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Sample records for code users manual

  1. SEVERO code - user's manual

    International Nuclear Information System (INIS)

    Sacramento, A.M. do.

    1989-01-01

    This user's manual contains all the necessary information concerning the use of SEVERO code. This computer code is related to the statistics of extremes = extreme winds, extreme precipitation and flooding hazard risk analysis. (A.C.A.S.)

  2. XSOR codes users manual

    International Nuclear Information System (INIS)

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

    1993-11-01

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

  3. User manual of UNF code

    International Nuclear Information System (INIS)

    Zhang Jingshang

    2001-01-01

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

  4. ARES: automated response function code. Users manual

    International Nuclear Information System (INIS)

    Maung, T.; Reynolds, G.M.

    1981-06-01

    This ARES user's manual provides detailed instructions for a general understanding of the Automated Response Function Code and gives step by step instructions for using the complete code package on a HP-1000 system. This code is designed to calculate response functions of NaI gamma-ray detectors, with cylindrical or rectangular geometries

  5. User's manual of Tokamak Simulation Code

    International Nuclear Information System (INIS)

    Nakamura, Yukiharu; Nishino, Tooru; Tsunematsu, Toshihide; Sugihara, Masayoshi.

    1992-12-01

    User's manual for use of Tokamak Simulation Code (TSC), which simulates the time-evolutional process of deformable motion of axisymmetric toroidal plasma, is summarized. For the use at JAERI computer system, the TSC is linked with the data management system GAEA. This manual is forcused on the procedure for the input and output by using the GAEA system. Model equations to give axisymmetric motion, outline of code system, optimal method to get the well converged solution are also described. (author)

  6. User's manual for the TMAD code

    International Nuclear Information System (INIS)

    Finfrock, S.H.

    1995-01-01

    This document serves as the User's Manual for the TMAD code system, which includes the TMAD code and the LIBMAKR code. The TMAD code was commissioned to make it easier to interpret moisture probe measurements in the Hanford Site waste tanks. In principle, the code is an interpolation routine that acts over a library of benchmark data based on two independent variables, typically anomaly size and moisture content. Two additional variables, anomaly type and detector type, also can be considered independent variables, but no interpolation is done over them. The dependent variable is detector response. The intent is to provide the code with measured detector responses from two or more detectors. The code then will interrogate (and interpolate upon) the benchmark data library and find the anomaly-type/anomaly-size/moisture-content combination that provides the closest match to the measured data

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-01-01

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

  8. User's manual for a measurement simulation code

    International Nuclear Information System (INIS)

    Kern, E.A.

    1982-07-01

    The MEASIM code has been developed primarily for modeling process measurements in materials processing facilities associated with the nuclear fuel cycle. In addition, the code computes materials balances and the summation of materials balances along with associated variances. The code has been used primarily in performance assessment of materials' accounting systems. This report provides the necessary information for a potential user to employ the code in these applications. A number of examples that demonstrate most of the capabilities of the code are provided

  9. User's manual for a process model code

    International Nuclear Information System (INIS)

    Kern, E.A.; Martinez, D.P.

    1981-03-01

    The MODEL code has been developed for computer modeling of materials processing facilities associated with the nuclear fuel cycle. However, it can also be used in other modeling applications. This report provides sufficient information for a potential user to apply the code to specific process modeling problems. Several examples that demonstrate most of the capabilities of the code are provided

  10. SSC-K code user's manual

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Y M; Lee, Y B; Chang, W P; Hahn, D

    2000-07-01

    , constitutive laws and correlations required to execute the SSC-K are described. It is noted that the user's manual will be revised later with the further development of SSC-K.

  11. ARES: automated response function code. Users manual. [HPGAM and LSQVM

    Energy Technology Data Exchange (ETDEWEB)

    Maung, T.; Reynolds, G.M.

    1981-06-01

    This ARES user's manual provides detailed instructions for a general understanding of the Automated Response Function Code and gives step by step instructions for using the complete code package on a HP-1000 system. This code is designed to calculate response functions of NaI gamma-ray detectors, with cylindrical or rectangular geometries.

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  13. Citham-2 computer code-User manual

    International Nuclear Information System (INIS)

    Batista, J.L.

    1984-01-01

    The procedures and the input data for the Citham-2 computer code are described. It is a subroutine that modifies the nuclide concentration taking in account its burn and prepares cross sections library in 2,3 or 4 energy groups, to the used for Citation program. (E.G.) [pt

  14. Colloid transport code-nuclear user's manual

    International Nuclear Information System (INIS)

    Jain, R.

    1992-01-01

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

  15. User's manual for the NEFTRAN II computer code

    International Nuclear Information System (INIS)

    Olague, N.E.; Campbell, J.E.; Leigh, C.D.; Longsine, D.E.

    1991-02-01

    This document describes the NEFTRAN II (NEtwork Flow and TRANsport in Time-Dependent Velocity Fields) computer code and is intended to provide the reader with sufficient information to use the code. NEFTRAN II was developed as part of a performance assessment methodology for storage of high-level nuclear waste in unsaturated, welded tuff. NEFTRAN II is a successor to the NEFTRAN and NWFT/DVM computer codes and contains several new capabilities. These capabilities include: (1) the ability to input pore velocities directly to the transport model and bypass the network fluid flow model, (2) the ability to transport radionuclides in time-dependent velocity fields, (3) the ability to account for the effect of time-dependent saturation changes on the retardation factor, and (4) the ability to account for time-dependent flow rates through the source regime. In addition to these changes, the input to NEFTRAN II has been modified to be more convenient for the user. This document is divided into four main sections consisting of (1) a description of all the models contained in the code, (2) a description of the program and subprograms in the code, (3) a data input guide and (4) verification and sample problems. Although NEFTRAN II is the fourth generation code, this document is a complete description of the code and reference to past user's manuals should not be necessary. 19 refs., 33 figs., 25 tabs

  16. User's manual for seismic analysis code 'SONATINA-2V'

    International Nuclear Information System (INIS)

    Hanawa, Satoshi; Iyoku, Tatsuo

    2001-08-01

    The seismic analysis code, SONATINA-2V, has been developed to analyze the behavior of the HTTR core graphite components under seismic excitation. The SONATINA-2V code is a two-dimensional computer program capable of analyzing the vertical arrangement of the HTTR graphite components, such as fuel blocks, replaceable reflector blocks, permanent reflector blocks, as well as their restraint structures. In the analytical model, each block is treated as rigid body and is restrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions between upper and lower blocks. Moreover, the SONATINA-2V code is capable of analyzing the core vibration behavior under both simultaneous excitations of vertical and horizontal directions. The SONATINA-2V code is composed of the main program, pri-processor for making the input data to SONATINA-2V and post-processor for data processing and making the graphics from analytical results. Though the SONATINA-2V code was developed in order to work in the MSP computer system of Japan Atomic Energy Research Institute (JAERI), the computer system was abolished with the technical progress of computer. Therefore, improvement of this analysis code was carried out in order to operate the code under the UNIX machine, SR8000 computer system, of the JAERI. The users manual for seismic analysis code, SONATINA-2V, including pri- and post-processor is given in the present report. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

  18. Users' manual for fault tree analysis code: CUT-TD

    International Nuclear Information System (INIS)

    Watanabe, Norio; Kiyota, Mikio.

    1992-06-01

    The CUT-TD code has been developed to find minimal cut sets for a given fault tree and to calculate the occurrence probability of its top event. This code uses an improved top-down algorithm which can enhance the efficiency in deriving minimal cut sets. The features in processing techniques incorporated into CUT-TD are as follows: (1) Consecutive OR gates or consecutive AND gates can be coalesced into a single gate. As a result, this processing directly produces cut sets for the redefined single gate with each gate not being developed. (2) The independent subtrees are automatically identified and their respective cut sets are separately found to enhance the efficiency in processing. (3) The minimal cut sets can be obtained for the top event of a fault tree by combining their respective minimal cut sets for several gates of the fault tree. (4) The user can reduce the computing time for finding minimal cut sets and control the size and significance of cut sets by inputting a minimum probability cut off and/or a maximum order cut off. (5) The user can select events that need not to be further developed in the process of obtaining minimal cut sets. This option can reduce the number of minimal cut sets, save the computing time and assists the user in reviewing the result. (6) Computing time is monitored by the CUT-TD code so that it can prevent the running job from abnormally ending due to excessive CPU time and produce an intermediate result. The CUT-TD code has the ability to restart the calculation with use of the intermediate result. This report provides a users' manual for the CUT-TD code. (author)

  19. Colloid transport code-nuclear user`s manual

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-04-03

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

  20. SSC-K code users manual (rev.1)

    International Nuclear Information System (INIS)

    Kwon, Y. M.; Lee, Y. B.; Chang, W. P.; Hahn, D.

    2002-01-01

    The Supper System Code of KAERI (SSC-K) is a best-estimate system code for analyzing a variety of off-normal or accidents in the heat transport system of a pool type LMR design. It is being developed at Korea Atomic Energy Research Institution (KAERI) on the basis of SSC-L, originally developed at BNL to analyze loop-type LMR transients. SSC-K can handle both designs of loop and pool type LMRs. SSC-K contains detailed mechanistic models of transient thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, coolant, fuel elements, and structures to accident conditions. This report provides a revised User's Manual (rev.1) of the SSC-K computer code, focusing on phenomenological model descriptions for new thermal, hydraulic, neutronic, and mechanical modules. A comprehensive description of the models for pool-type reactor is given in Chapters 2 and 3; the steady-state plant characterization, prior to the initiation of transient is described in Chapter 2 and their transient counterparts are discussed in Chapter 3. Discussions on the intermediate heat exchanger (IHX) and the electromagnetic (EM) pump are described in Chapter 4 and 5, respectively. A model of passive safety decay heat removal system (PSDRS) is discussed in Chapter 6, and models for various reactivity feedback effects are discussed in Chapter 7. In Chapter 8, constitutive laws and correlations required to execute the SSC-K are described. New models developed for SSC-K rev.1 are two dimensional hot pool model in Chapter 9, and long term cooling model in Chapter 10. Finally, a brief description of MINET code adopted to simulate BOP is presented in Chapter 11. Based on test runs for typical LMFBR accident analyses, it was found that the present version of SSC-K would be used for the safety analysis of KALIMER. However, the further validation of SSC-K is required for real applications. It is noted that the user's manual of SSC-K will be revised later with the

  1. MACSIS User's Manual and Code Description

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woan; Lee, Byoung Oon; Kim, Kyung Gun; Kim, Young Jin [Korea Atomic Energy Research Institute, Taejeon (Korea); Lee, Dong Uk [Hanyang Univ., Seoul (Korea)

    2000-03-01

    MACSIS is a computer program for simulating the behavior of metal fuel elements under normal operating conditions of a Liquid Metal Cooled Reactor. It computes the one-dimensional temperature distribution and the thermo-mechanical characteristics of fuel rod under the steady state operation condition, including the swelling and rod deformation. The amount of fission gas released during the irradiation of the fuel is also computed. The thermal expansion and the gas pressure inside the fuel element are then used to compute the stresses and strains in the cladding. This document is mainly intended as a user's manual for the MACSIS code. A short description of the capabilities of the code and detailed input instructions are supplied for this purpose. MACSIS is constructed of a series of modules with a single set of dimensional units used throughout to provide flexibility in model usage and ease of upgrading as models developed from future tests are finalized. Radial steady state heat transfer can be computed for 21 axial segments. The code computes all major quantities which affect in-reactor performances of fuel rod, such as, fission gas generation and retention, fission gas release, swelling, and deformation, etc. 37 refs., 24 figs., 3 tabs. (Author)

  2. AFTC Code for Automatic Fault Tree Construction: Users Manual

    International Nuclear Information System (INIS)

    Gopika Vinod; Saraf, R.K.; Babar, A.K.

    1999-04-01

    Fault Trees perform a predominant role in reliability and safety analysis of system. Manual construction of fault tree is a very time consuming task and moreover, it won't give a formalized result, since it relies highly on analysts experience and heuristics. This necessitates a computerised fault tree construction, which is still attracting interest of reliability analysts. AFTC software is a user friendly software model for constructing fault trees based on decision tables. Software is equipped with libraries of decision tables for components commonly used in various Nuclear Power Plant (NPP) systems. User is expected to make a nodal diagram of the system, for which fault tree is to be constructed, from the flow sheets available. The text nodal diagram goes as the sole input defining the system flow chart. AFTC software is a rule based expert system which draws the fault tree from the system flow chart and component decision tables. AFTC software gives fault tree in both text and graphic format. Help is provided as how to enter system flow chart and component decision tables. The software is developed in 'C' language. Software is verified with simplified version of the fire water system of an Indian PHWR. Code conversion will be undertaken to create a window based version. (author)

  3. User manual of FRAPCON-I computer code

    International Nuclear Information System (INIS)

    Chia, C.T.

    1985-11-01

    The manual for using the FRAPCON-I code implanted by Reactor Department of Brazilian-CNEN to convert IBM FORTRAN in FORTRAN 77 of Honeywell Bull computer is presented. The FRAPCON-I code describes the behaviour of fuel rods of PWR type reactors at stationary state during long periods of burnup. (M.C.K.)

  4. SSC-K code user's manual

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Y.M.; Lee, Y.B.; Chang, W.P.; Hahn, D

    2000-07-01

    , constitutive laws and correlations required to execute the SSC-K are described. It is noted that the user's manual will be revised later with the further development of SSC-K.

  5. User manual of FUNF code for fissile material data calculation

    International Nuclear Information System (INIS)

    Zhang, Jingshang

    2006-03-01

    The FUNF code (2005 version) is used to calculate fast neutron reaction data of fissile materials with incident energies from about 1 keV up to 20 MeV. The first version of the FUNF code was completed in 1994. the code has been developed continually since that time and has often been used as an evaluation tool for setting up CENDL and for analyzing the measurements of fissile materials. During these years many improvements have been made. In this manual, the format of the input parameter files and the output files, as well as the functions of flag used in FUNF code, are introduced in detail, and the examples of the format of input parameters files are given. FUNF code consists of the spherical optical model, the Hauser-Feshbach model, and the unified Hauser-Feshbach and exciton model. (authors)

  6. The TESS [Tandem Experiment Simulation Studies] computer code user's manual

    International Nuclear Information System (INIS)

    Procassini, R.J.

    1990-01-01

    TESS (Tandem Experiment Simulation Studies) is a one-dimensional, bounded particle-in-cell (PIC) simulation code designed to investigate the confinement and transport of plasma in a magnetic mirror device, including tandem mirror configurations. Mirror plasmas may be modeled in a system which includes an applied magnetic field and/or a self-consistent or applied electrostatic potential. The PIC code TESS is similar to the PIC code DIPSI (Direct Implicit Plasma Surface Interactions) which is designed to study plasma transport to and interaction with a solid surface. The codes TESS and DIPSI are direct descendants of the PIC code ES1 that was created by A. B. Langdon. This document provides the user with a brief description of the methods used in the code and a tutorial on the use of the code. 10 refs., 2 tabs

  7. User manual for the KfK code PCROSS

    International Nuclear Information System (INIS)

    Ravndal, S.; Oblozinsky, P.; Kelzenberg, S.; Cierjacks, S.

    1991-08-01

    The PCROSS code calculates the so-called 'pseudo' cross sections for sequential (x,n) reactions and merges them together with 'collapsed' cross sections for neutron induced reactions into one file of cross sections. The file is tailored to provide an input for the FISPACT inventory code that calculates the activation and related radiological quantities of material irradiated in a neutron flux. The present report describes the structure of the KfK code PCROSS, outlines the role of subroutines, and provides necessary information for a practical user of the code. (orig.) [de

  8. Code development and analysis program. RELAP4/MOD7 (Version 2): user's manual

    International Nuclear Information System (INIS)

    1978-08-01

    This manual describes RELAP4/MOD7 (Version 2), which is the latest version of the RELAP4 LPWR blowdown code. Version 2 is a precursor to the final version of RELAP4/MOD7, which will address LPWR LOCA analysis in integral fashion (i.e., blowdown, refill, and reflood in continuous fashion). This manual describes the new code models and provides application information required to utilize the code. It must be used in conjunction with the RELAP4/MOD5 User's Manual (ANCR-NUREG-1335, dated September 1976), and the RELAP4/MOD6 User's Manual

  9. BWR plant dynamic analysis code BWRDYN user's manual

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Yoshida, Kazuo; Fujiki, Kazuo

    1989-06-01

    Computer code BWRDYN has been developed for thermal-hydraulic analysis of a BWR plant. It can analyze the various types of transient caused by not only small but also large disturbances such as operating mode changes and/or system malfunctions. The verification of main analytical models of the BWRDYN code has been performed with measured data of actual BWR plant. Furthermore, the installation of BOP (Balance of Plant) model has made it possible to analyze the effect of BOP on reactor system. This report describes on analytical models and instructions for user of the BWRDYN code. (author)

  10. User's manual for the FLORA equilibrium and stability code

    International Nuclear Information System (INIS)

    Freis, R.P.; Cohen, B.I.

    1985-01-01

    This document provides a user's guide to the content and use of the two-dimensional axisymmetric equilibrium and stability code FLORA. FLORA addresses the low-frequency MHD stability of long-thin axisymmetric tandem mirror systems with finite pressure and finite-larmor-radius effects. FLORA solves an initial-value problem for interchange, rotational, and ballooning stability

  11. NASA Lewis steady-state heat pipe code users manual

    International Nuclear Information System (INIS)

    Tower, L.K.

    1992-06-01

    The NASA Lewis heat pipe code has been developed to predict the performance of heat pipes in the steady state. The code can be used as a design tool on a personal computer or, with a suitable calling routine, as a subroutine for a mainframe radiator code. A variety of wick structures, including a user input option, can be used. Heat pipes with multiple evaporators, condensers, and adiabatic sections in series and with wick structures that differ among sections can be modeled. Several working fluids can be chosen, including potassium, sodium, and lithium, for which the monomer-dimer equilibrium is considered. The code incorporates a vapor flow algorithm that treats compressibility and axially varying heat input. This code facilitates the determination of heat pipe operating temperatures and heat pipe limits that may be encountered at the specified heat input and environment temperature. Data are input to the computer through a user-interactive input subroutine. Output, such as liquid and vapor pressures and temperatures, is printed at equally spaced axial positions along the pipe as determined by the user

  12. User's manual for the radioactive decay and accumulation code RADAC

    International Nuclear Information System (INIS)

    Salmon, R.; Loghry, S.L.; Ashline, R.C.

    1995-11-01

    The RADAC computer code calculates radioactive decay and accumulation of decayed products using an algorithm based on the direct use of the Bateman equations and referred to here as the yield factor method. This report explains the yield factor method, gives an overview of the various modules in the RADAC code system, and describes the decay and accumulation code in detail. The RADAC code has capacity for two waste types and can accommodate up to 60 years of annual waste inputs. Decay times as high as 1 million years can be calculated. The user supplies the undecayed composition and radioactivity of the waste placed in storage each year. The code calculates the decayed composition, radioactivity, and thermal power of the accumulated waste at the end of each year and gives the results in terms of grams and curies of individual radionuclides. Calculations can be made for up to 19 waste storage sites in a single run. For each site and each waste type, calculations can be made by 1-year steps up to 60 years, by 10-year steps to 160 years, and by 6 discrete steps to 1 million years. Detailed outputs can be printed for each waste site and each time step by individual radionuclides. Summarized outputs are also available. Excluding data-preparation time, RADAC requires about 2 min to run 19 waste sites with two types of transuranic waste at each site, using a 486 DX computer with a clock speed of 33 MHz. Because RADAC uses a preselected set of decay times and does not make in-reactor calculations, it should not be viewed as a substitute for ORIGEN2. RADAC is intended for use in applications in which accumulations at the decay times provided by the code are sufficient for the user's purposes

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

    International Nuclear Information System (INIS)

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

    1997-03-01

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

  14. User's manual for the Oak Ridge Tokamak Transport Code

    International Nuclear Information System (INIS)

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

    1977-02-01

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

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

    Science.gov (United States)

    1975-07-01

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

  16. User manual for PACTOLUS: a code for computing power costs

    International Nuclear Information System (INIS)

    Huber, H.D.; Bloomster, C.H.

    1979-02-01

    PACTOLUS is a computer code for calculating the cost of generating electricity. Through appropriate definition of the input data, PACTOLUS can calculate the cost of generating electricity from a wide variety of power plants, including nuclear, fossil, geothermal, solar, and other types of advanced energy systems. The purpose of PACTOLUS is to develop cash flows and calculate the unit busbar power cost (mills/kWh) over the entire life of a power plant. The cash flow information is calculated by two principal models: the Fuel Model and the Discounted Cash Flow Model. The Fuel Model is an engineering cost model which calculates the cash flow for the fuel cycle costs over the project lifetime based on input data defining the fuel material requirements, the unit costs of fuel materials and processes, the process lead and lag times, and the schedule of the capacity factor for the plant. For nuclear plants, the Fuel Model calculates the cash flow for the entire nuclear fuel cycle. For fossil plants, the Fuel Model calculates the cash flow for the fossil fuel purchases. The Discounted Cash Flow Model combines the fuel costs generated by the Fuel Model with input data on the capital costs, capital structure, licensing time, construction time, rates of return on capital, tax rates, operating costs, and depreciation method of the plant to calculate the cash flow for the entire lifetime of the project. The financial and tax structure for both investor-owned utilities and municipal utilities can be simulated through varying the rates of return on equity and debt, the debt-equity ratios, and tax rates. The Discounted Cash Flow Model uses the principal that the present worth of the revenues will be equal to the present worth of the expenses including the return on investment over the economic life of the project. This manual explains how to prepare the input data, execute cases, and interpret the output results with the updated version of PACTOLUS. 11 figures, 2 tables

  17. User manual for PACTOLUS: a code for computing power costs.

    Energy Technology Data Exchange (ETDEWEB)

    Huber, H.D.; Bloomster, C.H.

    1979-02-01

    PACTOLUS is a computer code for calculating the cost of generating electricity. Through appropriate definition of the input data, PACTOLUS can calculate the cost of generating electricity from a wide variety of power plants, including nuclear, fossil, geothermal, solar, and other types of advanced energy systems. The purpose of PACTOLUS is to develop cash flows and calculate the unit busbar power cost (mills/kWh) over the entire life of a power plant. The cash flow information is calculated by two principal models: the Fuel Model and the Discounted Cash Flow Model. The Fuel Model is an engineering cost model which calculates the cash flow for the fuel cycle costs over the project lifetime based on input data defining the fuel material requirements, the unit costs of fuel materials and processes, the process lead and lag times, and the schedule of the capacity factor for the plant. For nuclear plants, the Fuel Model calculates the cash flow for the entire nuclear fuel cycle. For fossil plants, the Fuel Model calculates the cash flow for the fossil fuel purchases. The Discounted Cash Flow Model combines the fuel costs generated by the Fuel Model with input data on the capital costs, capital structure, licensing time, construction time, rates of return on capital, tax rates, operating costs, and depreciation method of the plant to calculate the cash flow for the entire lifetime of the project. The financial and tax structure for both investor-owned utilities and municipal utilities can be simulated through varying the rates of return on equity and debt, the debt-equity ratios, and tax rates. The Discounted Cash Flow Model uses the principal that the present worth of the revenues will be equal to the present worth of the expenses including the return on investment over the economic life of the project. This manual explains how to prepare the input data, execute cases, and interpret the output results. (RWR)

  18. CSTEM User Manual

    Science.gov (United States)

    Hartle, M.; McKnight, R. L.

    2000-01-01

    This manual is a combination of a user manual, theory manual, and programmer manual. The reader is assumed to have some previous exposure to the finite element method. This manual is written with the idea that the CSTEM (Coupled Structural Thermal Electromagnetic-Computer Code) user needs to have a basic understanding of what the code is actually doing in order to properly use the code. For that reason, the underlying theory and methods used in the code are described to a basic level of detail. The manual gives an overview of the CSTEM code: how the code came into existence, a basic description of what the code does, and the order in which it happens (a flowchart). Appendices provide a listing and very brief description of every file used by the CSTEM code, including the type of file it is, what routine regularly accesses the file, and what routine opens the file, as well as special features included in CSTEM.

  19. Users' manual for the FTDRAW (Fault Tree Draw) code

    International Nuclear Information System (INIS)

    Oikawa, Tetsukuni; Hikawa, Michihiro; Tanabe, Syuichi; Nakamura, Norihiro

    1985-02-01

    This report provides the information needed to use the FTDRAW (Fault Tree Draw) code, which is designed for drawing a fault tree. The FTDRAW code has several optional functions, such as the overview of a fault tree output, fault tree output in English description, fault tree output in Japanese description and summary tree output. Inputs for the FTDRAW code are component failure rate information and gate information which are filed out by a execution of the FTA-J (Fault Tree Analysis-JAERI) code system and option control data. Using the FTDRAW code, we can get drawings of fault trees which is easy to see, efficiently. (author)

  20. User manual for the probabilistic fuel performance code FRP

    International Nuclear Information System (INIS)

    Friis Jensen, J.; Misfeldt, I.

    1980-10-01

    This report describes the use of the probabilistic fuel performance code FRP. Detailed description of both input to and output from the program are given. The use of the program is illustrated by an example. (author)

  1. The computer code EURDYN-1M (release 2). User's manual

    International Nuclear Information System (INIS)

    1982-01-01

    EURDYN-1M is a finite element computer code developed at J.R.C. Ispra to compute the response of two-dimensional coupled fluid-structure configurations to transient dynamic loading for reactor safety studies. This report gives instructions for preparing input data to EURDYN-1M, release 2, and describes a test problem in order to illustrate both the input and the output of the code

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

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C

    1979-07-01

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

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

    International Nuclear Information System (INIS)

    Ahnert, C.

    1979-01-01

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

  4. The computer code Eurdyn - 1 M. (Release 1) Part 2: User's Manual

    International Nuclear Information System (INIS)

    Donea, J.; Giuliani, S.

    1979-01-01

    This report is the user's manual for the computer code Eurdyn-1 M developed at the J.R.C. Ispra for use in containment and fuel subassembly analyses for fast reactor safety studies. The input data are defined and a test problem is presented to illustrate both the input and the output of results

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

    International Nuclear Information System (INIS)

    Anhert, C.

    1979-01-01

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

  6. Users manual for CAFE-3D : a computational fluid dynamics fire code

    International Nuclear Information System (INIS)

    Khalil, Imane; Lopez, Carlos; Suo-Anttila, Ahti Jorma

    2005-01-01

    The Container Analysis Fire Environment (CAFE) computer code has been developed to model all relevant fire physics for predicting the thermal response of massive objects engulfed in large fires. It provides realistic fire thermal boundary conditions for use in design of radioactive material packages and in risk-based transportation studies. The CAFE code can be coupled to commercial finite-element codes such as MSC PATRAN/THERMAL and ANSYS. This coupled system of codes can be used to determine the internal thermal response of finite element models of packages to a range of fire environments. This document is a user manual describing how to use the three-dimensional version of CAFE, as well as a description of CAFE input and output parameters. Since this is a user manual, only a brief theoretical description of the equations and physical models is included

  7. User Manual for the NASA Glenn Ice Accretion Code LEWICE: Version 2.0

    Science.gov (United States)

    Wright, William B.

    1999-01-01

    A research project is underway at NASA Glenn to produce a computer code which can accurately predict ice growth under a wide range of meteorological conditions for any aircraft surface. This report will present a description of the code inputs and outputs from version 2.0 of this code, which is called LEWICE. This version differs from previous releases due to its robustness and its ability to reproduce results accurately for different spacing and time step criteria across computing platform. It also differs in the extensive effort undertaken to compare the results against the database of ice shapes which have been generated in the NASA Glenn Icing Research Tunnel (IRT) 1. This report will only describe the features of the code related to the use of the program. The report will not describe the inner working of the code or the physical models used. This information is available in the form of several unpublished documents which will be collectively referred to as a Programmers Manual for LEWICE 2 in this report. These reports are intended as an update/replacement for all previous user manuals of LEWICE. In addition to describing the changes and improvements made for this version, information from previous manuals may be duplicated so that the user will not need to consult previous manuals to use this code.

  8. User's manual for BINIAC: A computer code to translate APET bins

    International Nuclear Information System (INIS)

    Gough, S.T.

    1994-03-01

    This report serves as the user's manual for the FORTRAN code BINIAC. BINIAC is a utility code designed to format the output from the Defense Waste Processing Facility (DWPF) Accident Progression Event Tree (APET) methodology. BINIAC inputs the accident progression bins from the APET methodology, converts the frequency from occurrences per hour to occurrences per year, sorts the progression bins, and converts the individual dimension character codes into facility attributes. Without the use of BINIAC, this process would be done manually at great time expense. BINIAC was written under the quality assurance control of IQ34 QAP IV-1, revision 0, section 4.1.4. Configuration control is established through the use of a proprietor and a cognizant users list

  9. Integrated Fuel-Coolant Interaction (IFCI 7.0) Code User's Manual

    Energy Technology Data Exchange (ETDEWEB)

    Young, Michael F.

    1999-05-01

    The integrated fuel-coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, three-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a description of IFCI 7.0. The user's manual describes the hydrodynamic method and physical models used in IFCI 7.0. Appendix A is an input manual provided for the creation of working decks.

  10. Integrated Fuel-Coolant Interaction (IFCI 7.0) Code User's Manual

    International Nuclear Information System (INIS)

    Young, Michael F.

    1999-01-01

    The integrated fuel-coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, three-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a description of IFCI 7.0. The user's manual describes the hydrodynamic method and physical models used in IFCI 7.0. Appendix A is an input manual provided for the creation of working decks

  11. User's manual for the Heat Pipe Space Radiator design and analysis Code (HEPSPARC)

    Science.gov (United States)

    Hainley, Donald C.

    1991-01-01

    A heat pipe space radiatior code (HEPSPARC), was written for the NASA Lewis Research Center and is used for the design and analysis of a radiator that is constructed from a pumped fluid loop that transfers heat to the evaporative section of heat pipes. This manual is designed to familiarize the user with this new code and to serve as a reference for its use. This manual documents the completed work and is intended to be the first step towards verification of the HEPSPARC code. Details are furnished to provide a description of all the requirements and variables used in the design and analysis of a combined pumped loop/heat pipe radiator system. A description of the subroutines used in the program is furnished for those interested in understanding its detailed workings.

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

    International Nuclear Information System (INIS)

    Tilley, D.R.

    1983-03-01

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

  13. MINTEQ user's manual

    International Nuclear Information System (INIS)

    Peterson, S.R.; Hostetler, C.J.; Deutsch, W.J.; Cowan, C.E.

    1987-02-01

    This manual will aid the user in applying the MINTEQ geochemical computer code to model aqueous solutions and the interactions of aqueous solutions with hypothesized assemblages of solid phases. The manual will provide a basic understanding of how the MINTEQ computer code operates and the important principles that are incorporated into the code and instruct a user of the MINTEQ code on how to create input files to simulate a variety of geochemical problems. Chapters 2 through 8 are for the user who has some experience with or wishes to review the principles important to geochemical computer codes. These chapters include information on the methodology MINTEQ uses to incorporate these principles into the code. Chapters 9 through 11 are for the user who wants to know how to create input data files to model various types of problems. 35 refs., 2 figs., 5 tabs

  14. User Manual for the NASA Glenn Ice Accretion Code LEWICE. Version 2.2.2

    Science.gov (United States)

    Wright, William B.

    2002-01-01

    A research project is underway at NASA Glenn to produce a computer code which can accurately predict ice growth under a wide range of meteorological conditions for any aircraft surface. This report will present a description of the code inputs and outputs from version 2.2.2 of this code, which is called LEWICE. This version differs from release 2.0 due to the addition of advanced thermal analysis capabilities for de-icing and anti-icing applications using electrothermal heaters or bleed air applications. An extensive effort was also undertaken to compare the results against the database of electrothermal results which have been generated in the NASA Glenn Icing Research Tunnel (IRT) as was performed for the validation effort for version 2.0. This report will primarily describe the features of the software related to the use of the program. Appendix A of this report has been included to list some of the inner workings of the software or the physical models used. This information is also available in the form of several unpublished documents internal to NASA. This report is intended as a replacement for all previous user manuals of LEWICE. In addition to describing the changes and improvements made for this version, information from previous manuals may be duplicated so that the user will not need to consult previous manuals to use this code.

  15. ABAREX -- A neutron spherical optical-statistical-model code -- A user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Smith, A.B. [ed.; Lawson, R.D.

    1998-06-01

    The contemporary version of the neutron spherical optical-statistical-model code ABAREX is summarized with the objective of providing detailed operational guidance for the user. The physical concepts involved are very briefly outlined. The code is described in some detail and a number of explicit examples are given. With this document one should very quickly become fluent with the use of ABAREX. While the code has operated on a number of computing systems, this version is specifically tailored for the VAX/VMS work station and/or the IBM-compatible personal computer.

  16. Light water reactor fuel analysis code FEMAXI-IV(Ver.2). Detailed structure and user's manual

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Saitou, Hiroaki.

    1997-11-01

    A light water reactor fuel behavior analysis code FEMAXI-IV(Ver.2) was developed as an improved version of FEMAXI-IV. Development of FEMAXI-IV has been already finished in 1992, though a detailed structure and input manual of the code have not been open to users yet. Here, the basic theories and structure, the models and numerical solutions applied to FEMAXI-IV(Ver.2), and the material properties adopted in the code are described in detail. In FEMAXI-IV(Ver.2), programming bugs in previous FEMAXI-IV were eliminated, renewal of the pellet thermal conductivity was performed, and a model of thermal-stress restraint on FP gas release was incorporated. For facilitation of effective and wide-ranging application of the code, methods of input/output of the code are also described in detail, and sample output is included. (author)

  17. User's manual for SPLPLOT-2: a computer code for data plotting and editing in conversational mode

    International Nuclear Information System (INIS)

    Muramatsu, Ken; Matsumoto, Kiyoshi; Kohsaka, Atsuo; Maniwa, Masaki.

    1985-07-01

    The computer code SPLPLOT-2 for plotting and data editing has been developed as a part of the code package: SPLPACK-1. The SPLPLOT-2 code has capabilities of both conversational and batch processings. This report describes the user's manual for SPLPLOT-2. The following improvements have been made in the SPLPLOT-2. (1) It has capabilities of both conversational and batch processings, (2) function of conversion of files from the input SPL (Standard PLotter) files to internal work files have been implemented to reduce number of time consuming access to the input SPL files, (3) user supplied subroutines can be assigned for data editing from the SPL files, (4) in addition to the two-dimensional graphs, streamline graphs, contour line graphs and bird's-eye view graphs can be drawn. (author)

  18. ELCOS: the PSI code system for LWR core analysis. Part II: user's manual for the fuel assembly code BOXER

    International Nuclear Information System (INIS)

    Paratte, J.M.; Grimm, P.; Hollard, J.M.

    1996-02-01

    ELCOS is a flexible code system for the stationary simulation of light water reactor cores. It consists of the four computer codes ETOBOX, BOXER, CORCOD and SILWER. The user's manual of the second one is presented here. BOXER calculates the neutronics in cartesian geometry. The code can roughly be divided into four stages: - organisation: choice of the modules, file manipulations, reading and checking of input data, - fine group fluxes and condensation: one-dimensional calculation of fluxes and computation of the group constants of homogeneous materials and cells, - two-dimensional calculations: geometrically detailed simulation of the configuration in few energy groups, - burnup: evolution of the nuclide densities as a function of time. This manual shows all input commands which can be used while running the different modules of BOXER. (author) figs., tabs., refs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

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

    International Nuclear Information System (INIS)

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

    1985-04-01

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

  1. User's manual for computer code RIBD-II, a fission product inventory code

    International Nuclear Information System (INIS)

    Marr, D.R.

