fuel Tanks at Kennedy Space Center by Fast/Thermal Neutron Techniques

    Science.gov (United States)

    Livingston, R. A.; Schweitzer, J. S.; Parsons, A. M.; Arens, E. E.

    2010-01-01

    MCNP simulations have been run to evaluate the feasibility of using a combination of fast and thermal neutrons as a nondestructive method to measure of the compaction of the perlite insulation in the liquid hydrogen and oxygen cryogenic storage tanks at John F. Kennedy Space Center (KSC). Perlite is a feldspathic volcanic rock made up of the major elements Si, AI, Na, K and 0 along with some water. When heated it expands from four to twenty times its original volume which makes it very useful for thermal insulation. The cryogenic tanks at Kennedy Space Center are spherical with outer diameters of 69-70 feet and lined with a layer of expanded perlite with thicknesses on the order of 120 cm. There is evidence that some of the perlite has compacted over time since the tanks were built 1965, affecting the thermal properties and possibly also the structural integrity of the tanks. With commercially available portable neutron generators it is possible to produce simultaneously fluxes of neutrons in two energy ranges: fast (14 Me V) and thermal (25 me V). The two energy ranges produce complementary information. Fast neutrons produce gamma rays by inelastic scattering, which is sensitive to Fe and O. Thermal neutrons produce gamma rays by prompt gamma neutron activation (PGNA) and this is sensitive to Si, Al, Na, K and H. The compaction of the perlite can be measured by the change in gamma ray signal strength which is proportional to the atomic number densities of the constituent elements. The MCNP simulations were made to determine the magnitude of this change. The tank wall was approximated by a I-dimensional slab geometry with an 11/16" outer carbon steel wall, an inner stainless wall and 120 cm thick perlite zone. Runs were made for cases with expanded perlite, compacted perlite or with various void fractions. Runs were also made to simulate the effect of adding a moderator. Tallies were made for decay-time analysis from t=0 to 10 ms; total detected gamma

  17. OASIS User Manual

    CERN Document Server

    Bojtar, L

    2009-01-01

    The OASIS system has been operational for years now. After a long development the project has reached a state where the number of features it provides exceeds largely what most of its users knows about. The author felt it was time to write a user manual explaining all the functionality of the viewer application. This document is a user manual, concentrating on the functionality of the viewer from the user’s point of view. There are already documents available on the project’s web site about the technical aspects at http://project-oasis.web.cern.ch/project-oasis/presentations.htm . There was an attempt to produce a tutorial on the viewer, but it didn’t get much further than the table of contents, that however is well thought. The structure of this user manual follows the same principle, the basic and most often used features are grouped together. Advanced or less often used features are described in a separate chapter. There is a second organizational principle, features belong to different levels: chann...

  18. Users Office - Removal

    CERN Multimedia

    CERN Bulletin

    2010-01-01

    As of 8 December 2010 and until the end of February 2011, the Users Office will move from Bldg. 60. New Location : Bldg. 510-R-033 Opening Hours: Monday, Tuesday, Thursday, Friday : 08.30 – 12.30 Monday to Friday: 14.00 – 16.00 Closed Wednesday mornings.

  19. Educating the Music User

    Science.gov (United States)

    Adams, Mark C.

    2016-01-01

    To better serve students' evolving needs in music, music educators must connect classroom learning with how students use and interact with music in their daily lives. One way to accomplish this is by approaching classrooms with the music user in mind, which can open new possibilities for meaningful music making and remove students from the…

  20. Power User Interface

    Science.gov (United States)

    Pfister, Robin; McMahon, Joe

    2006-01-01

    Power User Interface 5.0 (PUI) is a system of middleware, written for expert users in the Earth-science community, PUI enables expedited ordering of data granules on the basis of specific granule-identifying information that the users already know or can assemble. PUI also enables expert users to perform quick searches for orderablegranule information for use in preparing orders. PUI 5.0 is available in two versions (note: PUI 6.0 has command-line mode only): a Web-based application program and a UNIX command-line- mode client program. Both versions include modules that perform data-granule-ordering functions in conjunction with external systems. The Web-based version works with Earth Observing System Clearing House (ECHO) metadata catalog and order-entry services and with an open-source order-service broker server component, called the Mercury Shopping Cart, that is provided separately by Oak Ridge National Laboratory through the Department of Energy. The command-line version works with the ECHO metadata and order-entry process service. Both versions of PUI ultimately use ECHO to process an order to be sent to a data provider. Ordered data are provided through means outside the PUI software system.

  1. Educating the Music User

    Science.gov (United States)

    Adams, Mark C.

    2016-01-01

    To better serve students' evolving needs in music, music educators must connect classroom learning with how students use and interact with music in their daily lives. One way to accomplish this is by approaching classrooms with the music user in mind, which can open new possibilities for meaningful music making and remove students from the…

  2. TO STORES USERS

    CERN Multimedia

    SPL Division

    2001-01-01

    Stores users are informed that the Stores (Central, Emergency window, Raw materials, Chemical products and Prévessin Self service stores) will be closed on Friday, 7 December owing to migration of the Stores computers to Windows 2000. Thank you for your understanding.

  3. The User Interface.

    Science.gov (United States)

    Lindeman, Martha J.; And Others

    1989-01-01

    The first of three articles on the design of user interfaces for information retrieval systems discusses the need to examine types of display, prompting, and input as separate entities. The second examines the use of artificial intelligence in creating natural language interfaces, and the third outlines standards for case studies in human computer…

  4. Users, Bystanders and Agents

    DEFF Research Database (Denmark)

    Krummheuer, Antonia Lina

    2015-01-01

    Human-agent interaction (HAI), especially in the field of embodied conversational agents (ECA), is mainly construed as dyadic communication between a human user and a virtual agent. This is despite the fact that many application scenarios for future ECAs involve the presence of others. This paper...

  5. Usability Testing of User Manuals

    DEFF Research Database (Denmark)

    Møller, Margrethe H.

    2013-01-01

    Many guidelines and several standards exist for the development of good user manuals. But even though technical writers comply with all guidelines, problems will typically arise when users apply the manual in practice. Therefore, it is useful to have real users test the manual before it is publis...... it is published. This article discusses user tests in the form of think-aloud tests, with examples from the research project ”User Manuals for older adults"....

  6. SHARP User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States); Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Mahadevan, Vijay S. [Argonne National Lab. (ANL), Argonne, IL (United States); Rahaman, Ronald O. [Argonne National Lab. (ANL), Argonne, IL (United States); Solberg, Jerome [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-03-31

    SHARP is an advanced modeling and simulation toolkit for the analysis of nuclear reactors. It is comprised of several components including physical modeling tools, tools to integrate the physics codes for multi-physics analyses, and a set of tools to couple the codes within the MOAB framework. Physics modules currently include the neutronics code PROTEUS, the thermal-hydraulics code Nek5000, and the structural mechanics code Diablo. This manual focuses on performing multi-physics calculations with the SHARP ToolKit. Manuals for the three individual physics modules are available with the SHARP distribution to help the user to either carry out the primary multi-physics calculation with basic knowledge or perform further advanced development with in-depth knowledge of these codes. This manual provides step-by-step instructions on employing SHARP, including how to download and install the code, how to build the drivers for a test case, how to perform a calculation and how to visualize the results. Since SHARP has some specific library and environment dependencies, it is highly recommended that the user read this manual prior to installing SHARP. Verification tests cases are included to check proper installation of each module. It is suggested that the new user should first follow the step-by-step instructions provided for a test problem in this manual to understand the basic procedure of using SHARP before using SHARP for his/her own analysis. Both reference output and scripts are provided along with the test cases in order to verify correct installation and execution of the SHARP package. At the end of this manual, detailed instructions are provided on how to create a new test case so that user can perform novel multi-physics calculations with SHARP. Frequently asked questions are listed at the end of this manual to help the user to troubleshoot issues.

  7. Use of the MCNP code for analysis of the attenuation of the radiation produced by radioactive sources used in radiotherapy in skin tumors; Uso do codigo MCNP para analise da atenuacao da radiacao produzida por fontes radioativas utilizadas em radioterapia em tumores de pele

    Energy Technology Data Exchange (ETDEWEB)

    Tada, A., E-mail: ariane.tada@gmail.co [Instituto de Pesquisas Energeticas e Nucleares (CEN/IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear; Instituto de Pesquisas Tecnologicas (IPT), Sao Paulo, SP (Brazil); Salles, T.; Yoriyaz, H., E-mail: hyoriyaz@ipen.b, E-mail: tasallesc@gmail.co [Instituto de Pesquisas Energeticas e Nucleares (CEN/IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear; Fernandes, M.A.R, E-mail: marfernandes@fmb.unesp.b [Universidade Estadual Paulista Julio de Mesquita Filho (UNESP), Botucatu, SP (Brazil). Fac. de Medicina. Dept. de Dermatologia e Radioterapia

    2010-07-01

    The present work had as objective to analyze the distribution profile of a therapeutic dose of radiation produced by radioactive sources used in radiotherapy procedures in superficial lesions on the skin. The experimental measurements for analysis of dosimetric radiation sources were compared with calculations obtained from the computer system based on the Monte Carlo Method. The results obtained by the computations calculations using the code MCNP-4C showed a good agreement with the experimental measurements. A comparison of different treatment modalities allows an indication of more appropriate procedures for each clinical case. (author)

  8. Acceleration of MCNP calculations for small pipes configurations by using Weigth Windows Importance cards created by the SN-3D ATTILA

    Science.gov (United States)

    Castanier, Eric; Paterne, Loic; Louis, Céline

    2017-09-01

    In the nuclear engineering, you have to manage time and precision. Especially in shielding design, you have to be more accurate and efficient to reduce cost (shielding thickness optimization), and for this, you use 3D codes. In this paper, we want to see if we can easily applicate the CADIS methods for design shielding of small pipes which go through large concrete walls. We assess the impact of the WW generated by the 3D-deterministic code ATTILA versus WW directly generated by MCNP (iterative and manual process). The comparison is based on the quality of the convergence (estimated relative error (σ), Variance of Variance (VOV) and Figure of Merit (FOM)), on time (computer time + modelling) and on the implement for the engineer.

  9. Calculations of the thermal and fast neutron fluxes in the Syrian miniature neutron source reactor using the MCNP-4C code.

    Science.gov (United States)

    Khattab, K; Sulieman, I

    2009-04-01

    The MCNP-4C code, based on the probabilistic approach, was used to model the 3D configuration of the core of the Syrian miniature neutron source reactor (MNSR). The continuous energy neutron cross sections from the ENDF/B-VI library were used to calculate the thermal and fast neutron fluxes in the inner and outer irradiation sites of MNSR. The thermal fluxes in the MNSR inner irradiation sites were also measured experimentally by the multiple foil activation method ((197)Au (n, gamma) (198)Au and (59)Co (n, gamma) (60)Co). The foils were irradiated simultaneously in each of the five MNSR inner irradiation sites to measure the thermal neutron flux and the epithermal index in each site. The calculated and measured results agree well.

  10. Observing the user experience a practitioner's guide to user research

    CERN Document Server

    Kuniavsky, Mike; Goodman, Elizabeth

    2012-01-01

    The gap between who designers and developers imagine their users are, and who those users really are can be the biggest problem with product development. Observing the User Experience will help you bridge that gap to understand what your users want and need from your product, and whether they'll be able to use what you've created. Filled with real-world experience and a wealth of practical information, this book presents a complete toolbox of techniques to help designers and developers see through the eyes of their users. It provides in-depth coverage of 13 user experience research techniques

  11. Bias estimates used in lieu of validation of fission products and minor actinides in MCNP Keff calculations for PWR burnup credit casks

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Don [ORNL; Marshall, William BJ J [ORNL; Wagner, John C [ORNL; Bowen, Douglas G [ORNL

    2015-09-01

    The U.S. Nuclear Regulatory Commission (NRC) Division of Spent Fuel Storage and Transportation recently issued Interim Staff Guidance (ISG) 8, Revision 3. This ISG provides guidance for burnup credit (BUC) analyses supporting transport and storage of PWR pressurized water reactor (PWR) fuel in casks. Revision 3 includes guidance for addressing validation of criticality (keff) calculations crediting the presence of a limited set of fission products and minor actinides (FP&MA). Based on previous work documented in NUREG/CR-7109, recommendation 4 of ISG-8, Rev. 3, includes a recommendation to use 1.5 or 3% of the FP&MA worth to conservatively cover the bias due to the specified FP&MAs. This bias is supplementary to the bias and bias uncertainty resulting from validation of keff calculations for the major actinides in SNF and does not address extension to actinides and fission products beyond those identified herein. The work described in this report involves comparison of FP&MA worths calculated using SCALE and MCNP with ENDF/B-V, -VI, and -VII based nuclear data and supports use of the 1.5% FP&MA worth bias when either SCALE or MCNP codes are used for criticality calculations, provided the other conditions of the recommendation 4 are met. The method used in this report may also be applied to demonstrate the applicability of the 1.5% FP&MA worth bias to other codes using ENDF/B V, VI or VII based nuclear data. The method involves use of the applicant s computational method to generate FP&MA worths for a reference SNF cask model using specified spent fuel compositions. The applicant s FP&MA worths are then compared to reference values provided in this report. The applicants FP&MA worths should not exceed the reference results by more than 1.5% of the reference FP&MA worths.

  12. IT User Community Survey

    CERN Document Server

    Peter Jones (IT-CDA-WF)

    2016-01-01

    IT-CDA is gathering information to more accurately form a snapshot of the CERN IT user community and we would appreciate you taking time to complete the following survey.   We want to use this survey to better understand how the user community uses their devices and our services, and how the delivery of those services could be improved. You will need to authenticate to complete the survey. However please note that your responses are confidential and will be compiled together and analysed as a group. You can also volunteer to offer additional information if you so wish. This survey should take no longer than 5 minutes. Thanks in advance for your collaboration.

  13. Trilinos users guide.

    Energy Technology Data Exchange (ETDEWEB)

    Willenbring, James M.; Heroux, Michael Allen

    2003-08-01

    The Trilinos Project is an effort to facilitate the design, development, integration and ongoing support of mathematical software libraries. A new software capability is introduced into Trilinos as a package. A Trilinos package is an integral unit usually developed by a small team of experts in a particular algorithms area such as algebraic preconditioners, nonlinear solvers, etc. The Trilinos Users Guide is a resource for new and existing Trilinos users. Topics covered include how to configure and build Trilinos, what is required to integrate an existing package into Trilinos and examples of how those requirements can be met, as well as what tools and services are available to Trilinos packages. Also discussed are some common practices that are followed by many Trilinos package developers. Finally, a snapshot of current Trilinos packages and their interoperability status is provided, along with a list of supported computer platforms.

  14. Portraying User Interface History

    DEFF Research Database (Denmark)

    Jørgensen, Anker Helms

    2008-01-01

    in that they largely address prevailing UI techno­logies, and thirdly history from above in that they focus on the great deeds of the visionaries. The paper then compares this state-of-art in UI history to the much more mature fields history of computing and history of technology. Based hereon, some speculations......The user interface is coming of age. Papers adressing UI history have appeared in fair amounts in the last 25 years. Most of them address particular aspects such as an in­novative interface paradigm or the contribution of a visionary or a research lab. Contrasting this, papers addres­sing UI...... history at large have been sparse. However, a small spate of publications appeared recently, so a reasonable number of papers are available. Hence this work-in-progress paints a portrait of the current history of user interfaces at large. The paper first describes a theoretical framework recruited from...

  15. Information for stores users

    CERN Multimedia

    Logistics Group - FI Department

    2005-01-01

    The Farnell catalogue can now be accessed from the Material Request form on EDH in addition to the CERN Stores catalogue. Users can order Farnell equipment as well as standard Stores equipment at the same time using a single document, the EDH Materials Request form. The Materials Request form offers users items from both the internal 'Stores' catalogue and the external 'Farnell' catalogue, all of which may be ordered on the same form. The system automatically forwards orders for standard Stores equipment to the CERN Stores and those for Farnell equipment to Farnell. The delivery time is 48 hours in both cases. Requests for materials are routed for approval in accordance with the standard EDH routing procedures. Logistics Group FI Department

  16. Information for stores users

    CERN Multimedia

    2005-01-01

    The Farnell catalogue can now be accessed from the Material Request form on EDH in addition to the CERN Stores catalogue. Users can order Farnell equipment as well as standard Stores equipment at the same time using a single document, the EDH Materials Request form. The Materials Request form offers users items from both the internal 'Stores' catalogue and the external 'Farnell' catalogue, all of which may be ordered on the same form. The system automatically forwards orders for standard Stores equipment to the CERN Stores and those for Farnell equipment to Farnell. The delivery time is 48 hours in both cases. Requests for materials are routed for approval in accordance with the standard EDH routing procedures. Logistics Group FI Department

  17. Internet user behaviour

    Directory of Open Access Journals (Sweden)

    Radbâță, A.

    2011-01-01

    Full Text Available Internet is a useful tool for everybody in a technologically advanced world. As Internet appears and develops, it creates a totally new network environment. The development of commerce on the Internet based on virtual communities has become one of the most successful business models in the world. After analyzing the concept of internet, the e-commerce market and its marketing mix and the benefits and limitations of the Internet, we have presented a few studies on Internet user behaviour. Furthermore, the paper looks at a representative sample of Romanian internet users. The results reveal that the Romanians are using the Internet especially for information gathering, e-mail, entertainment and social networking.

  18. Outside users payload model

    Science.gov (United States)

    1985-01-01

    The outside users payload model which is a continuation of documents and replaces and supersedes the July 1984 edition is presented. The time period covered by this model is 1985 through 2000. The following sections are included: (1) definition of the scope of the model; (2) discussion of the methodology used; (3) overview of total demand; (4) summary of the estimated market segmentation by launch vehicle; (5) summary of the estimated market segmentation by user type; (6) details of the STS market forecast; (7) summary of transponder trends; (8) model overview by mission category; and (9) detailed mission models. All known non-NASA, non-DOD reimbursable payloads forecast to be flown by non-Soviet-block countries are included in this model with the exception of Spacelab payloads and small self contained payloads. Certain DOD-sponsored or cosponsored payloads are included if they are reimbursable launches.

  19. MP users guide

    CERN Document Server

    Brent, Richard P

    2010-01-01

    MP is a package of ANSI Standard Fortran (ANS X3.9-1966) subroutines for performing multiple-precision floating-point arithmetic and evaluating elementary and special functions. The subroutines are machine independent and the precision is arbitrary, subject to storage limitations. The User's Guide describes the routines and their calling sequences, example and test programs, use of the Augment precompiler, and gives installation instructions for the package.

  20. 16. ESRF users meeting

    Energy Technology Data Exchange (ETDEWEB)

    Coraux, J.; Renevier, H.; Favre-Nicolin, V.; Daudin, B.; Proietti, M.G.; Renaud, G.; Fowler, B.; Mercer, D.L.; Omar, A.H.; Thompson, P.; Markovic, N.M.; Stamenkovic, V.; Lucas, C.A.; Andrejczuk, A.; Kwiatkowska, J.; Dobrzynski, L.; Zukowski, E.; Bellin, Ch.; Loupias, G.; Shukla, A.; Buslaps, Th.; Stankov, S.; Sladecek, M.; Slezak, T.; Korecki, J.; Spiridis, N.; Sepiol, B.; Vogl, G.; Chumakov, A.; Ruffer, R.; Hermann, R.P.; Grandjean, F.; Schweika, W.; Long, G.J.; Leupold, O.; Belrhall, H.; Caserotto, H.; Dauvergne, F.; Geoffroy, L.; Guljarro, M.; Launer, L.; Levault, B.; Walsh, M.; Beckers, M.; Schell, N.; Martins, R.M.S.; Mucklich, A.; Moller, W.; Silva, R.J.C.; Mahesh, K.K.; Braz Fernandes, F.M.; Tejas, Parikh; Neil, Fellows; Durodola, J.; Slawinski, W.; Przenioslo, R.; Sosnowska, I.; Suard, E

    2006-07-01

    This document gathers the posters that were presented during the poster session of this workshop. These posters highlight the results obtained by ESRF'users in different fields such as surface structure, Compton scattering studies, localized vibrational modes in thermoelectric materials, Ni-Ti thin films, residual stresses in superconducting wires, and changes in crystal and magnetic structure of NdFeO{sub 3}.

  1. Personal lifelong user model clouds

    DEFF Research Database (Denmark)

    Dolog, Peter; Kay, Judy; Kummerfeld, Bob

    This paper explores an architecture for very long term user modelling, based upon personal user model clouds. These ensure that the individual's applications can access their model whenever it is needed. At the same time, the user can control the use of their user model. So, they can ensure...... it is accessed only when and where they wish, by applications that they wish. We consider the challenges of representing user models so that they can be reused by multiple applications. We indicate potential synergies between distributed and centralised user modelling architectures, proposing an architecture...

  2. User Communities i Innovationsprocessen

    DEFF Research Database (Denmark)

    Brix, Jacob; Sejer Jakobsen, Henning; Jordansen, Inger

    to workshops uden fysisk tilstedeværelse af deltagerne med internettet som kommunikationskanal (online user communities via Skype & blogs). Empirien stammer fra BDI projektet Handivision1, hvor målgruppen og brugergruppen primært har været personer med fysiske funktionsnedsættelser. Vores analyse indikerer...... for at virke uvidende sammenlignet med andre meddeltagere. 3. At mødes fysisk under en innovationsworkshop har betydning for måden hvorpå deltagerne bliver inspireret af - og lærer af hinanden samt deres evnen til at blive konkrete 4. Online user communities, som styres af en konsulterende leder, resulterer i...... stor udstrækning i problemorienterede forslag og ideer, hvor dynamikken deltagerne imellem er svær at opretholde 5. Online user communities, der ikke styres eller ledes i processen, resulterer i større udstrækning end ved mere styrede forløb i problemerkendelse frem for forslag og ideer til nytænkning....

  3. Users in Persistant Action

    DEFF Research Database (Denmark)

    Christiansen, John K.; Gasparin, Marta; Varnes, Claus J.

    2012-01-01

    years before a 15 years-old- boy wanted the 1.5 litres back to the market, even though Coca-Cola resisted, he managed by the hybrid collective to struggle with Coca-Cola and convince them to re-introduce the 1.5 litres volume by various interessment devices, including buy-cot to frame the power relation......This study adds to the concept of lead users by investigating the role users´ post launch. The case of the 1.5 litre Urge bottle in Norway shows that what constitutes a ‘lead’ becomes an effect of the product displaced in a hybrid collective in time and space. The hybrid collective is an assumption...... in which realities are constructed in contrast to the assumption of diffusion in society, where reality is given and determined. The theory lead users is closely related to the product life cycle in the diffusion perspective, as they both progress linearly. The 1.5 litres was removed from the market 8...

  4. User interface inspection methods a user-centered design method

    CERN Document Server

    Wilson, Chauncey

    2014-01-01

    User Interface Inspection Methods succinctly covers five inspection methods: heuristic evaluation, perspective-based user interface inspection, cognitive walkthrough, pluralistic walkthrough, and formal usability inspections. Heuristic evaluation is perhaps the best-known inspection method, requiring a group of evaluators to review a product against a set of general principles. The perspective-based user interface inspection is based on the principle that different perspectives will find different problems in a user interface. In the related persona-based inspection, colleagues assume the

  5. Workflow User Interfaces Patterns

    Directory of Open Access Journals (Sweden)

    Jean Vanderdonckt

    2012-03-01

    Full Text Available Este trabajo presenta una colección de patrones de diseño de interfaces de usuario para sistemas de información para el flujo de trabajo; la colección incluye cuarenta y tres patrones clasificados en siete categorías identificados a partir de la lógica del ciclo de vida de la tarea sobre la base de la oferta y la asignación de tareas a los responsables de realizarlas (i. e. recursos humanos durante el flujo de trabajo. Cada patrón de la interfaz de usuario de flujo de trabajo (WUIP, por sus siglas en inglés se caracteriza por las propiedades expresadas en el lenguaje PLML para expresar patrones y complementado por otros atributos y modelos que se adjuntan a dicho modelo: la interfaz de usuario abstracta y el modelo de tareas correspondiente. Estos modelos se especifican en un lenguaje de descripción de interfaces de usuario. Todos los WUIPs se almacenan en una biblioteca y se pueden recuperar a través de un editor de flujo de trabajo que vincula a cada patrón de asignación de trabajo a su WUIP correspondiente.A collection of user interface design patterns for workflow information systems is presented that contains forty three resource patterns classified in seven categories. These categories and their corresponding patterns have been logically identified from the task life cycle based on offering and allocation operations. Each Workflow User Interface Pattern (WUIP is characterized by properties expressed in the PLML markup language for expressing patterns and augmented by additional attributes and models attached to the pattern: the abstract user interface and the corresponding task model. These models are specified in a User Interface Description Language. All WUIPs are stored in a library and can be retrieved within a workflow editor that links each workflow pattern to its corresponding WUIP, thus giving rise to a user interface for each workflow pattern.

  6. User constraints for reliable user-defined smart home scenarios

    DEFF Research Database (Denmark)

    Le Guilly, Thibaut; Nielsen, Michael Kvist; Pedersen, Thomas

    2016-01-01

    of constraints restricting the control commands that can be used inside user-defined scenarios. The system is based on timed automata model checking abstracted by event condition action rules. A prototype was implemented, including a user interface to interact with the user. The usability of the system...

  7. Adding and Removing Web Area Users, and Changing User Roles

    Science.gov (United States)

    Webmasters can add users to a web area, and assign or change roles, which define the actions a user is able to take in the web area. Non-webmasters must use a request form to add users and change roles.

  8. User producer interaction in context

    NARCIS (Netherlands)

    Nahuis, R.; Moors, E.H.M.; Smits, R.E.H.M.