    1975-01-01

    The computer code RIBD-II is used to calculate inventories, activities, decay powers, and energy releases for the fission products generated in a fuel irradiation. Changes from the earlier RIBD code are: the expansion to include up to 850 fission product isotopes, input in the user-oriented NAMELIST format, and run-time choice of fuels from an extensively enlarged library of nuclear data. The library that is included in the code package contains yield data for 818 fission product isotopes for each of fourteen different fissionable isotopes, together with fission product transmutation cross sections for fast and thermal systems. Calculational algorithms are little changed from those in RIBD. (U.S.)

  2. Detailed description and user`s manual of high burnup fuel analysis code EXBURN-I

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Saitou, Hiroaki

    1997-11-01

    EXBURN-I has been developed for the analysis of LWR high burnup fuel behavior in normal operation and power transient conditions. In the high burnup region, phenomena occur which are different in quality from those expected for the extension of behaviors in the mid-burnup region. To analyze these phenomena, EXBURN-I has been formed by the incorporation of such new models as pellet thermal conductivity change, burnup-dependent FP gas release rate, and cladding oxide layer growth to the basic structure of low- and mid-burnup fuel analysis code FEMAXI-IV. The present report describes in detail the whole structure of the code, models, and materials properties. Also, it includes a detailed input manual and sample output, etc. (author). 55 refs.

  3. 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.

  4. SHARP User Manual

    International Nuclear Information System (INIS)

    Yu, Y. Q.; Shemon, E. R.; Thomas, J. W.; Mahadevan, Vijay S.; Rahaman, Ronald O.; Solberg, Jerome

    2016-01-01

    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.

  5. Light water reactor fuel analysis code FEMAXI-IV(Ver.2). Detailed structure and user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Saitou, Hiroaki

    1997-11-01

    A light water reactor fuel behavior analysis code FEMAXI-IV(Ver.2) was developed as an improved version of FEMAXI-IV. Development of FEMAXI-IV has been already finished in 1992, though a detailed structure and input manual of the code have not been open to users yet. Here, the basic theories and structure, the models and numerical solutions applied to FEMAXI-IV(Ver.2), and the material properties adopted in the code are described in detail. In FEMAXI-IV(Ver.2), programming bugs in previous FEMAXI-IV were eliminated, renewal of the pellet thermal conductivity was performed, and a model of thermal-stress restraint on FP gas release was incorporated. For facilitation of effective and wide-ranging application of the code, methods of input/output of the code are also described in detail, and sample output is included. (author)

  6. Sandia National Laboratories environmental fluid dynamics code. Marine Hydrokinetic Module User's Manual

    Energy Technology Data Exchange (ETDEWEB)

    James, Scott Carlton; Roberts, Jesse D

    2014-03-01

    This document describes the marine hydrokinetic (MHK) input file and subroutines for the Sandia National Laboratories Environmental Fluid Dynamics Code (SNL-EFDC), which is a combined hydrodynamic, sediment transport, and water quality model based on the Environmental Fluid Dynamics Code (EFDC) developed by John Hamrick [1], formerly sponsored by the U.S. Environmental Protection Agency, and now maintained by Tetra Tech, Inc. SNL-EFDC has been previously enhanced with the incorporation of the SEDZLJ sediment dynamics model developed by Ziegler, Lick, and Jones [2-4]. SNL-EFDC has also been upgraded to more accurately simulate algae growth with specific application to optimizing biomass in an open-channel raceway for biofuels production [5]. A detailed description of the input file containing data describing the MHK device/array is provided, along with a description of the MHK FORTRAN routine. Both a theoretical description of the MHK dynamics as incorporated into SNL-EFDC and an explanation of the source code are provided. This user manual is meant to be used in conjunction with the original EFDC [6] and sediment dynamics SNL-EFDC manuals [7]. Through this document, the authors provide information for users who wish to model the effects of an MHK device (or array of devices) on a flow system with EFDC and who also seek a clear understanding of the source code, which is available from staff in the Water Power Technologies Department at Sandia National Laboratories, Albuquerque, New Mexico.

  7. SHEAT for PC. A computer code for probabilistic seismic hazard analysis for personal computer, user's manual

    International Nuclear Information System (INIS)

    Yamada, Hiroyuki; Tsutsumi, Hideaki; Ebisawa, Katsumi; Suzuki, Masahide

    2002-03-01

    The SHEAT code developed at Japan Atomic Energy Research Institute is for probabilistic seismic hazard analysis which is one of the tasks needed for seismic Probabilistic Safety Assessment (PSA) of a nuclear power plant. At first, SHEAT was developed as the large sized computer version. In addition, a personal computer version was provided to improve operation efficiency and generality of this code in 2001. It is possible to perform the earthquake hazard analysis, display and the print functions with the Graphical User Interface. With the SHEAT for PC code, seismic hazard which is defined as an annual exceedance frequency of occurrence of earthquake ground motions at various levels of intensity at a given site is calculated by the following two steps as is done with the large sized computer. One is the modeling of earthquake generation around a site. Future earthquake generation (locations, magnitudes and frequencies of postulated earthquake) is modeled based on the historical earthquake records, active fault data and expert judgment. Another is the calculation of probabilistic seismic hazard at the site. An earthquake ground motion is calculated for each postulated earthquake using an attenuation model taking into account its standard deviation. Then the seismic hazard at the site is calculated by summing the frequencies of ground motions by all the earthquakes. This document is the user's manual of the SHEAT for PC code. It includes: (1) Outline of the code, which include overall concept, logical process, code structure, data file used and special characteristics of code, (2) Functions of subprogram and analytical models in them, (3) Guidance of input and output data, (4) Sample run result, and (5) Operational manual. (author)

  8. User's manual for seismic analysis code 'SONATINA-2V'

    Energy Technology Data Exchange (ETDEWEB)

    Hanawa, Satoshi; Iyoku, Tatsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-08-01

    The seismic analysis code, SONATINA-2V, has been developed to analyze the behavior of the HTTR core graphite components under seismic excitation. The SONATINA-2V code is a two-dimensional computer program capable of analyzing the vertical arrangement of the HTTR graphite components, such as fuel blocks, replaceable reflector blocks, permanent reflector blocks, as well as their restraint structures. In the analytical model, each block is treated as rigid body and is restrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions between upper and lower blocks. Moreover, the SONATINA-2V code is capable of analyzing the core vibration behavior under both simultaneous excitations of vertical and horizontal directions. The SONATINA-2V code is composed of the main program, pri-processor for making the input data to SONATINA-2V and post-processor for data processing and making the graphics from analytical results. Though the SONATINA-2V code was developed in order to work in the MSP computer system of Japan Atomic Energy Research Institute (JAERI), the computer system was abolished with the technical progress of computer. Therefore, improvement of this analysis code was carried out in order to operate the code under the UNIX machine, SR8000 computer system, of the JAERI. The users manual for seismic analysis code, SONATINA-2V, including pri- and post-processor is given in the present report. (author)

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

    International Nuclear Information System (INIS)

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

    1983-03-01

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

  10. VIPRE-01: a thermal-hydraulic analysis code for reactor cores. Volume 2. User's manual

    International Nuclear Information System (INIS)

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.

    1983-04-01

    VIPRE (Versatile Internals and Component Program for Reactors; EPRI) has been developed for nuclear power utility thermal-hydraulic analysis applications. It is designed to help evaluate nuclear energy reactor core safety limits including minimum departure from nucleate boiling ratio (MDNBR), critical power ratio (CPR), fuel and clad temperatures, and coolant state in normal operation and assumed accident conditions. This volume (Volume 2: User's Manual) describes the input requirements of VIPRE and its auxiliary programs, SPECSET, ASP and DECCON, and lists the input instructions for each code

  11. In-vessel source term analysis code TRACER version 2.3. User's manual

    International Nuclear Information System (INIS)

    Toyohara, Daisuke; Ohno, Shuji; Hamada, Hirotsugu; Miyahara, Shinya

    2005-01-01

    A computer code TRACER (Transport Phenomena of Radionuclides for Accident Consequence Evaluation of Reactor) version 2.3 has been developed to evaluate species and quantities of fission products (FPs) released into cover gas during a fuel pin failure accident in an LMFBR. The TRACER version 2.3 includes new or modified models shown below. a) Both model: a new model for FPs release from fuel. b) Modified model for FPs transfer from fuel to bubbles or sodium coolant. c) Modified model for bubbles dynamics in coolant. Computational models, input data and output data of the TRACER version 2.3 are described in this user's manual. (author)

  12. PREREM: an interactive data preprocessing code for INREM II. Part I: user's manual. Part II: code structure

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, M.T.; Fields, D.E.

    1981-05-01

    PREREM is an interactive computer code developed as a data preprocessor for the INREM-II (Killough, Dunning, and Pleasant, 1978a) internal dose program. PREREM is intended to provide easy access to current and self-consistent nuclear decay and radionuclide-specific metabolic data sets. Provision is made for revision of metabolic data, and the code is intended for both production and research applications. Documentation for the code is in two parts. Part I is a user's manual which emphasizes interpretation of program prompts and choice of user input. Part II stresses internal structure and flow of program control and is intended to assist the researcher who wishes to revise or modify the code or add to its capabilities. PREREM is written for execution on a Digital Equipment Corporation PDP-10 System and much of the code will require revision before it can be run on other machines. The source program length is 950 lines (116 blocks) and computer core required for execution is 212 K bytes. The user must also have sufficient file space for metabolic and S-factor data sets. Further, 64 100 K byte blocks of computer storage space are required for the nuclear decay data file. Computer storage space must also be available for any output files produced during the PREREM execution. 9 refs., 8 tabs.

  13. A user's manual for the three-dimensional Monte Carlo transport code SPARTAN

    International Nuclear Information System (INIS)

    Bending, R.C.; Heffer, P.J.H.

    1975-09-01

    SPARTAN is a general-purpose Monte Carlo particle transport code intended for neutron or gamma transport problems in reactor physics, health physics, shielding, and safety studies. The code used a very general geometry system enabling a complex layout to be described and allows the user to obtain physics data from a number of different types of source library. Special tracking and scoring techniques are used to improve the quality of the results obtained. To enable users to run SPARTAN, brief descriptions of the facilities available in the code are given and full details of data input and job control language, as well as examples of complete calculations, are included. It is anticipated that changes may be made to SPARTAN from time to time, particularly in those parts of the code which deal with physics data processing. The load module is identified by a version number and implementation date, and updates of sections of this manual will be issued when significant changes are made to the code. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-01

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

  15. SHEAT: a computer code for probabilistic seismic hazard analysis, user's manual

    International Nuclear Information System (INIS)

    Ebisawa, Katsumi; Kondo, Masaaki; Abe, Kiyoharu; Tanaka, Toshiaki; Takani, Michio.

    1994-08-01

    The SHEAT code developed at Japan Atomic Energy Research Institute is for probabilistic seismic hazard analysis which is one of the tasks needed for seismic Probabilistic Safety Assessment (PSA) of a nuclear power plant. Seismic hazard is defined as an annual exceedance frequency of occurrence of earthquake ground motions at various levels of intensity at a given site. With the SHEAT code, seismic hazard is calculated by the following two steps: (1) Modeling of earthquake generation around a site. Future earthquake generation (locations, magnitudes and frequencies of postulated earthquakes) is modelled based on the historical earthquake records, active fault data and expert judgement. (2) Calculation of probabilistic seismic hazard at the site. An earthquake ground motion is calculated for each postulated earthquake using an attenuation model taking into account its standard deviation. Then the seismic hazard at the site is calculated by summing the frequencies of ground motions by all the earthquakes. This document is the user's manual of the SHEAT code. It includes: (1) Outlines of the code, which include overall concept, logical process, code structure, data file used and special characteristics of the code, (2) Functions of subprograms and analytical models in them, (3) Guidance of input and output data, and (4) Sample run results. The code has widely been used at JAERI to analyze seismic hazard at various nuclear power plant sites in japan. (author)

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

    International Nuclear Information System (INIS)

    Devary, J.L.

    1987-06-01

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

  17. User's manuals of probabilistic fracture mechanics analysis code for aged piping, PASCAL-SP

    International Nuclear Information System (INIS)

    Itoh, Hiroto; Nishikawa, Hiroyuki; Onizawa, Kunio; Kato, Daisuke; Osakabe, Kazuya

    2010-03-01

    As a part of research on the material degradation and structural integrity assessment for aged LWR components, a PFM (Probabilistic Fracture Mechanics) analysis code PASCAL-SP (PFM Analysis of Structural Components in Aging LWR - Stress Corrosion Cracking at Welded Joints of Piping) has been developed. This code evaluates the failure probabilities at welded joints of aged piping by a Monte Carlo method. PASCAL-SP treats stress corrosion cracking (SCC) and fatigue crack growth in piping, according to the approaches of NISA and JSME FFS Code. The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the latest knowledge in the SCC assessment and fracture criteria of piping. In addition, the accuracy of flaw detection and sizing at in-service inspection and residual stress distribution were modeled based on experimental data and introduced into PASCAL-SP. This code has been developed for a cross-check use by the regulatory body in Japan. In addition to this, this code can also be used for a research purpose by researchers in academia and industries. This report provides the user's manual and theoretical background of the code. (author)

  18. ELCOS: the PSI code system for LWR core analysis. Part II: user`s manual for the fuel assembly code BOXER

    Energy Technology Data Exchange (ETDEWEB)

    Paratte, J.M.; Grimm, P.; Hollard, J.M. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-02-01

    ELCOS is a flexible code system for the stationary simulation of light water reactor cores. It consists of the four computer codes ETOBOX, BOXER, CORCOD and SILWER. The user`s manual of the second one is presented here. BOXER calculates the neutronics in cartesian geometry. The code can roughly be divided into four stages: - organisation: choice of the modules, file manipulations, reading and checking of input data, - fine group fluxes and condensation: one-dimensional calculation of fluxes and computation of the group constants of homogeneous materials and cells, - two-dimensional calculations: geometrically detailed simulation of the configuration in few energy groups, - burnup: evolution of the nuclide densities as a function of time. This manual shows all input commands which can be used while running the different modules of BOXER. (author) figs., tabs., refs.

  19. User's manual for the BNW-II optimization code for dry/wet-cooled power plants

    International Nuclear Information System (INIS)

    Braun, D.J.; Bamberger, J.A.; Braun, D.J.; Faletti, D.W.; Wiles, L.E.

    1978-05-01

    The User's Manual describes how to operate BNW-II, a computer code developed by the Pacific Northwest Laboratory (PNL) as a part of its activities under the Department of Energy (DOE) Dry Cooling Enhancement Program. The computer program offers a comprehensive method of evaluating the cost savings potential of dry/wet-cooled heat rejection systems. Going beyond simple ''figure-of-merit'' cooling tower optimization, this method includes such items as the cost of annual replacement capacity, and the optimum split between plant scale-up and replacement capacity, as well as the purchase and operating costs of all major heat rejection components. Hence the BNW-II code is a useful tool for determining potential cost savings of new dry/wet surfaces, new piping, or other components as part of an optimized system for a dry/wet-cooled plant

  20. CANAL user's manual

    International Nuclear Information System (INIS)

    Faya, A.; Wolf, L.; Todreas, N.

    1979-11-01

    CANAL is a subchannel computer program for the steady-state and transient thermal hydraulic analysis of BWR fuel rod bundles. The purpose of this manual is to introduce the user into the mechanism of running the code by providing information about the input data and options

  1. GRSAC Users Manual

    International Nuclear Information System (INIS)

    Ball, S.J.; Nypaver, D.J.

    1999-01-01

    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

  2. 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.

  3. FISPACT 97: user manual

    International Nuclear Information System (INIS)

    Forrest, R.A.; Sublet, J.-Ch.

    1997-05-01

    FISPACT is the inventory code included in the European Activation System (EASY). A new version of FISPACT: FISPACT-97 has been developed and this report is the User manual for the code. It explains the use of all the code words used in the input file to specify a FISPACT run and describes how all the data files are connected. A series of appendices cover the working of the code and the physical and mathematical details. Background information on the data files and extensive examples of input files suitable for various applications are included. (Author)

  4. User's manual of SECOM2: a computer code for seismic system reliability analysis

    International Nuclear Information System (INIS)

    Uchiyama, Tomoaki; Oikawa, Tetsukuni; Kondo, Masaaki; Tamura, Kazuo

    2002-03-01

    This report is the user's manual of seismic system reliability analysis code SECOM2 (Seismic Core Melt Frequency Evaluation Code Ver.2) developed at the Japan Atomic Energy Research Institute for systems reliability analysis, which is one of the tasks of seismic probabilistic safety assessment (PSA) of nuclear power plants (NPPs). The SECOM2 code has many functions such as: Calculation of component failure probabilities based on the response factor method, Extraction of minimal cut sets (MCSs), Calculation of conditional system failure probabilities for given seismic motion levels at the site of an NPP, Calculation of accident sequence frequencies and the core damage frequency (CDF) with use of the seismic hazard curve, Importance analysis using various indicators, Uncertainty analysis, Calculation of the CDF taking into account the effect of the correlations of responses and capacities of components, and Efficient sensitivity analysis by changing parameters on responses and capacities of components. These analyses require the fault tree (FT) representing the occurrence condition of the system failures and core damage, information about response and capacity of components and seismic hazard curve for the NPP site as inputs. This report presents the models and methods applied in the SECOM2 code and how to use those functions. (author)

  5. User's manual of the REFLA-1D/MODE4 reflood thermo-hydrodynamic analysis code

    International Nuclear Information System (INIS)

    Hojo, Tsuneyuki; Iguchi, Tadashi; Okubo, Tsutomu; Murao, Yoshio; Sugimoto, Jun.

    1986-01-01

    REFLA-1D/MODE4 code has been developed by incorporating local power effect model and fuel temperature profile effect model into REFLA-1D/MODE3 code. This code can calculate the temperature transient of local rod by considering radial power profile effect in core and simulate the thermal characteristics of the nuclear fuel rod. This manual describes the outline of incorporated models, modification of the code with incorporating models and provides application information required to utilize the code. (author)

  6. Rotor Wake/Stator Interaction Noise Prediction Code Technical Documentation and User's Manual

    Science.gov (United States)

    Topol, David A.; Mathews, Douglas C.

    2010-01-01

    This report documents the improvements and enhancements made by Pratt & Whitney to two NASA programs which together will calculate noise from a rotor wake/stator interaction. The code is a combination of subroutines from two NASA programs with many new features added by Pratt & Whitney. To do a calculation V072 first uses a semi-empirical wake prediction to calculate the rotor wake characteristics at the stator leading edge. Results from the wake model are then automatically input into a rotor wake/stator interaction analytical noise prediction routine which calculates inlet aft sound power levels for the blade-passage-frequency tones and their harmonics, along with the complex radial mode amplitudes. The code allows for a noise calculation to be performed for a compressor rotor wake/stator interaction, a fan wake/FEGV interaction, or a fan wake/core stator interaction. This report is split into two parts, the first part discusses the technical documentation of the program as improved by Pratt & Whitney. The second part is a user's manual which describes how input files are created and how the code is run.

  7. GANDALF: users' manual

    International Nuclear Information System (INIS)

    Strout, R.E. II; Beach, J.L.

    1977-01-01

    The GANDALF computer code was written to calculate neutron dose equivalent given the pulse-height data obtained by using a Linear Energy Transfer (LET) proportional counter. The code also uses pre- and/or post-calibration spectra, from an alpha source, to determine a calibration factor in keV/μ/channel. Output from the code consists of the effective radius of the detection chamber in microns, a calibration factor in keV/μ/channel, and the total dose and dose equivalent in rad or rem between any two LET energies by using the equations by Attix and Roesch [Radiation Dosimetry, 1, 71 (1968)]. This report is a user's manual and is not intended as anything else, and assumes that the user has a basic knowledge of the LLL Octopus timesharing system. However, a very brief description of how the code operates is included

  8. Tripoli-3: monte Carlo transport code for neutral particles - version 3.5 - users manual

    International Nuclear Information System (INIS)

    Vergnaud, Th.; Nimal, J.C.; Chiron, M.

    2001-01-01

    The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)

  9. RELAP/MOD3 code manual: User`s guidelines. Volume 5, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Fletcher, C.D.; Schultz, R.R. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

    1995-08-01

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

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

    International Nuclear Information System (INIS)

    Ahnert, C.; Aragones, J. M.

    1981-01-01

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

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

    International Nuclear Information System (INIS)

    Ahnert, C.; Aragones, J.M.

    1981-01-01

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

  12. IMAGE User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Stehfest, E; De Waal, L; Oostenrijk, R.

    2010-09-15

    This user manual contains the basic information for running the simulation model IMAGE ('Integrated Model to Assess the Global Environment') of PBL. The motivation for this report was a substantial restructuring of the source code for IMAGE version 2.5. The document gives concise content information about the submodels, tells the user how to install the program, describes the directory structure of the run environment, shows how scenarios have to be prepared and run, and gives insight in the restart functionality.

  13. ASSERT-4 user's manual

    International Nuclear Information System (INIS)

    Judd, R.A.; Tahir, A.; Carver, M.B.; Stewart, D.G.; Thibeault, P.R.; Rowe, D.S.

    1984-09-01

    ASSERT-4 is an advanced subchannel code being developed primarily to model single- and two-phase flow and heat transfer in horizontal rod bundles. This manual is intended to facilitate the application of this code to the analysis of flow in reactor fuel channels. It contains a brief description of the thermalhydraulic model and ASSERT-4 solution scheme, and other information required by users. This other information includes a detailed discussion of input data requirements, a sample problem and solution, and information describing how to access and run ASSERT-4 on the Chalk River computers

  14. 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.

  15. A computer code to estimate accidental fire and radioactive airborne releases in nuclear fuel cycle facilities: User's manual for FIRIN

    International Nuclear Information System (INIS)

    Chan, M.K.; Ballinger, M.Y.; Owczarski, P.C.

    1989-02-01

    This manual describes the technical bases and use of the computer code FIRIN. This code was developed to estimate the source term release of smoke and radioactive particles from potential fires in nuclear fuel cycle facilities. FIRIN is a product of a broader study, Fuel Cycle Accident Analysis, which Pacific Northwest Laboratory conducted for the US Nuclear Regulatory Commission. The technical bases of FIRIN consist of a nonradioactive fire source term model, compartment effects modeling, and radioactive source term models. These three elements interact with each other in the code affecting the course of the fire. This report also serves as a complete FIRIN user's manual. Included are the FIRIN code description with methods/algorithms of calculation and subroutines, code operating instructions with input requirements, and output descriptions. 40 refs., 5 figs., 31 tabs

  16. MELCOR computer code manuals

    Energy Technology Data Exchange (ETDEWEB)

    Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A.; Hyman, C.R.; Sanders, R.L. [Oak Ridge National Lab., TN (United States)

    1995-03-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.

  17. MELCOR computer code manuals

    International Nuclear Information System (INIS)

    Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L.; Hodge, S.A.; Hyman, C.R.; Sanders, R.L.

    1995-03-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR's phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package

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

    International Nuclear Information System (INIS)

    Shirakawa, Toshihiko; Hatanaka, Koichiro

    2001-11-01

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

  19. WAM-E user's manual

    International Nuclear Information System (INIS)

    Rayes, L.G.; Riley, J.E.

    1986-07-01

    The WAM-E series of mainframe computer codes have been developed to efficiently analyze the large binary models (e.g., fault trees) used to represent the logic relationships within and between the systems of a nuclear power plant or other large, multisystem entity. These codes have found wide application in reliability and safety studies of nuclear power plant systems. There are now nine codes in the WAM-E series, with six (WAMBAM/WAMTAP, WAMCUT, WAMCUT-II, WAMFM, WAMMRG, and SPASM) classified as Type A Production codes and the other three (WAMFTP, WAMTOP, and WAMCONV) classified as Research codes. This document serves as a combined User's Guide, Programmer's Manual, and Theory Reference for the codes, with emphasis on the Production codes. To that end, the manual is divided into four parts: Part I, Introduction; Part II, Theory and Numerics; Part III, WAM-E User's Guide; and Part IV, WAMMRG Programmer's Manual

  20. LCS Users Manual

    International Nuclear Information System (INIS)

    Redd, A.J.; Ignat, D.W.

    1998-01-01

    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

  1. CALENDF-2010: user manual

    International Nuclear Information System (INIS)

    Sublet, Jean-Christophe; Ribon, Pierre; Coste-Delclaux, Mireille

    2011-09-01

    CALENDF-2010 represents a Fortran-95 update of the 1994, 2001 then 2005 code distribution with emphasise on programming quality and standards, physics and usage improvements. Devised to process multigroup cross-sections it relies on Gauss quadrature mathematical principle and strength. The followings processes can be handled by the code: moment probability table and effective cross-section calculation; pointwise cross section, probability table and effective cross-section regrouping; probability table condensation; probability table mix for several isotopes; probability table interpolation; effective cross section based probability table calculations; probability table calculations from effective cross-sections; cross-section comparison, complete energy pointwise cross-section processing and thickness dependent averaged transmission sample calculation. The CALENDF user manual, after having listed all principal code functions, describes sequentially each of them and gives comments on their associated output streams. Installation procedures, test cases and running time platform comparisons are given in the appendix. (authors)

  2. 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.

  3. User's manual for the G.T.M.-1 computer code

    International Nuclear Information System (INIS)

    Prado-Herrero, P.

    1992-01-01

    This document describes the GTM-1 ( Geosphere Transport Model, release-1) computer code and is intended to provide the reader with enough detailed information in order to use the code. GTM-1 was developed for the assessment of radionuclide migration by the ground water through geologic deposits whose properties can change along the pathway.GTM-1 solves the transport equation by the finite differences method ( Crank-Nicolson scheme ). It was developped for specific use within Probabilistic System Assessment (PSA) Monte Carlo Method codes; in this context the first application of GTM-1 was within the LISA (Long Term Isolation System Assessment) code. GTM-1 is also available as an independent model, which includes various submodels simulating a multi-barrier disposal system. The code has been tested with the PSACOIN ( Probabilistic System Assessment Codes intercomparison) benchmarks exercises from PSAC User Group (OECD/NEA). 10 refs., 6 Annex., 2 tabs

  4. Manually operated coded switch

    International Nuclear Information System (INIS)

    Barnette, J.H.

    1978-01-01

    The disclosure related to a manually operated recodable coded switch in which a code may be inserted, tried and used to actuate a lever controlling an external device. After attempting a code, the switch's code wheels must be returned to their zero positions before another try is made

  5. Probe code: a set of programs for processing and analysis of the left ventricular function - User's manual

    International Nuclear Information System (INIS)

    Piva, R.M.V.

    1987-01-01

    The User's Manual of the Probe Code is an addendum to the M.Sc. thesis entitled A Microcomputer System of Nuclear Probe to Check the Left Ventricular Function. The Probe Code is a software which was developed for processing and off-line analysis curves from the Left Ventricular Function, that were obtained in vivo. These curves are produced by means of an external scintigraph probe, which was collimated and put on the left ventricule, after a venous inoculation of Tc-99 m. (author)

  6. Langley Stability and Transition Analysis Code (LASTRAC) Version 1.2 User Manual

    Science.gov (United States)

    Chang, Chau-Lyan

    2004-01-01

    LASTRAC is a general-purposed, physics-based transition prediction code released by NASA for Laminar Flow Control studies and transition research. The design and development of the LASTRAC code is aimed at providing an engineering tool that is easy to use and yet capable of dealing with a broad range of transition related issues. It was written from scratch based on the state-of-the-art numerical methods for stability analysis and modern software technologies. At low fidelity, it allows users to perform linear stability analysis and N-factor transition correlation for a broad range of flow regimes and configurations by using either the linear stability theory or linear parabolized stability equations method. At high fidelity, users may use nonlinear PSE to track finite-amplitude disturbances until the skin friction rise. This document describes the governing equations, numerical methods, code development, detailed description of input/output parameters, and case studies for the current release of LASTRAC.

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

    International Nuclear Information System (INIS)

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

    1995-06-01

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

  8. ABAQUS-EPGEN: a general-purpose finite-element code. Volume 1. User's manual

    International Nuclear Information System (INIS)

    Hibbitt, H.D.; Karlsson, B.I.; Sorensen, E.P.

    1982-10-01

    This document is the User's Manual for ABAQUS/EPGEN, a general purpose finite element computer program, designed specifically to serve advanced structural analysis needs. The program contains very general libraries of elements, materials and analysis procedures, and is highly modular, so that complex combinations of features can be put together to model physical problems. The program is aimed at production analysis needs, and for this purpose aspects such as ease-of-use, reliability, flexibility and efficiency have received maximum attention. The input language is designed to make it straightforward to describe complicated models; the analysis procedures are highly automated with the program choosing time or load increments based on user supplied tolerances and controls; and the program offers a wide range of post-processing options for display of the analysis results

  9. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user's manual

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document is the User's Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code's capabilities and limitations; Chapter 2 describes the code's structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs

  10. User's manual for DSTAR MOD1: A comprehensive tokamak disruption code

    International Nuclear Information System (INIS)

    Merrill, B.J.; Jardin, S.J.

    1986-01-01

    A computer code, DSTAR, has recently been developed to quantify the surface erosion and induced forces that can occur during major tokamak plasma disruptions. The DSTAR code development effort has been accomplished by coupling a recently developed free boundary tokamak plasma transport computational model with other models developed to predict impurity transport and radiation, and the electromagnetic and thermal dynamic response of vacuum vessel components. The combined model, DSTAR, is a unique tool for predicting the consequences of tokamak disruptions. This informal report discusses the sequence of events of a resistive disruption, models developed to predict plasma transport and electromagnetic field evolution, the growth of the stochastic region of the plasma, the transport and nonequilibrium ionization/emitted radiation of the ablated vacuum vessel material, the vacuum vessel thermal and magnetic response, and user input and code output

  11. RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.; Mallen, A.N.; Neymotin, L.Y.

    1998-03-01

    This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARC and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.

  12. Users Manual for the FEHMN application

    International Nuclear Information System (INIS)

    Zyvoloski, G.A.; Robinson, B.A.; Dash, Z.V.; Trease, L.L.

    1996-01-01

    The user's manual documents the use of the Yucca Mountain Site Characterization Projects Finite element heat and mass transfer code (FEHMN) application. The manual covers: Program considerations, data files, input data, output, system interface, and examples

  13. JASMINE-pro: A computer code for the analysis of propagation process in steam explosions. User's manual

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yanhua; Nilsuwankosit, Sunchai; Moriyama, Kiyofumi; Maruyama, Yu; Nakamura, Hideo; Hashimoto, Kazuichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-12-01

    A steam explosion is a phenomenon where a high temperature liquid gives its internal energy very rapidly to another low temperature volatile liquid, causing very strong pressure build up due to rapid vaporization of the latter. In the field of light water reactor safety research, steam explosions caused by the contact of molten core and coolant has been recognized as a potential threat which could cause failure of the pressure vessel or the containment vessel during a severe accident. A numerical simulation code JASMINE was developed at Japan Atomic Energy Research Institute (JAERI) to evaluate the impact of steam explosions on the integrity of reactor boundaries. JASMINE code consists of two parts, JASMINE-pre and -pro, which handle the premixing and propagation phases in steam explosions, respectively. JASMINE-pro code simulates the thermo-hydrodynamics in the propagation phase of a steam explosion on the basis of the multi-fluid model for multiphase flow. This report, 'User's Manual', gives the usage of JASMINE-pro code as well as the information on the code structures which should be useful for users to understand how the code works. (author)

  14. JASMINE-pro: A computer code for the analysis of propagation process in steam explosions. User's manual

    International Nuclear Information System (INIS)

    Yang, Yanhua; Nilsuwankosit, Sunchai; Moriyama, Kiyofumi; Maruyama, Yu; Nakamura, Hideo; Hashimoto, Kazuichiro

    2000-12-01

    A steam explosion is a phenomenon where a high temperature liquid gives its internal energy very rapidly to another low temperature volatile liquid, causing very strong pressure build up due to rapid vaporization of the latter. In the field of light water reactor safety research, steam explosions caused by the contact of molten core and coolant has been recognized as a potential threat which could cause failure of the pressure vessel or the containment vessel during a severe accident. A numerical simulation code JASMINE was developed at Japan Atomic Energy Research Institute (JAERI) to evaluate the impact of steam explosions on the integrity of reactor boundaries. JASMINE code consists of two parts, JASMINE-pre and -pro, which handle the premixing and propagation phases in steam explosions, respectively. JASMINE-pro code simulates the thermo-hydrodynamics in the propagation phase of a steam explosion on the basis of the multi-fluid model for multiphase flow. This report, 'User's Manual', gives the usage of JASMINE-pro code as well as the information on the code structures which should be useful for users to understand how the code works. (author)

  15. User and reference manual for the KfK code INS

    International Nuclear Information System (INIS)

    Daum, E.

    1993-09-01

    The INS code (Intense Neutron Source) serves to calculate uncollided neutron flux contours, neutron flux volumes and spatial-dependent neutron flux spectra in the test cell of an intense neutron source, of the t-H 2 O or d-Li concept. With the information of the neutron flux spectra the neutron irradiation damage like displacements per atom (DPA), H- and He-production rates and the generation of foreign elements by transmutations can be calculated for any element at any position in the test cell. This manual gives an introduction into the theory of neutron flux calculation of thick targets and neutron irradiation damage calculations. It is explained how the code is working and the handling of the input and output parameters. For each application of the several code modules an example is given. The results like contours, spectra, flux volumes and damage rates are summarized in tabular form and graphically. Damage and element transmutation data have been calculated for 23 isotopes and compared with the DEMO 1st wall values. (orig./HP) [de

  16. User manual of Visual Balan V. 1.0 Interactive code for water balances and refueling estimation

    International Nuclear Information System (INIS)

    Samper, J.; Huguet, L.; Ares, J.; Garcia, M. A.

    1999-01-01

    This document contains the Users Manual of Visual Balan V1.0, an updated version of Visual Balan V0.0 (Samper et al., 1997). Visual Balan V1.0 performs daily water balances in the soil, the unsaturated zone and the aquifer in a user-friendly environment which facilitates both the input data process and the postprocessing of results. The main inputs of the balance are rainfall and irrigation while the outputs are surface runoff, evapotranspiration, interception, inter flow and groundwater flow. The code evaluates all these components in a sequential manner by starting with rainfall and irrigation, which must be provided by the user, and continuing with interception, surface runoff, evapotranspiration, and potential recharge (water flux crossing the bottom of the soil). This potential recharge is the input to the unsaturated zone where water can flow horizontally as subsurface flow (inter flow) or vertically as percolation into the aquifer. (Author)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  18. 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)

  19. 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)

  20. MELCOR computer code manuals: Primer and user's guides, Version 1.8.3 September 1994. Volume 1

    International Nuclear Information System (INIS)

    Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.; Stuart, D.S.; Thompson, S.L.; Hodge, S.A.; Hyman, C.R.; Sanders, R.L.