    2012-01-01

    User producer interaction (UPI) increases chances for successful innovations. It is not always clear, however, what type of interaction is necessary in a particular context. This article identifies seven different types of UPI: constructing linkages, broadening, characterizing users, upstream

  9. Accessible Capacity of Secondary Users

    CERN Document Server

    Ma, Xiao; Lin, Lei; Bai, Baoming

    2010-01-01

    A new problem formulation is presented for the Gaussian interference channels (GIFC) with two pairs of users, which are distinguished as primary users and secondary users, respectively. The primary users employ a pair of encoder and decoder that were originally designed to satisfy a given error performance requirement under the assumption that no interference exists from other users. In the case when the secondary users attempt to access the same medium, we are interested in the maximum transmission rate (defined as {\\em accessible capacity}) at which secondary users can communicate reliably without affecting the error performance requirement by the primary users under the constraint that the primary encoder (not the decoder) is kept unchanged. By modeling the primary encoder as a generalized trellis code (GTC), we are then able to treat the secondary link as a finite state channel (FSC). The relation of the accessible capacity to the capacity region of the GIFC is revealed. Upper and lower bounds on the acce...

  10. User Control Problems and Taking User Empowerment Further

    Science.gov (United States)

    Rodrigues, Rowena

    User control in identity management is beset with a number of problems, as outlined in this paper. It is argued that akin to traditional contexts, greater user control will result in greater user liability, which is demonstrated with the help of digital and non-digital examples. In this context, there is a critical need for greater user empowerment. This could be achieved in two ways-first, facilitating user awareness of identity management technologies, their scope and effects and second, through the implementation of proposed control-liability notices.

  11. Measuring user experience : what's new?

    NARCIS (Netherlands)

    Cremers, A.H.M.; Smets, N.; Vermeeren, A.; Kort, J.

    2007-01-01

    This paper proposes a short overview of characteristics of different user evaluation methods and a research framework to systematically compare these different methods. Comparisons will be carried out in the context of Freeband user experience studies. Results will provide more insight into how user

  12. Measuring user experience : what's new?

    NARCIS (Netherlands)

    Cremers, A.H.M.; Smets, N.; Vermeeren, A.; Kort, J.

    2007-01-01

    This paper proposes a short overview of characteristics of different user evaluation methods and a research framework to systematically compare these different methods. Comparisons will be carried out in the context of Freeband user experience studies. Results will provide more insight into how user

  13. What drives Users' Website Registration?

    NARCIS (Netherlands)

    T. Li (Ting); P.A. Pavlou (Paul)

    2013-01-01

    textabstractUser registration is an important prerequisite for the success of many websites by enabling users to gain access to domain information and personalized content. It is not always desirable for users, however, because they need to disclose personal information. This paper examines what dri

  14. Evaluation from a user perspective

    DEFF Research Database (Denmark)

    Krogstrup, Hanne Kathrine

    2004-01-01

    Over the last decade, user participation has been placed on the agenda in many contexts and also in relation to evaluation. The reasons for user participation in evaluation are based om several overlapping arguments. In this contexts four arguments for user participation are discussed: a control ...... argument, a democratic argument, a knowledge argument and an emancipatory argument...

  15. MOSS user's manual

    Science.gov (United States)

    Salmen, Larry; Gropper, James; Hamill, John; Gentry, Barbara

    1978-01-01

    The Map Overlay and Statistical System (MOSS) Users' Manual is specialized document has been designed for trained users of the MOSS interactive graphics software. Those totally unfamiliar with MOSS or Geographic Information Systems are referred elsewhere as described below: -- If you know nothing about MOSS or what it can do for you, and you wish introductory information on MOSS, or you want to deign an application and data entry process compatible with MOSS, or you want "hands-on" training, contact the WELUT Team Leader at the address below for a "hands-on" GIS training session. -- If you have been introduced to MOSS, have your application defined, data entered, and want to know how to use MOSS, start reading at Section 1 of this Manual. --If you are interested in the MOSS data structure, refer to Section 2 of this Manual. --If you have some experience in using MOSS and want to refer to the general types of MOSS commands, read Section 3 of this Manual. --If you are an experience MOSS user and want details on individual MOSS commands, refer to Section 4.3 of this Manual. --If you are interested in the Federation of Rocky Mountain States -- WELUT 02 Project contractual background results, turn to Appendices D and E of this Manual. MOSS has been operation for less than 3 months, and has received limited operational testing at the date of this printing (October 1978). Undiscovered software limitations and bugs may yet appear. All such bugs as well as documentation errors, obscurities, and inadequacies should be reported to: Team Leader

  16. Cliffs User Manual

    CERN Document Server

    Tolkova, Elena

    2014-01-01

    Cliffs is an open-source relative of MOST (Method Of Splitting Tsunamis) numerical model, implemented as described in (Tolkova, 2014, Pure and Appl. Geophys., 171(9), 2289-2314). Cliffs features: Shallow-Water approximation with an option to manipulate numerical dispersion; Use of Cartesian or spherical (lon/lat) coordinates; 1D and 2D configurations; Structured co-located grid with (optionally) varying spacing; Runup on land; Initial conditions or boundary forcing; Grid nesting with one-way coupling; Parallelized with OpenMP; NetCDF format of input/output data. This user manual accompanies Cliffs code distribution.

  17. 15. ESRF users meeting

    Energy Technology Data Exchange (ETDEWEB)

    Fotis C, Kafatos; Ulrich, K.U.; Weib, S.; Rossberg, A.; Scheinost, A.C.; Foerstendorf, H.; Zanker, H.; Meyerheim, H.L.; Sander, D.; Popescu, R.; Kirschner, J.; Robach, O.; Ferrer, S.; Lyman, P.F.; Shneerson, V.L.; Fung, R.; Harder, R.J.; Parihar, S.S.; Johnson-Steigelman, H.T.; Lu, E.D.; Saldin, D.K.; Eastwood, D.S.; Atkinson, D.; Tanner, B.K.; Hase, T.P.A.; Van Kampen, M.; Hjorvarsson, B.; Brown, S.; Thompson, P.; Konovalov, O.; Saint-Martin, E.; Daillant, J.; Luzet, D.; Szlachetko, J.; Barrett, R.; Berset, M.; Dousse, J.C.; Fennane, K.; Hoszowska, J.; Kubala-Kukus, A.; Pajek, M.; Szlachetko, M.; Monaco, A.; Chumakov, A.; Crichton, W.; Van Buerck, I.; Wortmann, G.; Meyer, A.; Ponkratz, U.; Ruffer, R.; Sakurai, Y.; Hiraoka, N.; Itou, M.; Buslaps, T.; Honkimki, V.; Maeno, Y.; Collart, E.; Shukla, A.; Rueff, J.P.; Leininger, Ph.; Ishii, H.; Cai, Y.; Cheong, S.W.; Martins, R.M.S.; Schell, N.; Beckers, M.; Silva, R.; Braz Fernandes, F.M.; Acapito, F.; Seta, M. de; Capelini, G.; Giorgi, M.; Schorr, G.; Geandier, G.; Alves Marques, M.; Barros Marquesa, M.I. de; Cabaco, M.I.; Gaspara, A.M.; Marques, M.P.M.; Amado, A.M.; Amorim da Costa, A.M.; Bruneseaux, F.; Weisbecker, P.; Brandao, M.J.; Aeby-Gautier, E.; Simmonds, H.; Lei, C.; Das, A.; Trolley, D.; Thomas, H.E.; Macdonald, J.E.; Wiegart, L.; Tolan, M.; Struth, B.; Petukhov, A.V.; Thijssen, J.H.J.; Hart, D.C.; Imhof, A.; Van Blaaderen, A.; Dolbnya, I.P.; Snigirev, A.; Mossaid, A.; Snigireva, I.; Reconditi, M.; Brunello, E

    2005-07-01

    This document gathers the posters presented on the one day and a half long plenary meeting workshop. This meeting workshop is a privileged forum where ESRF users can exchange their views on the latest scientific and technical development involving synchrotron radiation. One poster deals with the investigation of colloid composition and uranium bond structure to see whether the migration of contaminants from abandoned mines could be stimulated or attenuated by colloids. Another poster is dedicated to the investigation of the uranium speciation in covered mine tailings by a combination of micro-spectroscopic and wet chemical approaches. 2 posters deal with the contribution of synchrotron radiation to radiotherapy.

  18. User-Driven CHAOS

    DEFF Research Database (Denmark)

    Lykke, Marianne; Lund, Haakon; Skov, Mette

    2016-01-01

    CHAOS (Cultural Heritage Archive Open System) provides streaming access to more than 500.000 broad-casts by the Danish Broadcast Corporation from 1931 and onwards. The archive is part of the LARM project with the purpose of enabling researchers to search, annotate, and interact with recordings....... To optimally sup-port the researchers a user-centred approach was taken to develop the platform and related metadata scheme. Based on the requirements a three level metadata scheme was developed: (1) core archival metadata, (2) LARM metadata, and (3) project-specific metadata. The paper analyses how.......fm’s strength in providing streaming access to a large, shared corpus of broadcasts....

  19. User Ethnography as Theatre

    DEFF Research Database (Denmark)

    Torguet, Rosa; Friis, Preben; Buur, Jacob

    2013-01-01

    This paper discusses the potential of using theatre with professional actors to convey the outcome of ethnographic user studies to industry or academia. Framed in the ongoing discussion within design ethnography of how representations can support the communication of ethnographic findings more...... effectively. The use of theatre within innovation processes can help facilitate the provoking role that an ethnography often plays when presented to organizations. Live performances were used as part of a participatory innovation project in the field of indoor climate with industry partners and academic...

  20. XTV users guide

    Energy Technology Data Exchange (ETDEWEB)

    Dearing, J.F.; Johns, R.C. [Los Alamos National Lab., NM (United States). Technology and Safety Assessment Div.

    1996-09-01

    XTV is an X-Windows based Graphical User Interface for viewing results of Transient Reactor Analysis Code (TRAC) calculations. It provides static and animated color mapped visualizations of both thermal-hydraulic and heat conduction components in a TRAC model of a nuclear power plant, as well as both on-screen and hard copy two-dimensional plot capabilities. XTV is the successor to TRAP, the former TRAC postprocessor using the proprietary DISSPLA graphics library. This manual describes Version 2.0, which requires TRAC version 5.4.20 or later for full visualization capabilities.

  1. Raspberry Pi user guide

    CERN Document Server

    Upton, Eben

    2013-01-01

    The essential guide to getting started with the Raspberry Pi ® The Raspberry Pi has been a success beyond the dream of its creators. Their goal, to encourage a new generation of computer programmers who understand how computers work, is well under way. Raspberry Pi User Guide 2e is the newest edition of the runaway bestseller written by the Pi's co-creator, Eben Upton, and tech writer Gareth Halfacree. It contains everything you need to know to get the Pi up and running, including how to: Connect a keyboard, mouse, monitor and other peripheralsInstall software and configure your Raspberry

  2. Raspberry Pi user guide

    CERN Document Server

    Halfacree, Gareth

    2012-01-01

    Make the most out of the world’s first truly compact computer It's the size of a credit card, it can be charged like a smartphone, it runs on open-source Linux, and it holds the promise of bringing programming and playing to millions at low cost. And now you can learn how to use this amazing computer from its co-creator, Eben Upton, in Raspberry Pi User Guide. Cowritten with Gareth Halfacree, this guide gets you up and running on Raspberry Pi, whether you're an educator, hacker, hobbyist, or kid. Learn how to connect your Pi to other hardware, install software, write basic programs, an

  3. User and technical documentation

    Science.gov (United States)

    1988-01-01

    The program LP1 calculates outbound and return trajectories between low earth orbit (LEO) and libration point no. 1 (L1). Libration points (LP) are defined as locations in space that orbit the Earth such that they are always stationary with respect to the Earth-Moon line. L1 is located behind the Moon such that the pull of the Earth and Moon together just cancel the centrifugal acceleration associated with the libration point's orbit. The input required from the user to define the flight is described. The contents of the six reports produced as outputs are presented. Also included are the instructions needed to execute the program.

  4. Information for Stores users

    CERN Multimedia

    FI Department

    2008-01-01

    The DISTRELEC catalogue (IT) is now available in EDH in addition to the CERN Stores catalogue and the catalogues of existing suppliers. Using an EDH materials request form, users can now order DISTRELEC equipment from amongst the following product groups: peripherals, multimedia, PC components, data media, communication and data cables and adapters. Non-authorised materials will be clearly indicated. As a reminder, the system automatically manages the distribution of standard Stores equipment and punch out equipment ordered on the same request form. In both cases, delivery will take a maximum of 48 hours. The approval of the EDH document will follow the usual EDH routing procedures. Logistics Group FI Department

  5. Information for stores users

    CERN Multimedia

    2006-01-01

    The Radiospares Catalogue is now accessible from the Material Request page on EDH in the same way as the CERN Stores Catalogue. This means that users can order Radiospares equipment by completing an EDH Materials Request form. N.B.: The system will automatically forward orders for standard Stores equipment to the CERN Stores and those for Radiospares equipment to Radiospares. In both cases the delivery time will be a maximum of 48 hours. Requests for materials will be routed for approval in accordance with the standard EDH routing procedures. Logistics Group FI Department

  6. Information for stores users

    CERN Multimedia

    2007-01-01

    From next week, the SFS UNIMARKET (tooling) catalogue will be accessible using the Material Request form on EDH in addition to the CERN Stores catalogue and those of existing suppliers. Users will now be able to place orders from the SFS catalogue using the Material Request form on EDH. Note: The system automatically forwards orders for standard Stores equipment and those for SFS equipment, placed using the same Material Request form, to the CERN Stores and SFS respectively. In both cases, the maximum delivery time will be 48 hours. Requests for equipment will be routed for approval in accordance with standard EDH routing procedures. Logistics Group FI Department

  7. INFORMATION FOR STORES USERS

    CERN Multimedia

    2007-01-01

    From next week, the SFS UNIMARKET (tooling) catalogue will be accessible using the Material Request form on EDH in addition to the CERN Stores catalogue and those of existing suppliers. Users will now be able to place orders from the SFS catalogue using the Material Request form on EDH. Note: The system automatically forwards orders for standard Stores equipment and those for SFS equipment, placed using the same Material Request form, to the CERN Stores and SFS respectively. In both cases, the maximum delivery time will be 48 hours. Requests for equipment will be routed for approval in accordance with standard EDH routing procedures. Logistics Group FI Department

  8. Percept User Manual.

    Energy Technology Data Exchange (ETDEWEB)

    Carnes, Brian [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kennon, Stephen Ray [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-05-01

    This document is the main user guide for the Sierra/Percept capabilities including the mesh_adapt and mesh_transfer tools. Basic capabilities for uniform mesh refinement (UMR) and mesh transfers are discussed. Examples are used to provide illustration. Future versions of this manual will include more advanced features such as geometry and mesh smoothing. Additionally, all the options for the mesh_adapt code will be described in detail. Capabilities for local adaptivity in the context of offline adaptivity will also be included. This page intentionally left blank.

  9. XMGR5 users manual

    Energy Technology Data Exchange (ETDEWEB)

    Jones, K.R.; Fisher, J.E.

    1997-03-01

    ACE/gr is XY plotting tool for workstations or X-terminals using X. A few of its features are: User defined scaling, tick marks, labels, symbols, line styles, colors. Batch mode for unattended plotting. Read and write parameters used during a session. Polynomial regression, splines, running averages, DFT/FFT, cross/auto-correlation. Hardcopy support for PostScript, HP-GL, and FrameMaker.mif format. While ACE/gr has a convenient point-and-click interface, most parameter settings and operations are available through a command line interface (found in Files/Commands).

  10. Information for stores users

    CERN Multimedia

    2006-01-01

    The Radiospares Catalogue is now accessible from the Material Request page on EDH in the same way as the CERN Stores Catalogue. This means that users can order Radiospares equipment by completing an EDH Materials Request form. N.B.: The system will automatically forward orders for standard Stores equipment to the CERN Stores and those for Radiospares equipment to Radiospares. In both cases the delivery time will be a maximum of 48 hours. Requests for materials will be routed for approval in accordance with the standard EDH routing procedures. Logistics Group FI Department

  11. CDS User survey

    CERN Multimedia

    CERN Document Service

    2011-01-01

      The CERN Document Server is launching a user survey in order to collect information relative to its search engine, submission interfaces, collaborative features and content organisation. With the view of re-shaping its collections and interfaces and to better integrate with the new INSPIRE platform that serves all HEP literature, CERN Document Server team invites you to take part in the survey. Your input is essential to provide us with useful information before setting up the new service and improve your interactions with CDS. Thanks for participating !  

  12. Incident users of antipsychotics

    DEFF Research Database (Denmark)

    Baandrup, Lone; Kruse, Marie

    2016-01-01

    PURPOSE: In Denmark, as well as in many other countries, consumption of antipsychotics is on the rise, partly due to increasing off-label use. The aim of this study was to analyze and quantify the extent of off-label use and polypharmacy in incident users of antipsychotic medication, and to examine...... polypharmacy (HR 1.38; 95 % CI 1.32-1.45), whereas antipsychotic discontinuation was associated with decreased hospitalization risk in most off-label conditions. CONCLUSIONS: The brief duration of most antipsychotic prescriptions suggests that antipsychotics are prescribed more liberally than recommended...

  13. User Types in Online Applications

    Directory of Open Access Journals (Sweden)

    Ion IVAN

    2011-08-01

    Full Text Available Online applications are presented in the context of information society. Online applications characteristics are analyzed. Quality characteristics are presented in relation to online applications users. Types of users for AVIO application are presented. Use cases for AVIO application are identified. The limitations of AVIO application are defined. Types of users in online applications are identified. The threedimensional matrix of access to the online application resources is built. The user type-oriented database is structured. Access management of the fields related to the database tables is analyzed. The classification of online applications users is done.

  14. TRLAN User Guide

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Kesheng; Simon, H.

    1999-03-09

    TRLAN is a program designed to find a small number of extreme eigenvalues and their corresponding eigenvectors of a real symmetric matrix. Denote the matrix as A, the eigenvalue as {lambda}, and the corresponding eigenvector as x, they are defined by the following equation, Ax = {lambda}x. There are a number of different implementations of the Lanczos algorithm available. Why another one? Our main motivation is to develop a specialized version that only target the case where one wants both eigenvalues and eigenvectors of a large real symmetric eigenvalue problems that can not use the shift-and-invert scheme. In this case the standard non-restarted Lanczos algorithm requires one to store a large number of Lanczos vectors which can cause storage problem and make each iteration of the method very expensive. The underlying algorithm of TRLAN is a dynamic thick-restart Lanczos algorithm. Like all restarted methods, the user can choose how many vectors can be generated at once. Typically, th e user chooses a moderate size so that all Lanczos vectors can be stored in core. This allows the restarted methods to execute efficiently. This implementation of the thick-restart Lanczos method also uses the latest restarting technique, it is very effective in reducing the time required to compute a desired solutions compared to similar restarted Lanczos schemes, e.g., ARPACK.

  15. Photovoltaics information user study

    Energy Technology Data Exchange (ETDEWEB)

    Belew, W.W.; Wood, B.L.; Marie, T.L.; Reinhardt, C.L.

    1980-10-01

    The results of a series of telephone interviews with groups of users of information on photovoltaics (PV) are described. These results, part of a larger study on many different solar technologies, identify types of information each group needed and the best ways to get information to each group. The report is 1 of 10 discussing study results. The overall study provides baseline data about information needs in the solar community. It covers these technological areas: photovoltaics, passive solar heating and cooling, active solar heating and cooling, biomass energy, solar thermal electric power, solar industrial and agricultural process heat, wind energy, ocean energy, and advanced energy storage. An earlier study identified the information user groups in the solar community and the priority (to accelerate solar energy commercialization) of getting information to each group. In the current study only high-priority groups were examined. Results from seven PV groups respondents are analyzed in this report: DOE-Funded Researchers, Non-DOE-Funded Researchers, Researchers Working for Manufacturers, Representatives of Other Manufacturers, Representatives of Utilities, Electric Power Engineers, and Educators.

  16. Nephrolithiasis in topiramate users.

    Science.gov (United States)

    Maalouf, Naim M; Langston, Joshua P; Van Ness, Paul C; Moe, Orson W; Sakhaee, Khashayar

    2011-08-01

    Topiramate is a neuromodulatory agent increasingly prescribed for a number of neurological and non-neurological indications. Topiramate-treated patients are at risk for nephrolithiasis due to hypocitraturia and high urine pH. However, the prevalence of symptomatic stone disease in TPM users is generally perceived to be low. This study was undertaken to assess in topiramate-treated patients the prevalence of symptomatic nephrolithiasis (by history) and of asymptomatic nephrolithiasis by computed tomography (CT) scan. Topiramate users were identified from a database of patients with neurological disorders at a single university hospital. Among 75 topiramate-treated adult patients with a median daily dose of 300 mg and median treatment duration of 48 months, the prevalence of symptomatic nephrolithiasis was 10.7%. In a subset of topiramate-treated patients and no history of symptomatic stone disease, the prevalence of asymptomatic nephrolithiasis detected by CT scan was 20%. The prevalence of symptomatic nephrolithiasis with long-term topiramate use is higher than reported in short-term studies. Furthermore, clinical prevalence is underestimated due to asymptomatic nephrolithiasis.

  17. Electronic Commerce user manual

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-10

    This User Manual supports the Electronic Commerce Standard System. The Electronic Commerce Standard System is being developed for the Department of Defense of the Technology Information Systems Program at the Lawrence Livermore National Laboratory, operated by the University of California for the Department of Energy. The Electronic Commerce Standard System, or EC as it is known, provides the capability for organizations to conduct business electronically instead of through paper transactions. Electronic Commerce and Computer Aided Acquisition and Logistics Support, are two major projects under the DoD`s Corporate Information Management program, whose objective is to make DoD business transactions faster and less costly by using computer networks instead of paper forms and postage. EC runs on computers that use the UNIX operating system and provides a standard set of applications and tools that are bound together by a common command and menu system. These applications and tools may vary according to the requirements of the customer or location and may be customized to meet the specific needs of an organization. Local applications can be integrated into the menu system under the Special Databases & Applications option on the EC main menu. These local applications will be documented in the appendices of this manual. This integration capability provides users with a common environment of standard and customized applications.

  18. Electronic Commerce user manual

    Energy Technology Data Exchange (ETDEWEB)

    1992-04-10

    This User Manual supports the Electronic Commerce Standard System. The Electronic Commerce Standard System is being developed for the Department of Defense of the Technology Information Systems Program at the Lawrence Livermore National Laboratory, operated by the University of California for the Department of Energy. The Electronic Commerce Standard System, or EC as it is known, provides the capability for organizations to conduct business electronically instead of through paper transactions. Electronic Commerce and Computer Aided Acquisition and Logistics Support, are two major projects under the DoD's Corporate Information Management program, whose objective is to make DoD business transactions faster and less costly by using computer networks instead of paper forms and postage. EC runs on computers that use the UNIX operating system and provides a standard set of applications and tools that are bound together by a common command and menu system. These applications and tools may vary according to the requirements of the customer or location and may be customized to meet the specific needs of an organization. Local applications can be integrated into the menu system under the Special Databases Applications option on the EC main menu. These local applications will be documented in the appendices of this manual. This integration capability provides users with a common environment of standard and customized applications.

  19. Rivet user manual

    Science.gov (United States)

    Buckley, Andy; Butterworth, Jonathan; Grellscheid, David; Hoeth, Hendrik; Lönnblad, Leif; Monk, James; Schulz, Holger; Siegert, Frank

    2013-12-01

    This is the manual and user guide for the Rivet system for the validation and tuning of Monte Carlo event generators. As well as the core Rivet library, this manual describes the usage of the rivet program and the AGILe generator interface library. The depth and level of description is chosen for users of the system, starting with the basics of using validation code written by others, and then covering sufficient details to write new Rivet analyses and calculational components. Catalogue identifier: AEPS_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEPS_v1_0.html Program obtainable from: CPC Program Library, Queen’s University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 571126 No. of bytes in distributed program, including test data, etc.: 4717522 Distribution format: tar.gz Programming language: C++, Python. Computer: PC running Linux, Mac. Operating system: Linux, Mac OS. RAM: 20 MB Classification: 11.9, 11.2. External routines: HepMC (https://savannah.cern.ch/projects/hepmc/), GSL (http://www.gnu.org/software/gsl/manual/gsl-ref.html), FastJet (http://fastjet.fr/), Python (http://www.python.org/), Swig (http://www.swig.org/), Boost (http://www.boostsoftware.com/), YAML (http://www.yaml.org/spec/1.2/spec.html) Nature of problem: Experimental measurements from high-energy particle colliders should be defined and stored in a general framework such that it is simple to compare theory predictions to them. Rivet is such a framework, and contains at the same time a large collection of existing measurements. Solution method: Rivet is based on HepMC events, a standardised output format provided by many theory simulation tools. Events are processed by Rivet to generate histograms for the requested list of analyses, incorporating all experimental phase space cuts and histogram definitions. Restrictions: Cannot calculate

  20. 基于MCNP安全作业舱辐射屏蔽特性研究%Performance of Shielding Gamma Radiation Based on MCNP

    Institute of Scientific and Technical Information of China (English)

    邹树梁; 孙毅成; 唐德文; 徐守龙

    2015-01-01

    采用蒙特卡罗软件MCNP5程序建立挖掘机安全作业舱及内部驾驶人员实验模型,研究在辐射源(钴-60)类型不同和源距不同的情况下,不同材料及不同结构的安全作业舱屏蔽γ射线的能力.分析讨论人体各部位剂量率值以评价安全作业舱的屏蔽效果,并依据电离辐射防护与辐射源安全基本标准( GB18871—2002)规定的工作人员的照射水平应低于0.01 mSv/h的剂量限值,确定安全作业舱结构的设计方案.此次研究为挖掘机驾驶舱屏蔽设计提供必要的研究数据和设计依据,使驾驶舱在野外高放射环境下能够有效地屏蔽酌射线,保护驾驶人员的辐射屏蔽安全.%Excavator safety chamber for work and internal driver experimental models are constructed in a Monte Carlo program MCNP5 code, and the different types of radiation source and various of sources have effect on the dose ratios of different parts of human. In different situations, the different materials and different structure of security chamber for work can also affect the ability of shielding gamma rays performance. This paper analyzes the dose ratios of every part of human body to evaluate the shielding gamma rays perform-ance of safety chamber for work. On the basis of ionizing radiation protection and radioac-tive source security basic standards (GB18871—2002),the workers exposed to gamma ra-diation should absorb the dose ratios which is less than 0 . 01 mSv/h and the structure de-sign of chamber must meet the goal to protect the drivers. The study of excavator cockpit shielding design provides essential data for further study and provides the basis of design. Above all,the study makes the cockpit with exposures to the field of high gamma radiation safe and protect the driver from gamma radiation.