    1995-03-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR's phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users' Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package

  1. DIMAC program user's manual

    International Nuclear Information System (INIS)

    Lee, Byoung Oon; Song, Tae Young

    2003-11-01

    DIMAC (A DIspersion Metallic fuel performance Analysis Code) is a computer program for simulating the behavior of dispersion fuel rods under normal operating conditions of HYPER. It computes the one-dimensional temperature distribution and the thermo-mechanical characteristics of fuel rod under the steady state operation condition, including the swelling and rod deformation. DIMAC was developed based on the experience of research reactor fuel. DIMAC consists of the temperature calculation module, the mechanical swelling calculation module, and the fuel deformation calculation module in order to predict the deformation of a dispersion fuel as a function of power history. Because there are a little of available U-TRU-Zr or TRU-Zr characteristics, the material data of U-Pu-Zr or Pu-Zr are used for those of U-TRU-Zr or TRU-Zr. This report is mainly intended as a user's manual for the DIMAC code. The general description on this code, the description on input parameter, the description on each subroutine, the sample problem and the sample input and partial output are written in this report

  2. DIMAC program user's manual

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byoung Oon; Song, Tae Young

    2003-11-01

    DIMAC (A DIspersion Metallic fuel performance Analysis Code) is a computer program for simulating the behavior of dispersion fuel rods under normal operating conditions of HYPER. It computes the one-dimensional temperature distribution and the thermo-mechanical characteristics of fuel rod under the steady state operation condition, including the swelling and rod deformation. DIMAC was developed based on the experience of research reactor fuel. DIMAC consists of the temperature calculation module, the mechanical swelling calculation module, and the fuel deformation calculation module in order to predict the deformation of a dispersion fuel as a function of power history. Because there are a little of available U-TRU-Zr or TRU-Zr characteristics, the material data of U-Pu-Zr or Pu-Zr are used for those of U-TRU-Zr or TRU-Zr. This report is mainly intended as a user's manual for the DIMAC code. The general description on this code, the description on input parameter, the description on each subroutine, the sample problem and the sample input and partial output are written in this repo0008.

  3. User's manual for the Oak Ridge Tokamak Transport Code

    Energy Technology Data Exchange (ETDEWEB)

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

    1977-02-01

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

  4. EMAP Users Manual.

    Science.gov (United States)

    Kotz, Arnold; Redondo, Rory

    Presented is the user's manual for the Educational Manpower Information Sources Project (EMAP), an information file containing approximately 325 document abstracts related to the field of educational planning. (The EMAP file is described in document SP 006 747.) (JB)

  5. GERTS GQ User's Manual.

    Science.gov (United States)

    Akiba, Y.; And Others

    This user's manual for the simulation program Graphical Evaluation and Review Technique (GERT) GQ contains sections on nodes, branches, program input description and format, and program output, as well as examples. Also included is a programmer's manual which contains information on scheduling, subroutine descriptions, COMMON Variables, and…

  6. Dosimetry and health effects self-teaching curriculum: illustrative problems to supplement the user's manual for the Dosimetry and Health Effects Computer Code

    International Nuclear Information System (INIS)

    Runkle, G.E.; Finley, N.C.

    1983-03-01

    This document contains a series of sample problems for the Dosimetry and Health Effects Computer Code to be used in conjunction with the user's manual (Runkle and Cranwell, 1982) for the code. This code was developed at Sandia National Laboratories for the Risk Methodology for Geologic Disposal of Radioactive Waste program (NRC FIN A-1192). The purpose of this document is to familiarize the user with the code, its capabilities, and its limitations. When the user has finished reading this document, he or she should be able to prepare data input for the Dosimetry and Health Effects code and have some insights into interpretation of the model output

  7. A probabilistic assessment code system for derivation of clearance levels of radioactive materials. PASCLR user's manual

    International Nuclear Information System (INIS)

    Takahashi, Tomoyuki; Takeda, Seiji; Kimura, Hideo

    2001-01-01

    It is indicated that some types of radioactive material generating from the development and utilization of nuclear energy do not need to be subject regulatory control because they can only give rise to trivial radiation hazards. The process to remove such materials from regulatory control is called as 'clearance'. The corresponding levels of the concentration of radionuclides are called as 'clearance levels'. In the Nuclear Safety Commission's discussion, the deterministic approach was applied to derive the clearance levels, which are the concentrations of radionuclides in a cleared material equivalent to an individual dose criterion. Basically, realistic parameter values were selected for it. If the realistic values could not be defined, reasonably conservative values were selected. Additionally, the stochastic approaches were performed to validate the results which were obtained by the deterministic calculations. We have developed a computer code system PASCLR (Probabilistic Assessment code System for derivation of Clearance Levels of Radioactive materials) by using the Monte Carlo technique for carrying out the stochastic calculations. This report describes the structure and user information for execution of PASCLR code. (author)

  8. A User's Manual for MASH V1.5 - A Monte Carlo Adjoint Shielding Code System

    Energy Technology Data Exchange (ETDEWEB)

    C. O. Slater; J. M. Barnes; J. O. Johnson; J.D. Drischler

    1998-10-01

    The Monte Carlo ~djoint ~ielding Code System, MASH, calculates neutron and gamma- ray environments and radiation protection factors for armored military vehicles, structures, trenches, and other shielding configurations by coupling a forward discrete ordinates air- over-ground transport calculation with an adjoint Monte Carlo treatment of the shielding geometry. Efficiency and optimum use of computer time are emphasized. The code system includes the GRTUNCL and DORT codes for air-over-ground transport calculations, the MORSE code with the GIFT5 combinatorial geometry package for adjoint shielding calculations, and several peripheral codes that perform the required data preparations, transformations, and coupling functions. The current version, MASH v 1.5, is the successor to the original MASH v 1.0 code system initially developed at Oak Ridge National Laboratory (ORNL). The discrete ordinates calculation determines the fluence on a coupling surface surrounding the shielding geometry due to an external neutron/gamma-ray source. The Monte Carlo calculation determines the effectiveness of the fluence at that surface in causing a response in a detector within the shielding geometry, i.e., the "dose importance" of the coupling surface fluence. A coupling code folds the fluence together with the dose importance, giving the desired dose response. The coupling code can determine the dose response as a function of the shielding geometry orientation relative to the source, distance from the source, and energy response of the detector. This user's manual includes a short description of each code, the input required to execute the code along with some helpful input data notes, and a representative sample problem.

  9. User's manual for the BNW-II optimization code for dry/wet-cooled power plants

    Energy Technology Data Exchange (ETDEWEB)

    Braun, D.J.; Bamberger, J.A.; Braun, D.J.; Faletti, D.W.; Wiles, L.E.

    1978-05-01

    The User's Manual describes how to operate BNW-II, a computer code developed by the Pacific Northwest Laboratory (PNL) as a part of its activities under the Department of Energy (DOE) Dry Cooling Enhancement Program. The computer program offers a comprehensive method of evaluating the cost savings potential of dry/wet-cooled heat rejection systems. Going beyond simple ''figure-of-merit'' cooling tower optimization, this method includes such items as the cost of annual replacement capacity, and the optimum split between plant scale-up and replacement capacity, as well as the purchase and operating costs of all major heat rejection components. Hence the BNW-II code is a useful tool for determining potential cost savings of new dry/wet surfaces, new piping, or other components as part of an optimized system for a dry/wet-cooled plant.

  10. User's manual and analysis methodology of probabilistic fracture mechanics analysis code PASCAL Ver.2 for reactor pressure vessel (Contract research)

    International Nuclear Information System (INIS)

    Osakabe, Kazuya; Onizawa, Kunio; Shibata, Katsuyuki; Kato, Daisuke

    2006-09-01

    As a part of the aging structural integrity research for LWR components, the probabilistic fracture mechanics (PFM) analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed in JAEA. This code evaluates the conditional probabilities of crack initiation and fracture of a reactor pressure vessel (RPV) under transient conditions such as pressurized thermal shock (PTS). The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the recent development in the fracture mechanics and computer performance. PASCAL Ver.1 has functions of optimized sampling in the stratified Monte Carlo simulation, elastic-plastic fracture criterion of the R6 method, crack growth analysis models for a semi-elliptical crack, recovery of fracture toughness due to thermal annealing and so on. Since then, under the contract between the Ministry of Economy, Trading and Industry of Japan and JAEA, we have continued to develop and introduce new functions into PASCAL Ver.2 such as the evaluation method for an embedded crack, K I database for a semi-elliptical crack considering stress discontinuity at the base/cladding interface, PTS transient database, and others. A generalized analysis method is proposed on the basis of the development of PASCAL Ver.2 and results of sensitivity analyses. Graphical user interface (GUI) including a generalized method as default values has been also developed for PASCAL Ver.2. This report provides the user's manual and theoretical background of PASCAL Ver.2. (author)

  11. TRUBA User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Tereshchenko, M. A.; Castejon, F.; Cappa, A.

    2008-04-25

    The TRUBA (pipeline in Russian) code is a computational tool for studying the propagation of Gaussian-shaped microwave beams in a prescribed equilibrium plasma. This manual covers the basic material handed to use the implementation of TRUBA (version 3,4) interfaced with the numerical library of the TJ-II stellarator. The manual provides a concise theoretical background of the problem, specifications for setting up the input files and interpreting the output of the code, and some information useful in modifying TRUBA. (Author) 13 refs.

  12. TRUBA User Manual

    International Nuclear Information System (INIS)

    Tereshchenko, M. A.; Castejon, F.; Cappa, A.

    2008-01-01

    The TRUBA (pipeline in Russian) code is a computational tool for studying the propagation of Gaussian-shaped microwave beams in a prescribed equilibrium plasma. This manual covers the basic material handed to use the implementation of TRUBA (version 3,4) interfaced with the numerical library of the TJ-II stellarator. The manual provides a concise theoretical background of the problem, specifications for setting up the input files and interpreting the output of the code, and some information useful in modifying TRUBA. (Author) 13 refs

  13. AELIB user's manual

    International Nuclear Information System (INIS)

    Evans, L.E.; Klawitter, G.L.

    1981-05-01

    This report is an updatable manual for users of AELIB, the AECL Library of FORTRAN-callable routines for the CRNL CDC 6600/CYBER 170 system. It provides general advice on the use of this library and detailed information on the selection and usage of particular library routines

  14. User manual for semi-circular compact range reflector code: Version 2

    Science.gov (United States)

    Gupta, Inder J.; Burnside, Walter D.

    1987-01-01

    A computer code has been developed at the Ohio State University ElectroScience Laboratory to analyze a semi-circular paraboloidal reflector with or without a rolled edge at the top and a skirt at the bottom. The code can be used to compute the total near field of the reflector or its individual components at a given distance from the center of the paraboloid. The code computes the fields along a radial, horizontal, vertical or axial cut at that distance. Thus, it is very effective in computing the size of the sweet spot for a semi-circular compact range reflector. This report describes the operation of the code. Various input and output statements are explained. Some results obtained using the computer code are presented to illustrate the code's capability as well as being samples of input/output sets.

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

    International Nuclear Information System (INIS)

    Vonach, H.

    1982-01-01

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

  16. VIPRE-01: a thermal-hydraulic code for reactor cores. Volume 2: user's manual (Revision 2)

    International Nuclear Information System (INIS)

    Cuta, J.M.; Koontz, A.S.; Stewart, C.W.; Montgomery, S.D.; Nomura, K.K.

    1985-07-01

    Revisions to the VIPRE code documents for Volume 2 are presented. These revisions conform to the changes made to VIPRE-01, CYCLE-00 to produce the new version of the code denoted by VIPRE-01, CYCLE-01. The first pages of the revisions specify where the replacement pages are to be inserted and which pages of the original documents should be retained

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

    International Nuclear Information System (INIS)

    Lim, Doo-Hyun

    2006-03-01

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

  18. 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...

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

    International Nuclear Information System (INIS)

    1995-08-01

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

  20. Data Parallel Line Relaxation (DPLR) Code User Manual: Acadia - Version 4.01.1

    Science.gov (United States)

    Wright, Michael J.; White, Todd; Mangini, Nancy

    2009-01-01

    Data-Parallel Line Relaxation (DPLR) code is a computational fluid dynamic (CFD) solver that was developed at NASA Ames Research Center to help mission support teams generate high-value predictive solutions for hypersonic flow field problems. The DPLR Code Package is an MPI-based, parallel, full three-dimensional Navier-Stokes CFD solver with generalized models for finite-rate reaction kinetics, thermal and chemical non-equilibrium, accurate high-temperature transport coefficients, and ionized flow physics incorporated into the code. DPLR also includes a large selection of generalized realistic surface boundary conditions and links to enable loose coupling with external thermal protection system (TPS) material response and shock layer radiation codes.

  1. Light water reactor fuel analysis code. FEMAXI-6 (Ver.1). Detailed structure and user's manual

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Saitou, Hiroaki

    2006-02-01

    A light water reactor fuel analysis code FEMAXI-6 is an advanced version which has been produced by integrating the former version FEMAXI-V with numerous functional improvements and extensions. In particular, the FEMAXI-6 code has attained a complete coupled solution of thermal analysis and mechanical analysis, enabling an accurate prediction of pellet-clad gap size and PCMI in high burnup fuel rods. Also, such new models have been implemented as pellet-clad bonding and fission gas bubble swelling, and linkage function with detailed burning analysis code has been enhanced. Furthermore, a number of new materials properties and parameters have been introduced. With these advancements, the FEMAXI-6 code has been upgraded to a versatile analytical tool for high burnup fuel behavior not only in the normal operation but also in anticipated transient conditions. This report describes in detail the design, basic theory and structure, models and numerical method, improvements and extensions, and method of model modification. In order to facilitate effective and wide-ranging application of the code, formats and methods of input/output of the code are also described, and a sample output in an actual form is included. (author)

  2. AVRAM user's manual

    International Nuclear Information System (INIS)

    McGrady, P.W.

    1988-02-01

    This document details the use of the reliability code for the Atomic Vapor Laser Isotope Separation (AVLIS) project. This code was designed by Tom Anklam and John Harris. In late 1984 Patrick McGrady and Elena Koontz of C and TD/TA were assigned the task of improving the code and converting it for use on the DEC-10 system. In early 1986, Patric McGrady converted it to the CRAY. The AVRAM code is divided into distinct parts (often referred to as programs in this document). There is a COSMOS file that controls the execution of the FORTRAN code and controls the naming of output datasets and the deletion of temporary datasets created by the code. The FORTRAN code consists of a main program as a driver and of three main subroutines: EDIT, PARAM, and AVRAM. The EDIT program allows the user to create a new user defined system or add to an existing system or to change certain parameters. The PARAM program allows the user to alter system parameters and to select options such as economics run, criticality analysis or sensitivity studies. The AVRAM program does a reliability analysis of the system

  3. RELAP/MOD3 code manual: User's guidelines. Volume 5, Revision 1

    International Nuclear Information System (INIS)

    Fletcher, C.D.; Schultz, R.R.

    1995-08-01

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

  4. Description and user's manual of light water reactor fuel analysis code FEMAXI-IV (Ver.2)

    International Nuclear Information System (INIS)

    Suzuki, Motoe; Saitou, Hiroaki.

    1997-03-01

    FEMAXI-IV is an advanced version of FEMAXI-III, the analysis code of light water reactor fuel behavior in which various functions and improvements have been incorporated. The present report describes in detail the basic theories and structure, the models and numerical solutions applied, and the material properties adopted in the version 2 which is an improved version of the first version of FEMAXI-IV. In FEMAXI-IV (Ver.2), bugs have been fixed, pellet thermal conductivity properties have been updated, and thermal-stress-induced FP gas release model have been incorporated. In order to facilitate effective and wide-ranging application of the code, types and methods of input/output of the code are also described, and a sample output in an actual form is included. (author)

  5. IAC user manual

    Science.gov (United States)

    Vos, R. G.; Beste, D. L.; Gregg, J.

    1984-01-01

    The User Manual for the Integrated Analysis Capability (IAC) Level 1 system is presented. The IAC system currently supports the thermal, structures, controls and system dynamics technologies, and its development is influenced by the requirements for design/analysis of large space systems. The system has many features which make it applicable to general problems in engineering, and to management of data and software. Information includes basic IAC operation, executive commands, modules, solution paths, data organization and storage, IAC utilities, and module implementation.

  6. 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.

  7. TMAP/Mod 1: Tritium Migration Analysis Program code description and user's manual

    International Nuclear Information System (INIS)

    Merrill, B.J.; Jones, J.L.; Holland, D.F.

    1986-01-01

    The Tritium Migration Analysis Program (TMAP) has been developed by the Fusion Safety Program of EG and G Idaho, Inc., at the Idaho National Engineering Laboratory (INEL) as a safety analysis code to analyze tritium loss from fusion systems during normal operation and under accident conditions. TMAP is a one-dimensional code that determines tritium movement and inventories in a system of interconnected enclosures and wall structures. In addition, the thermal response of structures is modeled to provide temperature information required for calculations of tritium movement. The program is written in FORTRAN 4 and has been implemented on the National Magnetic Fusion Energy Computing Center (NMFECC)

  8. Users manual for the FORSS sensitivity and uncertainty analysis code system

    International Nuclear Information System (INIS)

    Lucius, J.L.; Weisbin, C.R.; Marable, J.H.; Drischler, J.D.; Wright, R.Q.; White, J.E.

    1981-01-01

    FORSS is a code system used to study relationships between nuclear reaction cross sections, integral experiments, reactor performance parameter predictions and associated uncertainties. This report describes the computing environment and the modules currently used to implement FORSS Sensitivity and Uncertainty Methodology

  9. User's manual for the vertical axis wind turbine performance computer code darter

    Energy Technology Data Exchange (ETDEWEB)

    Klimas, P. C.; French, R. E.

    1980-05-01

    The computer code DARTER (DARrieus, Turbine, Elemental Reynolds number) is an aerodynamic performance/loads prediction scheme based upon the conservation of momentum principle. It is the latest evolution in a sequence which began with a model developed by Templin of NRC, Canada and progressed through the Sandia National Laboratories-developed SIMOSS (SSImple MOmentum, Single Streamtube) and DART (SARrieus Turbine) to DARTER.

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

    International Nuclear Information System (INIS)

    Anhert, C.

    1980-01-01

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

  11. Users manual for the FORSS sensitivity and uncertainty analysis code system

    Energy Technology Data Exchange (ETDEWEB)

    Lucius, J.L.; Weisbin, C.R.; Marable, J.H.; Drischler, J.D.; Wright, R.Q.; White, J.E.

    1981-01-01

    FORSS is a code system used to study relationships between nuclear reaction cross sections, integral experiments, reactor performance parameter predictions and associated uncertainties. This report describes the computing environment and the modules currently used to implement FORSS Sensitivity and Uncertainty Methodology.

  12. User's manual for the FEHM application - A finite-element heat- and mass-transfer code

    International Nuclear Information System (INIS)

    Zyvoloski, G.A.; Robinson, B.A.; Dash, Z.V.; Trease, L.L.

    1997-07-01

    The use of this code is applicable to natural-state studies of geothermal systems and groundwater flow. A primary use of the FEHM application will be to assist in the understanding of flow fields and mass transport in the saturated and unsaturated zones below the proposed Yucca Mountain nuclear waste repository in Nevada. The equations of heat and mass transfer for multiphase flow in porous and permeable media are solved in the FEHM application by using the finite-element method. The permeability and porosity of the medium are allowed to depend on pressure and temperature. The code also has provisions for movable air and water phases and noncoupled tracers; that is, tracer solutions that do not affect the heat- and mass-transfer solutions. The tracers can be passive or reactive. The code can simulate two-dimensional, two-dimensional radial, or three-dimensional geometries. In fact, FEHM is capable of describing flow that is dominated in many areas by fracture and fault flow, including the inherently three-dimensional flow that results from permeation to and from faults and fractures. The code can handle coupled heat and mass-transfer effects, such as boiling, dryout, and condensation that can occur in the near-field region surrounding the potential repository and the natural convection that occurs through Yucca Mountain due to seasonal temperature changes. This report outlines the uses and capabilities of the FEHM application, initialization of code variables, restart procedures, and error processing. The report describes all the data files, the input data, including individual input records or parameters, and the various output files. The system interface is described, including the software environment and installation instructions

  13. A User’s Manual for: Electromagnetic Surface Patch Code (ESP).

    Science.gov (United States)

    1981-07-01

    common e equivalent to say physical contact, that the two (or The parameter IOVL thi 5 code intersect ove rlap, ce contin dge Inse ing...cross polarizations is shown. The phase of field field is also shown (the incident wave at the origon). For backscattering STPM = both are printed...scattered fields phi polarized. STPM = scattering cross section with incident field theta polarized and scattered field phi polarized. SPTM

  14. BBC users manual

    International Nuclear Information System (INIS)

    Ltterst, R.F.; Sutcliffe, W.G.; Warshaw, S.I.

    1977-11-01

    BBC is a two-dimensional, multifluid Eulerian hydro-radiation code based on KRAKEN and some subsequent ideas. It was developed in the explosion group in T-Division as a basic two-dimensional code to which various types of physics can be added. For this reason BBC is a FORTRAN (LRLTRAN) code. In order to gain the 2-to-1 to 4-to-1 speed advantage of the STACKLIB software on the 7600's and to be able to execute at high speed on the STAR, the vector extensions of LRLTRAN (STARTRAN) are used throughout the code. Either cylindrical- or slab-type problems can be run on BBC. The grid is bounded by a rectangular band of boundary zones. The interfaces between the regular and boundary zones can be selected to be either rigid or nonrigid. The setup for BBC problems is described in the KEG Manual and LEG Manual. The difference equations are described in BBC Hydrodynamics. Basic input and output for BBC are described

  15. STAIRS User's Manual

    Energy Technology Data Exchange (ETDEWEB)

    Gadjokov, V; Dragulev, V; Gove, N; Schmid, H

    1976-10-15

    The STorage And Information Retrieval System (STAIRS) of IBM is described from the user's point of view. The description is based on the experimental use of STAIRS at the IAEA computer, with INIS and AGRIS data bases, from June 1975 to May 1976. Special attention is paid to what may be termed the hierarchical approach to retrieval in STAIRS. Such an approach allows for better use of the intrinsic data-base structure and, hence, contributes to higher recall and/or relevance ratios in retrieval. The functions carried out by STAIRS are explained and the communication language between the user and the system outlined. Details are given of the specific structure of the INIS and AGRIS data bases for STAIRS. The manual should enable an inexperienced user to start his first on-line dialogues by means of a CRT or teletype terminal. (author)

  16. STAIRS User's Manual

    International Nuclear Information System (INIS)

    Gadjokov, V.; Dragulev, V.; Gove, N.; Schmid, H.

    1976-10-01

    The STorage And Information Retrieval System (STAIRS) of IBM is described from the user's point of view. The description is based on the experimental use of STAIRS at the IAEA computer, with INIS and AGRIS data bases, from June 1975 to May 1976. Special attention is paid to what may be termed the hierarchical approach to retrieval in STAIRS. Such an approach allows for better use of the intrinsic data-base structure and, hence, contributes to higher recall and/or relevance ratios in retrieval. The functions carried out by STAIRS are explained and the communication language between the user and the system outlined. Details are given of the specific structure of the INIS and AGRIS data bases for STAIRS. The manual should enable an inexperienced user to start his first on-line dialogues by means of a CRT or teletype terminal. (author)

  17. Plasma Separation Process: Betacell (BCELL) code: User's manual. [Bipolar barrier junction

    Energy Technology Data Exchange (ETDEWEB)

    Taherzadeh, M.

    1987-11-13

    The emergence of clearly defined applications for (small or large) amounts of long-life and reliable power sources has given the design and production of betavoltaic systems a new life. Moreover, because of the availability of the plasma separation program, (PSP) at TRW, it is now possible to separate the most desirable radioisotopes for betacell power generating devices. A computer code, named BCELL, has been developed to model the betavoltaic concept by utilizing the available up-to-date source/cell parameters. In this program, attempts have been made to determine the betacell energy device maximum efficiency, degradation due to the emitting source radiation and source/cell lifetime power reduction processes. Additionally, comparison is made between the Schottky and PN junction devices for betacell battery design purposes. Certain computer code runs have been made to determine the JV distribution function and the upper limit of the betacell generated power for specified energy sources. A Ni beta emitting radioisotope was used for the energy source and certain semiconductors were used for the converter subsystem of the betacell system. Some results for a Promethium source are also given here for comparison. 16 refs.

  18. FORIG: a computer code for calculating radionuclide generation and depletion in fusion and fission reactors. User's manual

    International Nuclear Information System (INIS)

    Blink, J.A.

    1985-03-01

    In this manual we describe the use of the FORIG computer code to solve isotope-generation and depletion problems in fusion and fission reactors. FORIG runs on a Cray-1 computer and accepts more extensive activation cross sections than ORIGEN2 from which it was adapted. This report is an updated and a combined version of the previous ORIGEN2 and FORIG manuals. 7 refs., 15 figs., 13 tabs

  19. User's manual for computer code SOLTES-1 (simulator of large thermal energy systems)

    International Nuclear Information System (INIS)

    Fewell, M.E.; Grandjean, N.R.; Dunn, J.C.; Edenburn, M.W.

    1978-09-01

    SOLTES simulates the steady-state response of thermal energy systems to time-varying data such as weather and loads. Thermal energy system models of both simple and complex systems can easily be modularly constructed from a library of routines. These routines mathematically model solar collectors, pumps, switches, thermal energy storage, thermal boilers, auxiliary boilers, heat exchangers, extraction turbines, extraction turbine/generators, condensers, regenerative heaters, air conditioners, heating and cooling of buildings, process vapor, etc.; SOLTES also allows user-supplied routines. The analyst need only specify fluid names to obtain readout of property data for heat-transfer fluids and constants that characterize power-cycle working fluids from a fluid property data bank. A load management capability allows SOLTES to simulate total energy systems that simultaneously follow heat and power loads and demands. Generalized energy accounting is available, and values for system performance parameters may be automatically determined by SOLTES. Because of its modularity and flexibility, SOLTES can be used to simulate a wide variety of thermal energy systems such as solar power/total energy, fossil fuel power plants/total energy, nuclear power plants/total energy, solar energy heating and cooling, geothermal energy, and solar hot water heaters

  20. User`s manual for the FEHM application -- A finite-element heat- and mass-transfer code

    Energy Technology Data Exchange (ETDEWEB)

    Zyvoloski, G.A.; Robinson, B.A.; Dash, Z.V.; Trease, L.L.

    1997-07-01

    The use of this code is applicable to natural-state studies of geothermal systems and groundwater flow. A primary use of the FEHM application will be to assist in the understanding of flow fields and mass transport in the saturated and unsaturated zones below the proposed Yucca Mountain nuclear waste repository in Nevada. The equations of heat and mass transfer for multiphase flow in porous and permeable media are solved in the FEHM application by using the finite-element method. The permeability and porosity of the medium are allowed to depend on pressure and temperature. The code also has provisions for movable air and water phases and noncoupled tracers; that is, tracer solutions that do not affect the heat- and mass-transfer solutions. The tracers can be passive or reactive. The code can simulate two-dimensional, two-dimensional radial, or three-dimensional geometries. In fact, FEHM is capable of describing flow that is dominated in many areas by fracture and fault flow, including the inherently three-dimensional flow that results from permeation to and from faults and fractures. The code can handle coupled heat and mass-transfer effects, such as boiling, dryout, and condensation that can occur in the near-field region surrounding the potential repository and the natural convection that occurs through Yucca Mountain due to seasonal temperature changes. This report outlines the uses and capabilities of the FEHM application, initialization of code variables, restart procedures, and error processing. The report describes all the data files, the input data, including individual input records or parameters, and the various output files. The system interface is described, including the software environment and installation instructions.

  1. User's manual for ASTERIX-2: A two-dimensional modular code system for the steady state and xenon transient analysis of a pebble bed high temperature reactor

    International Nuclear Information System (INIS)

    Wu, T.; Cowan, C.L.; Lauer, A.; Schwiegk, H.J.

    1982-03-01

    The ASTERIX modular code package was developed at KFA Laboratory-Juelich for the steady state and xenon transient analysis of a pebble bed high temperature reactor. The code package was implemented on the Stanford Linear Accelerator Center Computer in August, 1980, and a user's manual for the current version of the code, identified as ASTERIX-2, was prepared as a cooperative effort by KFA Laboratory and GE-ARSD. The material in the manual includes the requirements for accessing the program, a description of the major subroutines, a listing of the input options, and a listing of the input data for a sample problem. The material is provided in sufficient detail for the user to carry out a wide range of analysis from steady state operations to the xenon induced power transients in which the local xenon, temperature, buckling and control feedback effects have been incorporated in the problem solution. (orig.)

  2. Four-D propagation code for high-energy laser beams: a user's manual

    Energy Technology Data Exchange (ETDEWEB)

    Morris, J.R.

    1976-08-05

    This manual describes the use and structure of the June 30, 1976 version of the Four-D propagation code for high energy laser beams. It provides selected sample output from a typical run and from several debug runs. The Four-D code now includes the important noncoplanar scenario feature. Many problems that required excessive computer time can now be meaningfully simulated as steady-state noncoplanar problems with short run times.

  3. ETAP user's manual

    International Nuclear Information System (INIS)

    Watanabe, Norio; Higuchi, Suminori.

    1990-11-01

    The event tree analysis technique has been used in Probabilistic Safety Assessment for LWRs to delineate various accident scenarios leading to core melt or containment failure and to evaluate their frequencies. This technique often requires manual preparation of event trees with iterative process and time-consuming work in data handling. For the purpose of reducing manual efforts in event tree analysis, we developed a new software package named ETAP (Event Tree Analysis Supporting Program) for event tree analysis. ETAP is an interactive PC-based program which has the ability to construct, update, document, and quantify event trees. Because of its fast running capability to quantify event trees, use of the EATP program can make it easy to perform the sensitivity studies on a variety of system/containment performance issues. This report provides a user's manual for ETAP, which describes the structure, installation, and use of EATP. This software runs on NEC/PC-9800 or compatible PCs that have a 640 KB memory and MS-DOS 2.11 or higher. (author)

  4. ARDS User Manual

    Science.gov (United States)

    Fleming, David P.

    2001-01-01

    Personal computers (PCs) are now used extensively for engineering analysis. their capability exceeds that of mainframe computers of only a few years ago. Programs originally written for mainframes have been ported to PCs to make their use easier. One of these programs is ARDS (Analysis of Rotor Dynamic Systems) which was developed at Arizona State University (ASU) by Nelson et al. to quickly and accurately analyze rotor steady state and transient response using the method of component mode synthesis. The original ARDS program was ported to the PC in 1995. Several extensions were made at ASU to increase the capability of mainframe ARDS. These extensions have also been incorporated into the PC version of ARDS. Each mainframe extension had its own user manual generally covering only that extension. Thus to exploit the full capability of ARDS required a large set of user manuals. Moreover, necessary changes and enhancements for PC ARDS were undocumented. The present document is intended to remedy those problems by combining all pertinent information needed for the use of PC ARDS into one volume.

  5. 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.

  6. 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.

  7. MARS CODE MANUAL VOLUME III - Programmer's Manual

    International Nuclear Information System (INIS)

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

    2010-02-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This programmer's manual provides a complete list of overall information of code structure and input/output function of MARS. In addition, brief descriptions for each subroutine and major variables used in MARS are also included in this report, so that this report would be very useful for the code maintenance. The overall structure of the manual is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

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

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C; Aragones, J M

    1981-07-01

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

  9. NJOY nuclear data processing system: user's manual

    International Nuclear Information System (INIS)

    MacFarlane, R.E.; Barrett, R.J.; Muir, D.W.; Boicourt, R.M.

    1978-12-01

    The NJOY nuclear data processing system is a comprehensive computer code package for producing cross sections for neutron and photon transport calculations from ENDF/B-IV and -V evaluated nuclear data. This user's manual provides a concise description of the code, input instructions, sample problems, and installation instructions. 1 figure, 3 tables

  10. THREETRAN (hex, z) users' manual

    International Nuclear Information System (INIS)

    Walters, W.F.; O'Dell, R.D.; Brinkley, F.W. Jr.

    1979-10-01

    THREETRAN (hex,z) is a three-dimensional, multigroup, discrete-ordinates neutral-particle transport code for use in solving problems in hexagonal, z geometries. An efficient and flexible data management strategy is incorporated and uses three hierarchies of storage: fast core (or small core memory), extended core (or large core memory), and random access disk. Both isotropic (P 0 ) and linearly anisotropic (P 1 ) scattering can be treated. This manual is intended to be a guide for the users of THREETRAN (hex,z) in setting up problem input and in interpreting the output. It is not intended to provide a description of code theory or architecture. 5 figures, 4 tables

  11. Task 7: ADPAC User's Manual

    Science.gov (United States)

    Hall, E. J.; Topp, D. A.; Delaney, R. A.

    1996-01-01

    The overall objective of this study was to develop a 3-D numerical analysis for compressor casing treatment flowfields. The current version of the computer code resulting from this study is referred to as ADPAC (Advanced Ducted Propfan Analysis Codes-Version 7). This report is intended to serve as a computer program user's manual for the ADPAC code developed under Tasks 6 and 7 of the NASA Contract. The ADPAC program is based on a flexible multiple- block grid discretization scheme permitting coupled 2-D/3-D mesh block solutions with application to a wide variety of geometries. Aerodynamic calculations are based on a four-stage Runge-Kutta time-marching finite volume solution technique with added numerical dissipation. Steady flow predictions are accelerated by a multigrid procedure. An iterative implicit algorithm is available for rapid time-dependent flow calculations, and an advanced two equation turbulence model is incorporated to predict complex turbulent flows. The consolidated code generated during this study is capable of executing in either a serial or parallel computing mode from a single source code. Numerous examples are given in the form of test cases to demonstrate the utility of this approach for predicting the aerodynamics of modem turbomachinery configurations.

  12. COBRA-SFS [Spent Fuel Storage]: A thermal-hydraulic analysis computer code: Volume 2, User's manual

    International Nuclear Information System (INIS)

    Rector, D.R.; Cuta, J.M.; Lombardo, N.J.; Michener, T.E.; Wheeler, C.L.

    1986-11-01

    COBRA-SFS (Spent Fuel Storage) is a general thermal-hydraulic analysis computer code used to predict temperatures and velocities in a wide variety of systems. The code was refined and specialized for spent fuel storage system analyses for the US Department of Energy's Commercial Spent Fuel Management Program. The finite-volume equations governing mass, momentum, and energy conservation are written for an incompressible, single-phase fluid. The flow equations model a wide range of conditions including natural circulation. The energy equations include the effects of solid and fluid conduction, natural convection, and thermal radiation. The COBRA-SFS code is structured to perform both steady-state and transient calculations; however, the transient capability has not yet been validated. This volume contains the input instructions for COBRA-SFS and an auxiliary radiation exchange factor code, RADX-1. It is intended to aid the user in becoming familiar with the capabilities and modeling conventions of the code

  13. TMAP7 User Manual

    International Nuclear Information System (INIS)

    Longhurst, Glen R.

    2008-01-01

    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

  14. AMBER User's Manual

    International Nuclear Information System (INIS)

    Vay, J.L.; Fawley, W.