  1. PARALLELIZATION AND PERFECTION OF MCNP MONTE CARLO PARTICLE TRANSPORT CODE IN MPI%粒子输运蒙特卡罗程序MCNP在MPI下的并行化及完善

    Institute of Scientific and Technical Information of China (English)

    邓力; 刘杰; 张文勇

    2003-01-01

    The particle transport Monte Carlo code MCNP had been realized the paral-lelization in MPI (Message Passing Interface) in 1999. But due to adopting the leap random number producer, some differences were existed between the parallel result and the serial result. Now the same results have been achieved by using the segment random number. The speedup of the applied problem is the liner ups to 53 in 64-Processors and the parallel efficiencv is up to 83% in 64-Processors.

  2. Validation of MCNP6 Version 1.0 with the ENDF/B-VII.1 Cross Section Library for Plutonium Metals, Oxides, and Solutions on the High Performance Computing Platform Moonlight

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, Bryan Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gough, Sean T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-12-05

    This report documents a validation of the MCNP6 Version 1.0 computer code on the high performance computing platform Moonlight, for operations at Los Alamos National Laboratory (LANL) that involve plutonium metals, oxides, and solutions. The validation is conducted using the ENDF/B-VII.1 continuous energy group cross section library at room temperature. The results are for use by nuclear criticality safety personnel in performing analysis and evaluation of various facility activities involving plutonium materials.

  3. User Experience Dimensions

    DEFF Research Database (Denmark)

    Lykke, Marianne; Jantzen, Christian

    2016-01-01

    The present study develops a set of 10 dimensions based on a systematic understanding of the concept of experience as a holistic psychological. Seven of these are derived from a psychological conception of what experiencing and experiences are. Three supplementary dimensions spring from...... the observation that experiences apparently have become especially valuable phenomena in Western societies. The 10 dimensions are tried out in a field study at the Center for Art and Media (ZKM) in Germany with the purpose to study their applicability in the evaluation of interactive sound archives. 29 walk......-alongs were carried out with 58 museums visitors. Our analysis showed that it was possible to identify the 10 experience dimensions in the study material. Some dimensions were expressed more frequently than others. The distribution of expressed dimensions and the content of the user comments provided a clear...

  4. Information for stores users

    CERN Multimedia

    2006-01-01

    The Bossard catalogue is now accessible alongside the CERN Stores catalogue from the Material Request form on EDH. Users will thus be able to order Bossard equipment using the EDH Materials Request form. As a reminder, the system automatically forwards orders for standard Stores equipment to the CERN Stores and those for Bossard equipment to Bossard. In both cases the delivery time will be a maximum of 48 hours. Requests for materials will be routed for approval in accordance with the standard EDH routing procedures. Some items will remain available from the emergency desk in the event of urgent requests. These items will be visible in the Stores catalogue even if they cannot be purchased via the EDH material request form. Logistics Group FI Department

  5. INFORMATION FOR GAS USERS

    CERN Multimedia

    Logistics Group

    2001-01-01

    The contractor for the supply and distribution of pressurised gases has drawn our attention to the large number of gas bottles and banks being stored on the site for increasingly long periods. Users are reminded that the rental charges for gas bottles and banks are based on a progressive rate depending on their period of use. To assist CERN in its efforts to optimise its operations in this field, you are kindly requested : to return empty or unused containers to the official gas distribution points as soon as possible, to try to limit reserve stocks, bearing in mind that standardised gases can be delivered within 36 hours. This will result in a higher turnover rate and in increased safety and will improve the availability of the gases. For all further enquiries, please contact Gas.Store@cern.ch by e-mail or call 72265. Thank you for your co-operation.

  6. Communication to Linux users

    CERN Multimedia

    IT Department

    We would like to inform you that the aging “phone” Linux command will stop working: On lxplus on 30 November 2009, On lxbatch on 4 January 2010, and is replaced by the new “phonebook” command, currently available on SLC4 and SLC5 Linux. As the new “phonebook” command has different syntax and output formats from the “phone” command, please update and test all scripts currently using “phone” before the above dates. You can refer to the article published on the IT Service Status Board, under the Service Changes section. Please send any comments to it-dep-phonebook-feedback@cern.ch Best regards, IT-UDS User Support Section

  7. Information for gas users

    CERN Multimedia

    2003-01-01

    The contractor for the supply and distribution of pressurised gases has drawn our attention to the large number of gas bottles and banks being stored on the site for increasingly long periods. Users are reminded that the rental charges for gas bottles and banks are based on a progressive rate depending on their period of use. To assist CERN in its efforts to optimise its operations in this field, you are kindly requested: - to return empty or unused containers to the official gas distribution points as soon as possible - to try to limit reserve stocks, bearing in mind that standardised gases can be delivered within 36 hours. This will result in a higher turnover rate and in increased safety and will improve the availability of the gases. For all further enquiries, please contact "Gas store" by e-mail. Thank you for your co-operation. Logistics Group SPL Division

  8. INFORMATION FOR GAS USERS

    CERN Multimedia

    Logistics Group

    2001-01-01

    The contractor for the supply and distribution of pressurised gases has drawn our attention to the large number of gas bottles and banks being stored on the site for increasingly long periods. Users are reminded that the rental charges for gas bottles and banks are based on a progressive rate depending on their period of use. To assist CERN in its efforts to optimise its operations in this field, you are kindly requested : to return empty or unused containers to the official gas distribution points as soon as possible, to try to limit reserve stocks, bearing in mind that standardised gases can be delivered within 36 hours. This will result in a higher turnover rate and in increased safety and will improve the availability of the gases. For all further enquiries, please contact Gas.Store@cern.ch by e-mail or call 72265. Thank you for your co-operation.

  9. User involvement in care work

    DEFF Research Database (Denmark)

    Dybbroe, Betina; Kamp, Annette

    effectiveness and shared responsibility for care pathways. While NPM position users as consumers making their free choice, the user involvement paradigm underlines the users’ active participation in the mastering of their problems and disease. Research is scarce on this theme, and has until now primarily......In recent years user involvement has become a paradigm for transforming the health and social care sector. This development–also labelled empowerment, co-creation, partnership, patient-centeredness - is seen as a means to reform organizations in ways that enhance quality, economic cost...... addressed the way this paradigm affects the users, in specific sectors. However user involvement also affects working life. It may imply change and redistribution of tasks and identities between users and professionals, and may also transform the relations of care. In this paper we explore the possible...

  10. Time, Attitude, and User Participation

    DEFF Research Database (Denmark)

    Pries-Heje, Lene

    2008-01-01

    , equivocation, resistance and rejection depending on three things: (1) the dynamic between user and consultants, (2) the dynamic between different user groups, and (3) the understanding of technical, organizational and socio-technical options. When relating the empirical findings to existing theory on user...... be that the perception of usefulness of the system in any given phase of the implementation is heavily dependent on preceding events—the process. A process model analysis identifies eight episodes and nine encounters in the case showing that the user’s attitude towards the ERP system changes between acceptance...... participation, it is argued that the changes could be explained as a slide from influential user participation toward pseudo participation and back to influential participation, and that user participation in the context of ERP implementations raises new issues regarding user participation. Thus further...

  11. Evaluation of the thermal neutron flux in the core of IPEN/MB-01 reactor using the code Monte Carlo (MCNP)

    Energy Technology Data Exchange (ETDEWEB)

    Salome, Jean A.D.; Cardoso, Fabiano; Faria, Rochkhudson B.; Pereira, Claubia, E-mail: jadsalome@yahoo.com.br, E-mail: fabinuclear@yahoo.com.br, E-mail: rockdefaria@yahoo.com.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2015-07-01

    The IPEN/MB-01 reactor, located in the city of Sao Paulo - Brazil, reached its first criticality on the year of 1988. The reactor is characterized by a low output power of 100 W only, even because its purpose is to produce knowledge about nuclear power plants on a smaller geometric scale without the requirement of an extremely complex cooling system. The use of devices such as this it is very interesting because it achieves the demands of nuclear engineering about the neutronic parameters needed in the design of large nuclear plants through relatively simple and inexpensive methods. In this paper, the computational mathematical code MCNP5 is used to perform the calculation of the thermal neutron flux in the core of the IPEN/MB-01 reactor. To do this is used an experiment from the LEU-COMP-THERM-077 benchmark that represents the standard rectangular configuration of the IPEN/MB-01 reactor. The thermal neutron flux is calculated at some axial planes of different heights and, after that, axial profiles of the thermal neutron flux are done and compared to experimental results issued previously. The experimental values used as reference refer to a cylindrical configuration of the core of the reactor. Finally, the pertinence and relevance of the results are checked. With this work is expected to produce more knowledge about the dynamics of neutron flux in the core of the IPEN/MB-01 reactor. (author)

  12. Verification of TG-61 dose for synchrotron-produced monochromatic x-ray beams using fluence-normalized MCNP5 calculations

    CERN Document Server

    Brown, Thomas A D; Alvarez, Diane; Matthews, Kenneth L; Ham, Kyungmin; 10.1118/1.4761870

    2012-01-01

    Ion chamber dosimetry is being used to calibrate dose for cell irradiations designed to investigate photoactivated Auger electron therapy at the Louisiana State University CAMD synchrotron facility. This study performed a dosimetry intercomparison for synchrotron-produced monochromatic x-ray beams at 25 and 35 keV. Ion chamber depth-dose measurements in a PMMA phantom were compared with the product of MCNP5 Monte Carlo calculations of dose per fluence and measured incident fluence. Monochromatic beams of 25 and 35 keV were generated on the tomography beamline at CAMD. A cylindrical, air-equivalent ion chamber was used to measure the ionization created in a 10x10x10-cm3 PMMA phantom for depths from 0.6 to 7.7 cm. The American Association of Physicists in Medicine TG-61 protocol was applied to convert measured ionization into dose. Photon fluence was determined using a NaI detector to make scattering measurements of the beam from a thin polyethylene target at angles 30 degrees to 60 degrees. Differential Compto...

  13. Study of the source-detector system geometry using the MCNP-X code in the flowrate measurement with radioactive tracers

    Energy Technology Data Exchange (ETDEWEB)

    Avilan Puertas, Eddie, E-mail: epuertas@nuclear.ufrj.br [Universidad Central de Venezuela (UCV), Facultad de Ingenieria, Departamento de Fisica Aplicada, Caracas (Venezuela, Bolivarian Republic of); Braz, Delson, E-mail: delson@lin.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Brandao, Luis E.; Salgado, Cesar M., E-mail: brandao@ien.gov.br, E-mail: otero@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The use radioactive tracers for flow rate measurement is applied to a great variety of situations, however the accuracy of the technique is highly dependent of the adequate choice of the experimental measurement conditions. To measure flow rate of fluids in ducts partially filled, is necessary to measure the fluid flow velocity and the fluid height. The flow velocity can be measured with the cross correlation function and the fluid level, with a fluid level meter system. One of the error factors when measuring flow rate, is on the correct setting of the source-detector of the fluid level meter system. The goal of the present work is to establish by mean of MCNP-X code simulations the experimental parameters to measure the fluid level. The experimental tests will be realized in a flow rate system of 10 mm of diameter of acrylic tube for water and oil as fluids. The radioactive tracer to be used is the {sup 82}Br and for the detection will be employed two 1″ NaI(Tl) scintillator detectors, shielded with collimators of 0.5 cm and 1 cm of circular aperture diameter. (author)

  14. Calculation of Absorbed Dose in Target Tissue and Equivalent Dose in Sensitive Tissues of Patients Treated by BNCT Using MCNP4C

    Science.gov (United States)

    Zamani, M.; Kasesaz, Y.; Khalafi, H.; Pooya, S. M. Hosseini

    Boron Neutron Capture Therapy (BNCT) is used for treatment of many diseases, including brain tumors, in many medical centers. In this method, a target area (e.g., head of patient) is irradiated by some optimized and suitable neutron fields such as research nuclear reactors. Aiming at protection of healthy tissues which are located in the vicinity of irradiated tissue, and based on the ALARA principle, it is required to prevent unnecessary exposure of these vital organs. In this study, by using numerical simulation method (MCNP4C Code), the absorbed dose in target tissue and the equiavalent dose in different sensitive tissues of a patiant treated by BNCT, are calculated. For this purpose, we have used the parameters of MIRD Standard Phantom. Equiavelent dose in 11 sensitive organs, located in the vicinity of target, and total equivalent dose in whole body, have been calculated. The results show that the absorbed dose in tumor and normal tissue of brain equal to 30.35 Gy and 0.19 Gy, respectively. Also, total equivalent dose in 11 sensitive organs, other than tumor and normal tissue of brain, is equal to 14 mGy. The maximum equivalent doses in organs, other than brain and tumor, appear to the tissues of lungs and thyroid and are equal to 7.35 mSv and 3.00 mSv, respectively.