    2000-01-01

    AMBER is a Particle-In-Cell (PIC) code which models the evolution of a representative slice of a relativistic electron beam in a linear accelerator. The beam is modeled as a steady flow and therefore no electromagnetic waves: all the fields (external and self-fields) are electrostatic and magnetostatic fields (for a complete description, see chapter 5). The possible elements describing the accelerator lattice are solenoids, accelerating gaps, pipes and apertures. Several kinds of beam distribution can be loaded: KV, gaussian, semi-gaussian, etc. Alternatively, the user can reconstruct (or load) a distribution from the output of another codefile, for example, an interface generating the beam distribution from output produced from EGUN or LSP codes is available as an option. This documentation first describes in detail the input files needed to run AMBER and the procedure to start the executable. The possible data files and graphical output are explained in the two following chapters. The last chapter describes the physics model and numerical techniques used. An example of input files and the result obtained with these inputs are also given in the Appendix

  15. GASFLOW: A Computational Fluid Dynamics Code for Gases, Aerosols, and Combustion, Volume 2: User's Manual

    Energy Technology Data Exchange (ETDEWEB)

    Nichols, B. D.; Mueller, C.; Necker, G. A.; Travis, J. R.; Spore, J. W.; Lam, K. L.; Royl, P.; Wilson, T. L.

    1998-10-01

    Los Alamos National Laboratory (LANL) and Forschungszentrum Karlsruhe (FzK) are developing GASFLOW, a three-dimensional (3D) fluid dynamics field code as a best-estimate tool to characterize local phenomena within a flow field. Examples of 3D phenomena include circulation patterns; flow stratification; hydrogen distribution mixing and stratification; combustion and flame propagation; effects of noncondensable gas distribution on local condensation and evaporation; and aerosol entrainment, transport, and deposition. An analysis with GASFLOW will result in a prediction of the gas composition and discrete particle distribution in space and time throughout the facility and the resulting pressure and temperature loadings on the walls and internal structures with or without combustion. A major application of GASFLOW is for predicting the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containment and other facilities. It has been applied to situations involving transporting and distributing combustible gas mixtures. It has been used to study gas dynamic behavior in low-speed, buoyancy-driven flows, as well as sonic flows or diffusion dominated flows; and during chemically reacting flows, including deflagrations. The effects of controlling such mixtures by safety systems can be analyzed. The code version described in this manual is designated GASFLOW 2.1, which combines previous versions of the United States Nuclear Regulatory Commission code HMS (for Hydrogen Mixing Studies) and the Department of Energy and FzK versions of GASFLOW. The code was written in standard Fortran 90. This manual comprises three volumes. Volume I describes the governing physical equations and computational model. Volume II describes how to use the code to set up a model geometry, specify gas species and material properties, define initial and boundary conditions, and specify different outputs, especially graphical displays. Sample problems are included. Volume III

  16. Fission product data for thermal reactors. Final report. Part 2. Users manual for EPRI-CINDER code and data

    International Nuclear Information System (INIS)

    England, T.R.; Wilson, W.B.; Stamatelatos, M.G.

    1976-12-01

    A four-group fission-product absorption chain library using ENDF/B-IV decay data and cross sections processed with a typical light water reactor spectrum for a modified version of the original CINDER code has been developed as described in Part 1. CINDER is a general point-depletion and fission product code based on an analytical solution of the equations describing nuclides coupled in any linear sequence of radioactive decays and neutron absorptions. The basic code has been in wide use for a number of years. Previously, the user was required to specify all physical data. This report describes the chain library in detail and a modified version of the basic CINDER code (EPRI-CINDER) that is still compatible with existing libraries

  17. User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code

    International Nuclear Information System (INIS)

    Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B.

    2003-01-01

    This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k eff (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

  18. FAMIAS User Manual

    Science.gov (United States)

    Zima, Wolfgang

    2008-10-01

    The excitation of pulsation modes in Beta Cephei and Slowly Pulsating B stars is known to be very sensitive to opacity changes in the stellar interior where T ~ 2 x 10E5 K. In this region differences in opacity up to ~ 50% can be induced by the choice between OPAL and OP opacity tables, and between two different metal mixtures (Grevesse & Noels 1993 and Asplund et al. 2005). We have extended the non-adiabatic computations presented in Miglio et al. (2007) towards models of higher mass and pulsation modes of degree l = 3, and we present here the instability domains in the HR- and log P-log Teff diagrams resulting from different choices of opacity tables, and for three different metallicities. FAMIAS (Frequency Analysis and Mode Identification for AsteroSeismology) is a collection of state-of-the-art software tools for the analysis of photometric and spectroscopic time series data. It is one of the deliverables of the Work Package NA5: Asteroseismology of the European Coordination Action in Helio-and Asteroseismology (HELAS). Two main sets of tools are incorporated in FAMIAS. The first set allows to search for periodicities in the data using Fourier and non-linear least-squares fitting algorithms. The other set allows to carry out a mode identification for the detected pulsation frequencies to determine their pulsational quantum numbers, the harmonic degree, m. The types of stars to which famias is applicable are main-sequence pulsators hotter than the Sun. This includes the Gamma Dor stars, Delta Sct stars, the slowly pulsating B stars and the Beta Cep stars - basically all pulsating main-sequence stars, for which empirical mode identification is required to successfully carry out asteroseismology. This user manual describes how to use the different features of FAMIAS and provides two tutorials that demonstrate the usage of FAMIAS for spectroscopic and photometric mode identification.

  19. How 2 HAWC2, the user's manual

    DEFF Research Database (Denmark)

    Larsen, Torben J.; Hansen, Anders Melchior

    The report contains the user's manual for the aeroleastic code HAWC2. The code is intended for calculating wind turbine response in time domain and has a structural formulation based on multi-body dynamics. The aerodynamic part of the code is based on the blade element momentum theory, but extended...... from the classic approach to handle dynamic inflow, dynamic stall, skew inflow, shear effects on the induction and effects from large deflections. It has been developed within the years 2003-2006 at the aeroelastic design research programme at Risoe, National laboratory Denmark. This manual is updated...

  20. TESHOM: a users manual

    International Nuclear Information System (INIS)

    Griffiths, J.

    1978-10-01

    The TESHOM code and its associated data format is widely used to provide input data for lattice codes. Through TESHOM one data deck can be used to provide input to the LATREP, WIMS and POWDERPUFS lattice codes. The data format is described in detail, some of the principles behind the design of the code and its limitations are noted. (author)

  1. User's Manual for the FEHM Application-A Finite-Element Heat- and Mass-Transfer Code

    Energy Technology Data Exchange (ETDEWEB)

    George A. Zyvoloski; Bruce A. Robinson; Zora V. Dash; Lynn L. Trease

    1997-07-07

    This document is a manual for the use of the FEHM application, a finite-element heat- and mass-transfer computer code that can simulate nonisothermal multiphase multicomponent flow in porous media. The use of this code is applicable to natural-state studies of geothermal systems and groundwater flow. A primary use of the FEHM application will be to assist in the understanding of flow fields and mass transport in the saturated and unsaturated zones below the proposed Yucca Mountain nuclear waste repository in Nevada. The equations of heat and mass transfer for multiphase flow in porous and permeable media are solved in the FEHM application by using the finite-element method. The permeability and porosity of the medium are allowed to depend on pressure and temperature. The code also has provisions for movable air and water phases and noncoupled tracers; that is, tracer solutions that do not affect the heat- and mass-transfer solutions. The tracers can be passive or reactive. The code can simulate two-dimensional, two-dimensional radial, or three-dimensional geometries. In fact, FEHM is capable of describing flow that is dominated in many areas by fracture and fault flow, including the inherently three-dimensional flow that results from permeation to and from faults and fractures. The code can handle coupled heat and mass-transfer effects, such as boiling, dryout, and condensation that can occur in the near-field region surrounding the potential repository and the natural convection that occurs through Yucca Mountain due to seasonal temperature changes. The code is also capable of incorporating the various adsorption mechanisms, ranging from simple linear relations to nonlinear isotherms, needed to describe the very complex transport processes at Yucca Mountain. This report outlines the uses and capabilities of the FEHM application, initialization of code variables, restart procedures, and error processing. The report describes all the data files, the input data

  2. TASS code topical report. V.1 TASS code technical manual

    International Nuclear Information System (INIS)

    Sim, Suk K.; Chang, W. P.; Kim, K. D.; Kim, H. C.; Yoon, H. Y.

    1997-02-01

    TASS 1.0 code has been developed at KAERI for the initial and reload non-LOCA safety analysis for the operating PWRs as well as the PWRs under construction in Korea. TASS code will replace various vendor's non-LOCA safety analysis codes currently used for the Westinghouse and ABB-CE type PWRs in Korea. This can be achieved through TASS code input modifications specific to each reactor type. The TASS code can be run interactively through the keyboard operation. A simimodular configuration used in developing the TASS code enables the user easily implement new models. TASS code has been programmed using FORTRAN77 which makes it easy to install and port for different computer environments. The TASS code can be utilized for the steady state simulation as well as the non-LOCA transient simulations such as power excursions, reactor coolant pump trips, load rejections, loss of feedwater, steam line breaks, steam generator tube ruptures, rod withdrawal and drop, and anticipated transients without scram (ATWS). The malfunctions of the control systems, components, operator actions and the transients caused by the malfunctions can be easily simulated using the TASS code. This technical report describes the TASS 1.0 code models including reactor thermal hydraulic, reactor core and control models. This TASS code models including reactor thermal hydraulic, reactor core and control models. This TASS code technical manual has been prepared as a part of the TASS code manual which includes TASS code user's manual and TASS code validation report, and will be submitted to the regulatory body as a TASS code topical report for a licensing non-LOCA safety analysis for the Westinghouse and ABB-CE type PWRs operating and under construction in Korea. (author). 42 refs., 29 tabs., 32 figs

  3. PROTEUS-SN User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Shemon, Emily R. [Argonne National Lab. (ANL), Argonne, IL (United States); Smith, Micheal A. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, Changho [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-16

    is a part of the SHARP multi-physics suite for coupled multi-physics analysis of nuclear reactors. This user manual describes how to set up a neutron transport simulation with the PROTEUS-SN code. A companion methodology manual describes the theory and algorithms within PROTEUS-SN.

  4. COSIS User's Manual

    International Nuclear Information System (INIS)

    Cho, J. Y.; Lee, K. B.; Koo, B. S.; Lee, W. K.; Lee, C. C.; Zee, S. Q.

    2006-02-01

    COSIS (COre State Indication System) which implemented in the SMART research reactor plays a role to supply the core state parameters or graphs for the operator to recognize the core state effectively. The followings are the main functions of COSIS. Validity Check for the Process Signals and Determination of the COSIS Inputs (SIGVAL), Coolant Flow Rate Calculation (FLOW), Core Thermal Power Calculation (COREPOW), In-core 3-Dimensional Power Distribution Calculation and Peaking Parameters Generation (POWER3D), Azimuthal Tilt Calculation (AZITILT). This report describes the structures of the I/O files that are essential for the users to run COSIS. COSIS handles the following 4 input files. DATABASE: The base input file, COSIS.INP: The signal input file, CCS.DAT: File required for the in-core detector signal processing and the 3-D power distribution calculation, TPFH2O: Steam table for the water properties The DATABASE file contains the base information for a nuclear power plant and is read at the first COSIS calculation. The COSIS.INP file contains the process input and detector signals, and is read by COSIS at every second. CCS.DAT file, that is produced by the COSISMAS code, contains the information for the in-core detector signal processing and the 3-D power distribution calculation. TPFH2O file is a steam table and is written in binary format. COSIS produces the following 4 output files. DATABASE.OUT: The output file for the DATABASE input file, COSIS.OUT: The normal output file produced after the COSIS calculation, COSIS.SUM: File for the operator to recognize the core state effectively, MAS S IG: File to run the COSISMAS code The DATABASE.OUT file is produced right after finishing DATABASE processing. The COSIS.OUT file is produced after finishing the input signal processing and the main COSIS calculation. The COSIS.SUM file is the summary file of the COSIS results for the operator to recognize the core state effectively. The MAS S IG file is the COSISMAS input

  5. TARTNP user's manual

    International Nuclear Information System (INIS)

    Kimlinger, J.; Plechaty, E.

    1976-01-01

    The operating instructions and input definitions are given for the TARTNP Monte Carlo neutron/photon transport code. It also describes the limitations of the code and the pitfalls that may be encountered in using it

  6. Tokamak simulation code manual

    International Nuclear Information System (INIS)

    Chung, Moon Kyoo; Oh, Byung Hoon; Hong, Bong Keun; Lee, Kwang Won

    1995-01-01

    The method to use TSC (Tokamak Simulation Code) developed by Princeton plasma physics laboratory is illustrated. In KT-2 tokamak, time dependent simulation of axisymmetric toroidal plasma and vertical stability have to be taken into account in design phase using TSC. In this report physical modelling of TSC are described and examples of application in JAERI and SERI are illustrated, which will be useful when TSC is installed KAERI computer system. (Author) 15 refs., 6 figs., 3 tabs

  7. SES2D user's manual

    International Nuclear Information System (INIS)

    Johnson, J.D.; Lyon, S.P.

    1982-04-01

    SES2D is an interactive graphics code designed to generate plots of equation of state data from the Los Alamos National Laboratory Group T-4 computer libraries. This manual discusses the capabilities of the code. It describes the prompts and commands and illustrates their use with a sample run

  8. SAFIRE user's manual

    International Nuclear Information System (INIS)

    Glasgow, B.B.; Meier, W.R.

    1987-01-01

    Analytical models for scaling the cost and performance of inertial confinement fusion (ICF) electric-power plants have been developed and incorporated into the SAFIRE code. SAFIRE denotes systems analysis for ICF reactor economics. This volume, Volume 2, describes the subroutines, COMMON blocks, file handling, and input and output variables of the code

  9. A probabilistic assessment code system for derivation of clearance levels of radioactive materials. PASCLR user's manual

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Tomoyuki [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst; Takeda, Seiji; Kimura, Hideo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-01-01

    It is indicated that some types of radioactive material generating from the development and utilization of nuclear energy do not need to be subject regulatory control because they can only give rise to trivial radiation hazards. The process to remove such materials from regulatory control is called as 'clearance'. The corresponding levels of the concentration of radionuclides are called as 'clearance levels'. In the Nuclear Safety Commission's discussion, the deterministic approach was applied to derive the clearance levels, which are the concentrations of radionuclides in a cleared material equivalent to an individual dose criterion. Basically, realistic parameter values were selected for it. If the realistic values could not be defined, reasonably conservative values were selected. Additionally, the stochastic approaches were performed to validate the results which were obtained by the deterministic calculations. We have developed a computer code system PASCLR (Probabilistic Assessment code System for derivation of Clearance Levels of Radioactive materials) by using the Monte Carlo technique for carrying out the stochastic calculations. This report describes the structure and user information for execution of PASCLR code. (author)

  10. AVERT User Manual

    Science.gov (United States)

    AVERT is a flexible modeling framework with a simple user interface designed specifically to meet the needs of state air quality planners and other interested stakeholders. Use this guide to get started.

  11. CONHOR. Code system for determination of power distribution and burnup for the HOR reactor. Version 1.0.. User's manual

    International Nuclear Information System (INIS)

    Serov, I.V.; Hoogenboom, J.E.

    1993-07-01

    The main calculational tool is the CITATION code. CITATION is used for both static and burnup calculations. The pointwise flux density and power distributions obtained from these calculations are used to obtain the values of the desired quantities at the beginning of a burnup cycle. To obtain the most trustful values of the desired quantities CONHOR employs experimental information together with the CITATION calculated flux distributions. Axially averaged foil activation rates are obtained based on both CITATION pointwise flux density distributions and measured foil activity counts. These two sets of activation rates are called the distributions of auxiliary quantities and are compared with each other in order to pick up the corrections to the U-235 number densities in fuel containing elements. The methodical corrections to the calculational auxiliary quantities are obtained on this basis as well. They are used to obtain the methodical corrections to the desired quantities. The corrected desired quantities are the recommended ones. The correction procedure requires the knowledge of the sensitivity coefficients of the average foil activation rates with respect to the U-235 number densities (through the text of this manual U-235 is denoted also and especially in the input-output description sections as a BUrning-COrrected material, or 'BuCo' material). These sensitivity coefficients are calculated by the CONHOR SENS module. CITATION is employed to perform the calculations with perturbed values of U-235 number densities. Burnup calculations can be performed being based on either corrected or uncorrected U-235 number densities. Through the text of this manual XXXX means a 4-symbol identification of the burnup cycle to be studied. XX-1 and XX+1 mean correspondingly the previous and the following cycles. (orig./HP)

  12. User's manual of MANYCASK code for calculation of spatial distributions of radiation dose rates in a system composed of many spent-fuel-shipping casks

    International Nuclear Information System (INIS)

    Yamakoshi, Hisao

    1986-01-01

    A calculation code MANYCASK is designed for evaluation of spatial distributions of radiation dose rates in ships loaded with a lot of spent fuel shipping casks. Principle of the calculation method adopted in this code is different from that of ordinary codes, and is advantageous for calculating highly reliable dose rate distributions with a very short calculation time. Basic concept of the principle has been described in other reports in detail. A brief description of the principle will be included in the present report along with a technique named Shadow Technique in this report, in addition to format descriptions of output data as well as input data. Results of sample calculations are compared with measured results in figures so as to show how the calculation method adopted is valid. For the purpose of making this code popular among many people, the author writes the user's manual in the present report in Japanese for domestic users, and in English in another report for people in abroad. (author)

  13. MAP user's manual copyright

    International Nuclear Information System (INIS)

    Pillsbury, R.D. Jr.

    1991-12-01

    The program MITMAP represents a set of general purpose, two- dimensional, finite element programs for the calculation of magnetic fields. It consists of the program MAP and MAP2DJ. The two programs are used to solve different electromagnetic problems, but they have a common set of subrountines for pre- and postprocessing. Originally separate programs, they have been combined to make modification easier. The manuals, however, will remain separate. The program MAP is described in this manual. MAP is applicable to the class of problems with two-dimensional-planar or axisymmetric - geometries, in which the current density and the magnetic vector potential have only a single nonvanishing component. The single component is associated with the direction that is perpendicular to the plane of the problem and is invariant with respect to that direction. Maxwell's equations can be reduced to a solver diffusion equation in terms of the single, nonvanishing component of the magnetic vector potential for planar problems and to a single component of a vector potential for planar problems and to a single component of a vector diffusion equation for axisymmetric problems. The magnetic permeability appears in the governing equation. The permeability may be a function of the magnetic flux density. In addition, any electrically conducting material present will have eddy currents induced by a time varying magnetic field. These eddy currents must be included in the solution process. This manual provides a description of the structure of the input data and output for the program. There are several example problems presented that illustrate the major program features. Appendices are included that contain a derivation of the governing equations and the application of the finite element method to the solution of the equations

  14. APUD - User's manual

    International Nuclear Information System (INIS)

    Vamanu, D.

    1989-04-01

    APUD is a computed code designed for rough, expeditious assessments of the dispersal of airborne radioactivity discharged, either normally or accidentally, from nuclear facilities. In particular codes such as APUD may aptly complement the tool-kit required in the adoption of informed contingency plans in case of abnormal nuclear occurences that are accompanied by atmospheric releases of radioactivity. Apart from such extreme circumstances, APUD may profitable be used a a simulation/drill facility. Given adequate inputs, APUD may also work on any sort of industrial atmospheric emissions.(author)

  15. TOTAL user manual

    Science.gov (United States)

    Johnson, Sally C.; Boerschlein, David P.

    1994-01-01

    Semi-Markov models can be used to analyze the reliability of virtually any fault-tolerant system. However, the process of delineating all of the states and transitions in the model of a complex system can be devastatingly tedious and error-prone. Even with tools such as the Abstract Semi-Markov Specification Interface to the SURE Tool (ASSIST), the user must describe a system by specifying the rules governing the behavior of the system in order to generate the model. With the Table Oriented Translator to the ASSIST Language (TOTAL), the user can specify the components of a typical system and their attributes in the form of a table. The conditions that lead to system failure are also listed in a tabular form. The user can also abstractly specify dependencies with causes and effects. The level of information required is appropriate for system designers with little or no background in the details of reliability calculations. A menu-driven interface guides the user through the system description process, and the program updates the tables as new information is entered. The TOTAL program automatically generates an ASSIST input description to match the system description.

  16. RAVEN User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cogliati, Joshua Joseph [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinoshita, Robert Arthur [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sen, Ramazan Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Congjian [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

    RAVEN is a generic software framework to perform parametric and probabilistic analysis based on the response of complex system codes. The initial development was aimed to provide dynamic risk analysis capabilities to the Thermo-Hydraulic code RELAP-7, currently under development at the Idaho National Laboratory (INL). Although the initial goal has been fully accomplished, RAVEN is now a multi-purpose probabilistic and uncertainty quantification platform, capable to agnostically communicate with any system code. This agnosticism includes providing Application Programming Interfaces (APIs). These APIs are used to allow RAVEN to interact with any code as long as all the parameters that need to be perturbed are accessible by input files or via python interfaces. RAVEN is capable of investigating the system response, and investigating the input space using Monte Carlo, Grid, or Latin Hyper Cube sampling schemes, but its strength is focused toward system feature discovery, such as limit surfaces, separating regions of the input space leading to system failure, using dynamic supervised learning techniques. The development of RAVEN started in 2012, when, within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the need to provide a modern risk evaluation framework became stronger. RAVEN principal assignment is to provide the necessary software and algorithms in order to employ the concept developed by the Risk Informed Safety Margin Characterization (RISMC) program. RISMC is one of the pathways defined within the Light Water Reactor Sustainability (LWRS) program. In the RISMC approach, the goal is not just the individuation of the frequency of an event potentially leading to a system failure, but the closeness (or not) to key safety-related events. Hence, the approach is interested in identifying and increasing the safety margins related to those events. A safety margin is a numerical value quantifying the probability that a safety metric (e.g. for an

  17. RAVEN User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cogliati, Joshua Joseph [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinoshita, Robert Arthur [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sen, Ramazan Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    RAVEN is a generic software framework to perform parametric and probabilistic analysis based on the response of complex system codes. The initial development was aimed to provide dynamic risk analysis capabilities to the Thermo-Hydraulic code RELAP-7, currently under development at the Idaho National Laboratory (INL). Although the initial goal has been fully accomplished, RAVEN is now a multi-purpose probabilistic and uncertainty quantification platform, capable to agnostically communicate with any system code. This agnosticism includes providing Application Programming Interfaces (APIs). These APIs are used to allow RAVEN to interact with any code as long as all the parameters that need to be perturbed are accessible by inputs files or via python interfaces. RAVEN is capable of investigating the system response, and investigating the input space using Monte Carlo, Grid, or Latin Hyper Cube sampling schemes, but its strength is focused toward system feature discovery, such as limit surfaces, separating regions of the input space leading to system failure, using dynamic supervised learning techniques. The development of RAVEN has started in 2012, when, within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the need to provide a modern risk evaluation framework became stronger. RAVEN principal assignment is to provide the necessary software and algorithms in order to employ the concept developed by the Risk Informed Safety Margin Characterization (RISMC) program. RISMC is one of the pathways defined within the Light Water Reactor Sustainability (LWRS) program. In the RISMC approach, the goal is not just the individuation of the frequency of an event potentially leading to a system failure, but the closeness (or not) to key safety-related events. Hence, the approach is interested in identifying and increasing the safety margins related to those events. A safety margin is a numerical value quantifying the probability that a safety metric (e.g. for

  18. RAVEN User Manual

    International Nuclear Information System (INIS)

    Mandelli, Diego; Rabiti, Cristian; Cogliati, Joshua Joseph; Kinoshita, Robert Arthur; Alfonsi, Andrea; Sen, Ramazan Sonat

    2015-01-01

    RAVEN is a generic software framework to perform parametric and probabilistic analysis based on the response of complex system codes. The initial development was aimed to provide dynamic risk analysis capabilities to the Thermo-Hydraulic code RELAP-7, currently under development at the Idaho National Laboratory (INL). Although the initial goal has been fully accomplished, RAVEN is now a multi-purpose probabilistic and uncertainty quantification platform, capable to agnostically communicate with any system code. This agnosticism includes providing Application Programming Interfaces (APIs). These APIs are used to allow RAVEN to interact with any code as long as all the parameters that need to be perturbed are accessible by inputs files or via python interfaces. RAVEN is capable of investigating the system response, and investigating the input space using Monte Carlo, Grid, or Latin Hyper Cube sampling schemes, but its strength is focused toward system feature discovery, such as limit surfaces, separating regions of the input space leading to system failure, using dynamic supervised learning techniques. The development of RAVEN has started in 2012, when, within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the need to provide a modern risk evaluation framework became stronger. RAVEN principal assignment is to provide the necessary software and algorithms in order to employ the concept developed by the Risk Informed Safety Margin Characterization (RISMC) program. RISMC is one of the pathways defined within the Light Water Reactor Sustainability (LWRS) program. In the RISMC approach, the goal is not just the individuation of the frequency of an event potentially leading to a system failure, but the closeness (or not) to key safety-related events. Hence, the approach is interested in identifying and increasing the safety margins related to those events. A safety margin is a numerical value quantifying the probability that a safety metric (e.g. for

  19. RAVEN User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cogliati, Joshua Joseph [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinoshita, Robert Arthur [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sen, Ramazan Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Congjian [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-03-01

    RAVEN is a generic software framework to perform parametric and probabilistic analy- sis based on the response of complex system codes. The initial development was aimed to provide dynamic risk analysis capabilities to the Thermo-Hydraulic code RELAP-7, currently under development at the Idaho National Laboratory (INL). Although the initial goal has been fully accomplished, RAVEN is now a multi-purpose probabilistic and uncer- tainty quantification platform, capable to agnostically communicate with any system code. This agnosticism includes providing Application Programming Interfaces (APIs). These APIs are used to allow RAVEN to interact with any code as long as all the parameters that need to be perturbed are accessible by inputs files or via python interfaces. RAVEN is capable of investigating the system response, and investigating the input space using Monte Carlo, Grid, or Latin Hyper Cube sampling schemes, but its strength is focused to- ward system feature discovery, such as limit surfaces, separating regions of the input space leading to system failure, using dynamic supervised learning techniques. The development of RAVEN has started in 2012, when, within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the need to provide a modern risk evaluation framework became stronger. RAVEN principal assignment is to provide the necessary software and algorithms in order to employ the concept developed by the Risk Informed Safety Margin Characterization (RISMC) program. RISMC is one of the pathways defined within the Light Water Reactor Sustainability (LWRS) program. In the RISMC approach, the goal is not just the individuation of the frequency of an event potentially leading to a system failure, but the closeness (or not) to key safety-related events. Hence, the approach is in- terested in identifying and increasing the safety margins related to those events. A safety margin is a numerical value quantifying the probability that a safety metric (e

  20. ASSIST user manual

    Science.gov (United States)

    Johnson, Sally C.; Boerschlein, David P.

    1995-01-01

    Semi-Markov models can be used to analyze the reliability of virtually any fault-tolerant system. However, the process of delineating all the states and transitions in a complex system model can be devastatingly tedious and error prone. The Abstract Semi-Markov Specification Interface to the SURE Tool (ASSIST) computer program allows the user to describe the semi-Markov model in a high-level language. Instead of listing the individual model states, the user specifies the rules governing the behavior of the system, and these are used to generate the model automatically. A few statements in the abstract language can describe a very large, complex model. Because no assumptions are made about the system being modeled, ASSIST can be used to generate models describing the behavior of any system. The ASSIST program and its input language are described and illustrated by examples.

  1. ASTRA Plus User Manual

    OpenAIRE

    MATUZAS Vaidas; CONTINI SERGIO

    2012-01-01

    This report describes the user interface and the main commands to perform system dependability analysis by means of ASTRA Plus. This package implements the analysis methods developed at the Institute for the Protection and Security of the Citizen from mid-2008. ASTRA Plus is composed of the Fault Tree Analysis (FTA) module and of the Concurrent Importance and Sensitivity Analysis (CISA) module. The FTA module contains three different methods for solving a fault tree; all are based on the...

  2. AXAIRQ User's Manual

    International Nuclear Information System (INIS)

    Simpkins, A.A.

    1995-10-01

    AXAIRQ is the primary dose assessment code used at the Savannah River Site to predict downwind doses following a hypothetical atmospheric release of relatively short duration. Purpose is to perform calculations for safety-related documentation, and AXAIRQ strictly adheres to the guidance in US NRC Regulatory Guide 1.145 (USNRC 1982) entitled Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants. Doses are determined for the plume shine, ground shine, and inhalation pathways. Ingestion is not considered

  3. 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).

  4. CONCEPT-5 user's manual

    International Nuclear Information System (INIS)

    Hudson, C.R. II.

    1979-01-01

    The CONCEPT computer code package was developed to provide conceptual capital cost estimates for nuclear-fueled and fossil-fired power plants. Cost estimates can be made as a function of plant type, size, location, and date of initial operation. The output includes a detailed breakdown of the estimate into direct and indirect costs similar to the accounting system described in document NUS--531. Cost models are currently provided in CONCEPT 5 for single- and multiunit pressurized-water reactors, boiling-water reactors, and cost-fired plants with and without flue gas desulfurization equipment

  5. Dynamic link: user's manual

    International Nuclear Information System (INIS)

    Harada, Hiroo; Asai, Kiyoshi; Kihara, Kazuhisa.

    1981-09-01

    The purpose of dynamic link facility is to link a load module dynamically only when it is used in execution time. The facility is very useful for development, execution and maintenance of a large scale computer program which is too big to be saved as one load module in main memory, or it is poor economy to save it due to many unused subroutines depending on an input. It is also useful for standardization and common utilization of programs. Standard usage of dynamic link facility of FACOM M-200 computer system, a software tool which analyzes the effect of dynamic link facility and application of dynamic link to nuclear codes are described. (author)

  6. Slycat™ User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Crossno, Patricia J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gittinger, Jaxon [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hunt, Warren L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Letter, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Martin, Shawn [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sielicki, Milosz Aleksander [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2018-01-01

    Slycat™ is a web-based system for performing data analysis and visualization of potentially large quantities of remote, high-dimensional data. Slycat™ specializes in working with ensemble data. An ensemble is a group of related data sets, which typically consists of a set of simulation runs exploring the same problem space. An ensemble can be thought of as a set of samples within a multi-variate domain, where each sample is a vector whose value defines a point in high-dimensional space. To understand and describe the underlying problem being modeled in the simulations, ensemble analysis looks for shared behaviors and common features across the group of runs. Additionally, ensemble analysis tries to quantify differences found in any members that deviate from the rest of the group. The Slycat™ system integrates data management, scalable analysis, and visualization. Results are viewed remotely on a user’s desktop via commodity web clients using a multi-tiered hierarchy of computation and data storage, as shown in Figure 1. Our goal is to operate on data as close to the source as possible, thereby reducing time and storage costs associated with data movement. Consequently, we are working to develop parallel analysis capabilities that operate on High Performance Computing (HPC) platforms, to explore approaches for reducing data size, and to implement strategies for staging computation across the Slycat™ hierarchy. Within Slycat™, data and visual analysis are organized around projects, which are shared by a project team. Project members are explicitly added, each with a designated set of permissions. Although users sign-in to access Slycat™, individual accounts are not maintained. Instead, authentication is used to determine project access. Within projects, Slycat™ models capture analysis results and enable data exploration through various visual representations. Although for scientists each simulation run is a model of real-world phenomena given certain

  7. User's manual for DYNA2D: an explicit two-dimensional hydrodynamic finite-element code with interactive rezoning

    Energy Technology Data Exchange (ETDEWEB)

    Hallquist, J.O.

    1982-02-01

    This revised report provides an updated user's manual for DYNA2D, an explicit two-dimensional axisymmetric and plane strain finite element code for analyzing the large deformation dynamic and hydrodynamic response of inelastic solids. A contact-impact algorithm permits gaps and sliding along material interfaces. By a specialization of this algorithm, such interfaces can be rigidly tied to admit variable zoning without the need of transition regions. Spatial discretization is achieved by the use of 4-node solid elements, and the equations-of motion are integrated by the central difference method. An interactive rezoner eliminates the need to terminate the calculation when the mesh becomes too distorted. Rather, the mesh can be rezoned and the calculation continued. The command structure for the rezoner is described and illustrated by an example.

  8. BLOCKAGE 2.5 user's manual

    International Nuclear Information System (INIS)

    Rao, D.V.; Brideau, J.; Shaffer, C.; Souto, F.; Bernahl, W.

    1996-12-01

    The BLOCKAGE 2.5 code described in this User's Manual was developed by the US Nuclear Regulatory Commission (NRC) as a tool to evaluate licensee compliance with NRC Bulletin 96-03, ''Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling Water Reactors.'' As such, BLOCKAGE 2.5 provides a generalized framework into which a user can input plant-specific and insulation-specific data for performing analyses in accordance with Regulatory Guide 1.82, Rev. 2. This user's manual describes the capabilities of BLOCKAGE 2.5 along with a description of the graphics user's interface provided for data entry. Each input/output dialog is described in detail along with special considerations related to developing and executing BLOCKAGE. Also, several sample problems are provided such that user can easily modify them to suit a particular plant of interest. The models used in BLOCKAGE 2.5 and their validation are presented in the accompanying NUREG/CR-6371. The BLOCKAGE models were designed to be parametric in nature, allowing the user flexibility to examine the impact of several modeling assumptions and to conduct sensitivity analyses. As a result, BLOCKAGE 2.5 results are known to be very sensitive to the user provided input. It is therefore strongly recommended that users become thoroughly familiar with BLOCKAGE models and their limitations as described in NUREG/CR-6224

  9. CONHOR. Code system for determination of power distribution and burnup for the HOR reactor. Version 1.0.. User`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Serov, I V; Hoogenboom, J E

    1993-07-01

    The main calculational tool is the CITATION code. CITATION is used for both static and burnup calculations. The pointwise flux density and power distributions obtained from these calculations are used to obtain the values of the desired quantities at the beginning of a burnup cycle. To obtain the most trustful values of the desired quantities CONHOR employs experimental information together with the CITATION calculated flux distributions. Axially averaged foil activation rates are obtained based on both CITATION pointwise flux density distributions and measured foil activity counts. These two sets of activation rates are called the distributions of auxiliary quantities and are compared with each other in order to pick up the corrections to the U-235 number densities in fuel containing elements. The methodical corrections to the calculational auxiliary quantities are obtained on this basis as well. They are used to obtain the methodical corrections to the desired quantities. The corrected desired quantities are the recommended ones. The correction procedure requires the knowledge of the sensitivity coefficients of the average foil activation rates with respect to the U-235 number densities (through the text of this manual U-235 is denoted also and especially in the input-output description sections as a BUrning-COrrected material, or `BuCo` material). These sensitivity coefficients are calculated by the CONHOR SENS module. CITATION is employed to perform the calculations with perturbed values of U-235 number densities. Burnup calculations can be performed being based on either corrected or uncorrected U-235 number densities. Through the text of this manual XXXX means a 4-symbol identification of the burnup cycle to be studied. XX-1 and XX+1 mean correspondingly the previous and the following cycles. (orig./HP).

  10. CMGTooL user's manual

    Science.gov (United States)

    Xu, Jingping; Lightsom, Fran; Noble, Marlene A.; Denham, Charles

    2002-01-01

    of MATLAB). The GUI and some of the library routines call low-level NetCDF file I/O, variable and attribute functions. These NetCDF exclusive functions are supported by a MATLAB toolbox named NetCDF, created by Dr. Charles Denham . This toolbox has to be installed in order to use the CMGTooL GUI. The CMGTooL GUI calls several routines that were initially developed by others. The authors would like to acknowledge the following scientists for their ideas and codes: Dr. Rich Signell (USGS), Dr. Chris Sherwood (USGS), and Dr. Bob Beardsley (WHOI). Many special terms that carry special meanings in either MATLAB or the NetCDF Toolbox are used in this manual. Users are encouraged to read the documents of MATLAB and NetCDF for references.

  11. User's manual for the BNW-II optimization code for dry/wet-cooled power plants

    Energy Technology Data Exchange (ETDEWEB)

    Braun, D.J.; Bamberger, J.A.; Braun, D.J.; Faletti, D.W.; Wiles, L.E.

    1978-05-01

    This volume provides a listing of the BNW-II dry/wet ammonia heat rejection optimization code and is an appendix to Volume I which gives a narrative description of the code's algorithms as well as logic, input and output information.

  12. FDTD-ANT User Manual

    Science.gov (United States)

    Zimmerman, Martin L.

    1995-01-01

    This manual explains the theory and operation of the finite-difference time domain code FDTD-ANT developed by Analex Corporation at the NASA Lewis Research Center in Cleveland, Ohio. This code can be used for solving electromagnetic problems that are electrically small or medium (on the order of 1 to 50 cubic wavelengths). Calculated parameters include transmission line impedance, relative effective permittivity, antenna input impedance, and far-field patterns in both the time and frequency domains. The maximum problem size may be adjusted according to the computer used. This code has been run on the DEC VAX and 486 PC's and on workstations such as the Sun Sparc and the IBM RS/6000.

  13. Wien Automatic System Planning (WASP) Package. A computer code for power generating system expansion planning. Version WASP-IV. User's manual

    International Nuclear Information System (INIS)

    2001-01-01

    generated by each plant and the user specified characteristics of fuels used. Expanded dimensions for handling up to 90 types of plants and a larger number of configurations (up to 500 per year and 5000 for the study period). The present manual allows us to support the use of the WASP-IV version and to illustrate the capabilities of the model. This manual contains 13 chapters. Chapter 1 gives a summary description of WASP-IV Computer Code and its Modules and file system. Chapter 2 explains the hardware requirement and the installation of the package. The sequence of the execution of WASP-IV is also briefly introduced in this chapter. Chapters 3 to 9 explains, in detail, how to execute each of the module of WASP-IV package, the organisation of input files and output from the run of the model. Special attention was paid to the description of the linkage of modules. Chapter 10 specially guides the users on how to effectively search for an optimal solution. Chapter 11 describes the execution of sensitivity analyses that can be (recommend to be) performed with WASP-IV. To ease the debugging during the running of the software, Chapter 12 provides technical details of the new features incorporated in this version. Chapter 13 provides a list of error and warning messages produced for each module of WASP. The reader of this manual is assumed to have experience in the field of power generation expansion planning and to be familiar with all concepts related to such type of analysis, therefore these aspects are not treated in this manual. Additional information on power generation expansion planning can be found in the IAEA publication 'Expansion Planning for Electrical Generating Systems, A Guidebook', Technical Reports Series No. 241 (1984) or User's Manual of WASP-IV Plus, Computer Manual Series No. 8, (1995)

  14. Elemental ABAREX -- a user's manual

    International Nuclear Information System (INIS)

    Smith, A.B.

    1999-01-01

    ELEMENTAL ABAREX is an extended version of the spherical optical-statistical model code ABAREX, designed for the interpretation of neutron interactions with elemental targets consisting of up to ten isotopes. The contributions from each of the isotopes of the element are explicitly dealt with, and combined for comparison with the elemental observables. Calculations and statistical fitting of experimental data are considered. The code is written in FORTRAN-77 and arranged for use on the IBM-compatible personal computer (PC), but it should operate effectively on a number of other systems, particularly VAX/VMS and IBM work stations. Effort is taken to make the code user friendly. With this document a reasonably skilled individual should become fluent with the use of the code in a brief period of time

  15. HTGR Cost Model Users' Manual

    International Nuclear Information System (INIS)

    Gandrik, A.M.

    2012-01-01

    The High Temperature Gas-Cooler Reactor (HTGR) Cost Model was developed at the Idaho National Laboratory for the Next Generation Nuclear Plant Project. The HTGR Cost Model calculates an estimate of the capital costs, annual operating and maintenance costs, and decommissioning costs for a high-temperature gas-cooled reactor. The user can generate these costs for multiple reactor outlet temperatures; with and without power cycles, including either a Brayton or Rankine cycle; for the demonstration plant, first of a kind, or nth of a kind project phases; for a single or four-pack configuration; and for a reactor size of 350 or 600 MWt. This users manual contains the mathematical models and operating instructions for the HTGR Cost Model. Instructions, screenshots, and examples are provided to guide the user through the HTGR Cost Model. This model was design for users who are familiar with the HTGR design and Excel. Modification of the HTGR Cost Model should only be performed by users familiar with Excel and Visual Basic.

  16. ADPAC v1.0: User's Manual

    Science.gov (United States)

    Hall, Edward J.; Heidegger, Nathan J.; Delaney, Robert A.

    1999-01-01

    The overall objective of this study was to evaluate the effects of turbulence models in a 3-D numerical analysis on the wake prediction capability. The current version of the computer code resulting from this study is referred to as ADPAC v7 (Advanced Ducted Propfan Analysis Codes -Version 7). This report is intended to serve as a computer program user's manual for the ADPAC code used and modified under Task 15 of NASA Contract NAS3-27394. The ADPAC program is based on a flexible multiple-block and discretization scheme permitting coupled 2-D/3-D mesh block solutions with application to a wide variety of geometries. Aerodynamic calculations are based on a four-stage Runge-Kutta time-marching finite volume solution technique with added numerical dissipation. Steady flow predictions are accelerated by a multigrid procedure. Turbulence models now available in the ADPAC code are: a simple mixing-length model, the algebraic Baldwin-Lomax model with user defined coefficients, the one-equation Spalart-Allmaras model, and a two-equation k-R model. The consolidated ADPAC code is capable of executing in either a serial or parallel computing mode from a single source code.

  17. CalendF-2005: user manual

    International Nuclear Information System (INIS)

    Sublet, J.Ch.; Ribon, P.; Coste-Delclaux, M.

    2006-01-01

    CALENDF-2005 represents a Fortran-95 update of the 1994 code distribution with emphasize on programming quality and standard, physics and usage improvements. Devised to process multigroup cross-sections it relies on Gauss quadratures mathematical principle and strength. The followings processes can be handled by the code: moment probability tables and effective cross-sections calculation; regroups pointwise cross sections, probability tables and effective cross-sections; probability table condensation; probability table mix for several isotopes; probability table interpolation; effective cross section based probability table calculations; probability table calculations from effective cross-sections; cross-section comparison, complete energy pointwise cross-section processing and thickness dependant averaged transmission sample calculation, The CALENDF user manual, after having listed all principal code functions, describes sequentially each of them and gives comments on their associated output streams. Installation procedures, test cases and running time platforms comparisons are given in the appendix. (authors)

  18. LAURA Users Manual: 5.2-43231

    Science.gov (United States)

    Mazaheri, Alireza; Gnoffo, Peter A.; Johnston, Christopher O.; Kleb, Bil

    2009-01-01

    This users manual provides in-depth information concerning installation and execution of LAURA, version 5. LAURA is a structured, multi-block, computational aerothermodynamic simulation code. Version 5 represents a major refactoring of the original Fortran 77 LAURA code toward a modular structure afforded by Fortran 95. The refactoring improved usability and maintainability by eliminating the requirement for problem-dependent re-compilations, providing more intuitive distribution of functionality, and simplifying interfaces required for multiphysics coupling. As a result, LAURA now shares gas-physics modules, MPI modules, and other low-level modules with the FUN3D unstructured-grid code. In addition to internal refactoring, several new features and capabilities have been added, e.g., a GNU-standard installation process, parallel load balancing, automatic trajectory point sequencing, free-energy minimization, and coupled ablation and flowfield radiation.

  19. Laura Users Manual: 5.1-41601

    Science.gov (United States)

    Mazaheri, Alireza; Gnoffo, Peter A.; Johnston, Christopher O.; Kleb, Bil

    2009-01-01

    This users manual provides in-depth information concerning installation and execution of LAURA, version 5. LAURA is a structured, multi-block, computational aerothermodynamic simulation code. Version 5 represents a major refactoring of the original Fortran 77 LAURA code toward a modular structure afforded by Fortran 95. The refactoring improved usability and maintainability by eliminating the requirement for problem-dependent re-compilations, providing more intuitive distribution of functionality, and simplifying interfaces required for multiphysics coupling. As a result, LAURA now shares gas-physics modules, MPI modules, and other low-level modules with the FUN3D unstructured-grid code. In addition to internal refactoring, several new features and capabilities have been added, e.g., a GNU-standard installation process, parallel load balancing, automatic trajectory point sequencing, free-energy minimization, and coupled ablation and flowfield radiation.

  20. LAURA Users Manual: 5.3-48528

    Science.gov (United States)

    Mazaheri, Alireza; Gnoffo, Peter A.; Johnston, Chirstopher O.; Kleb, Bil

    2010-01-01

    This users manual provides in-depth information concerning installation and execution of LAURA, version 5. LAURA is a structured, multi-block, computational aerothermodynamic simulation code. Version 5 represents a major refactoring of the original Fortran 77 LAURA code toward a modular structure afforded by Fortran 95. The refactoring improved usability and maintainability by eliminating the requirement for problem-dependent re-compilations, providing more intuitive distribution of functionality, and simplifying interfaces required for multi-physics coupling. As a result, LAURA now shares gas-physics modules, MPI modules, and other low-level modules with the FUN3D unstructured-grid code. In addition to internal refactoring, several new features and capabilities have been added, e.g., a GNU-standard installation process, parallel load balancing, automatic trajectory point sequencing, free-energy minimization, and coupled ablation and flowfield radiation.

  1. LAURA Users Manual: 5.5-64987

    Science.gov (United States)

    Mazaheri, Alireza; Gnoffo, Peter A.; Johnston, Christopher O.; Kleb, William L.

    2013-01-01

    This users manual provides in-depth information concerning installation and execution of LAURA, version 5. LAURA is a structured, multi-block, computational aerothermodynamic simulation code. Version 5 represents a major refactoring of the original Fortran 77 LAURA code toward a modular structure afforded by Fortran 95. The refactoring improved usability and maintain ability by eliminating the requirement for problem dependent recompilations, providing more intuitive distribution of functionality, and simplifying interfaces required for multi-physics coupling. As a result, LAURA now shares gas-physics modules, MPI modules, and other low-level modules with the Fun3D unstructured-grid code. In addition to internal refactoring, several new features and capabilities have been added, e.g., a GNU standard installation process, parallel load balancing, automatic trajectory point sequencing, free-energy minimization, and coupled ablation and flowfield radiation.

  2. LAURA Users Manual: 5.4-54166

    Science.gov (United States)

    Mazaheri, Alireza; Gnoffo, Peter A.; Johnston, Christopher O.; Kleb, Bil

    2011-01-01

    This users manual provides in-depth information concerning installation and execution of Laura, version 5. Laura is a structured, multi-block, computational aerothermodynamic simulation code. Version 5 represents a major refactoring of the original Fortran 77 Laura code toward a modular structure afforded by Fortran 95. The refactoring improved usability and maintainability by eliminating the requirement for problem dependent re-compilations, providing more intuitive distribution of functionality, and simplifying interfaces required for multi-physics coupling. As a result, Laura now shares gas-physics modules, MPI modules, and other low-level modules with the Fun3D unstructured-grid code. In addition to internal refactoring, several new features and capabilities have been added, e.g., a GNU-standard installation process, parallel load balancing, automatic trajectory point sequencing, free-energy minimization, and coupled ablation and flowfield radiation.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Yamano, N.; Brockmann, J.E.

    1989-05-01

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

  5. HEFF---A user's manual and guide for the HEFF code for thermal-mechanical analysis using the boundary-element method

    International Nuclear Information System (INIS)

    St John, C.M.; Sanjeevan, K.

    1991-12-01

    The HEFF Code combines a simple boundary-element method of stress analysis with the closed form solutions for constant or exponentially decaying heat sources in an infinite elastic body to obtain an approximate method for analysis of underground excavations in a rock mass with heat generation. This manual describes the theoretical basis for the code, the code structure, model preparation, and step taken to assure that the code correctly performs its intended functions. The material contained within the report addresses the Software Quality Assurance Requirements for the Yucca Mountain Site Characterization Project. 13 refs., 26 figs., 14 tabs

  6. User's manual for ASTERIX-2: a two-dimensional modular-code system for the steady-state and xenon-transient analysis of a pebble-bed high-temperature reactor

    International Nuclear Information System (INIS)

    Lauer, A.; Schwiegk, H.J.; Wu, T.; Cowan, C.L.

    1982-03-01

    The ASTERIX modular code package was developed at KFA Laboratory-Juelich for the steady state and xenon transient analysis of a pebble bed high temperature reactor. The code package was implemented on the Stanford Linear Accelerator Center Computer in August, 1980, and a user's manual for the current version of the code, identified as ASTERIX-2, was prepared as a cooperative effort by KFA Laboratory and GE-ARSD. The material in the manual includes the requirements for accessing the program, a description of the major subroutines, a listing of the input options, and a listing of the input data for a sample problem. The material is provided in sufficient detail for the user to carry out a wide range of analyses from steady state operations to the xenon induced power transients in which the local xenon, temperature, buckling and control feedback effects have been incorporated in the problem solution

  7. Generic communications index: User's manual

    International Nuclear Information System (INIS)

    Dean, R.S.; Steinbrecher, D.H.; Hennick, A.

    1987-12-01

    This report is a manual for providing information required to use a special computer program developed by the NRC for indexing generic communications. The program is written in a user-friendly menu driven form using dBASE III programming language. It facilitates use of the required dBASE III search and sort capabilities to access records in a database called Generic Communications Index. This index is made up of one record each for all bulletins, circulars, and information notices, including revisions and supplements, from 1971, when such documentation started, through 1986 (or to the latest update). The program is designed for use by anyone modestly acquainted with the general use of IBM-compatible personal computers. The manual contains both a brief overview and a detailed description of the program, as well as detailed instructions for getting started using the program on a personal computer with either a two-floppy disk or a hard disk system. Included at the end are a brief description of how to handle problems which might occur, and notes on the makeup of the program and database files for help in adding records of communications for future years

  8. PRETART/TARTV user's manual

    International Nuclear Information System (INIS)

    Dubois, P.F.

    1977-01-01

    PRETART and TARTV are preprocessors for input decks for TARTNP, a coupled neutron--photon Monte Carlo transport code. They accept the normal TARTNP input decks. The purpose of these routines is to aid users in checking their geometry before making a TARTNP run. PRETART finds the volume of each zone, or of any selected subset of zones. When the volumes of all zones are found, the user may select a percentage of random checking for gaps or overlaps between the zones. If gaps or overlaps are found, a diagnostic routine describes the difficulty as best it can. TARTV allows the user to see a picture on TMDS (with a fiche copy) of a plane slice, parallel to the x-y plane or parallel to the x-z plane, through space, with an arbitrary picture window. A complete description of how to use these routines follows. It needs to be emphasized that PRETART cannot be used naively. It requires an active user who understands what it is trying to do, and its limitations

  9. Nuclear science references coding manual

    International Nuclear Information System (INIS)

    Ramavataram, S.; Dunford, C.L.

    1996-08-01

    This manual is intended as a guide to Nuclear Science References (NSR) compilers. The basic conventions followed at the National Nuclear Data Center (NNDC), which are compatible with the maintenance and updating of and retrieval from the Nuclear Science References (NSR) file, are outlined. In Section H, the structure of the NSR file such as the valid record identifiers, record contents, text fields as well as the major TOPICS for which are prepared are enumerated. Relevant comments regarding a new entry into the NSR file, assignment of , generation of and linkage characteristics are also given in Section II. In Section III, a brief definition of the Keyword abstract is given followed by specific examples; for each TOPIC, the criteria for inclusion of an article as an entry into the NSR file as well as coding procedures are described. Authors preparing Keyword abstracts either to be published in a Journal (e.g., Nucl. Phys. A) or to be sent directly to NNDC (e.g., Phys. Rev. C) should follow the illustrations in Section III. The scope of the literature covered at the NNDC, the categorization into Primary and Secondary sources, etc., is discussed in Section IV. Useful information regarding permitted character sets, recommended abbreviations, etc., is given under Section V as Appendices

  10. Nuclear structure references coding manual

    International Nuclear Information System (INIS)

    Ramavataram, S.; Dunford, C.L.

    1984-02-01

    This manual is intended as a guide to Nuclear Structure References (NSR) compilers. The basic conventions followed at the National Nuclear Data Center (NNDC), which are compatible with the maintenance and updating of and retrieval from the Nuclear Structure References (NSR) file, are outlined. The structure of the NSR file such as the valid record identifiers, record contents, text fields as well as the major topics for which [KEYWORDS] are prepared are ennumerated. Relevant comments regarding a new entry into the NSR file, assignment of [KEYNO ], generation of [SELECTRS] and linkage characteristics are also given. A brief definition of the Keyword abstract is given followed by specific examples; for each TOPIC, the criteria for inclusion of an article as an entry into the NSR file as well as coding procedures are described. Authors submitting articles to Journals which require Keyword abstracts should follow the illustrations. The scope of the literature covered at NNDC, the categorization into Primary and Secondary sources, etc. is discussed. Useful information regarding permitted character sets, recommended abbreviations, etc. is given

  11. User Manual for the PROTEUS Mesh Tools

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Micheal A. [Argonne National Lab. (ANL), Argonne, IL (United States); Shemon, Emily R. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-06-01

    This report describes the various mesh tools that are provided with the PROTEUS code giving both descriptions of the input and output. In many cases the examples are provided with a regression test of the mesh tools. The most important mesh tools for any user to consider using are the MT_MeshToMesh.x and the MT_RadialLattice.x codes. The former allows the conversion between most mesh types handled by PROTEUS while the second allows the merging of multiple (assembly) meshes into a radial structured grid. Note that the mesh generation process is recursive in nature and that each input specific for a given mesh tool (such as .axial or .merge) can be used as “mesh” input for any of the mesh tools discussed in this manual.

  12. UCB-NE-107 user's manual

    International Nuclear Information System (INIS)

    Lee, W.W.L.

    1989-03-01

    The purpose of this manual is to provide users of UCB-NE-107 with the information necessary to use UCB-NE-107 effectively. UCB-NE-107 is a computer code for calculating the fractional rate of readily soluble radionuclides that are released from nuclear waste emplaced in water-saturated porous media. Waste placed in such environments will gradually dissolve. For many species such as actinides and rare earths, the process of dissolution is governed by the exterior flow field, and the chemical reaction rate or leaching rate. However, for readily soluble species such as 135 Cs, 137 Cs, and 129 I, it has been observed that their dissolution rates are rapid. UCB-NE-107 is a code for calculating the release rate at the waste/rock interface, to check compliance with the US Nuclear Regulatory Commission's (USNRC) subsystem performance objective. It is an implementation of the analytic solution given below. 5 refs., 2 figs

  13. User Manual for the PROTEUS Mesh Tools

    International Nuclear Information System (INIS)

    Smith, Micheal A.; Shemon, Emily R.

    2015-01-01

    This report describes the various mesh tools that are provided with the PROTEUS code giving both descriptions of the input and output. In many cases the examples are provided with a regression test of the mesh tools. The most important mesh tools for any user to consider using are the MT M eshToMesh.x and the MT R adialLattice.x codes. The former allows the conversion between most mesh types handled by PROTEUS while the second allows the merging of multiple (assembly) meshes into a radial structured grid. Note that the mesh generation process is recursive in nature and that each input specific for a given mesh tool (such as .axial or .merge) can be used as ''mesh'' input for any of the mesh tools discussed in this manual.

  14. User's manual for EXALPHA (a code for calculating electronic properties of molecules). [Muscatel code, multiply scattered electron approximation

    Energy Technology Data Exchange (ETDEWEB)

    Jones, H.D.

    1976-06-01

    The EXALPHA procedures provide a simplified method for running the MUSCATEL computer code, which in turn is used for calculating electronic properties of simple molecules and atomic clusters, based on the multiply scattered electron approximation for the wave equations. The use of the EXALPHA procedures to set up a run of MUSCATEL is described.

  15. AFM-CMM integrated instrument user manual

    DEFF Research Database (Denmark)

    Marinello, Francesco; Bariani, Paolo

    This manual gives general important guidelines for a proper use of the integrated AFM-CMM instrument. More information can be collected reading: • N. Kofod Ph.D thesis [1]; • P. Bariani Ph.D thesis [2]; • Dualscope DME 95-200 operation manuals [3]; • SPIP help [4] • Stitching software user manual...

  16. THEAP-I: A computer program for thermal hydraulic analysis of a thermally interacting channel bundle of complex geometry. Code description and user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Bartzis, J G; Megaritou, A; Belessiotis, V

    1987-09-01

    THEAP-I is a computer code developed in NRCPS `DEMOCRITUS` with the aim to contribute to the safety analysis of the open pool research reactors. THEAP-I is designed for three dimensional, transient thermal/hydraulic analysis of a thermally interacting channel bundle totally immersed into water or air, such as the reactor core. In the present report the mathematical and physical models and methods of the solution are given as well as the code description and the input data. A sample problem is also included, refering to the Greek Research Reactor analysis, under an hypothetical severe loss of coolant accident.

  17. A user's manual to the PMBOK guide

    CERN Document Server

    Stackpole Snyder, Cynthia

    2013-01-01

    The must-have manual to understand and use the latest edition of the Fifth Edition The professional standard in the field of project management, A Guide to the Project Management Body of Knowledge (PMBOK® Guide-Fifth Edition) published by the Project Management Institute (PMI) serves as the ultimate resource for professionals and as a valuable studying and training device for students taking the PMP® Exam. A User''s Manual to the PMBOK® Guide takes the next logical step to act as a true user''s manual. With an accessible format and easy-to-understand language, it helps to not only distill es

  18. Important wheelchair skills for new manual wheelchair users: health care professional and wheelchair user perspectives.

    Science.gov (United States)

    Morgan, Kerri A; Engsberg, Jack R; Gray, David B

    2017-01-01

    The purpose of this project was to identify wheelchair skills currently being taught to new manual wheelchair users, identify areas of importance for manual wheelchair skills' training during initial rehabilitation, identify similarities and differences between the perspectives of health care professionals and manual wheelchair users and use the ICF to organize themes related to rehabilitation and learning how to use a manual wheelchair. Focus groups were conducted with health care professionals and experienced manual wheelchair users. ICF codes were used to identify focus group themes. The Activities and Participation codes were more frequently used than Structure, Function and Environment codes. Wheelchair skills identified as important for new manual wheelchair users included propulsion techniques, transfers in an out of the wheelchair, providing maintenance to the wheelchair and navigating barriers such as curbs, ramps and rough terrain. Health care professionals and manual wheelchair users identified the need to incorporate the environment (home and community) into the wheelchair training program. Identifying essential components for training the proper propulsion mechanics and wheelchair skills in new manual wheelchair users is an important step in preventing future health and participation restrictions. Implications for Rehabilitation Wheelchair skills are being addressed frequently during rehabilitation at the activity-dependent level. Propulsion techniques, transfers in an out of the wheelchair, providing maintenance to the wheelchair and navigating barriers such as curbs, ramps and rough terrain are important skills to address during wheelchair training. Environment factors (in the home and community) are important to incorporate into wheelchair training to maximize safe and multiple-environmental-setting uses of manual wheelchairs. The ICF has application to understanding manual wheelchair rehabilitation for wheelchair users and therapists for improving

  19. Tripoli-3: monte Carlo transport code for neutral particles - version 3.5 - users manual; Tripoli-3: code de transport des particules neutres par la methode de monte carlo - version 3.5 - manuel d'utilisation

    Energy Technology Data Exchange (ETDEWEB)

    Vergnaud, Th.; Nimal, J.C.; Chiron, M

    2001-07-01

    The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)

  20. Tripoli-3: monte Carlo transport code for neutral particles - version 3.5 - users manual; Tripoli-3: code de transport des particules neutres par la methode de monte carlo - version 3.5 - manuel d'utilisation

    Energy Technology Data Exchange (ETDEWEB)

    Vergnaud, Th; Nimal, J C; Chiron, M

    2001-07-01

    The TRIPOLI-3 code applies the Monte Carlo method to neutron, gamma-ray and coupled neutron and gamma-ray transport calculations in three-dimensional geometries, either in steady-state conditions or having a time dependence. It can be used to study problems where there is a high flux attenuation between the source zone and the result zone (studies of shielding configurations or source driven sub-critical systems, with fission being taken into account), as well as problems where there is a low flux attenuation (neutronic calculations -- in a fuel lattice cell, for example -- where fission is taken into account, usually with the calculation on the effective multiplication factor, fine structure studies, numerical experiments to investigate methods approximations, etc). TRIPOLI-3 has been operational since 1995 and is the version of the TRIPOLI code that follows on from TRIPOLI-2; it can be used on SUN, RISC600 and HP workstations and on PC using the Linux or Windows/NT operating systems. The code uses nuclear data libraries generated using the THEMIS/NJOY system. The current libraries were derived from ENDF/B6 and JEF2. There is also a response function library based on a number of evaluations, notably the dosimetry libraries IRDF/85, IRDF/90 and also evaluations from JEF2. The treatment of particle transport is the same in version 3.5 as in version 3.4 of the TRIPOLI code; but the version 3.5 is more convenient for preparing the input data and for reading the output. The french version of the user's manual exists. (authors)

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  2. How 2 HAWC2, the user's manual

    Energy Technology Data Exchange (ETDEWEB)

    Juul Larsen, T.; Melchior Hansen, A.

    2007-12-15

    The report contains the user's manual for the aeroelastic code HAWC2. The code is intended for calculating wind turbine response in time domain and has a structural formulation based on multi-body dynamics. The aerodynamic part of the code is based on the blade element momentum theory, but extended from the classic approach to handle dynamic inflow, dynamic stall, skew inflow, shear effects on the induction and effects from large deflections. It has been developed within the years 2003-2006 at the aeroelastic design research programme at Risoe National Laboratory, Denmark. This manual is updated for HAWC2 version 6.4. (au)

  3. ERIC User Services Manual. Revised Edition.

    Science.gov (United States)

    Wagner, Judith O., Comp.

    This manual explains how the user services functions, usually performed by a User Services Coordinator, can be conducted in the 16 ERIC (Educational Resources Information Center) Clearinghouses and the various adjunct ERIC Clearinghouses. It provides guidelines, suggestions, and examples of how ERIC components currently perform the user services…

  4. User's Manual for FEMOM3DR. Version 1.0

    Science.gov (United States)

    Reddy, C. J.

    1998-01-01

    FEMoM3DR is a computer code written in FORTRAN 77 to compute radiation characteristics of antennas on 3D body using combined Finite Element Method (FEM)/Method of Moments (MoM) technique. The code is written to handle different feeding structures like coaxial line, rectangular waveguide, and circular waveguide. This code uses the tetrahedral elements, with vector edge basis functions for FEM and triangular elements with roof-top basis functions for MoM. By virtue of FEM, this code can handle any arbitrary shaped three dimensional bodies with inhomogeneous lossy materials; and due to MoM the computational domain can be terminated in any arbitrary shape. The User's Manual is written to make the user acquainted with the operation of the code. The user is assumed to be familiar with the FORTRAN 77 language and the operating environment of the computers on which the code is intended to run.

  5. Wien Automatic System Planning (WASP) Package. A computer code for power generating system expansion planning. Version WASP-III Plus. User's manual. Volume 1: Chapters 1-11

    International Nuclear Information System (INIS)

    1995-01-01

    (FIXSYS plants); user control of the distribution of capital cost expenditures during the construction period (if required to be different from the general 'S' curve distribution used as default). The present document has been produced to support use of the WASP-Ill Plus computer code and to illustrate the capabilities of the program. This Manual is organized in two separate volumes. This first one includes 11 main chapters describing how to use the WASP-Ill Plus computer program. Chapter 1 gives a summary description and some background information about the program. Chapter 2 introduces some concepts, mainly related to the computer requirements imposed by the program, that are used throughout the Manual. Chapters 3 to 9 describe how to execute each of the various programs (or modules) of the WASP-Ill Plus package. The description for each module shows the user how to prepare the Job Control statements and input data needed to execute the module and how to interpret the printed output produced. The iterative process that should be followed in order to obtain the 'optimal solution' for a WASP case study is covered in Chapters 6 to 8. Chapter 10 explains the use of an auxiliary program of the WASP package which is mainly intended for saving computer time. Lastly, Chapter 11 recapitulates the use of WASP-Ill Plus for executing a generation expansion planning study; describes the several phases normally involved in this type of study; and provides the user with practical hints about the most important aspects that need to be verified at each phase while executing the various WASP modules

  6. T-3 cask users' manual. Revision 1

    International Nuclear Information System (INIS)

    1986-06-01

    This user's manual for the T-C spent fuel cask provides information on: operating procedures; inspection and maintenance procedures; criticality evaluation; shielding evaluation; thermal evaluation; structural evaluation; and limitations

  7. User's Manual for CoEAT Tool

    Science.gov (United States)

    The Co-EAT users manual is designed to help the anaerobic digestion system operators evaluate the costs and benefits of accepting and processing wasted food, fats, oils and greases (FOG) or other organic materials.

  8. Los Alamos waste drum shufflers users manual

    International Nuclear Information System (INIS)

    Rinard, P.M.; Adams, E.L.; Painter, J.

    1993-01-01

    This user manual describes the Los Alamos waste drum shufflers. The primary purpose of the instruments is to assay the mass of 235 U (or other fissile materials) in drums of assorted waste. It can perform passive assays for isotopes that spontaneously emit neutrons or active assays using the shuffler technique as described on this manual

  9. SPRAY code user's report

    International Nuclear Information System (INIS)

    Shire, P.R.

    1977-03-01

    The SPRAY computer code has been developed to model the effects of postulated sodium spray release from LMFBR piping within containment chambers. The calculation method utilizes gas convection, heat transfer and droplet combustion theory to calculate the pressure and temperature effects within the enclosure. The applicable range is 0-21 mol percent oxygen and .02-.30 inch droplets with or without humidity. Droplet motion and large sodium surface area combine to produce rapid heat release and pressure rise within the enclosed volume

  10. The ISIS Open GENIE user manual

    International Nuclear Information System (INIS)

    Akeroyd, F.A.; Ashworth, R.L.; Campbell, S.I.; Johnston, S.D.; Martin, J.M.; Moreton-Smith, C.M.; Sivia, D.S.

    2000-01-01

    This manual should enable you to become familiar with Open GENIE quickly and easily. It therefore complements the Open GENIE Reference Manual which should be used to understand the full meaning of Open GENIE commands. The reference manual is accessible on the ISIS web server at: http://www.isis.rl.ac.uk/GENIEReferenceManual/ and the user manual is planned to be accessible by January '98 at http://www.isis.rl.ac.ukIGENIEUserManual/. The Open GENIE User Manual is separated into two parts: Part A. The User Manual. An introduction to the use of Open GENIE. Part B. The Installation Guide. General information on how to install and run Open GENIE. Assuming that you are new to Open GENIE and have just downloaded a copy you will need to consult the Installation Guide to get Open GENIE installed on your machine. After this we recommend you experiment with some of the example files to get an idea of the capabilities of Open GENIE. For further information, comments, additions of routines that you feel should be included, please contact us at genie at isise.rl.ac.uk

  11. User's manual for CBS3DS, version 1.0

    Science.gov (United States)

    Reddy, C. J.; Deshpande, M. D.

    1995-10-01

    CBS3DS is a computer code written in FORTRAN 77 to compute the backscattering radar cross section of cavity backed apertures in infinite ground plane and slots in thick infinite ground plane. CBS3DS implements the hybrid Finite Element Method (FEM) and Method of Moments (MoM) techniques. This code uses the tetrahedral elements, with vector edge basis functions for FEM in the volume of the cavity/slot and the triangular elements with the basis functions for MoM at the apertures. By virtue of FEM, this code can handle any arbitrarily shaped three-dimensional cavities filled with inhomogeneous lossy materials; due to MoM, the apertures can be of any arbitrary shape. The User's Manual is written to make the user acquainted with the operation of the code. The user is assumed to be familiar with the FORTRAN 77 language and the operating environment of the computer the code is intended to run.

  12. User's Manual for FEMOM3DS. Version 1.0

    Science.gov (United States)

    Reddy, C.J.; Deshpande, M. D.

    1997-01-01

    FEMOM3DS is a computer code written in FORTRAN 77 to compute electromagnetic(EM) scattering characteristics of a three dimensional object with complex materials using combined Finite Element Method (FEM)/Method of Moments (MoM) technique. This code uses the tetrahedral elements, with vector edge basis functions for FEM in the volume of the cavity and the triangular elements with the basis functions similar to that described for MoM at the outer boundary. By virtue of FEM, this code can handle any arbitrarily shaped three-dimensional cavities filled with inhomogeneous lossy materials. The User's Manual is written to make the user acquainted with the operation of the code. The user is assumed to be familiar with the FORTRAN 77 language and the operating environment of the computers on which the code is intended to run.

  13. User's Manual for FEM-BEM Method. 1.0

    Science.gov (United States)

    Butler, Theresa; Deshpande, M. D. (Technical Monitor)

    2002-01-01

    A user's manual for using FORTRAN code to perform electromagnetic analysis of arbitrarily shaped material cylinders using a hybrid method that combines the finite element method (FEM) and the boundary element method (BEM). In this method, the material cylinder is enclosed by a fictitious boundary and the Maxwell's equations are solved by FEM inside the boundary and by BEM outside the boundary. The electromagnetic scattering on several arbitrarily shaped material cylinders using this FORTRAN code is computed to as examples.

  14. User's manual of a computer code for seismic hazard evaluation for assessing the threat to a facility by fault model. SHEAT-FM

    International Nuclear Information System (INIS)

    Sugino, Hideharu; Onizawa, Kunio; Suzuki, Masahide

    2005-09-01

    To establish the reliability evaluation method for aged structural component, we developed a probabilistic seismic hazard evaluation code SHEAT-FM (Seismic Hazard Evaluation for Assessing the Threat to a facility site - Fault Model) using a seismic motion prediction method based on fault model. In order to improve the seismic hazard evaluation, this code takes the latest knowledge in the field of earthquake engineering into account. For example, the code involves a group delay time of observed records and an update process model of active fault. This report describes the user's guide of SHEAT-FM, including the outline of the seismic hazard evaluation, specification of input data, sample problem for a model site, system information and execution method. (author)

  15. MCNP-DSP users manual

    International Nuclear Information System (INIS)

    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 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

  16. SANDIA-ORIGEN user's manual

    International Nuclear Information System (INIS)

    Bennett, D.E.