  15. Investigation of the Effects of Tissue Inhomogeneities on the Dosimetric Parameters of a Cs-137 Brachytherapy Source using the MCNP4C Code

    Directory of Open Access Journals (Sweden)

    Mehdi Zehtabian

    2010-09-01

    Full Text Available Introduction: Brachytherapy is the use of small encapsulated radioactive sources in close vicinity of tumors. Various methods are used to obtain the dose distribution around brachytherapy sources. TG-43 is a dosimetry protocol proposed by the AAPM for determining dose distributions around brachytherapy sources. The goal of this study is to update this protocol for presence of bone and air inhomogenities.  Material and Methods: To update the dose rate constant parameter of the TG-43 formalism, the MCNP4C simulations were performed in phantoms composed of water-bone and water-air combinations. The values of dose at different distances from the source in both homogeneous and inhomogeneous phantoms were estimated in spherical tally cells of 0.5 mm radius using the F6 tally. Results: The percentages of dose reductions in presence of air and bone inhomogenities for the Cs-137 source were found to be 4% and 10%, respectively. Therefore, the updated dose rate constant (Λ will also decrease by the same percentages.   Discussion and Conclusion: It can be easily concluded that such dose variations are more noticeable when using lower energy sources such as Pd-103 or I-125.

  16. Modeling the effect in of criticality from changes in key parameters for small High Temperature Nuclear Reactor (U-BatteryTM) using MCNP4C

    Science.gov (United States)

    Pauzi, A. M.

    2013-06-01

    The neutron transport code, Monte Carlo N-Particle (MCNP) which was wellkown as the gold standard in predicting nuclear reaction was used to model the small nuclear reactor core called "U-batteryTM", which was develop by the University of Manchester and Delft Institute of Technology. The paper introduces on the concept of modeling the small reactor core, a high temperature reactor (HTR) type with small coated TRISO fuel particle in graphite matrix using the MCNPv4C software. The criticality of the core were calculated using the software and analysed by changing key parameters such coolant type, fuel type and enrichment levels, cladding materials, and control rod type. The criticality results from the simulation were validated using the SCALE 5.1 software by [1] M Ding and J L Kloosterman, 2010. The data produced from these analyses would be used as part of the process of proposing initial core layout and a provisional list of materials for newly design reactor core. In the future, the criticality study would be continued with different core configurations and geometries.

  17. Monte Carlo calculation for the development of a BNCT neutron source (1eV-10KeV) using MCNP code.

    Science.gov (United States)

    El Moussaoui, F; El Bardouni, T; Azahra, M; Kamili, A; Boukhal, H

    2008-09-01

    Different materials have been studied in order to produce the epithermal neutron beam between 1eV and 10KeV, which are extensively used to irradiate patients with brain tumors such as GBM. For this purpose, we have studied three different neutrons moderators (H(2)O, D(2)O and BeO) and their combinations, four reflectors (Al(2)O(3), C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation showed that the best obtained assembly configuration corresponds to the combination of the three moderators H(2)O, BeO and D(2)O jointly to Al(2)O(3) reflector and two filter Cd+Bi optimize the spectrum of the epithermal neutron at 72%, and minimize the thermal neutron to 4% and thus it can be used to treat the deep tumor brain. The calculations have been performed by means of the Monte Carlo N (particle code MCNP 5C). Our results strongly encourage further studying of irradiation of the head with epithermal neutron fields.

  18. User interface for personal accounting

    OpenAIRE

    Femec, Vasilij

    2008-01-01

    This diploma work describes a method for user interface development for an application Bilanca that is intended for a review of personal financial flows. It is a simple application that subtracts outcome from income and shows the current financial state. The work begins with a detailed analysis of the best possible user interface options that give the most comfortable user experience. This is followed by the implementation in a Delphi environment. The results show that even a simple applicati...

  19. Profiting from innovative user communities

    DEFF Research Database (Denmark)

    Jeppesen, Lars Bo

    platforms. This article explains how manufacturers can profit from their abilities to organize and facilitate a process of innovation by user communities and capture the value of the innovations produced in such communities. When managed strategically, two distinct, but not mutually exclusive business......, a manufacturer can incorporate and commercialize the best complements found in the user communities. Keywords: innovation, modding, user communities, software platform, business model. JEL code(s): L21; L23; O31; O32...

  20. Library user metaphors and services

    DEFF Research Database (Denmark)

    Johannsen, Carl Gustav

    How do library professionals talk about and refer to library users, and how is this significant? In recent decades, the library profession has conceived of users in at least five different ways, viewing them alternatively as citizens, clients, customers, guests, or partners. This book argues...... that these user metaphors crucially inform librarians' interactions with the public, and, by extension, determine the quality and content of the services received. The ultimate aim of the book is to provide library professionals with insights and tools for avoiding common pitfalls associated with false...... or professionally inadequate conceptions of library users....

  1. Library user metaphors and services

    DEFF Research Database (Denmark)

    Johannsen, Carl Gustav

    How do library professionals talk about and refer to library users, and how is this significant? In recent decades, the library profession has conceived of users in at least five different ways, viewing them alternatively as citizens, clients, customers, guests, or partners. This book argues...... that these user metaphors crucially inform librarians' interactions with the public, and, by extension, determine the quality and content of the services received. The ultimate aim of the book is to provide library professionals with insights and tools for avoiding common pitfalls associated with false...... or professionally inadequate conceptions of library users....

  2. User acquaintance with mobile interfaces.

    Science.gov (United States)

    Ehrler, Frederic; Walesa, Magali; Sarrey, Evelyne; Wipfli, Rolf; Lovis, Christian

    2013-01-01

    Handheld technology finds slowly its place in the healthcare world. Some clinicians already use intensively dedicated mobile applications to consult clinical references. However, handheld technology hasn't still broadly embraced to the core of the healthcare business, the hospitals. The weak penetration of handheld technology in the hospitals can be partly explained by the caution of stakeholders that must be convinced about the efficiency of these tools before going forward. In a domain where temporal constraints are increasingly strong, caregivers cannot loose time on playing with gadgets. All users are not comfortable with tactile manipulations and the lack of dedicated peripheral complicates entering data for novices. Stakeholders must be convinced that caregivers will be able to master handheld devices. In this paper, we make the assumption that the proper design of an interface may influence users' performances to record information. We are also interested to find out whether users increase their efficiency when using handheld tools repeatedly. To answer these questions, we have set up a field study to compare users' performances on three different user interfaces while recording vital signs. Some user interfaces were familiar to users, and others were totally innovative. Results showed that users' familiarity with smartphone influences their performances and that users improve their performances by repeating a task.

  3. DIRAC: Secure web user interface

    Energy Technology Data Exchange (ETDEWEB)

    Casajus Ramo, A [University of Barcelona, Diagonal 647, ES-08028 Barcelona (Spain); Sapunov, M, E-mail: sapunov@in2p3.f [Centre de Physique des Particules de Marseille, 163 Av de Luminy Case 902 13288 Marseille (France)

    2010-04-01

    Traditionally the interaction between users and the Grid is done with command line tools. However, these tools are difficult to use by non-expert users providing minimal help and generating outputs not always easy to understand especially in case of errors. Graphical User Interfaces are typically limited to providing access to the monitoring or accounting information and concentrate on some particular aspects failing to cover the full spectrum of grid control tasks. To make the Grid more user friendly more complete graphical interfaces are needed. Within the DIRAC project we have attempted to construct a Web based User Interface that provides means not only for monitoring the system behavior but also allows to steer the main user activities on the grid. Using DIRAC's web interface a user can easily track jobs and data. It provides access to job information and allows performing actions on jobs such as killing or deleting. Data managers can define and monitor file transfer activity as well as check requests set by jobs. Production managers can define and follow large data productions and react if necessary by stopping or starting them. The Web Portal is build following all the grid security standards and using modern Web 2.0 technologies which allow to achieve the user experience similar to the desktop applications. Details of the DIRAC Web Portal architecture and User Interface will be presented and discussed.

  4. TMAP7 User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Glen R. Longhurst

    2006-09-01

    The TMAP Code was written at the Idaho National Engineering and Environmental Laboratory by Brad Merrill and James Jones in the late 1980s as a tool for safety analysis of systems involving tritium. Since then it has been upgraded to TMAP4 and has been used in numerous applications including experiments supporting fusion safety, predictions for advanced systems such as the International Thermonuclear Experimental Reactor (ITER), and estimates involving tritium production technologies. Its further upgrade to TMAP2000 and now to TMAP7 was accomplished in response to several needs. TMAP and TMAP4 had the capacity to deal with only a single trap for diffusing gaseous species in solid structures. TMAP7 includes up to three separate traps and up to 10 diffusing species. The original code had difficulty dealing with heteronuclear molecule formation such as HD and DT. That has been removed. Under pre-specified boundary enclosure conditions and solution-law dependent diffusion boundary conditions, such as Sieverts' law, TMAP7 automatically generates heteronuclear molecular partial pressures when solubilities and partial pressures of the homonuclear molecular species are provided for law-dependent diffusion boundary conditions. A further sophistication is the addition of non-diffusing surface species. Atoms such as oxygen or nitrogen or formation and decay or combination of hydroxyl radicals on metal surfaces are sometimes important in reactions with diffusing hydrogen isotopes but do not themselves diffuse appreciably in the material. TMAP7 will accommodate up to 30 such surface species, allowing the user to specify relationships between those surface concentrations and partial pressures of gaseous species above the surfaces or to form them dynamically by combining diffusion species or other surface species. Additionally, TMAP7 allows the user to include a surface binding energy and an adsorption barrier energy. The code includes asymmetrical diffusion between the surface

  5. TMAP7 User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Glen R. Longhurst

    2008-12-01

    The TMAP Code was written at the Idaho National Engineering and Environmental Laboratory by Brad Merrill and James Jones in the late 1980s as a tool for safety analysis of systems involving tritium. Since then it was upgraded to TMAP4 and has been used in numerous applications including experiments supporting fusion safety, predictions for advanced systems such as the International Thermonuclear Experimental Reactor (ITER), and estimates involving tritium production technologies. Its further upgrade to TMAP2000 and now to TMAP7 was accomplished in response to several needs. TMAP and TMAP4 had the capacity to deal with only a single trap for diffusing gaseous species in solid structures. TMAP7 includes up to three separate traps and up to 10 diffusing species. The original code had difficulty dealing with heteronuclear molecule formation such as HD and DT under solution-law dependent diffusion boundary conditions. That difficulty has been overcome. TMAP7 automatically generates heteronuclear molecular partial pressures when solubilities and partial pressures of the homonuclear molecular species are provided for law-dependent diffusion boundary conditions. A further sophistication is the addition of non-diffusing surface species. Atoms such as oxygen or nitrogen or formation and decay or combination of hydroxyl radicals on metal surfaces are sometimes important in reactions with diffusing hydrogen isotopes but do not themselves diffuse appreciably in the material. TMAP7 will accommodate up to 30 such surface species, allowing the user to specify relationships between those surface concentrations and partial pressures of gaseous species above the surfaces or to form them dynamically by combining diffusion species or other surface species. Additionally, TMAP7 allows the user to include a surface binding energy and an adsorption barrier energy. The code includes asymmetrical diffusion between the surface sites and regular diffusion sites in the bulk. All of the

  6. Non-professional user`s understanding of Geographic Information

    DEFF Research Database (Denmark)

    Arleth, Mette

    2003-01-01

    -based online services and comprehend the information contents? Using the Gi-based online services qualitatively in the participatory process obviously requires knowledge of the non-professional user`s understanding and use of GI. This paper discusses the needs for research into this field as well as relevant...... research methods....

  7. Biodigester User Survey Report

    Energy Technology Data Exchange (ETDEWEB)

    Chandararot, K.; Dannet, L.

    2007-06-15

    In May 2005, SNV and the Ministry of Agriculture, Forestry and Fisheries (MAFF) agreed to a joint development of a National Biodigester Programme (NBP) in Cambodia as a way to create an indigenous, sustainable energy source in the country and to utilize the potential of biogas in the country. The overall objective of the first phase of the National Biodigester Programme is 'The dissemination of domestic biodigesters as an indigenous, sustainable energy source through the development of a commercial, market oriented, biodigester sector in selected provinces of Cambodia'. The program aims to support the construction of 17,500 biodigesters in at least 6 provinces over the period of 2006 to 2009. To gain insights and feedbacks on the impacts of their activities to date, NBP commissioned the Cambodia Institute of Development Study (CIDS) to carry out a Biodigester User Survey in January 2007. The purpose of the survey is to evaluate the effects of domestic biodigester installations, as supported by the program, on 100 households in 3 provinces in Cambodia- Kampong Cham, Kandal and Svay Rieng.

  8. Echo™ User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Harvey, Dustin Yewell [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-06-06

    Echo™ is a MATLAB-based software package designed for robust and scalable analysis of complex data workflows. An alternative to tedious, error-prone conventional processes, Echo is based on three transformative principles for data analysis: self-describing data, name-based indexing, and dynamic resource allocation. The software takes an object-oriented approach to data analysis, intimately connecting measurement data with associated metadata. Echo operations in an analysis workflow automatically track and merge metadata and computation parameters to provide a complete history of the process used to generate final results, while automated figure and report generation tools eliminate the potential to mislabel those results. History reporting and visualization methods provide straightforward auditability of analysis processes. Furthermore, name-based indexing on metadata greatly improves code readability for analyst collaboration and reduces opportunities for errors to occur. Echo efficiently manages large data sets using a framework that seamlessly allocates resources such that only the necessary computations to produce a given result are executed. Echo provides a versatile and extensible framework, allowing advanced users to add their own tools and data classes tailored to their own specific needs. Applying these transformative principles and powerful features, Echo greatly improves analyst efficiency and quality of results in many application areas.