    1979-10-01

    The SANDIA-ORIGEN code calculates the detailed isotopic composition as a function of time in nuclear reactor fuel irradiation and radioactive decay problems. This code was developed specifically for Control Data Corporation computers from the original Oak Ridge National Laboratory ORIGEN code. The nuclear data file used with the code at Sandia Laboratories contains 1063 isotopes (254 structural materials, 101 actinides, and 708 fission products). SANDIA-ORIGEN is oriented toward simple, easy use and includes NAMELIST input, convenient control of the output, and versatile options for the blending and reprocessing of reactor fuel. System operating instructions and the input decks for numerous sample problems are also presented. 13 references, 14 figures

  17. Wien Automatic System Planning (WASP) Package. A computer code for power generating system expansion planning. Version WASP-III Plus. User's manual. Volume 1: Chapters 1-11

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    (FIXSYS plants); user control of the distribution of capital cost expenditures during the construction period (if required to be different from the general 'S' curve distribution used as default). The present document has been produced to support use of the WASP-Ill Plus computer code and to illustrate the capabilities of the program. This Manual is organized in two separate volumes. This first one includes 11 main chapters describing how to use the WASP-Ill Plus computer program. Chapter 1 gives a summary description and some background information about the program. Chapter 2 introduces some concepts, mainly related to the computer requirements imposed by the program, that are used throughout the Manual. Chapters 3 to 9 describe how to execute each of the various programs (or modules) of the WASP-Ill Plus package. (abstract truncated)

  18. PETSc Users Manual Revision 3.3

    Energy Technology Data Exchange (ETDEWEB)

    Balay, S. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Brown, J. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Buschelman, K. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Eijkhout, V. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Gropp, W. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Kaushik, D. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Knepley, M. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; McInnes, L. Curfman [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Smith, B. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Zhang, H. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division

    2013-05-11

    This manual describes the use of PETSc for the numerical solution of partial differential equations and related problems on high-performance computers. The Portable, Extensible Toolkit for Scientific Computation (PETSc) is a suite of data structures and routines that provide the building blocks for the implementation of large-scale application codes on parallel (and serial) computers. PETSc uses the MPI standard for all message-passing communication. PETSc includes an expanding suite of parallel linear, nonlinear equation solvers and time integrators that may be used in application codes written in Fortran, C, C++, Python, and MATLAB (sequential). PETSc provides many of the mechanisms needed within parallel application codes, such as parallel matrix and vector assembly routines. The library is organized hierarchically, enabling users to employ the level of abstraction that is most appropriate for a particular problem. By using techniques of object-oriented programming, PETSc provides enormous flexibility for users. PETSc is a sophisticated set of software tools; as such, for some users it initially has a much steeper learning curve than a simple subroutine library. In particular, for individuals without some computer science background, experience programming in C, C++ or Fortran and experience using a debugger such as gdb or dbx, it may require a significant amount of time to take full advantage of the features that enable efficient software use. However, the power of the PETSc design and the algorithms it incorporates may make the efficient implementation of many application codes simpler than “rolling them” yourself; For many tasks a package such as MATLAB is often the best tool; PETSc is not intended for the classes of problems for which effective MATLAB code can be written. PETSc also has a MATLAB interface, so portions of your code can be written in MATLAB to “try out” the PETSc solvers. The resulting code will not be scalable however because

  19. PETSc Users Manual Revision 3.5

    Energy Technology Data Exchange (ETDEWEB)

    Balay, S. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Abhyankar, S. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Adams, M. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Brown, J. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Brune, P. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Buschelman, K. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Eijkhout, V. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Gropp, W. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Kaushik, D. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Knepley, M. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; McInnes, L. Curfman [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Rupp, K. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Smith, B. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Zhang, H. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division

    2014-09-08

    This manual describes the use of PETSc for the numerical solution of partial differential equations and related problems on high-performance computers. The Portable, Extensible Toolkit for Scientific Computation (PETSc) is a suite of data structures and routines that provide the building blocks for the implementation of large-scale application codes on parallel (and serial) computers. PETSc uses the MPI standard for all message-passing communication. PETSc includes an expanding suite of parallel linear, nonlinear equation solvers and time integrators that may be used in application codes written in Fortran, C, C++, Python, and MATLAB (sequential). PETSc provides many of the mechanisms needed within parallel application codes, such as parallel matrix and vector assembly routines. The library is organized hierarchically, enabling users to employ the level of abstraction that is most appropriate for a particular problem. By using techniques of object-oriented programming, PETSc provides enormous flexibility for users. PETSc is a sophisticated set of software tools; as such, for some users it initially has a much steeper learning curve than a simple subroutine library. In particular, for individuals without some computer science background, experience programming in C, C++ or Fortran and experience using a debugger such as gdb or dbx, it may require a significant amount of time to take full advantage of the features that enable efficient software use. However, the power of the PETSc design and the algorithms it incorporates may make the efficient implementation of many application codes simpler than “rolling them” yourself. ;For many tasks a package such as MATLAB is often the best tool; PETSc is not intended for the classes of problems for which effective MATLAB code can be written. PETSc also has a MATLAB interface, so portions of your code can be written in MATLAB to “try out” the PETSc solvers. The resulting code will not be scalable however because

  20. PETSc Users Manual Revision 3.4

    Energy Technology Data Exchange (ETDEWEB)

    Balay, S. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Brown, J. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Buschelman, K. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Eijkhout, V. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Gropp, W. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Kaushik, D. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Knepley, M. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; McInnes, L. Curfman [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Smith, B. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division; Zhang, H. [Argonne National Lab. (ANL), Argonne, IL (United States). Mathematics and Computer Science Division

    2014-06-29

    This manual describes the use of PETSc for the numerical solution of partial differential equations and related problems on high-performance computers. The Portable, Extensible Toolkit for Scientific Computation (PETSc) is a suite of data structures and routines that provide the building blocks for the implementation of large-scale application codes on parallel (and serial) computers. PETSc uses the MPI standard for all message-passing communication. PETSc includes an expanding suite of parallel linear, nonlinear equation solvers and time integrators that may be used in application codes written in Fortran, C, C++, Python, and MATLAB (sequential). PETSc provides many of the mechanisms needed within parallel application codes, such as parallel matrix and vector assembly routines. The library is organized hierarchically, enabling users to employ the level of abstraction that is most appropriate for a particular problem. By using techniques of object-oriented programming, PETSc provides enormous flexibility for users. PETSc is a sophisticated set of software tools; as such, for some users it initially has a much steeper learning curve than a simple subroutine library. In particular, for individuals without some computer science background, experience programming in C, C++ or Fortran and experience using a debugger such as gdb or dbx, it may require a significant amount of time to take full advantage of the features that enable efficient software use. However, the power of the PETSc design and the algorithms it incorporates may make the efficient implementation of many application codes simpler than “rolling them” yourself; For many tasks a package such as MATLAB is often the best tool; PETSc is not intended for the classes of problems for which effective MATLAB code can be written. PETSc also has a MATLAB interface, so portions of your code can be written in MATLAB to “try out” the PETSc solvers. The resulting code will not be scalable however because

  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. User's Manual for the Langley Aerothermodynamic Upwind Relaxation Algorithm (LAURA)

    Science.gov (United States)

    Gnoffo, Peter A.; Cheatwood, F. McNeil

    1996-01-01

    This user's manual provides detailed instructions for the installation and the application of version 4.1 of the Langley Aerothermodynamic Upwind Relaxation Algorithm (LAURA). Also provides simulation of flow field in thermochemical nonequilibrium around vehicles traveling at hypersonic velocities through the atmosphere. Earlier versions of LAURA were predominantly research codes, and they had minimal (or no) documentation. This manual describes UNIX-based utilities for customizing the code for special applications that also minimize system resource requirements. The algorithm is reviewed, and the various program options are related to specific equations and variables in the theoretical development.

  3. ESP-TIMOC code manual

    International Nuclear Information System (INIS)

    Jaarsma, R.; Perlado, J.M.; Rief, H.

    1978-01-01

    ESP-TIMOC is an 'Event Scanning Program' to analyse the events (collision or boundary crossing parameters) of Monte Carlo particle transport problems. It is a modular program and belongs to the TIMOC code system. ESP-TIMOC is primarily designed to calculate the time dependent response functions such as energy dependent fluxes and currents at interfaces. An eventual extension to other quantities is simple and straight forward

  4. TMAP4 User`s Manual

    Energy Technology Data Exchange (ETDEWEB)

    Longhurst, G.R.; Holland, D.F.; Jones, J.L.; Merrill, B.J.

    1992-06-12

    The Tritium Migration Analysis Program, Version 4 (TMAP4) has been developed by the Fusion Safety Program at the Idaho National Engineering Laboratory (INEL) as a safety analysis code, mainly to analyze tritium retention and loss in fusion reactor structures and systems during normal operation and accident conditions. TMAP4 incorporates one-dimensional thermal- and mass-diffusive transport and trapping calculations through structures and zero dimensional fluid transport between enclosures and across the interface between enclosures and structures. A key feature is the ability to input problem definition parameters as constants, interpolation tables, or FORTRAN equations. The code is specifically intended for use under a DOS operating system on PC-type mini-computers, but it has also been run successfully on workstations and mainframe computer systems. Use of the equation-input feature requires access to a FORTRAN-77 compiler and a linker program.

  5. MARS CODE MANUAL VOLUME V: Models and Correlations

    International Nuclear Information System (INIS)

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

    2010-02-01

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

  6. Smart roadside initiative : user manual.

    Science.gov (United States)

    2015-09-01

    This document provides the user instructions for the Smart Roadside Initiative (SRI) applications including mobile and web-based SRI applications. These applications include smartphone-enabled information exchange and notification, and software compo...

  7. TNG-GENOA User's Manual

    International Nuclear Information System (INIS)

    Guimaraes, F.B.; Fu, C.Y.

    2000-01-01

    The aim of this work is to describe the basic aspects of the codes TNG and GENOA. These codes have been developed and used at ORNL in the last decades for the analysis and evaluation of neutron induced nuclear data. In the energy region of the unresolved resonances range and high energies range. These evaluations have been performed in support of various projects and were included in the ENDF/B library. The implementation of these codes into the code SAMMY has been performed as part of a program of the creation of a general Nuclear Data evaluation tool for the analysis of reactions in a broad energy range, from few eV up to about 150 MeV

  8. User's manual for DELSOL2: a computer code for calculating the optical performance and optimal system design for solar-thermal central-receiver plants

    Energy Technology Data Exchange (ETDEWEB)

    Dellin, T.A.; Fish, M.J.; Yang, C.L.

    1981-08-01

    DELSOL2 is a revised and substantially extended version of the DELSOL computer program for calculating collector field performance and layout, and optimal system design for solar thermal central receiver plants. The code consists of a detailed model of the optical performance, a simpler model of the non-optical performance, an algorithm for field layout, and a searching algorithm to find the best system design. The latter two features are coupled to a cost model of central receiver components and an economic model for calculating energy costs. The code can handle flat, focused and/or canted heliostats, and external cylindrical, multi-aperture cavity, and flat plate receivers. The program optimizes the tower height, receiver size, field layout, heliostat spacings, and tower position at user specified power levels subject to flux limits on the receiver and land constraints for field layout. The advantages of speed and accuracy characteristic of Version I are maintained in DELSOL2.

  9. ADFStealthViewer User Manual

    Science.gov (United States)

    2013-05-01

    environmental objects are used to immerse the user in a 3D visualisation of the simulated war game. ADFStealthViewer has several ADF produced 3D models...OpenGL, audio , and networking devices. Some advanced functionality of the engine relies on modern graphics pixel and vertex shaders. These advanced

  10. RAVEN User Manual Revision 5

    International Nuclear Information System (INIS)

    Mandelli, Diego; Rabiti, Cristian; Cogliati, Joshua Joseph; Kinoshita, Robert Arthur; Alfonsi, Andrea; Sen, Ramazan Sonat; Wang, Congjian

    2017-01-01

    RAVEN is a generic software framework to perform parametric and probabilistic analy- sis based on the response of complex system codes. The initial development was aimed to provide dynamic risk analysis capabilities to the Thermo-Hydraulic code RELAP-7, currently under development at the Idaho National Laboratory (INL). Although the initial goal has been fully accomplished, RAVEN is now a multi-purpose probabilistic and uncer- tainty quantification platform, capable to agnostically communicate with any system code. This agnosticism includes providing Application Programming Interfaces (APIs). These APIs are used to allow RAVEN to interact with any code as long as all the parameters that need to be perturbed are accessible by inputs files or via python interfaces. RAVEN is capable of investigating the system response, and investigating the input space using Monte Carlo, Grid, or Latin Hyper Cube sampling schemes, but its strength is focused to- ward system feature discovery, such as limit surfaces, separating regions of the input space leading to system failure, using dynamic supervised learning techniques. The development of RAVEN has started in 2012, when, within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the need to provide a modern risk evaluation framework became stronger. RAVEN principal assignment is to provide the necessary software and algorithms in order to employ the concept developed by the Risk Informed Safety Margin Characterization (RISMC) program. RISMC is one of the pathways defined within the Light Water Reactor Sustainability (LWRS) program. In the RISMC approach, the goal is not just the individuation of the frequency of an event potentially leading to a system failure, but the closeness (or not) to key safety-related events. Hence, the approach is in- terested in identifying and increasing the safety margins related to those events. A safety margin is a numerical value quantifying the probability that a safety metric (e

  11. High School and Beyond: Twins and Siblings' File Users' Manual, User's Manual for Teacher Comment File, Friends File Users' Manual.

    Science.gov (United States)

    National Center for Education Statistics (ED), Washington, DC.

    These three users' manuals are for specific files of the High School and Beyond Study, a national longitudinal study of high school sophomores and seniors in 1980. The three files are computerized databases that are available on magnetic tape. As one component of base year data collection, information identifying twins, triplets, and some non-twin…

  12. AXAIR89Q User's Manual

    International Nuclear Information System (INIS)

    Simpkins, A.A.

    1994-04-01

    AXAIR89Q is the primary dose assessment code used at the Savannah River Site to predict downwind doses following a hypothetical atmospheric release of relatively short duration. The primary purpose of the code is to perform calculations for safety-related documentation, and there is strict adherence in AXAIR89Q to the guidance in USNRC Regulatory Guide 1.145 (USNRC, 1983) entitled Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants. Doses are determined for the cloud shine, and inhalation pathways. Ingestion and ground shine are not considered. Future modifications are expected to include these features. This report discusses the information needed to execute the program and interpret the results

  13. User Manual for the PROTEUS Mesh Tools

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Micheal A. [Argonne National Lab. (ANL), Argonne, IL (United States); Shemon, Emily R [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-19

    PROTEUS is built around a finite element representation of the geometry for visualization. In addition, the PROTEUS-SN solver was built to solve the even-parity transport equation on a finite element mesh provided as input. Similarly, PROTEUS-MOC and PROTEUS-NEMO were built to apply the method of characteristics on unstructured finite element meshes. Given the complexity of real world problems, experience has shown that using commercial mesh generator to create rather simple input geometries is overly complex and slow. As a consequence, significant effort has been put into place to create multiple codes that help assist in the mesh generation and manipulation. There are three input means to create a mesh in PROTEUS: UFMESH, GRID, and NEMESH. At present, the UFMESH is a simple way to generate two-dimensional Cartesian and hexagonal fuel assembly geometries. The UFmesh input allows for simple assembly mesh generation while the GRID input allows the generation of Cartesian, hexagonal, and regular triangular structured grid geometry options. The NEMESH is a way for the user to create their own mesh or convert another mesh file format into a PROTEUS input format. Given that one has an input mesh format acceptable for PROTEUS, we have constructed several tools which allow further mesh and geometry construction (i.e. mesh extrusion and merging). This report describes the various mesh tools that are provided with the PROTEUS code giving both descriptions of the input and output. In many cases the examples are provided with a regression test of the mesh tools. The most important mesh tools for any user to consider using are the MT_MeshToMesh.x and the MT_RadialLattice.x codes. The former allows the conversion between most mesh types handled by PROTEUS while the second allows the merging of multiple (assembly) meshes into a radial structured grid. Note that the mesh generation process is recursive in nature and that each input specific for a given mesh tool (such as .axial

  14. Hanford inventory program user's manual

    International Nuclear Information System (INIS)

    Hinkelman, K.C.

    1994-01-01

    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

  15. IDSE Version 1 User's Manual

    Science.gov (United States)

    Mayer, Richard

    1988-01-01

    The integrated development support environment (IDSE) is a suite of integrated software tools that provide intelligent support for information modelling. These tools assist in function, information, and process modeling. Additional tools exist to assist in gathering and analyzing information to be modeled. This is a user's guide to application of the IDSE. Sections covering the requirements and design of each of the tools are presented. There are currently three integrated computer aided manufacturing definition (IDEF) modeling methodologies: IDEF0, IDEF1, and IDEF2. Also, four appendices exist to describe hardware and software requirements, installation procedures, and basic hardware usage.

  16. MARS code manual volume I: code structure, system models, and solution methods

    International Nuclear Information System (INIS)

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

    2010-02-01

    Korea Advanced Energy Research Institute (KAERI) conceived and started the development of MARS code with the main objective of producing a state-of-the-art realistic thermal hydraulic systems analysis code with multi-dimensional analysis capability. MARS achieves this objective by very tightly integrating the one dimensional RELAP5/MOD3 with the multi-dimensional COBRA-TF codes. The method of integration of the two codes is based on the dynamic link library techniques, and the system pressure equation matrices of both codes are implicitly integrated and solved simultaneously. In addition, the Equation-Of-State (EOS) for the light water was unified by replacing the EOS of COBRA-TF by that of the RELAP5. This theory manual provides a complete list of overall information of code structure and major function of MARS including code architecture, hydrodynamic model, heat structure, trip / control system and point reactor kinetics model. Therefore, this report would be very useful for the code users. The overall structure of the manual is modeled on the structure of the RELAP5 and as such the layout of the manual is very similar to that of the RELAP. This similitude to RELAP5 input is intentional as this input scheme will allow minimum modification between the inputs of RELAP5 and MARS3.1. MARS3.1 development team would like to express its appreciation to the RELAP5 Development Team and the USNRC for making this manual possible

  17. 46 CFR 160.176-21 - User manuals.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 6 2010-10-01 2010-10-01 false User manuals. 160.176-21 Section 160.176-21 Shipping...: SPECIFICATIONS AND APPROVAL LIFESAVING EQUIPMENT Inflatable Lifejackets § 160.176-21 User manuals. (a) The manufacturer must develop a user's manual for each model of inflatable lifejacket. The content of the manual...

  18. MARS code manual volume II: input requirements

    International Nuclear Information System (INIS)

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

    2010-02-01

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

  19. User Manual and Supporting Information for Library of Codes for Centroidal Voronoi Point Placement and Associated Zeroth, First, and Second Moment Determination; TOPICAL

    International Nuclear Information System (INIS)

    BURKARDT, JOHN; GUNZBURGER, MAX; PETERSON, JANET; BRANNON, REBECCA M.

    2002-01-01

    The theory, numerical algorithm, and user documentation are provided for a new ''Centroidal Voronoi Tessellation (CVT)'' method of filling a region of space (2D or 3D) with particles at any desired particle density. ''Clumping'' is entirely avoided and the boundary is optimally resolved. This particle placement capability is needed for any so-called ''mesh-free'' method in which physical fields are discretized via arbitrary-connectivity discrete points. CVT exploits efficient statistical methods to avoid expensive generation of Voronoi diagrams. Nevertheless, if a CVT particle's Voronoi cell were to be explicitly computed, then it would have a centroid that coincides with the particle itself and a minimized rotational moment. The CVT code provides each particle's volume and centroid, and also the rotational moment matrix needed to approximate a particle by an ellipsoid (instead of a simple sphere). DIATOM region specification is supported

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

    Energy Technology Data Exchange (ETDEWEB)

    Nakka, B W; Chan, T

    1994-12-01

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

  1. C-TIC Console Operator's User Manual

    Science.gov (United States)

    1996-07-01

    The C-TIC Console Operator's User Manual is designed to assist the operator at : the Corridor Transportation Information Center with the navigation and use of : the application programs in the C-TIC. This document will concentrate solely on : the ext...

  2. National Radiobiology Archives Distributed Access user's manual

    International Nuclear Information System (INIS)

    Watson, C.; Smith, S.; Prather, J.

    1991-11-01

    This User's Manual describes installation and use of the National Radiobiology Archives (NRA) Distributed Access package. The package consists of a distributed subset of information representative of the NRA databases and database access software which provide an introduction to the scope and style of the NRA Information Systems

  3. User's manual about visualizing debugger, vdebug

    International Nuclear Information System (INIS)

    Matsuda, Katsuyuki; Takemiya, Hiroshi; Kawasaki, Takuji

    1998-12-01

    We developed a visualizing debugger, vdebug, which works in cooperation with the visualization system AVS and source level debuggers (gdb, dbx, and ladebug), whereby we can visualize any array data without recompiling, and find errors in programs efficiently. This report is user's manual about a visualizing debugger, vdebug, and describes its usage and examples. (author)

  4. Interlibrary Loan Communications Subsystem: Users Manual.

    Science.gov (United States)

    OCLC Online Computer Library Center, Inc., Dublin, OH.

    The OCLC Interlibrary Loan (ILL) Communications Subsystem provides participating libraries with on-line control of ILL transactions. This user manual includes a glossary of terms related to the procedures in using the system. Sections describe computer entry, searching, loan request form, loan response form, ILL procedures, the special message…

  5. User manual Dieka PreProcessor

    NARCIS (Netherlands)

    Valkering, Kasper

    2000-01-01

    This is the user manual belonging to the Dieka-PreProcessor. This application was written by Wenhua Cao and revised and expanded by Kasper Valkering. The aim of this preproccesor is to be able to draw and mesh extrusion dies in ProEngineer, and do the FE-calculation in Dieka. The preprocessor makes

  6. Sample problem manual for benchmarking of cask analysis codes

    International Nuclear Information System (INIS)

    Glass, R.E.

    1988-02-01

    A series of problems have been defined to evaluate structural and thermal codes. These problems were designed to simulate the hypothetical accident conditions given in Title 10 of the Code of Federal Regulation, Part 71 (10CFR71) while retaining simple geometries. This produced a problem set that exercises the ability of the codes to model pertinent physical phenomena without requiring extensive use of computer resources. The solutions that are presented are consensus solutions based on computer analyses done by both national laboratories and industry in the United States, United Kingdom, France, Italy, Sweden, and Japan. The intent of this manual is to provide code users with a set of standard structural and thermal problems and solutions which can be used to evaluate individual codes. 19 refs., 19 figs., 14 tabs

  7. Hanford Environmental Information System (HEIS) user's manual

    International Nuclear Information System (INIS)

    1991-10-01

    The Hanford Environmental Information System (HEIS) is a consolidated set of automated resources that effectively manage the data gathered during environmental monitoring and restoration of the Hanford Site. The HEIS includes an integrated database that provides consistent and current data to all users and promotes sharing of data by the entire user community. Data stored in the HEIS are collected under several regulatory programs. Currently these include the Comprehensive Environmental Response, Compensation and Liability Act of 1980 (CERCLA); the Resource Conservation and Recovery Act of 1976 (RCRA); and the Ground-Water Environmental Surveillance Project, managed by the Pacific Northwest Laboratory. The HEIS is an information system with an inclusive database. The manual, the HEIS User's Manual, describes the facilities available to the scientist, engineer, or manager who uses the system for environmental monitoring, assessment, and restoration planning; and to the regulator who is responsible for reviewing Hanford Site operations against regulatory requirements and guidelines

  8. Wien Automatic System Package (WASP). A computer code for power generating system expansion planning. Version WASP-III Plus. User's manual. Volume 2: Appendices

    International Nuclear Information System (INIS)

    1995-01-01

    With several Member States, the IAEA has completed a new version of the WASP program, which has been called WASP-Ill Plus since it follows quite closely the methodology of the WASP-Ill model. The major enhancements in WASP-Ill Plus with respect to the WASP-Ill version are: increase in the number of thermal fuel types (from 5 to 10); verification of which configurations generated by CONGEN have already been simulated in previous iterations with MERSIM; direct calculation of combined Loading Order of FIXSYS and VARSYS plants; simulation of system operation includes consideration of physical constraints imposed on some fuel types (i.e., fuel availability for electricity generation); extended output of the resimulation of the optimal solution; generation of a file that can be used for graphical representation of the results of the resimulation of the optimal solution and cash flows of the investment costs; calculation of cash flows allows to include the capital costs of plants firmly committed or in construction (FIXSYS plants); user control of the distribution of capital cost expenditures during the construction period (if required to be different from the general 'S' curve distribution used as default). This second volume of the document to support use of the WASP-Ill Plus computer code consists of 5 appendices giving some additional information about the WASP-Ill Plus program. Appendix A is mainly addressed to the WASP-Ill Plus system analyst and supplies some information which could help in the implementation of the program on the user computer facilities. This appendix also includes some aspects about WASP-Ill Plus that could not be treated in detail in Chapters 1 to 11. Appendix B identifies all error and warning messages that may appear in the WASP printouts and advises the user how to overcome the problem. Appendix C presents the flow charts of the programs along with a brief description of the objectives and structure of each module. Appendix D describes the

  9. Hydrogen Mixing Studies (HMS), user's manual

    International Nuclear Information System (INIS)

    Lam, K.L.; Wilson, T.L.; Travis, J.R.

    1994-12-01

    Hydrogen Mixing Studies (HMS) is a best-estimate analysis tool for predicting the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containments and other facilities. It can model geometrically complex facilities having multiple compartments and internal structures. The code can simulate the effects of steam condensation, heat transfer to walls and internal structures, chemical kinetics, and fluid turbulence. The gas mixture may consist of components included in a built-in library of 20 species. HMS is a finite-volume computer code that solves the time-dependent, three-dimensional (3D) compressible Navier Stokes equations. Both Cartesian and cylindrical coordinate systems are available. Transport equations for the fluid internal energy and for gas species densities are also solved. HMS was originally developed to run on Cray-type supercomputers with vector-processing units that greatly improve the computational speed, especially for large, complex problems. Recently the code has been converted to run on Sun workstations. Both the Cray and Sun versions have the same built-in graphics capabilities that allow 1D, 2D, 3D, and time-history plots of all solution variables. Other code features include a restart capability and flexible definitions of initial and time-dependent boundary conditions. This manual describes how to use the code. It explains how to set up the model geometry, define walls and obstacles, and specify gas species and material properties. Definitions of initial and boundary conditions are also described. The manual also describes various physical model and numerical procedure options, as well as how to turn them on. The reader also learns how to specify different outputs, especially graphical display of solution variables. Finally sample problems are included to illustrate some applications of the code. An input deck that illustrates the minimum required data to run HMS is given at the end of this manual

  10. Rooftop Unit Comparison Calculator User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Miller, James D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-04-30

    This document serves as a user manual for the Packaged rooftop air conditioners and heat pump units comparison calculator (RTUCC) and is an aggregation of the calculator’s website documentation. Content ranges from new-user guide material like the “Quick Start” to the more technical/algorithmic descriptions of the “Methods Pages.” There is also a section listing all the context-help topics that support the features on the “Controls” page. The appendix has a discussion of the EnergyPlus runs that supported the development of the building-response models.

  11. PETSc Users Manual Revision 3.7

    Energy Technology Data Exchange (ETDEWEB)

    Balay, S.; Brune, P.; Buschelman, K.; Gropp, W.; Karpeyev, D.; Kaushik, D.; Knepley, M.; McInnes, L. Curfman; Rupp, K.; Smith, B.; Zhang, H.; Abhyankar, S.; Adams, M.; Dalcin, L.; Zampini, S.; Zhang, H.

    2016-04-01

    This manual describes the use of PETSc for the numerical solution of partial differential equations and related problems on high-performance computers. The Portable, Extensible Toolkit for Scientific Computation (PETSc) is a suite of data structures and routines that provide the building blocks for the implementation of large-scale application codes on parallel (and serial) computers. PETSc uses the MPI standard for all message-passing communication.

  12. PETSc Users Manual Revision 3.8

    Energy Technology Data Exchange (ETDEWEB)

    Balay, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Abhyankar, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Adams, M. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Brown, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Brune, P. [Argonne National Lab. (ANL), Argonne, IL (United States); Buschelman, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Dalcin, L. D. [King Abdullah Univ. of Science and Technology, Thuwal (Saudi Arabia); Eijkhout, V. [Univ. of Texas, Austin, TX (United States); Gropp, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Kaushik, D. [Argonne National Lab. (ANL), Argonne, IL (United States); Knepley, M. [Argonne National Lab. (ANL), Argonne, IL (United States); May, D. [ETH Zurich (Switzerland); McInnes, L. Curfman [Argonne National Lab. (ANL), Argonne, IL (United States); Munson, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Rupp, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Sanan, P. [Univ. of Italian Switzerland, Lugano (Switzerland); Smith, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Zampini, S. [King Abdullah Univ. of Science and Technology, Thuwal (Saudi Arabia); Zhang, H. [Illinois Inst. of Technology, Chicago, IL (United States); Zhang, H. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-09-01

    This manual describes the use of PETSc for the numerical solution of partial differential equations and related problems on high-performance computers. The Portable, Extensible Toolkit for Scientific Computation (PETSc) is a suite of data structures and routines that provide the building blocks for the implementation of large-scale application codes on parallel (and serial) computers. PETSc uses the MPI standard for all message-passing communication.

  13. SIMone user's manual. V. 2.1

    International Nuclear Information System (INIS)

    Silk, M.

    1990-03-01

    Simone is a computer application program which provides an environment in which to run the nuclear reactor thermal hydraulic simulation code RELAP5. It enables users to run a RELAP5 calculation on-line on a SUN4 workstation, and provides an interactive means of executing operator-like actions. In addition, a version of the graphical post-processing tool ISOVU is incorporated, allowing a real-time display of the RELAP5 calculation to be presented in the form of a system mimic diagram. This report comprises a description of the facilities available within the Simone application, and a user guide describing the user interface. (author)

  14. A deterministic-probabilistic model for contaminant transport. User manual

    Energy Technology Data Exchange (ETDEWEB)

    Schwartz, F W; Crowe, A

    1980-08-01

    This manual describes a deterministic-probabilistic contaminant transport (DPCT) computer model designed to simulate mass transfer by ground-water movement in a vertical section of the earth's crust. The model can account for convection, dispersion, radioactive decay, and cation exchange for a single component. A velocity is calculated from the convective transport of the ground water for each reference particle in the modeled region; dispersion is accounted for in the particle motion by adding a readorn component to the deterministic motion. The model is sufficiently general to enable the user to specify virtually any type of water table or geologic configuration, and a variety of boundary conditions. A major emphasis in the model development has been placed on making the model simple to use, and information provided in the User Manual will permit changes to the computer code to be made relatively easily for those that might be required for specific applications. (author)

  15. A-THREE: a user's manual

    International Nuclear Information System (INIS)

    Auerbach, E.H.

    1977-06-01

    A-THREE is an optical-model code for elastic scattering, specifically designed for the requirements of heavy-ion projectiles. Fast search procedures for fitting model parameters to experimental data are provided. Because of structural similarity, provision for bound-state calculation is also included. In a sense, this code is ''grandson of ABACUS-2;'' it has, however, been written completely anew. Principal features include the ability to calculate to high partial waves (up to l = 800), refinement of integration meshes for large wave number, and a variety of forms for the optical potentials. The very superior Manchester Coulomb Wave Functions Routine, good in all parts of the (rho, eta) plane is used. This code, written in Fortran-extended, has been in use at Brookhaven on its CDC-7600 for the last two years, during which a number of minor improvements have been added. A general overview of the principal features of the code is given first, with some recommendations for input. This overview is intended to guide the user, and is not intended as a detailed description of the code. Then follow sections describing the input cards and the output generated by the code. These sections are sufficiently detailed to permit a user to prepare input without reference to a code listing. Finally, material necessary for writing special potential well subroutines and tables of control and parameter numbers are appended

  16. User's manual for THYDE-P1

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    1982-04-01

    THYDE-P1 is a computer code applicable to LWR (light water reactor) plant dynamics in response to various disturbances. This work is the user's mannual of THYDE-P1 (version SV02L03). The input requirements, steady state adjustment, execution of runs and output specifications are described. (author)

  17. User manual of Visual Balan V. 1.0 Interactive code for water balances and refueling estimation; Manual del usuario del programa Visual Balan V. 1.0. Codigo interactivo para la realizacion de balances hidrologicos y la estimacion de la recarga

    Energy Technology Data Exchange (ETDEWEB)

    Samper, J.; Huguer, L.; Ares, J.; Garcia, M. A. [Universidad de La Coruna (Spain)

    1999-07-01

    This document contains the Users Manual of Visual Balan V1.0, an updated version of Visual Balan V0.0 (Samper et al., 1997). Visual Balan V1.0 performs daily water balances in the soil, the unsaturated zone and the aquifer in a user-friendly environment which facilitates both the input data process and the postprocessing of results. The main inputs of the balance are rainfall and irrigation while the outputs are surface runoff, evapotranspiration, interception, inter flow and groundwater flow. The code evaluates all these components in a sequential manner by starting with rainfall and irrigation, which must be provided by the user, and continuing with interception, surface runoff, evapotranspiration, and potential recharge (water flux crossing the bottom of the soil). This potential recharge is the input to the unsaturated zone where water can flow horizontally as subsurface flow (inter flow) or vertically as percolation into the aquifer. (Author)

  18. PANTHERMIX (PANTHER-THERMIX). User manual

    International Nuclear Information System (INIS)

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

    1998-06-01

    The PANTHER code calculates steady-state or time-dependent power distribution in a reactor with a given temperature distribution. The THERMIX-DIREKT code calculates temperature and coolant flow distributions, in steady-state or transient mode, in a system with a given power distribution. It is described how to use the combination of the general purpose modular reactor code PANTHER and the HTR thermal hydraulics code THERMIX-DIREKT. An earlier version of PANTHERMIX consisted of THERMIX-DIREKT plus 2 conversion programs. The jobs and scripts to be edited by the user were very complex in their interactions. Therefore this version of PANTHERMIX has been extended with macros that take care of all these interactions, so the interaction parts of the jobs become much less complex. 6 refs

  19. User's Manual for LEWICE Version 3.2

    Science.gov (United States)

    Wright, William

    2008-01-01

    A research project is underway at NASA Glenn to produce a computer code which can accurately predict ice growth under a wide range of meteorological conditions for any aircraft surface. This report will present a description of the code inputs and outputs from version 3.2 of this software, which is called LEWICE. This version differs from release 2.0 due to the addition of advanced thermal analysis capabilities for de-icing and anti-icing applications using electrothermal heaters or bleed air applications, the addition of automated Navier-Stokes analysis, an empirical model for supercooled large droplets (SLD) and a pneumatic boot option. An extensive effort was also undertaken to compare the results against the database of electrothermal results which have been generated in the NASA Glenn Icing Research Tunnel (IRT) as was performed for the validation effort for version 2.0. This report will primarily describe the features of the software related to the use of the program. Appendix A has been included to list some of the inner workings of the software or the physical models used. This information is also available in the form of several unpublished documents internal to NASA. This report is intended as a replacement for all previous user manuals of LEWICE. In addition to describing the changes and improvements made for this version, information from previous manuals may be duplicated so that the user will not need to consult previous manuals to use this software.

  20. HTGR Application Economic Model Users' Manual

    International Nuclear Information System (INIS)

    Gandrik, A.M.

    2012-01-01

    The High Temperature Gas-Cooled Reactor (HTGR) Application Economic Model was developed at the Idaho National Laboratory for the Next Generation Nuclear Plant Project. The HTGR Application Economic Model calculates either the required selling price of power and/or heat for a given internal rate of return (IRR) or the IRR for power and/or heat being sold at the market price. The user can generate these economic results for a range of reactor outlet temperatures; with and without power cycles, including either a Brayton or Rankine cycle; for the demonstration plant, first of a kind, or nth of a kind project phases; for up to 16 reactor modules; and for module ratings of 200, 350, or 600 MWt. This users manual contains the mathematical models and operating instructions for the HTGR Application Economic Model. Instructions, screenshots, and examples are provided to guide the user through the HTGR Application Economic Model. This model was designed for users who are familiar with the HTGR design and Excel and engineering economics. Modification of the HTGR Application Economic Model should only be performed by users familiar with the HTGR and its applications, Excel, and Visual Basic.

  1. HOPSPACK 2.0 user manual.

    Energy Technology Data Exchange (ETDEWEB)

    Plantenga, Todd D.

    2009-09-01

    HOPSPACK (Hybrid Optimization Parallel Search PACKage) solves derivative-free optimization problems using an open source, C++ software framework. The framework enables parallel operation using MPI or multithreading, and allows multiple solvers to run simultaneously and interact to find solution points. HOPSPACK comes with an asynchronous pattern search solver that handles general optimization problems with linear and nonlinear constraints, and continuous and integer-valued variables. This user manual explains how to install and use HOPSPACK to solve problems, and how to create custom solvers within the framework.

  2. GASFLOW-MPI. A scalable computational fluid dynamics code for gases, aerosols and combustion. Vol. 2. Users' manual (Revision 1.0)

    Energy Technology Data Exchange (ETDEWEB)

    Xiao, Jianjun; Travis, Jack; Royl, Peter; Necker, Gottfried; Svishchev, Anatoly; Jordan, Thomas

    2016-07-01

    Karlsruhe Institute of Technology (KIT) is developing the parallel computational fluid dynamics code GASFLOW-MPI as a best-estimate tool for predicting transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containments and other facility buildings. GASFLOW-MPI is a finite-volume code based on proven computational fluid dynamics methodology that solves the compressible Navier-Stokes equations for three-dimensional volumes in Cartesian or cylindrical coordinates.

  3. STAT, GAPS, STRAIN, DRWDIM: a system of computer codes for analyzing HTGR fuel test element metrology data. User's manual

    Energy Technology Data Exchange (ETDEWEB)

    Saurwein, J.J.

    1977-08-01

    A system of computer codes has been developed to statistically reduce Peach Bottom fuel test element metrology data and to compare the material strains and fuel rod-fuel hole gaps computed from these data with HTGR design code predictions. The codes included in this system are STAT, STRAIN, GAPS, and DRWDIM. STAT statistically evaluates test element metrology data yielding fuel rod, fuel body, and sleeve irradiation-induced strains; fuel rod anisotropy; and additional data characterizing each analyzed fuel element. STRAIN compares test element fuel rod and fuel body irradiation-induced strains computed from metrology data with the corresponding design code predictions. GAPS compares test element fuel rod, fuel hole heat transfer gaps computed from metrology data with the corresponding design code predictions. DRWDIM plots the measured and predicted gaps and strains. Although specifically developed to expedite the analysis of Peach Bottom fuel test elements, this system can be applied, without extensive modification, to the analysis of Fort St. Vrain or other HTGR-type fuel test elements.

  4. User's manual to the ICRP Code: a series of computer programs to perform dosimetric calculations for the ICRP Committee 2 report

    International Nuclear Information System (INIS)

    Watson, S.B.; Ford, M.R.

    1980-02-01

    A computer code has been developed that implements the recommendations of ICRP Committee 2 for computing limits for occupational exposure of radionuclides. The purpose of this report is to describe the various modules of the computer code and to present a description of the methods and criteria used to compute the tables published in the Committee 2 report. The computer code contains three modules of which: (1) one computes specific effective energy; (2) one calculates cumulated activity; and (3) one computes dose and the series of ICRP tables. The description of the first two modules emphasizes the new ICRP Committee 2 recommendations in computing specific effective energy and cumulated activity. For the third module, the complex criteria are discussed for calculating the tables of committed dose equivalent, weighted committed dose equivalents, annual limit of intake, and derived air concentration

  5. User's manual to the ICRP Code: a series of computer programs to perform dosimetric calculations for the ICRP Committee 2 report

    Energy Technology Data Exchange (ETDEWEB)

    Watson, S.B.; Ford, M.R.

    1980-02-01

    A computer code has been developed that implements the recommendations of ICRP Committee 2 for computing limits for occupational exposure of radionuclides. The purpose of this report is to describe the various modules of the computer code and to present a description of the methods and criteria used to compute the tables published in the Committee 2 report. The computer code contains three modules of which: (1) one computes specific effective energy; (2) one calculates cumulated activity; and (3) one computes dose and the series of ICRP tables. The description of the first two modules emphasizes the new ICRP Committee 2 recommendations in computing specific effective energy and cumulated activity. For the third module, the complex criteria are discussed for calculating the tables of committed dose equivalent, weighted committed dose equivalents, annual limit of intake, and derived air concentration.

  6. Clean Lead Facility Inventory System user's manual

    International Nuclear Information System (INIS)

    Garcia, J.F.

    1994-12-01

    The purpose of this user's manual is to provide instruction and guidance needed to enter and maintain inventory information for the Clean Lead Facility (CLF), PER-612. Individuals responsible for maintaining and using the system should study and understand the information provided. The user's manual describes how to properly use and maintain the CLF Inventory System. Annual, quarterly, monthly, and current inventory reports may be printed from the Inventory System for reporting purposes. Profile reports of each shipment of lead may also be printed for verification and documentation of lead transactions. The CLF Inventory System was designed on Microsoft Access version 2.0. Similar inventory systems are in use at the Idaho National Engineering Laboratory (INEL) to facilitate site-wide compilations of mixed waste data. The CLF Inventory System was designed for inventorying the clean or non-radioactive contaminated lead stored at the CLF. This data, along with the mixed waste data, will be compiled into the Idaho Mixed Waste Information (IMWI) system for reporting to the Department of Energy Idaho Office, Department of Energy Headquarters, and/or the State of Idaho

  7. Explosives Classifications Tracking System User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Genoni, R.P.

    1993-10-01

    The Explosives Classification Tracking System (ECTS) presents information and data for U.S. Department of Energy (DOE) explosives classifications of interest to EM-561, Transportation Management Division, other DOE facilities, and contractors. It is intended to be useful to the scientist, engineer, and transportation professional, who needs to classify or transport explosives. This release of the ECTS reflects upgrading of the software which provides the user with an environment that makes comprehensive retrieval of explosives related information quick and easy. Quarterly updates will be provided to the ECTS throughout its development in FY 1993 and thereafter. The ECTS is a stand alone, single user system that contains unclassified, publicly available information, and administrative information (contractor names, product descriptions, transmittal dates, EX-Numbers, etc.) information from many sources for non-decisional engineering and shipping activities. The data is the most up-to-date and accurate available to the knowledge of the system developer. The system is designed to permit easy revision and updating as new information and data become available. These, additions and corrections are welcomed by the developer. This user manual is intended to help the user install, understand, and operate the system so that the desired information may be readily obtained, reviewed, and reported.

  8. The WECHSL-Mod3 code: A computer program for the interaction of a core melt with concrete including the long term behavior. Model description and user's manual

    International Nuclear Information System (INIS)

    Foit, J.J.; Adroguer, B.; Cenerino, G.; Stiefel, S.

    1995-02-01

    The WECHSL-Mod3 code is a mechanistic computer code developed for the analysis of the thermal and chemical interaction of initially molten reactor materials with concrete in a two-dimensional as well as in a one-dimensional, axisymmetrical concrete cavity. The code performs calculations from the time of initial contact of a hot molten pool over start of solidification processes until long term basemat erosion over several days with the possibility of basemat penetration. It is assumed that an underlying metallic layer exists covered by an oxidic layer or that only one oxidic layer is present which can contain a homogeneously dispersed metallic phase. Heat generation in the melt is by decay heat and chemical reactions from metal oxidation. Energy is lost to the melting concrete and to the upper containment by radiation or evaporation of sumpwater possibly flooding the surface of the melt. Thermodynamic and transport properties as well as criteria for heat transfer and solidification processes are internally calculated for each time step. Heat transfer is modelled taking into account the high gas flux from the decomposing concrete and the heat conduction in the crusts possibly forming in the long term at the melt/concrete interface. The CALTHER code (developed at CEA, France) which models the radiative heat transfer from the upper surface of the corium melt to the surrounding cavity is implemented in the present WECHSL version. The WECHSL code in its present version was validated by the BETA, ACE and SURC experiments. The test samples include a BETA and the SURC2 post test calculations and a WECHSL application to a reactor accident. (orig.) [de

  9. Maintenance Personnel Performance Simulation (MAPPS) model. Users' Manual

    International Nuclear Information System (INIS)

    Kopstein, F.F.; Wolf, J.J.

    1985-09-01

    This report (MAPPS User's Manual) is the last report to be published from this program and provides detailed guidelines for utilization of the MAPPS model. Although the model has been developed to be highly user-friendly and provides interactive means for controlling and running of the model, the user's manual is provided as a guide for the user in the event clarification or direction is required. The user will find that in general the model requires primarily user input that is self explanatory. Once initial familiarization with the model has been achieved by the user, the amount of interaction between the user's manual and the computer model will be minimal. It is suggested however that even the experienced user keep the user's manual handy for quick reference. 5 refs., 10 figs., 7 tabs

  10. SPARK Version 1.1 user manual

    International Nuclear Information System (INIS)

    Weissenburger, D.W.

    1988-01-01

    This manual describes the input required to use Version 1.1 of the SPARK computer code. SPARK 1.1 is a library of FORTRAN main programs and subprograms designed to calculate eddy currents on conducting surfaces where current flow is assumed zero in the direction normal to the surface. Surfaces are modeled with triangular and/or quadrilateral elements. Lorentz forces produced by the interaction of eddy currents with background magnetic fields can be output at element nodes in a form compatible with most structural analysis codes. In addition, magnetic fields due to eddy currents can be determined at points off the surface. Version 1.1 features eddy current streamline plotting with optional hidden-surface-removal graphics and topological enhancements that allow essentially any orientable surface to be modeled. SPARK also has extensive symmetry specification options. In order to make the manual as self-contained as possible, six appendices are included that present summaries of the symmetry options, topological options, coil options and code algorithms, with input and output examples. An edition of SPARK 1.1 is available on the Cray computers at the National Magnetic Fusion Energy Computer Center at Livermore, California. Another more generic edition is operational on the VAX computers at the Princeton Plasma Physics Laboratory and is available on magnetic tape by request. The generic edition requires either the GKS or PLOT10 graphics package and the IMSL or NAG mathematical package. Requests from outside the United States will be subject to applicable federal regulations regarding dissemination of computer programs. 22 refs

  11. SPARK Version 1. 1 user manual

    Energy Technology Data Exchange (ETDEWEB)

    Weissenburger, D.W.

    1988-01-01

    This manual describes the input required to use Version 1.1 of the SPARK computer code. SPARK 1.1 is a library of FORTRAN main programs and subprograms designed to calculate eddy currents on conducting surfaces where current flow is assumed zero in the direction normal to the surface. Surfaces are modeled with triangular and/or quadrilateral elements. Lorentz forces produced by the interaction of eddy currents with background magnetic fields can be output at element nodes in a form compatible with most structural analysis codes. In addition, magnetic fields due to eddy currents can be determined at points off the surface. Version 1.1 features eddy current streamline plotting with optional hidden-surface-removal graphics and topological enhancements that allow essentially any orientable surface to be modeled. SPARK also has extensive symmetry specification options. In order to make the manual as self-contained as possible, six appendices are included that present summaries of the symmetry options, topological options, coil options and code algorithms, with input and output examples. An edition of SPARK 1.1 is available on the Cray computers at the National Magnetic Fusion Energy Computer Center at Livermore, California. Another more generic edition is operational on the VAX computers at the Princeton Plasma Physics Laboratory and is available on magnetic tape by request. The generic edition requires either the GKS or PLOT10 graphics package and the IMSL or NAG mathematical package. Requests from outside the United States will be subject to applicable federal regulations regarding dissemination of computer programs. 22 refs.

  12. Electronic manual of the nuclear characteristics analysis code-set for FBR

    International Nuclear Information System (INIS)

    Makino, Tohru

    2001-03-01

    Reactor Physics Gr., System Engineering Technology Division, O-arai Engineering Center has consolidated the nuclear design database to improve analytical methods and prediction accuracy for large fast breeder cores such as demonstration or commercial FBRs from the previous research. The up-to-date information about usage of the nuclear characteristics analysis code-set was compiled as a part of the improvement of basic design data base for FBR core. The outlines of the electronic manual are as follows; (1) The electronic manual includes explanations of following codes: JOINT : Code Interface Program. SLAROM, CASUP : Effective Cross Section Calculation Code. CITATION-FBR : Diffusion Analysis Code. PERKY : Perturbative Diffusion Analysis Code. SNPERT, SNPERT-3D : Perturbative Transport Analysis Code. SAGEP, SAGEP-3D : Sensitivity Coefficient Calculation Code. NSHEX : Transport Analysis Code using Nodal Method. ABLE : Cross Section Adjustment Calculation Code. ACCEPT : Predicting Accuracy Evaluation Code. (2) The electronic manual is described using HTML file format and PDF file for easy maintenance, updating and for easy referring through JNC Intranet. User can refer manual pages by usual Web browser software without any special setup. (3) Many of manual pages include link-tags to jump to related pages. String search is available in both HTML and PDF documents. (4) User can download source code, sample input data and shell script files to carry out each analysis from download page of each code (JNC inside only). (5) Usage of the electronic manual and maintenance/updating process are described in this report and it makes possible to enroll new code or new information in the electronic manual. Since the information has been taken into account about modifications and error fixings, added to each code after the last consolidation in 1994, the electronic manual would cover most recent status of the nuclear characteristics analysis code-set. One of other advantages of use

  13. TOUGH+HYDRATE v1.2 User's Manual: A Code for the Simulation of System Behavior in Hydrate-Bearing Geologic Media

    Energy Technology Data Exchange (ETDEWEB)

    Moridis, George J. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Kowalsky, Michael B. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Pruess, Karsten [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2012-08-01

    TOUGH+HYDRATE v1.2 is a code for the simulation of the behavior of hydratebearing geologic systems, and represents the second update of the code since its first release [Moridis et al., 2008]. By solving the coupled equations of mass and heat balance, TOUGH+HYDRATE can model the non-isothermal gas release, phase behavior and flow of fluids and heat under conditions typical of common natural CH4-hydrate deposits (i.e., in the permafrost and in deep ocean sediments) in complex geological media at any scale (from laboratory to reservoir) at which Darcy’s law is valid. TOUGH+HYDRATE v1.2 includes both an equilibrium and a kinetic model of hydrate formation and dissociation. The model accounts for heat and up to four mass components, i.e., water, CH4, hydrate, and water-soluble inhibitors such as salts or alcohols. These are partitioned among four possible phases (gas phase, liquid phase, ice phase and hydrate phase). Hydrate dissociation or formation, phase changes and the corresponding thermal effects are fully described, as are the effects of inhibitors. The model can describe all possible hydrate dissociation mechanisms, i.e., depressurization, thermal stimulation, salting-out effects and inhibitor-induced effects. TOUGH+HYDRATE is a member of TOUGH+, the successor to the TOUGH2 [Pruess et al., 1991] family of codes for multi-component, multiphase fluid and heat flow developed at the Lawrence Berkeley National Laboratory. It is written in standard FORTRAN 95/2003, and can be run on any computational platform (workstation, PC, Macintosh) for which such compilers are available.

  14. PC/FRAM, Version 3.2 User Manual

    International Nuclear Information System (INIS)

    Kelley, T.A.; Sampson, T.E.

    1999-01-01

    This manual describes the use of version 3.2 of the PC/FRAM plutonium isotopic analysis software developed in the Safeguards Science and Technology Group, NE-5, Nonproliferation and International Security Division Los Alamos National Laboratory. The software analyzes the gamma ray spectrum from plutonium-bearing items and determines the isotopic distribution of the plutonium 241Am content and concentration of other isotopes in the item. The software can also determine the isotopic distribution of uranium isotopes in items containing only uranium. The body of this manual descenies the generic version of the code. Special facility-specific enhancements, if they apply, will be described in the appendices. The information in this manual applies equally well to version 3.3, which has been licensed to ORTEC. The software can analyze data that is stored in a file on disk. It understands several storage formats including Canberra's S1OO format, ORTEC'S 'chn' and 'SPC' formats, and several ASCII text formats. The software can also control data acquisition using an MCA and then store the results in a file on disk for later analysis or analyze the spectrum directly after the acquisition. The software currently only supports the control of ORTEC MCB'S. Support for Canbema's Genie-2000 Spectroscopy Systems will be added in the future. Support for reading and writing CAM files will also be forthcoming. A versatile parameter fde database structure governs all facets of the data analysis. User editing of the parameter sets allows great flexibility in handling data with different isotopic distributions, interfering isotopes, and different acquisition parameters such as energy calibration, and detector type. This manual is intended for the system supervisor or the local user who is to be the resident expert. Excerpts from this manual may also be appropriate for the system operator who will routinely use the instrument

  15. Finite element program Lamcal. (User's manual)

    International Nuclear Information System (INIS)

    Lamain, L.G.; Blanckenburg, J.F.G.

    1982-01-01

    The present user's manual gives the input formats, job control and an input example for the finite element part of the Lamcal program. The input data have been organized in a more or less self explaining way, using keywords and standard input formats and is printed at the beginning of every run. To simplify the use of the whole program and to avoid unecessary data handling, all three parts of the Lamcal program, meshgeneration, plotting and, FE, are combined into one load module. This setup allows to do all calculations in one single run. However, preprocessing, postprocessing and restarts can be made in separate runs as well. The same reserved space for the dynamic core storage is used in all three parts, if the available space is not sufficient the FE program will stop

  16. TOUGH+Hydrate v1.0 User's Manual: A Code for the Simulation of System Behavior in Hydrate-Bearing Geologic Media

    Energy Technology Data Exchange (ETDEWEB)

    Moridis, George; Moridis, George J.; Kowalsky, Michael B.; Pruess, Karsten

    2008-03-01

    TOUGH+HYDRATE v1.0 is a new code for the simulation of the behavior of hydrate-bearing geologic systems. By solving the coupled equations of mass and heat balance, TOUGH+HYDRATE can model the non-isothermal gas release, phase behavior and flow of fluids and heat under conditions typical of common natural CH{sub 4}-hydrate deposits (i.e., in the permafrost and in deep ocean sediments) in complex geological media at any scale (from laboratory to reservoir) at which Darcy's law is valid. TOUGH+HYDRATE v1.0 includes both an equilibrium and a kinetic model of hydrate formation and dissociation. The model accounts for heat and up to four mass components, i.e., water, CH{sub 4}, hydrate, and water-soluble inhibitors such as salts or alcohols. These are partitioned among four possible phases (gas phase, liquid phase, ice phase and hydrate phase). Hydrate dissociation or formation, phase changes and the corresponding thermal effects are fully described, as are the effects of inhibitors. The model can describe all possible hydrate dissociation mechanisms, i.e., depressurization, thermal stimulation, salting-out effects and inhibitor-induced effects. TOUGH+HYDRATE is the first member of TOUGH+, the successor to the TOUGH2 [Pruess et al., 1991] family of codes for multi-component, multiphase fluid and heat flow developed at the Lawrence Berkeley National Laboratory. It is written in standard FORTRAN 95, and can be run on any computational platform (workstation, PC, Macintosh) for which such compilers are available.

  17. AIMSsim Version 2.3.4 - User Manual

    National Research Council Canada - National Science Library

    Schoenborn, Oliver; Lachance, Patrick; Bahramifarid, Nima

    2008-01-01

    This user manual provides an overview of the software functionality developed to support the empirical investigation of a simulated user interface for an Advanced Integrated Multi-sensor Surveillance (AIMS) system...

  18. AIMSsim Version 2.2.1, User Manual

    National Research Council Canada - National Science Library

    Schoenborn, Oliver

    2007-01-01

    This user manual provides an overview of how to use the software developed to support the empirical investigation of a simulated user interface for an Advanced Integrated Multi sensor Surveillance (AIMS) system...

  19. Probability of pipe fracture in the primary coolant loop of a PWR plant. Volume 9: PRAISE computer code user's manual. Final report

    International Nuclear Information System (INIS)

    Lim, E.Y.

    1981-08-01

    The PRAISE (Piping Reliability Analysis Including Seismic Events) computer code estimates the influence of earthquakes on the probability of failure at a weld joint in the primary coolant system of a pressurized water reactor. Failure, either a through-wall defect (leak) or a complete pipe severance (a large-LOCA), is assumed to be caused by fatigue crack growth of an as-fabricated interior surface circumferential defect. These defects are assumed to be two-dimensional and semi-elliptical in shape. The distribution of initial crack sizes is a function of crack depth and aspect ratio. Crack propagation rates are governed by a Paris-type relationship with separate RMS cyclic stress intensity factors for the depth and length. Both uniform through the wall and radial gradient thermal stresses are included in the calculation of the stress intensity factors. The failure probabilities are estimated by applying Monte Carlo methods to simulate the life histories of the selected weld joint. In order to maximize computational efficiency, a stratified sampling procedure is used to select the initial crack size. Hydrostatic proof test, pre-service inspection, and in-service inspection can be simulated. PRAISE treats the inter-arrival times of operating transients either as a constant or exponentially distributed according to observed or postulated rates. Leak rate and leak detection models are also included. The criterion for complete pipe severance is exceedance of a net section critical stress. Earthquakes of various intensity and arbitrary occurrence times can be modeled. PRAISE presently assumes that exactly one initial defect exists in the weld and that the earthquake of interest is the first earthquake experienced at the reactor

  20. Model for Analysis of Energy Demand (MAED-2). User's manual

    International Nuclear Information System (INIS)

    2007-01-01

    The IAEA has been supporting its Member States in the area of energy planning for sustainable development. Development and dissemination of appropriate methodologies and their computer codes are important parts of this support. This manual has been produced to facilitate the use of the MAED model: Model for Analysis of Energy Demand. The methodology of the MAED model was originally developed by. B. Chateau and B. Lapillonne of the Institute Economique et Juridique de l'Energie (IEJE) of the University of Grenoble, France, and was presented as the MEDEE model. Since then the MEDEE model has been developed and adopted to be appropriate for modelling of various energy demand system. The IAEA adopted MEDEE-2 model and incorporated important modifications to make it more suitable for application in the developing countries, and it was named as the MAED model. The first version of the MAED model was designed for the DOS based system, which was later on converted for the Windows system. This manual presents the latest version of the MAED model. The most prominent feature of this version is its flexibility for representing structure of energy consumption. The model now allows country-specific representations of energy consumption patterns using the MAED methodology. The user can now disaggregate energy consumption according to the needs and/or data availability in her/his country. As such, MAED has now become a powerful tool for modelling widely diverse energy consumption patterns. This manual presents the model in details and provides guidelines for its application

  1. Model for Analysis of Energy Demand (MAED-2). User's manual

    International Nuclear Information System (INIS)

    2006-01-01

    The IAEA has been supporting its Member States in the area of energy planning for sustainable development. Development and dissemination of appropriate methodologies and their computer codes are important parts of this support. This manual has been produced to facilitate the use of the MAED model: Model for Analysis of Energy Demand. The methodology of the MAED model was originally developed by. B. Chateau and B. Lapillonne of the Institute Economique et Juridique de l'Energie (IEJE) of the University of Grenoble, France, and was presented as the MEDEE model. Since then the MEDEE model has been developed and adopted to be appropriate for modelling of various energy demand system. The IAEA adopted MEDEE-2 model and incorporated important modifications to make it more suitable for application in the developing countries, and it was named as the MAED model. The first version of the MAED model was designed for the DOS based system, which was later on converted for the Windows system. This manual presents the latest version of the MAED model. The most prominent feature of this version is its flexibility for representing structure of energy consumption. The model now allows country-specific representations of energy consumption patterns using the MAED methodology. The user can now disaggregate energy consumption according to the needs and/or data availability in her/his country. As such, MAED has now become a powerful tool for modelling widely diverse energy consumption patterns. This manual presents the model in details and provides guidelines for its application

  2. UCB-NE-101 user's manual

    International Nuclear Information System (INIS)

    Lee, W.W.L.

    1989-02-01

    The purpose of this manual is to provide users of UCB-NE-101 with the information necessary to use UCB-NE-101 effectively. UCB-NE-101 calculates the concentration of solubility-limited species as a function of space and time and its mass flux rates from a waste sphere buried in a nuclear waste repository in water-saturated rock. The waste is surrounded by one type of rock, and some distance away, there is another type of rock. The inner layer of rock can be a backfill around a nuclear waste package and the outer layer the natural rock. The mass flux calculated is at the interface of the two layers. The species concentration calculated is in the inner layer. A constant concentration of the species, usually the solubility, is specified at the waste sphere/inner layer interface. Dissolution and transport is governed by the solubility of the species, and diffusion in the porous media. 1 ref., 1 fig

  3. ROMUSE 2.0 User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Khuwaileh, Bassam [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Turinsky, Paul [North Carolina State Univ., Raleigh, NC (United States); Williams, Brian J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-10-04

    ROMUSE (Reduced Order Modeling Based Uncertainty/Sensitivity Estimator) is an effort within the Consortium for Advanced Simulation of Light water reactors (CASL) to provide an analysis tool to be used in conjunction with reactor core simulators, especially the Virtual Environment for Reactor Applications (VERA). ROMUSE is written in C++ and is currently capable of performing various types of parameters perturbations, uncertainty quantification, surrogate models construction and subspace analysis. Version 2.0 has the capability to interface with DAKOTA which gives ROMUSE access to the various algorithms implemented within DAKOTA. ROMUSE is mainly designed to interface with VERA and the Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design (SCALE) [1,2,3], however, ROMUSE can interface with any general model (e.g. python and matlab) with Input/Output (I/O) format that follows the Hierarchical Data Format 5 (HDF5). In this brief user manual, the use of ROMUSE will be overviewed and example problems will be presented and briefly discussed. The algorithms provided here range from algorithms inspired by those discussed in Ref.[4] to nuclear-specific algorithms discussed in Ref. [3].

  4. Database on wind characteristics. Users manual

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, G.C.; Hansen, K.S.

    2001-11-01

    The main objective of IEA R and D Wind Annex XVII - Database on Wind Characteristics - is to provide wind energy planners and designers, as well as the international wind engineering community in general, with easy access to quality controlled measured wind field time series observed in a wide range of environments. The project partners are Sweden, Norway, U.S.A., The Netherlands, Japan and Denmark, with Denmark as the Operating Agent. The reporting of IEA R and D Annex XVII falls in three separate parts. Part one deals with the overall structure and philosophy behind the database (including the applied data quality control procedures), part two accounts in details for the available data in the established database bank and part three is the Users Manual describing the various ways to access and analyse the data. The present report constitutes part three of the Annex XVII reporting and contains a trough description of the available online facilities for identifying, selecting, downloading and handling measured wind field time series and resource data from 'Database on Wind Characteristics'. (au)

  5. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    Energy Technology Data Exchange (ETDEWEB)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.

  6. Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

    International Nuclear Information System (INIS)

    Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.

    1997-12-01

    The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions

  7. Element Load Data Processor (ELDAP) Users Manual

    Science.gov (United States)

    Ramsey, John K., Jr.; Ramsey, John K., Sr.

    2015-01-01

    Often, the shear and tensile forces and moments are extracted from finite element analyses to be used in off-line calculations for evaluating the integrity of structural connections involving bolts, rivets, and welds. Usually the maximum forces and moments are desired for use in the calculations. In situations where there are numerous structural connections of interest for numerous load cases, the effort in finding the true maximum force and/or moment combinations among all fasteners and welds and load cases becomes difficult. The Element Load Data Processor (ELDAP) software described herein makes this effort manageable. This software eliminates the possibility of overlooking the worst-case forces and moments that could result in erroneous positive margins of safety and/or selecting inconsistent combinations of forces and moments resulting in false negative margins of safety. In addition to forces and moments, any scalar quantity output in a PATRAN report file may be evaluated with this software. This software was originally written to fill an urgent need during the structural analysis of the Ares I-X Interstage segment. As such, this software was coded in a straightforward manner with no effort made to optimize or minimize code or to develop a graphical user interface.

  8. Microgravity computing codes. User's guide

    Science.gov (United States)

    1982-01-01

    Codes used in microgravity experiments to compute fluid parameters and to obtain data graphically are introduced. The computer programs are stored on two diskettes, compatible with the floppy disk drives of the Apple 2. Two versions of both disks are available (DOS-2 and DOS-3). The codes are written in BASIC and are structured as interactive programs. Interaction takes place through the keyboard of any Apple 2-48K standard system with single floppy disk drive. The programs are protected against wrong commands given by the operator. The programs are described step by step in the same order as the instructions displayed on the monitor. Most of these instructions are shown, with samples of computation and of graphics.

  9. Manual for IRS Coding. Joint IAEA/NEA International Reporting System for Operating Experience

    International Nuclear Information System (INIS)

    2011-01-01

    The International Reporting System for Operating Experience (IRS) is jointly operated by the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA). In early 2010, the IAEA and OECD/NEA jointly issued the IRS Guidelines, which described the reporting system and process and gave users the necessary elements to enable them to produce IRS reports to a high standard of quality while retaining the effectiveness of the system expected by all Member States operating nuclear power plants. The purpose of the present Manual for IRS Coding is to provide supplementary guidance specifically on the coding element of IRS reports to ensure uniform coding of events that are reported through IRS. This Coding Manual does not supersede the IRS Guidelines, but rather, supports users and preparers in achieving a consistent and high level of quality in their IRS reports. Consistency and high quality in the IRS reports allow stakeholders to search and retrieve specific event information with ease. In addition, well-structured reports also enhance the efficient management of the IRS database. This Coding Manual will give specific guidance on the application of each section of the IRS codes, with examples where necessary, of when and how these codes are to be applied. As this reporting system is owned by the Member States, this manual has been developed and approved by the IRS National Coordinators with the assistance of the IAEA and NEA secretariats

  10. Coding and decoding for code division multiple user communication systems

    Science.gov (United States)

    Healy, T. J.

    1985-01-01

    A new algorithm is introduced which decodes code division multiple user communication signals. The algorithm makes use of the distinctive form or pattern of each signal to separate it from the composite signal created by the multiple users. Although the algorithm is presented in terms of frequency-hopped signals, the actual transmitter modulator can use any of the existing digital modulation techniques. The algorithm is applicable to error-free codes or to codes where controlled interference is permitted. It can be used when block synchronization is assumed, and in some cases when it is not. The paper also discusses briefly some of the codes which can be used in connection with the algorithm, and relates the algorithm to past studies which use other approaches to the same problem.

  11. Equipment Qualification Data Base user manual

    International Nuclear Information System (INIS)

    Decker, Q.R.; Fackrell, L.J.; Fitch, L.R.; Meeky, O.B.

    1985-09-01

    This manual details the Equipment Qualification Data Base (EQDB), its usage, and contents. The EQDB consists of two files; the Plant Qualification File (PQF) and the Equipment Qualification File (EQF). The PQF contains plant specific environmental data and the EQF contains summaries of various test results. Two data management systems are used to manipulate the data and are discussed in this manual. SAS Institute System 2000 (S2K) is the management system for the PQF and Query Update (QU) is the operating system for the EQF. Each management system contains report writers. These writers and how to use them are discussed in detail in this manual

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-15

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

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

    International Nuclear Information System (INIS)

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

    2012-07-01

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

  14. PAVECHECK : training material updated user's manual including GPS.

    Science.gov (United States)

    2009-01-01

    PAVECHECK is a software package used to integrate nondestructive test data from various testing systems to provide the pavement engineer with a comprehensive evaluation of both surface and subsurface conditions. This User's Manual is intended to demo...

  15. Prototype road weather performance management tool : installation instructions & user manual.

    Science.gov (United States)

    2016-07-20

    This document is the Installation Instructions and User Manual for the Road Weather Performance Management (RW-PM) Tool developed for the project on Development and Demonstration of a Prototype Road Weather Performance Management Application that Use...

  16. Users Manual for TMY3 Data Sets (Revised)

    Energy Technology Data Exchange (ETDEWEB)

    Wilcox, S.; Marion, W.

    2008-05-01

    This users manual describes how to obtain and interpret the data in the Typical Meteorological Year version 3 (TMY3) data sets. These data sets are an update to the TMY2 data released by NREL in 1994.

  17. Applying Minimal Manual Principles for Documentation of Graphical User Interfaces.

    Science.gov (United States)

    Nowaczyk, Ronald H.; James, E. Christopher

    1993-01-01

    Investigates the need to include computer screens in documentation for software using a graphical user interface. Describes the uses and purposes of "minimal manuals" and their principles. Studies student reaction to their use of one of three on-screen manuals: screens, icon, and button. Finds some benefit for including icon and button…

  18. COSIS User's Manual

    Energy Technology Data Exchange (ETDEWEB)

    Cho, J. Y.; Lee, K. B.; Koo, B. S.; Lee, W. K.; Lee, C. C.; Zee, S. Q

    2006-02-15

    COSIS (COre State Indication System) which implemented in the SMART research reactor plays a role to supply the core state parameters or graphs for the operator to recognize the core state effectively. The followings are the main functions of COSIS. Validity Check for the Process Signals and Determination of the COSIS Inputs (SIGVAL), Coolant Flow Rate Calculation (FLOW), Core Thermal Power Calculation (COREPOW), In-core 3-Dimensional Power Distribution Calculation and Peaking Parameters Generation (POWER3D), Azimuthal Tilt Calculation (AZITILT). This report describes the structures of the I/O files that are essential for the users to run COSIS. COSIS handles the following 4 input files. DATABASE: The base input file, COSIS.INP: The signal input file, CCS.DAT: File required for the in-core detector signal processing and the 3-D power distribution calculation, TPFH2O: Steam table for the water properties The DATABASE file contains the base information for a nuclear power plant and is read at the first COSIS calculation. The COSIS.INP file contains the process input and detector signals, and is read by COSIS at every second. CCS.DAT file, that is produced by the COSISMAS code, contains the information for the in-core detector signal processing and the 3-D power distribution calculation. TPFH2O file is a steam table and is written in binary format. COSIS produces the following 4 output files. DATABASE.OUT: The output file for the DATABASE input file, COSIS.OUT: The normal output file produced after the COSIS calculation, COSIS.SUM: File for the operator to recognize the core state effectively, MAS{sub S}IG: File to run the COSISMAS code The DATABASE.OUT file is produced right after finishing DATABASE processing. The COSIS.OUT file is produced after finishing the input signal processing and the main COSIS calculation. The COSIS.SUM file is the summary file of the COSIS results for the operator to recognize the core state effectively. The MAS{sub S}IG file is the

  19. Graphic user interface for COSMOS code

    International Nuclear Information System (INIS)

    Oh, Je Yong; Koo, Yang Hyun; Lee, Byung Ho; Cheon, Jin Sik; Sohn, Dong Seong

    2003-06-01

    The Graphic User Interface (GUI) - which consisted of graphical elements such as windows, menu, button, icon, and so on - made it possible that the computer could be easily used for common users. Hence, the GUI was introduced to improve the efficiency to input parameters in COSMOS code. The functions to output graphs on the screen and postscript files were also added. And the graph library can be applied to the other codes. The details of principles of GUI and graphic library were described in the report

  20. User's manual, version 1.00 for Monteburns, version 3.01

    International Nuclear Information System (INIS)

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

    1998-06-01

    Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN2. Monteburns produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. The program processes input from the user that specifies the system geometry, initial material compositions, feed/removal specifications, and other code-specific parameters. Various results from MCNP, ORIGEN2, and other calculations are then output successively as the code runs. The principle function of monteburns is to transfer one-group cross section and flux values from MCNP to ORIGEN2, and then transfer the resulting material compositions (after irradiation and/or decay) from ORIGEN2 back to MCNP in a repeated, cyclic fashion. The basic requirement of the code is that the user have a working MCNP input file and other input parameters; all interaction with ORIGEN2 and other calculations are performed by monteburns. This report serves as a user's manual for monteburns. It describes how the code functions, what input the user must provide, the calculations performed by the code, and it presents the format required for input files, as well as samples of these files. Monteburns is still in a developmental stage; thus, additions and/or changes may be made over time, and the user's manual will change as well. This is the first version of the user's manual (valid for monteburns version 3.01); users should contact the authors to inquire if a more recent version is available

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    International Nuclear Information System (INIS)

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

    2013-10-01

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

  3. User's guide to the biosphere code ECOS

    International Nuclear Information System (INIS)

    Kane, P.; Thorne, M.C.