  9. LCS Users Manual

    Energy Technology Data Exchange (ETDEWEB)

    A.J. Redd; D.W. Ignat

    1998-02-01

    The Lower Hybrid Simulation Code (LSC) is a computational model of lower hybrid current drive in the presence of an electric field. Details of geometry, plasma profiles, and circuit equations are treated. Two-dimensional velocity space effects are approximated in a one-dimensional Fokker-Planck treatment. The LSC was originally written to be a module for lower hybrid current drive called by the Tokamak Simulation Code (TSC), which is a numerical model of an axisymmetric tokamak plasma and the associated control systems. The TSC simulates the time evolution of a free boundary plasma by solving the MHD equations on a rectangular computational grid. The MHD equations are coupled to the external circuits (representing poloidal field coils) through the boundary conditions. The code includes provisions for modeling the control system, external heating, and fusion heating. The LSC module can also be called by the TRANSP code. TRANSP represents the plasma with an axisymmetric, fixed-boundary model and focuses on calculation of plasma transport to determine transport coefficients from data on power inputs and parameters reached. This manual covers the basic material needed to use the LSC. If run in conjunction with TSC, the "TSC Users Manual" should be consulted. If run in conjunction with TRANSP, on-line documentation will be helpful. A theoretical background of the governing equations and numerical methods is given. Information on obtaining, compiling, and running the code is also provided.

  10. Users page feedback

    CERN Multimedia

    2010-01-01

    In October last year the Communication Group proposed an interim redesign of the users’ web pages in order to improve the visibility of key news items, events and announcements to the CERN community. The proposed update to the users' page (right), and the current version (left, behind) This proposed redesign was seen as a small step on the way to much wider reforms of the CERN web landscape proposed in the group’s web communication plan.   The results are available here. Some of the key points: - the balance between news / events / announcements and access to links on the users’ pages was not right - many people asked to see a reversal of the order so that links appeared first, news/events/announcements last; - many people felt that we should keep the primary function of the users’ pages as an index to other CERN websites; - many people found the sections of the front page to be poorly delineated; - people do not like scrolling; - there were performance...

  11. Multisensor user authentication

    Science.gov (United States)

    Colombi, John M.; Krepp, D.; Rogers, Steven K.; Ruck, Dennis W.; Oxley, Mark E.

    1993-09-01

    User recognition is examined using neural and conventional techniques for processing speech and face images. This article for the first time attempts to overcome this significant problem of distortions inherently captured over multiple sessions (days). Speaker recognition uses both Linear Predictive Coding (LPC) cepstral and auditory neural model representations with speaker dependent codebook designs. For facial imagery, recognition is developed on a neural network that consists of a single hidden layer multilayer perceptron backpropagation network using either the raw data as inputs or principal components of the raw data computed using the Karhunen-Loeve Transform as inputs. The data consists of 10 subjects; each subject recorded utterances and had images collected for 10 days. The utterances collected were 400 rich phonetic sentences (4 sec), 200 subject name recordings (3 sec), and 100 imposter name recordings (3 sec). Face data consists of over 2000, 32 X 32 pixel, 8 bit gray scale images of the 10 subjects. Each subsystem attains over 90% verification accuracy individually using test data gathered on days following the training data.

  12. Policies to Promote User Innovation

    DEFF Research Database (Denmark)

    Svensson, Peter; Hartmann, Rasmus Koss

    As it becomes apparent that users are an important source in innovation in society and in individual organizations, scholars are realizing that user-directed innovation policy may contribute to improving social welfare. How such policy might be designed, however, is uncertain, as are the costs an...

  13. User-Centered Design Gymkhana

    OpenAIRE

    Garreta Domingo, Muriel; Almirall Hill, Magí; Mor Pera, Enric

    2007-01-01

    The User-centered design (UCD) Gymkhana is a tool for human-computer interaction practitioners to demonstrate through a game the key user-centered design methods and how they interrelate in the design process.The target audiences are other organizational departments unfamiliar with UCD but whose work is related to the definition, cretaion, and update of a product service.

  14. Scientific customer needs - NASA user

    Science.gov (United States)

    Black, David C.

    1987-01-01

    Some requirements for scientific users of the Space Station are considered. The use of testbeds to evaluate design concepts for information systems, and for interfacing between designers and builders of systems is examined. The need for an information system that provides an effective interaction between ground-based users and their space-based equipment is discussed.

  15. Mental models and user training

    Directory of Open Access Journals (Sweden)

    Saša Zupanič

    1997-01-01

    Full Text Available One of the functions of the reference service is user training which means teaching users how to use the library and it's information sorces (nowadays mainly computerized systems. While the scientific understanding of teaching/learning process is shifting, changes also affect the methods of user training in libraries.Human-computer interaction (HCI is an interdisciplinary and a very active research area which studies how humans use computers - their mental and behavioral characteristics. The application of psychological theories to HCI are especially great on three areas: psychological (mental, conceptual models, individual differences, and error behavior.The mental models theory is powerful tool for understanding the ways in which users interact with an information system. Claims, based on this theory can affect the methods (conceptualization of user training and the overall design of information systems.

  16. The Users Office turns 20

    CERN Multimedia

    2009-01-01

    20 years ago, in the summer of 1989, an office was created to assist the thousands of users who come to CERN each year, working over the broad range of projects and collaborations. Chris Onions (right), head of the Users’ Office, with Bryan Pattison (left), the Office’s founder.Before the inception of the Users Office, it was common for users to spend at least an entire day moving from office to office in search of necessary documentation and information in order to make their stay official. "Though the Office has undergone various changes throughout its lifetime, it has persisted in being a welcoming bridge to facilitate the installation of visitors coming from all over the world", says Chris Onions, head of the Users Office. This September, the Office will celebrate its 20-year anniversary with a drink offered to representatives of the User community, the CERN management and staff members from the services with whom the Office is involved. &...

  17. User Involvement And Entrepreneurial Action

    Directory of Open Access Journals (Sweden)

    Eva Heiskanen

    2007-01-01

    Full Text Available Involving users in the innovation process is a subject of much research, experimentation, and debate. Less attention has been given to the limits to user involvement that ensue from specific organizational characteristics. This article explores barriers to the utilization of users’ input in two small companies developing interactive digital applications. We contrast our findings to earlier research involving large companies to identify features of entrepreneurial sensemaking and action that influence the utilization of users’ input. We find that the small companies follow a distinct action rationality, leading to rapid implementation of some user inputs, and defensiveness toward others. Both sets of data also reveal common features that are often overlooked in the literature. We reconceptualize user involvement as a form of interaction between users and innovating companies that is facilitated and constrained by micro-sociological processes, on the one hand, and the nature of the competitive environment, on the other.

  18. Identifying online user reputation of user-object bipartite networks

    Science.gov (United States)

    Liu, Xiao-Lu; Liu, Jian-Guo; Yang, Kai; Guo, Qiang; Han, Jing-Ti

    2017-02-01

    Identifying online user reputation based on the rating information of the user-object bipartite networks is important for understanding online user collective behaviors. Based on the Bayesian analysis, we present a parameter-free algorithm for ranking online user reputation, where the user reputation is calculated based on the probability that their ratings are consistent with the main part of all user opinions. The experimental results show that the AUC values of the presented algorithm could reach 0.8929 and 0.8483 for the MovieLens and Netflix data sets, respectively, which is better than the results generated by the CR and IARR methods. Furthermore, the experimental results for different user groups indicate that the presented algorithm outperforms the iterative ranking methods in both ranking accuracy and computation complexity. Moreover, the results for the synthetic networks show that the computation complexity of the presented algorithm is a linear function of the network size, which suggests that the presented algorithm is very effective and efficient for the large scale dynamic online systems.

  19. Audit result and its users

    Directory of Open Access Journals (Sweden)

    Shalimova Nataliya S.

    2014-01-01

    Full Text Available The article identifies essence of the “audit result” and “users of audit result” notions and characteristics of the key audit results user. It shows that in order to give a wide characteristic of users it is expedient to unite all objects, which could be used (audit report, fact of refusal to conduct audit and information that is submitted to managers in the process of audit with the term “audit result” and classify it depending on the terms of submission by final and intermediate result. The article offers to define audit results user as a person, persons or category of persons for whom the auditor prepares the audit report and, in cases, envisaged by international standards of the audit and domestic legislative and regulatory acts, provides other additional information concerning audit issues. In order to identify the key audit results user the article distributes all audit tasks into two groups depending on possibilities of identification of users. The article proves that the key user should be identified especially in cases of a mandatory audit and this process should go in interconnection with the mechanism of allocation of a key user of financial reports. It offers to consider external users with direct financial interests, who cannot request economic subjects directly to provide information and who should rely on general financial reports and audit report when receiving significant portion of information they need, as the key user. The article makes proposals on specification of the categorical mechanism in the sphere of audit, which are the basis for audit quality assessment, identification of possibilities and conditions of appearance of the necessary and sufficient trust to the auditor opinion.

  20. Peak experiences of psilocybin users and non-users.

    Science.gov (United States)

    Cummins, Christina; Lyke, Jennifer

    2013-01-01

    Maslow (1970) defined peak experiences as the most wonderful experiences of a person's life, which may include a sense of awe, well-being, or transcendence. Furthermore, recent research has suggested that psilocybin can produce experiences subjectively rated as uniquely meaningful and significant (Griffiths et al. 2006). It is therefore possible that psilocybin may facilitate or change the nature of peak experiences in users compared to non-users. This study was designed to compare the peak experiences of psilocybin users and non-users, to evaluate the frequency of peak experiences while under the influence of psilocybin, and to assess the perceived degree of alteration of consciousness during these experiences. Participants were recruited through convenience and snowball sampling from undergraduate classes and at a musical event. Participants were divided into three groups, those who reported a peak experience while under the influence of psilocybin (psilocybin peak experience: PPE), participants who had used psilocybin but reported their peak experiences did not occur while they were under the influence of psilocybin (non-psilocybin peak experience: NPPE), and participants who had never used psilocybin (non-user: NU). A total of 101 participants were asked to think about their peak experiences and complete a measure evaluating the degree of alteration of consciousness during that experience. Results indicated that 47% of psilocybin users reported their peak experience occurred while using psilocybin. In addition, there were significant differences among the three groups on all dimensions of alteration of consciousness. Future research is necessary to identify factors that influence the peak experiences of psilocybin users in naturalistic settings and contribute to the different characteristics of peak experiences of psilocybin users and non-users.

  1. Designing for User Engagment Aesthetic and Attractive User Interfaces

    CERN Document Server

    Sutcliffe, Alistair

    2009-01-01

    This book explores the design process for user experience and engagement, which expands the traditional concept of usability and utility in design to include aesthetics, fun and excitement. User experience has evolved as a new area of Human Computer Interaction research, motivated by non-work oriented applications such as games, education and emerging interactive Web 2.0. The chapter starts by examining the phenomena of user engagement and experience and setting them in the perspective of cognitive psychology, in particular motivation, emotion and mood. The perspective of aesthetics is expande

  2. Solar walls in tsbi3 user`s guide

    Energy Technology Data Exchange (ETDEWEB)

    Wittchen, K.B.

    1997-12-01

    Tsbi3 is a user-friendly and flexible computer program, which provides support to the design team in the analysis of the indoor climate and the energy performance of buildings. The solar wall module gives tsbi3 the capability of simulating solar walls and their interaction with the building. This version, C, of tsbi3 is capable of simulating five types of solar walls say: mass-walls, Trombe-walls, double Trombe-walls, internally ventilated walls and solar walls for preheating ventilation air. The user`s guide gives a description of the capabilities and how to simulate solar walls in tsbi3. (au)

  3. Challenges and Opportunities for Enhancing User Privacy and User Empowerment

    DEFF Research Database (Denmark)

    Dhotre, Prashant Shantaram; Olesen, Henning; Khajuria, Samant

    2015-01-01

    Big data techniques allow service providers to collect data on a massive scale. User data has turned into an important asset to service providers. The current business model of service providers gathers too much information for knowledge extraction. In the era of Google and Facebook, privacy...... information and current business models as privacy issues. The important identified issues include that users are missing privacy awareness tools and unavailability to visualize of personal information flow. We also present some recent advances in these areas to address concern privacy issues like User...

  4. Comparison and physical interpretation of MCNP and TART neutron and {gamma} Monte Carlo shielding calculations for a heavy-ion ICF system

    Energy Technology Data Exchange (ETDEWEB)

    Mainardi, E. E-mail: enrico@nuc.berkeley.edu; Premuda, F.; Lee, E

    2004-01-01

    Transport Code, Lawrence Livermore National Laboratory, UCRL-ID-126455, Rev. 1, November, 1997] and MCNP4B [MCNP - A General Monte Carlo N-Particle Transport Code, Version 4B, La-12625-m, March 1997, Los Alamos National Laboratory] for two different configurations of the system is discussed, separating the n and {gamma} contributions, in the light of the physical interpretation of the results in terms of first flight and of scattered neutron fluxes, of primary {gamma} and of secondary {gamma} generated by inelastically scattered or radiatively captured neutrons. The final conclusions indicate some guidelines and suggest possible improvements for the future neutronic shielding design for a HIF facility.

  5. Design of Light Multi-layered Shields for Use in Diagnostic Radiology and Nuclear Medicine via MCNP5 Monte Carlo Code

    Directory of Open Access Journals (Sweden)

    Mehdi Zehtabian

    2015-09-01

    Full Text Available Introduction Lead-based shields are the most widely used attenuators in X-ray and gamma ray fields. The heavy weight, toxicity and corrosion of lead have led researchers towards the development of non-lead shields. Materials and Methods The purpose of this study was to design multi-layered shields for protection against X-rays and gamma rays in diagnostic radiology and nuclear medicine. In this study, cubic slabs composed of several materials with high atomic numbers, i.e., lead, barium, bismuth, gadolinium, tin and tungsten, were simulated, using MCNP5 Monte Carlo code. Cubic slabs (30×30×0.05 cm3 were simulated at a 50 cm distance from the point photon source. The X-ray spectra of 80 kVp and 120 kVp were obtained, using IPEM Report 78. The photon flux following the use of each shield was obtained inside cubic tally cells (1×1×0.5 cm3 at a 5 cm distance from the shields. The photon attenuation properties of multi-layered shields (i.e., two, three, four and five layers, composed of non-lead radiation materials, were also obtained via Monte Carlo simulations. Results Among different shield designs proposed in this study, the three-layered shield, composed of tungsten, bismuth and gadolinium, showed the most significant attenuation properties in radiology, with acceptable shielding at 140 keV energy in nuclear medicine. Conclusion According to the results, materials with k-edges equal to energies common to diagnostic radiology can be proper substitutes for lead shields.

  6. Determination of the dead layer and full-energy peak efficiency of an HPGe detector using the MCNP code and experimental results

    Directory of Open Access Journals (Sweden)

    M Moeinifar

    2017-02-01

    Full Text Available One important factor in using an High Purity Germanium (HPGe detector is its efficiency that highly depends on the geometry and absorption factors, so that when the configuration of source-detector geometry is changed, the detector efficiency must be re-measured. The best way of determining the efficiency of a detector is measuring the efficiency of standard sources. But considering the fact that standard sources are hardly available and it is time consuming to find them, determinig the efficiency by simulation which gives enough efficiency in less time, is important. In this study, the dead layer thickness and the full-energy peak efficiency of an HPGe detector was obtained by Monte Carlo simulation, using MCNPX code. For this, we first measured gamma–ray spectra for different sources placed at various distances from the detector and stored the measured spectra obtained. Then the obtained spectra were simulated under similar conditions in vitro.At first, the whole volume of germanium was regarded as active, and the obtaind spectra from calculation were compared with the corresponding experimental spectra. Comparison of the calculated spectra with the measured spectra showed considerable differences. By making small variations in the dead layer thickness of the detector (about a few hundredths of a millimeter in the simulation program, we tried to remove these differences and in this way a dead layer of 0.57 mm was obtained for the detector. By incorporating this value for the dead layer in the simulating program, the full-energy peak efficiency of the detector was then obtained both by experiment and by simulation, for various sources at various distances from the detector, and both methods showed good agreements. Then, using MCNP code and considering the exact measurement system, one can conclude that the efficiency of an HPGe detector for various source-detector geometries can be calculated with rather good accuracy by simulation method

  7. Biological Shielding Design Effectiveness of the Brachytherapy Unit at the Korle Bu Teaching Hospital in Ghana Using Mcnp5 Monte Carlo Code

    Directory of Open Access Journals (Sweden)

    C.C. Arwui

    2011-05-01

    Full Text Available Design objectives for brachytherapy treatment facilities require sufficient shielding to reduce primary and scatter radiation to design limit in order to limit exposure to patients, staff and the general public. The primary aim of this study is to verify whether shielding of the brachytherapy unit at the Korle Bu teaching Hospital in Ghana provides adequate protection in order to assess any radiological health and safety impact and also test the suitability of other available sources. The study evaluates the effectiveness of the biological shielding design of a Cs-137 brachytherapy unit at the Korle-Bu Teaching Hospital in Ghana using MCNP5. The facility was modeled based on the design specifications for LDR Cs-137, MDR Cs-137, HDR Co-60 and HDR Ir-192 treatment modalities. The estimated dose rate ranged from (0.01-0.15 μSv/h and (0.37-3.05 μSv/h for the existing initial and decayed activities of LDR Cs-137 for the public and controlled areas respectively, (0.03-0.57 μSv/h and (1.53-8.06 μSv/h for MDR Cs-137, (7.47-59.46 μSv/h and (144.87-178.74 μSv/h for HDR Co- 60, (0.13-6.95 μSv/h and (19.47-242.98 μSv/h for HDR Ir-192 for the public and controlled areas respectively. The results were verified by dose rates measurement for the current LDR setup at the Brachytherapy unit and agreed quiet well. It was also compared with the reference values of 0.5 μSv/h for public areas and 7.5 μSv/h for controlled areas respectively. It can be concluded that the shielding is adequate for the existing source.

  8. User Experimentation with Terminological Ontologies

    DEFF Research Database (Denmark)

    Pram Nielsen, Louise

    This paper outlines work-in-progress research suggesting that domain-specific knowledge in terminological resources can be transferred efficiently to end-users across different levels of expertise and by means of different information modes including articles (written mode) and concept diagrams...... (graph mode). An experimental approach is applied in an eye-tracking laboratory, where a natural user situation is replicated for Danish professional potential end-users of a ter-minology and knowledge bank in a chosen pilot domain (taxation)....

  9. Practical speech user interface design

    CERN Document Server

    Lewis, James R

    2010-01-01

    Although speech is the most natural form of communication between humans, most people find using speech to communicate with machines anything but natural. Drawing from psychology, human-computer interaction, linguistics, and communication theory, Practical Speech User Interface Design provides a comprehensive yet concise survey of practical speech user interface (SUI) design. It offers practice-based and research-based guidance on how to design effective, efficient, and pleasant speech applications that people can really use. Focusing on the design of speech user interfaces for IVR application

  10. User-Centered Agile Methods

    CERN Document Server

    Beyer, Hugh

    2010-01-01

    With the introduction and popularization of Agile methods of software development, existing relationships and working agreements between user experience groups and developers are being disrupted. Agile methods introduce new concepts: the Product Owner, the Customer (but not the user), short iterations, User Stories. Where do UX professionals fit in this new world? Agile methods also bring a new mindset -- no big design, no specifications, minimal planning -- which conflict with the needs of UX design. This lecture discusses the key elements of Agile for the UX community and describes strategie

  11. Defining and Measuring User Experience

    DEFF Research Database (Denmark)

    Stage, Jan

    2006-01-01

    User experience is being used to denote what a user goes through while using a computerized system. The concept has gained momentum as a means to distinguish new types of applications such as games and entertainment software from more traditional work-related applications. This paper focuses...... on the intrinsic relation between definition and measurement. In the area of usability, this relation has been developed over several years. It is described how usability is defined and measured in contemporary approaches. Based on that, it is discussed to what extent we can employ experience from the conceptual...... definition of usability to develop the notion of user experience....

  12. User interface design and evaluation

    CERN Document Server

    Stone, Debbie; Woodroffe, Mark

    2005-01-01

    Whether you are a professional new to the user-centered design field, or an experienced designer who needs to learn the fundamentals of user interface design and evaluation, this book can lead the way.What will you get from this book? Based on a course from the Open University, UK which has been taught to over a thousand professionals and students, this book presents an overview of the field. It illustrates the benefits of a user-centered approach to the design of software, computer systems, and web sites, and provides a clear and practical discussion of requirements gathering; develop

  13. Search-User Interface Design

    CERN Document Server

    Wilson, Max

    2011-01-01

    Search User Interfaces (SUIs) represent the gateway between people who have a task to complete, and the repositories of information and data stored around the world. Not surprisingly, therefore, there are many communities who have a vested interest in the way SUIs are designed. There are people who study how humans search for information, and people who study how humans use computers. There are people who study good user interface design, and people who design aesthetically pleasing user interfaces. There are also people who curate and manage valuable information resources, and people who desi

  14. CAPTCHA: Impact on User Experience of Users with Learning Disabilities

    Directory of Open Access Journals (Sweden)

    Ruti Gafni

    2016-12-01

    Full Text Available CAPTCHA is one of the most common solutions to check if the user trying to enter a Website is a real person or an automated piece of software. This challenge-response test, implemented in many Internet Websites, emphasizes the gaps between accessibility and security on the Internet, as it poses an obstacle for the learning-impaired in the reading and comprehension of what is presented in the test. Various types of CAPTCHA tests have been developed in order to address accessibility and security issues. The objective of this study is to investigate how the differences between various CAPTCHA tests affect user experience among populations with and without learning disabilities. A questionnaire accompanied by experiencing five different tests was administered to 212 users, 60 of them with learning disabilities. Response rates for each test and levels of success were collected automatically. Findings suggest that users with learning disabilities have more difficulties in solving the tests, especially those with distorted texts, have more negative attitudes towards the CAPTCHA tests, but the response time has no statistical difference from users without learning disabilities. These insights can help to develop and implement solutions suitable for many users and especially for population with learning disabilities.

  15. Family health program user: knowledge and satisfaction about user embracement

    Directory of Open Access Journals (Sweden)

    Saulo Lacerda Borges de Sá

    2012-06-01

    Full Text Available Objective: To assess the knowledge and satisfaction of users of a Basic Health Unit about the strategy of embracement. Methods: Descriptive study with qualitative approach, carried out in a Basic Health Unit, Fortaleza, Brazil, where practical activities of the Education Program of Work for Health of the University of Fortaleza were performed. Fifty eight service users were involved, following inclusion criteria: being present during the data collection, age over 18, regardless of sex, and voluntary participation. Data collection occurred in December 2009, through semi-structured interview. The data associated with the identification of users were processed in Microsoft Office Excel 2007, being organizedstatistically in table. Data related to qualitative aspects were analyzed according to the technique of content analysis. Results: 56 (97% were women, with ages ranging between 21 and 40 years, 34 (59% were married and 53 (91% are literate. On family income, 55 (95%received less than two minimum salaries per month. In order to facilitate understanding the speech of users, these were evaluated from the perspective of two categories: knowledge about embracement and satisfaction with embracement. Conclusion: Users have a limited view of the significance and magnitude of the embracement to provide the care. Although satisfied with the service, respondents report as negative aspects: the shortage of professionals, the professional relationship with user impaired due to constant delays of the professional, and the dehumanization of care.

  16. Unobtrusive user modeling for adaptive hypermedia

    NARCIS (Netherlands)

    Holz, H.J.; Hofmann, K.; Reed, C.; Uchyigit, G.; Ma, M.Y.

    2008-01-01

    We propose a technique for user modeling in Adaptive Hypermedia (AH) that is unobtrusive at both the level of observable behavior and that of cognition. Unobtrusive user modeling is complementary to transparent user modeling. Unobtrusive user modeling induces user models appropriate for Educational

  17. The Exploitation of Drug Users.

    Science.gov (United States)

    Stallings, Shirley; Montagne, Michael

    2015-01-01

    Drug users have been exploited in research studies and clinical practice. We explore ways in which exploitation has occurred and strategies to help patients, research subjects and communities to prevent or avoid exploitation.

  18. Gender Stereotypes among Road Users

    Directory of Open Access Journals (Sweden)

    Kabalevskaya, Alexandra I.

    2013-09-01

    Full Text Available This article analyzes the mechanism of stereotyping as exemplified by gender stereotypes of road users. Gender stereotypes are not only viewed as an a priori image of a percept, but also examined ‘in action’ — at the very moment of their actualization with road users. In the paper we have identified the content of road users’ gender stereotypes; analyzed the behaviour of male and female drivers, pinpointing a number of gender-specific behavioural features; demonstrated that male and female driving differ from each other in terms of speed, intensity and roughness; and identified the conditions and mechanisms underlying the actualization of gender stereotypes. Based on video and audio materials, we have found that drivers’ gender-specific behavioural features are perceivable to road users: such features trigger the actualization of gender stereotypes as attributive schemes, which determine the interaction between road users, while also laying the foundation for gender stereotypes.

  19. User perspectives on query difficulty

    DEFF Research Database (Denmark)

    Lioma, Christina; Larsen, Birger; Schütze, Hinrich

    2011-01-01

    , or to statistical and linguistic features of the queries that may render them difficult. This work addresses query difficulty from a different angle, namely the users’ own perspectives on query difficulty. Two research questions are asked: (1) Are users aware that the query they submit to an IR system may......The difficulty of a user query can affect the performance of Information Retrieval (IR) systems. What makes a query difficult and how one may predict this is an active research area, focusing mainly on factors relating to the retrieval algorithm, to the properties of the retrieval data...... for synthesising the user-assessed causes of query difficulty through opinion fusion into an overall assessment of query difficulty. The resulting assessments of query difficulty are found to agree notably more to the TREC categories than the direct user assessments....

  20. OpenEIS. Users Guide

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Woohyun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lutes, Robert G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Katipamula, Srinivas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Haack, Jereme N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Carpenter, Brandon J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Akyol, Bora A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Monson, Kyle E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Allwardt, Craig H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Kang, Timothy [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sharma, Poorva [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-02-28

    This document is a users guide for OpenEIS, a software code designed to provide standard methods for authoring, sharing, testing, using and improving algorithms for operational building energy efficiency.