    1984-10-01

    This report constitutes the user's guide to the biosphere model ECOS and provides a detailed description of the processes modelled and mathematical formulations used. The FORTRAN code ECOS is an equilibrium-type compartmental biosphere code. ECOS was designed with the objective of producing a general but comprehensive code for use in the assessment of the radiological impact of unspecified geological repositories for radioactive waste. ECOS transforms the rate of release of activity from the geosphere to the rate of accumulation of weighted committed effective dose equivalent (dose). Both maximum individual dose (critical group dose) and collective dose rates may be computed. (author)

  4. YAQUI user's manual for fireball calculations

    International Nuclear Information System (INIS)

    Norton, J.L.; Ruppel, H.M.

    1976-12-01

    Recent modifications and additions made to the YAQUI code are described. This code, which was written to simulate nuclear explosions in the atmosphere, has been improved to include the effects of turbulence. In addition, it now allows input data to be obtained by direct interpolation in the one-dimensional results of early time radiation codes. This new version also makes most of the input free-format (namelist) and is composed of modules for easier modification and isolation of computer system dependence

  5. Rad Chem data acquisition chassis users manual

    International Nuclear Information System (INIS)

    Jones, B.A.

    1980-01-01

    The Shiva Laser at LLL requires many forms of diagnostics to measure and analyze fusion experiments. This manual describes the operation of a Micro-Processor controlled data acquisition system designed at LLL to measure Neutron Activation during fusion experiments on the Shiva Laser

  6. User Manual for SSG Power Simulation 2

    DEFF Research Database (Denmark)

    Jensen, Palle Meinert; Gilling, Lasse; Kofoed, Jens Peter

    This manual gives a detailed description of the use of the computer program SSG Power Simulation 2. Furthermore, the underlying mathematics and algorithms are briefly described. The program is based on experimental data from model testing of Seawave Slot-Cone Generator (SSG) presented in Kofoed...

  7. Steam explosion simulation code JASMINE v.3 user's guide

    International Nuclear Information System (INIS)

    Moriyama, Kiyofumi; Maruyama, Yu; Nakamura, Hideo

    2008-07-01

    A steam explosion occurs when hot liquid contacts with cold volatile liquid. In this phenomenon, fine fragmentation of the hot liquid causes extremely rapid heat transfer from the hot liquid to the cold volatile liquid, and explosive vaporization, bringing shock waves and destructive forces. The steam explosion due to the contact of the molten core material and coolant water during severe accidents of light water reactors has been regarded as a potential threat to the integrity of the containment vessel. We developed a mechanistic steam explosion simulation code, JASMINE, that is applicable to plant scale assessment of the steam explosion loads. This document, as a manual for users of JASMINE code, describes the models, numerical solution methods, and also some verification and example calculations, as well as practical instructions for input preparation and usage of the code. (author)

  8. Object Toolkit Version 4.3 User’s Manual

    Science.gov (United States)

    2016-12-31

    and with Nascap-2k. See the EPIC and Nascap-2k manuals for instructions. Most of the difficulties that users have encountered with Object Toolkit are...4/icond). 12.3 Importing Components From a NX I-DEAS TMG ASCII VUFF File Users of the NX I-DEAS TMG thermal analysis program can import the ASCII...2k user interface. The meaning of these properties is discussed in the Nascap-2k User’s Manual . Figure 36. Detector Properties Dialog Box. 15.5

  9. TOUGH2-GRS version 1. User manual

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Martin; Eckel, Jens

    2016-07-15

    TOUGH2 is a code for the simulation of multi-phase flow processes in porous media that has been developed by the Lawrence Berkeley National Laboratory, California, USA. Since 1991, GRS has been using the code for process analyses and safety assessments for deep geological repositories and has extended the code by several processes that are relevant for repository systems. The TOUGH2 source code that has been developed further by GRS is referred to as TOUGH2-GRS. The present report presents code version 1.1.g, which was developed in project UM13 A 03400 sponsored by the German Federal Ministry for the Environment, Nature Conservation, Building and Nuclear Safety (BMUB).

  10. User's manual for the master equipment list. Phase 1

    International Nuclear Information System (INIS)

    Sandoval, J.D.

    1997-01-01

    This manual is intended to provide a user with enough detailed instruction to guide them through the Master Equipment List Phase 1 (MEL Phase 1) application system operations. The MEL Phase 1 application is a database system that stores Equipment Identification Number (EIN) information to support equipment tracking in the 200E and 200W Tank Farms for the Tank Waste Remediation System Division. The MEL Phase 1 application supports both the user application and administrative control functions. The user application functions include: viewing by Folder, reporting, performing queries, and editing specific data. The administrative control functions include: maintaining valid user identifications, passwords, privileges, defining drop-down lists, and review of the change log relating to EIN data entries, additions, deletions, and editing. The scope of this User's Manual is to discuss these functions and is intended to guide users and answer questions regarding the MEL Phase 1 application

  11. A User's Manual for the NRN Shield Design Method

    Energy Technology Data Exchange (ETDEWEB)

    Hjaerne, Leif [ed.; Aalto, E; Fraeki, R; Leimdoerfer, M; Lindblom, K; Linde, S; Malen, K; Nyman, K

    1964-06-15

    This report describes a code system for bulk shield design written for a Ferranti Mercury computer and is intended as a manual for those using the programme. The idea of an 'almost direct' flux, as in the removal theory serves as a basis for further development of the theory. An important aspiration has been to minimize the manual work of administering the codes. The codes concerned are: NECO, computing necessary group constants from primary data, REFUSE and REBOX (infinite plane or cylindrical, and box geometry, respectively), computing removal flux, NEDI a one-dimensional (plane, spherical, cylindrical) diffusion multigroup code, and SALOME a Monte Carlo code computing the gamma flux. Output tapes are constructed for direct use as input tapes, when required, for a following code.

  12. A User's Manual for the NRN Shield Design Method

    International Nuclear Information System (INIS)

    Hjaerne, Leif; Aalto, E.; Fraeki, R.; Leimdoerfer, M.; Lindblom, K.; Linde, S.; Malen, K.; Nyman, K.

    1964-06-01

    This report describes a code system for bulk shield design written for a Ferranti Mercury computer and is intended as a manual for those using the programme. The idea of an 'almost direct' flux, as in the removal theory serves as a basis for further development of the theory. An important aspiration has been to minimize the manual work of administering the codes. The codes concerned are: NECO, computing necessary group constants from primary data, REFUSE and REBOX (infinite plane or cylindrical, and box geometry, respectively), computing removal flux, NEDI a one-dimensional (plane, spherical, cylindrical) diffusion multigroup code, and SALOME a Monte Carlo code computing the gamma flux. Output tapes are constructed for direct use as input tapes, when required, for a following code

  13. Infrastructure Management Information System User Manual

    Science.gov (United States)

    1998-10-01

    This publication describes and explains the user interface for the Infrastructure Management Information System (IMIS). The IMIS is designed to answer questions regarding public water supply, wastewater treatment, and census information. This publica...

  14. UQTk version 2.0 user manual

    Energy Technology Data Exchange (ETDEWEB)

    Debusschere, Bert J. [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Sargsyan, Khachik [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Safta, Cosmin [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

    2013-10-01

    The UQ Toolkit (UQTk) is a collection of libraries and tools for the quantification of uncertainty in numerical model predictions. Version 2.0 ffers intrusive and non-intrusive methods for propagating input uncertainties through computational models, tools for sensitivity analysis, methods for sparse surrogate construction, and Bayesian inference tools for inferring parameters from experimental data. This manual discusses the download and installation process for UQTk, provides pointers to the UQ methods used in the toolkit, and describes some of the examples provided with the toolkit.

  15. TALENT user's manual.

    Energy Technology Data Exchange (ETDEWEB)

    Merchant, Bion John

    2012-01-01

    The Ground-Based Monitoring R and E Component Evaluation project performs testing on the hardware components that make up Seismic and Infrasound monitoring systems. The majority of the testing is focused on the Digital Waveform Recorder (DWR), Seismic Sensor, and Infrasound Sensor. The software tool used to capture and analyze the data collected from testing is called TALENT: Test and Analysis Evaluation Tool. This document is the manual for using TALENT. Other reports document the testing procedures that are in place (Kromer, 2007) and the algorithms employed in the test analysis (Merchant, 2011).

  16. Design Optimization Toolkit: Users' Manual

    Energy Technology Data Exchange (ETDEWEB)

    Aguilo Valentin, Miguel Alejandro [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Computational Solid Mechanics and Structural Dynamics

    2014-07-01

    The Design Optimization Toolkit (DOTk) is a stand-alone C++ software package intended to solve complex design optimization problems. DOTk software package provides a range of solution methods that are suited for gradient/nongradient-based optimization, large scale constrained optimization, and topology optimization. DOTk was design to have a flexible user interface to allow easy access to DOTk solution methods from external engineering software packages. This inherent flexibility makes DOTk barely intrusive to other engineering software packages. As part of this inherent flexibility, DOTk software package provides an easy-to-use MATLAB interface that enables users to call DOTk solution methods directly from the MATLAB command window.

  17. User manual for version 4.3 of the Tripoli-4 Monte-Carlo method particle transport computer code; Notice d'utilisation du code Tripoli-4, version 4.3: code de transport de particules par la methode de Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Both, J.P.; Mazzolo, A.; Peneliau, Y.; Petit, O.; Roesslinger, B

    2003-07-01

    This manual relates to Version 4.3 TRIPOLI-4 code. TRIPOLI-4 is a computer code simulating the transport of neutrons, photons, electrons and positrons. It can be used for radiation shielding calculations (long-distance propagation with flux attenuation in non-multiplying media) and neutronic calculations (fissile medium, criticality or sub-criticality basis). This makes it possible to calculate k{sub eff} (for criticality), flux, currents, reaction rates and multi-group cross-sections. TRIPOLI-4 is a three-dimensional code that uses the Monte-Carlo method. It allows for point-wise description in terms of energy of cross-sections and multi-group homogenized cross-sections and features two modes of geometrical representation: surface and combinatorial. The code uses cross-section libraries in ENDF/B format (such as JEF2-2, ENDF/B-VI and JENDL) for point-wise description cross-sections in APOTRIM format (from the APOLLO2 code) or a format specific to TRIPOLI-4 for multi-group description. (authors)

  18. BPACK -- A computer model package for boiler reburning/co-firing performance evaluations. User`s manual, Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Wu, K.T.; Li, B.; Payne, R.

    1992-06-01

    This manual presents and describes a package of computer models uniquely developed for boiler thermal performance and emissions evaluations by the Energy and Environmental Research Corporation. The model package permits boiler heat transfer, fuels combustion, and pollutant emissions predictions related to a number of practical boiler operations such as fuel-switching, fuels co-firing, and reburning NO{sub x} reductions. The models are adaptable to most boiler/combustor designs and can handle burner fuels in solid, liquid, gaseous, and slurried forms. The models are also capable of performing predictions for combustion applications involving gaseous-fuel reburning, and co-firing of solid/gas, liquid/gas, gas/gas, slurry/gas fuels. The model package is conveniently named as BPACK (Boiler Package) and consists of six computer codes, of which three of them are main computational codes and the other three are input codes. The three main codes are: (a) a two-dimensional furnace heat-transfer and combustion code: (b) a detailed chemical-kinetics code; and (c) a boiler convective passage code. This user`s manual presents the computer model package in two volumes. Volume 1 describes in detail a number of topics which are of general users` interest, including the physical and chemical basis of the models, a complete description of the model applicability, options, input/output, and the default inputs. Volume 2 contains a detailed record of the worked examples to assist users in applying the models, and to illustrate the versatility of the codes.

  19. Manual phased arrays for weld inspections using North American codes

    International Nuclear Information System (INIS)

    Moles, Michael

    2008-01-01

    Phased arrays are primarily a method of generating and receiving ultrasound, not a new technology. In addition, the physics of ultrasound generated by phased arrays is identical to that from conventional monocrystals. Not surprisingly, all the major North American (and some European) codes accept phased arrays, either explicitly or implicitly. However, the technique and procedures needs to be proven, typically by a Performance Demonstration. The ASME (AmeicanSociety for Mechanical Engineers) Section V and API RP2X explicitly accept phased arrays. Three ASME code cases have been written specifically fo manual phased array: Code Cases 2541. 2557 and 2558. Over and above the general requirements of Article 4, these Code Cases require full waveform calibration. This is echoed in ASTM E-2491, a Standard Guide for setting up phased arrays. In addition. details such as focusing and reporting are addressed. The American Petroleum Institute QUTE procedure did not need any modifications to be compatible with manual phased arrays. The American Welding Society (AWS) Structural Welding Code D1.1 implicitly accepts phased arrays. New technologies such as phased arrays can be proven using Annex K. Nonetheless, a manual phased array unit using the standard AWS probe and displaying 45, 60 and 70degrees waveforms would be acceptable for D1.1 a s is . Overall, most major North American codes accept phased arrays, however, the technique and procedures must be proven, often using a Performance Demonstration. (author)

  20. Database on wind characteristics. Users manual

    DEFF Research Database (Denmark)

    Larsen, Gunner Chr.; Hansen, K.S.

    2001-01-01

    quality control procedures), part two accounts in details for the available data in the established database bank and part three is the Users Manualdescribing the various ways to access and analyse the data. The present report constitutes part three of the Annex XVII reporting and contains a trough...

  1. 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.

  2. Simplified Linear Equation Solvers users manual

    Energy Technology Data Exchange (ETDEWEB)

    Gropp, W. [Argonne National Lab., IL (United States); Smith, B. [California Univ., Los Angeles, CA (United States)

    1993-02-01

    The solution of large sparse systems of linear equations is at the heart of many algorithms in scientific computing. The SLES package is a set of easy-to-use yet powerful and extensible routines for solving large sparse linear systems. The design of the package allows new techniques to be used in existing applications without any source code changes in the applications.

  3. HANSF 1.3 user's manual

    International Nuclear Information System (INIS)

    PLYS, M.G.

    1999-01-01

    The HANSF analysis tool is an integrated model considering phenomena inside a multi-canister overpack (MCO) spent nuclear fuel container such as fuel oxidation, convective and radiative heat transfer, and the potential for fission product release. It may be used for all phases of spent fuel disposition including cold vacuum drying, transportation, and storage. This manual reflects HANSF version 1.3, a revised version of version 1.2a. HANSF 1.3 was written to add new models for axial nodalization, add new features for ease of usage, and correct errors. HANSF 1.3 is intended for use on personal computers such as IBM-compatible machines with Intel processors running under a DOS-type operating system. HANSF 1.3 is known to compile under Lahey TI and Digital Visual FORTRAN, Version 6.0, but this does not preclude operation in other environments

  4. HANSF 1.3.2 User's Manual

    International Nuclear Information System (INIS)

    DUNCAN, D.R.

    1999-01-01

    The HANSF analysis tool is an integrated model considering phenomena inside a multi-canister overpack (MCO) spent nuclear fuel container such as fuel oxidation, convective and radiative heat transfer, and the potential for fission product release. This manual reflects the HANSF version 1.3.2, a revised version of 1.3.1. HANSF 1.3.2 was written to correct minor errors and to allow modeling of condensate flow on the MCO inner surface. HANSF 1.3.2 is intended for use on personal computers such as IBM-compatible machines with Intel processors running under Lahey TI or digital Visual FORTRAN, Version 6.0, but this does not preclude operation in other environments

  5. HANSF 1.3 user's manual; TOPICAL

    International Nuclear Information System (INIS)

    PLYS, M.G.

    1999-01-01

    The HANSF analysis tool is an integrated model considering phenomena inside a multi-canister overpack (MCO) spent nuclear fuel container such as fuel oxidation, convective and radiative heat transfer, and the potential for fission product release. It may be used for all phases of spent fuel disposition including cold vacuum drying, transportation, and storage. This manual reflects HANSF version 1.3, a revised version of version 1.2a. HANSF 1.3 was written to add new models for axial nodalization, add new features for ease of usage, and correct errors. HANSF 1.3 is intended for use on personal computers such as IBM-compatible machines with Intel processors running under a DOS-type operating system. HANSF 1.3 is known to compile under Lahey TI and Digital Visual FORTRAN, Version 6.0, but this does not preclude operation in other environments

  6. TMAP4 User's Manual

    Energy Technology Data Exchange (ETDEWEB)

    Longhurst, G.R.; Holland, D.F.; Jones, J.L.; Merrill, B.J.

    1992-06-12

    The Tritium Migration Analysis Program, Version 4 (TMAP4) has been developed by the Fusion Safety Program at the Idaho National Engineering Laboratory (INEL) as a safety analysis code, mainly to analyze tritium retention and loss in fusion reactor structures and systems during normal operation and accident conditions. TMAP4 incorporates one-dimensional thermal- and mass-diffusive transport and trapping calculations through structures and zero dimensional fluid transport between enclosures and across the interface between enclosures and structures. A key feature is the ability to input problem definition parameters as constants, interpolation tables, or FORTRAN equations. The code is specifically intended for use under a DOS operating system on PC-type mini-computers, but it has also been run successfully on workstations and mainframe computer systems. Use of the equation-input feature requires access to a FORTRAN-77 compiler and a linker program.

  7. Desalination Economic Evaluation Program (DEEP). User's manual

    International Nuclear Information System (INIS)

    2000-01-01

    DEEP (formerly named ''Co-generation and Desalination Economic Evaluation'' Spreadsheet, CDEE) has been developed originally by General Atomics under contract, and has been used in the IAEA's feasibility studies. For further confidence in the software, it was validated in March 1998. After that, a user friendly version has been issued under the name of DEEP at the end of 1998. DEEP output includes the levelised cost of water and power, a breakdown of cost components, energy consumption and net saleable power for each selected option. Specific power plants can be modelled by adjustment of input data including design power, power cycle parameters and costs

  8. Module type plant system dynamics analysis code (MSG-COPD). Code manual

    International Nuclear Information System (INIS)

    Sakai, Takaaki

    2002-11-01

    MSG-COPD is a module type plant system dynamics analysis code which involves a multi-dimensional thermal-hydraulics calculation module to analyze pool type of fast breeder reactors. Explanations of each module and the methods for the input data are described in this code manual. (author)

  9. HECTR Version 1.5 user's manual

    International Nuclear Information System (INIS)

    Dingman, S.E.; Camp, A.L.; Wong, C.C.; King, D.B.; Gasser, R.D.

    1986-04-01

    This report describes the use and features of HECTR Version 1.5. HECTR is a relatively fast-running, lumped-volume containment analysis computer program that is most useful for performing parametric studies. The main purpose of HECTR is to analyze nuclear reactor accidents involving the transport and combustion of hydrogen, but HECTR can also function as an experiment analysis tool and can solve a limited set of other types of containment problems. New models added to HECTR Version 1.5 include fan coolers, containment leakage, continuous burning, and the capability to treat carbon monoxide and carbon dioxide. Models for the ice condenser, sumps, and Mark III suppression pool were upgraded. HECTR is designed for flexibility and provides for user control of many important parameters, particularly those related to hydrogen combustion. Built-in correlations and default values of key parameters are also provided

  10. The users manual and concepts of nuclear materials accounting system

    International Nuclear Information System (INIS)

    Lee, Byung Du; Jeon, In

    1996-03-01

    This report is to describe the concepts, operation status and user's manuals of nuclear materials accounting system which was developed to not only make out, report and manage the IAEA accounting reports but also maintain the accounting information. Therefore, facility operator could effectively make use of the accounting system without a special training by using this report. 3 tabs., 15 figs., (Author) .new

  11. National Radiobiology Archives Distributed Access user's manual

    Energy Technology Data Exchange (ETDEWEB)

    Watson, C.; Smith, S. (Pacific Northwest Lab., Richland, WA (United States)); Prather, J. (Linfield Coll., McMinnville, OR (United States))

    1991-11-01

    This User's Manual describes installation and use of the National Radiobiology Archives (NRA) Distributed Access package. The package consists of a distributed subset of information representative of the NRA databases and database access software which provide an introduction to the scope and style of the NRA Information Systems.

  12. 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...

  13. User's Manual. Vocational Education Readiness Test (VERT). Revised Edition.

    Science.gov (United States)

    Thomas, Edward L.; And Others

    This user's manual provides information suggesting various ways in which educational personnel may wish to utilize the eight modules included in the Vocational Education Readiness Test (VERT). The introductory section presents questions which can be used by school personnel to determine whether or not VERT will be useful in their school system.…

  14. CHEMCON User's Manual, Version 3.1

    International Nuclear Information System (INIS)

    Gaeta, M.J.; Merrill, B.J.

    1995-09-01

    CHEMCON is a computer program developed to analyze thermal transients of tokamak fusion reactors. It contains a one dimensional, cylindrical geometry, conduction model that allows a variety of heat transfer modes within nodes and at node boundaries. Solid regions can be grouped into segments that communicate at their boundaries through a radiation enclosure model. CHEMCON includes a single volume, pressurization/condensation model that is used to include the effects of an in-vessel LOCA and the resulting heat transfer between hot surfaces and cold surfaces in contact with this volume. The code includes properties for 11 solid materials and two gases. CHEMCON also contains specialized models for modeling chemical reactions of node boundaries with air and steam including the gases produced from these reactions. In addition, a model treating the collapse of radiation shields within a gap is also included. CHEMCON is used mainly to simulate the thermal transient for post-blowdown loss-of-coolant-accidents

  15. A Walsh Function Module Users' Manual

    Science.gov (United States)

    Gnoffo, Peter A.

    2014-01-01

    The solution of partial differential equations (PDEs) with Walsh functions offers new opportunities to simulate many challenging problems in mathematical physics. The approach was developed to better simulate hypersonic flows with shocks on unstructured grids. It is unique in that integrals and derivatives are computed using simple matrix multiplication of series representations of functions without the need for divided differences. The product of any two Walsh functions is another Walsh function - a feature that radically changes an algorithm for solving PDEs. A FORTRAN module for supporting Walsh function simulations is documented. A FORTRAN code is also documented with options for solving time-dependent problems: an advection equation, a Burgers equation, and a Riemann problem. The sample problems demonstrate the usage of the Walsh function module including such features as operator overloading, Fast Walsh Transforms in multi-dimensions, and a Fast Walsh reciprocal.

  16. The SWAN coupling code: user's guide

    International Nuclear Information System (INIS)

    Litaudon, X.; Moreau, D.

    1988-11-01

    Coupling of slow waves in a plasma near the lower hybrid frequency is well known and linear theory with density step followed by a constant gradient can be used with some confidence. With the aid of the computer code SWAN, which stands for 'Slow Wave Antenna', the following parameters can be numerically calculated: n parallel power spectrum, directivity (weighted by the current drive efficiency), reflection coefficients (amplitude and phase) both before and after the E-plane junctions, scattering matrix at the plasma interface, scattering matrix at the E-plane junctions, maximum electric fields in secondary waveguides and location where it occurs, effect of passive waveguides on each side of the antenna, and the effect of a finite magnetic field in front of the antenna (for homogeneous plasma). This manual gives the basic information on the main assumptions of the coupling theory and on the use and general structure of the code itself. It answers the questions what are the main assumptions of the physical model? how to execute a job? what are the input parameters of the code? and what are the output results and where are they written? (author)

  17. User's Manual for PCSMS (Parallel Complex Sparse Matrix Solver). Version 1.

    Science.gov (United States)

    Reddy, C. J.

    2000-01-01

    PCSMS (Parallel Complex Sparse Matrix Solver) is a computer code written to make use of the existing real sparse direct solvers to solve complex, sparse matrix linear equations. PCSMS converts complex matrices into real matrices and use real, sparse direct matrix solvers to factor and solve the real matrices. The solution vector is reconverted to complex numbers. Though, this utility is written for Silicon Graphics (SGI) real sparse matrix solution routines, it is general in nature and can be easily modified to work with any real sparse matrix solver. The User's Manual is written to make the user acquainted with the installation and operation of the code. Driver routines are given to aid the users to integrate PCSMS routines in their own codes.

  18. HORECA. Hoger onderwijs reactor elementary core analysis system. User's manual

    International Nuclear Information System (INIS)

    Battum, E. van; Serov, I.V.

    1993-07-01

    HORECA is developed at IRI Delft for quick analysis of power distribution, burnup and safety for the HOR. It can be used for the manual search of a better loading of the reactor. HORECA is based on the Penn State Fuel Management Package and uses the MCRAC code included in this package as a calculation engine. (orig./HP)

  19. Mars 2.2 code manual: input requirements

    International Nuclear Information System (INIS)

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

    2003-07-01

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

  20. The Image of User Instructions: Comparing Users' Expectations of and Experiences with an Official and a Commercial Software Manual

    NARCIS (Netherlands)

    de Jong, Menno D.T.; Karreman, Joyce

    2017-01-01

    Purpose: The market for (paid-for) commercial software manuals is flourishing, while (free) official manuals are often assumed to be neglected by users. To investigate differences in user perceptions of commercial and official manuals, we conducted two studies: one focusing on user expectations and

  1. 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.

  2. User Manual for Whisper-1.1

    International Nuclear Information System (INIS)

    Brown, Forrest B.; Rising, Michael Evan; Alwin, Jennifer Louise

    2017-01-01

    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.

  3. MXS cross-section preprocessor user's manual

    International Nuclear Information System (INIS)

    Parker, F.; Ishikawa, M.; Luck, L.

    1987-03-01

    The MXS preprocessor has been designed to reduce the execution time of programs using isotopic cross-section data and to both reduce the execution time and improve the accuracy of shielding-factor interpolation in the SIMMER-II accident analysis program. MXS is a dual-purpose preprocessing code to: (1) mix isotopes into materials and (2) fit analytic functions to the shelf-shielding data. The program uses the isotope microscopic neutron cross-section data from the CCCC standard interface file ISOTXS and the isotope Bondarenko self-shielding data from the CCCC standard interface file BRKOXS to generate cross-section and self-shielding data for materials. The materials may be a mixture of several isotopes. The self-shielding data for the materials may be the actual shielding factors or a set of coefficients for functions representing the background dependence of the shielding factors. A set of additional data is given to describe the functions necessary to interpolate the shielding factors over temperature

  4. EPRI fuel performance data base: user's manual. Final report

    International Nuclear Information System (INIS)

    Simpson, J.; Lee, S.; Rumble, E.

    1980-10-01

    This user's manual provides instructions for accessing the data in the EPRI fuel performance data base (FPDB) and manipulating that data to solve specific problems that the user wishes to specify. The user interacts with the FPDB through the Relational Information Management System (RIMS) computer program. The structure and format of the FPDB and the general syntax of the data base commands are described. Instructions follow for the use of each command. Appendixes provide more detailed information about the FPDB and its software. The FPDB currently resides on a PRIME-750 computer

  5. 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.

  6. Miro V3.0: user guide and reference manual

    International Nuclear Information System (INIS)

    Donnat, Ph.; Treimany, C.; Morice, O.; Ribeyre, X.

    1998-06-01

    This paper contains the user's guide and reference manual of Miro software. This software is used for simulating propagation and amplification of laser beams in laser devices as Megajoules or NIF. The physical effects taken into account ar essentially: saturated amplification, absorption, Kerr effect, birefringence and aberrations. The models of propagation are either geometrical optics of parallel beams, or Fresnel diffraction. A graphic user interface as been included to allow interactive management of optical devices and results. A Unix environment with X-Window and Motif is required to run Miro. The user's guide gives a short insight of the software. The reference manual details the physical models and the way they are implanted in Miro. (author)

  7. European correlation counter (ECC) VER. 1.0 users manual

    International Nuclear Information System (INIS)

    Vocino, V.

    1989-01-01

    The users manual is conceived for the utilization of personal computers linked to a shift register for the assay of Pu containing fuel. It is based on an MS DOS system and permits the users to store calibration curves, constants, measurement and assay data on floppy disks for ulterior utilisation. The interpretation models use the latest theories of neutron multiplication and dead time effects. The inspector is guided by the screen display step by step from the instrument test via calibration to the assay and data storage. All test, calibration and assay data can be printed out on paper tape and serve to the inspector to prepare the respective inspection report. The users are invited to make suggestions to the authors, such that the manual's use can be adapted to further requirements

  8. Facility Interface Capability Assessment (FICA) user manual

    International Nuclear Information System (INIS)

    Pope, R.B.; MacDonald, R.R.; Massaglia, J.L.; Williamson, D.A.; Viebrock, J.M.; Mote, N.

    1995-09-01

    The US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is responsible for developing the Civilian Radioactive Waste Management System (CRWMS) to accept spent nuclear fuel from commercial facilities. The objective of the Facility Interface Capability Assessment (FICA) project was to assess the capability of each commercial spent nuclear fuel (SNF) storage facility, at which SNF is stored, to handle various SNF shipping casks. The purpose of this report is describe the FICA computer software and to provide the FICA user with a guide on how to use the FICA system. The FICA computer software consists of two executable programs: the FICA Reactor Report program and the FICA Summary Report program (written in the Ca-Clipper version 5.2 development system). The complete FICA software system is contained on either a 3.5 in. (double density) or a 5.25 in. (high density) diskette and consists of the two FICA programs and all the database files (generated using dBASE III). The FICA programs are provided as ''stand alone'' systems and neither the Ca-Clipper compiler nor dBASE III is required to run the FICA programs. The steps for installing the FICA software system and executing the FICA programs are described in this report. Instructions are given on how to install the FICA software system onto the hard drive of the PC and how to execute the FICA programs from the FICA subdirectory on the hard drive. Both FICA programs are menu driven with the up-arrow and down-arrow keys used to move the cursor to the desired selection

  9. User's manual for computer program BASEPLOT

    Science.gov (United States)

    Sanders, Curtis L.

    2002-01-01

    The checking and reviewing of daily records of streamflow within the U.S. Geological Survey is traditionally accomplished by hand-plotting and mentally collating tables of data. The process is time consuming, difficult to standardize, and subject to errors in computation, data entry, and logic. In addition, the presentation of flow data on the internet requires more timely and accurate computation of daily flow records. BASEPLOT was developed for checking and review of primary streamflow records within the U.S. Geological Survey. Use of BASEPLOT enables users to (1) provide efficiencies during the record checking and review process, (2) improve quality control, (3) achieve uniformity of checking and review techniques of simple stage-discharge relations, and (4) provide a tool for teaching streamflow computation techniques. The BASEPLOT program produces tables of quality control checks and produces plots of rating curves and discharge measurements; variable shift (V-shift) diagrams; and V-shifts converted to stage-discharge plots, using data stored in the U.S. Geological Survey Automatic Data Processing System database. In addition, the program plots unit-value hydrographs that show unit-value stages, shifts, and datum corrections; input shifts, datum corrections, and effective dates; discharge measurements; effective dates for rating tables; and numeric quality control checks. Checklist/tutorial forms are provided for reviewers to ensure completeness of review and standardize the review process. The program was written for the U.S. Geological Survey SUN computer using the Statistical Analysis System (SAS) software produced by SAS Institute, Incorporated.

  10. M3 User's Manual. Version 3.0

    International Nuclear Information System (INIS)

    Laaksoharju, Marcus; Skaarman, Erik; Gomez, Javier B.

    2009-11-01

    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 of the

  11. Software Users Manual (SUM): Extended Testability Analysis (ETA) Tool

    Science.gov (United States)

    Maul, William A.; Fulton, Christopher E.

    2011-01-01

    This software user manual describes the implementation and use the Extended Testability Analysis (ETA) Tool. The ETA Tool is a software program that augments the analysis and reporting capabilities of a commercial-off-the-shelf (COTS) testability analysis software package called the Testability Engineering And Maintenance System (TEAMS) Designer. An initial diagnostic assessment is performed by the TEAMS Designer software using a qualitative, directed-graph model of the system being analyzed. The ETA Tool utilizes system design information captured within the diagnostic model and testability analysis output from the TEAMS Designer software to create a series of six reports for various system engineering needs. The ETA Tool allows the user to perform additional studies on the testability analysis results by determining the detection sensitivity to the loss of certain sensors or tests. The ETA Tool was developed to support design and development of the NASA Ares I Crew Launch Vehicle. The diagnostic analysis provided by the ETA Tool was proven to be valuable system engineering output that provided consistency in the verification of system engineering requirements. This software user manual provides a description of each output report generated by the ETA Tool. The manual also describes the example diagnostic model and supporting documentation - also provided with the ETA Tool software release package - that were used to generate the reports presented in the manual

  12. The Radiological Safety Analysis Computer Program (RSAC-5) user's manual

    International Nuclear Information System (INIS)

    Wenzel, D.R.

    1994-02-01

    The Radiological Safety Analysis Computer Program (RSAC-5) calculates the consequences of the release of radionuclides to the atmosphere. Using a personal computer, a user can generate a fission product inventory from either reactor operating history or nuclear criticalities. RSAC-5 models the effects of high-efficiency particulate air filters or other cleanup systems and calculates decay and ingrowth during transport through processes, facilities, and the environment. Doses are calculated through the inhalation, immersion, ground surface, and ingestion pathways. RSAC+, a menu-driven companion program to RSAC-5, assists users in creating and running RSAC-5 input files. This user's manual contains the mathematical models and operating instructions for RSAC-5 and RSAC+. Instructions, screens, and examples are provided to guide the user through the functions provided by RSAC-5 and RSAC+. These programs are designed for users who are familiar with radiological dose assessment methods

  13. Automated Transportation Management System (ATMS) user's manual. Revision 1

    International Nuclear Information System (INIS)

    Smith, P.D.

    1994-01-01

    The Automated Transportation Management System (ATMS) Software User Guide (SUG) constitutes the user procedures for the ATMS System. Information in this document will be used by the user to operate the automated system. It is intended to be used as a reference manual to guide and direct the user(s) through the ATMS software product and its environment. The objectives of ATMS are as follows: to better support the Procurement function with freight rate information; to free Transportation Logistics personnel from routine activities such as the auditing and input of freight billing information; to comply with Headquarters Department of Energy-Inspector General (DOE-IG) audit findings to automate transportation management functions; to reduce the keying of data into the Shipment Mobility Accountability Collection (SMAC) database; and to provide automation for the preparing of Bill of Lading, Declaration of Dangerous Goods, Emergency Response Guide and shipping Labels using HM181 Retrieval of hazardous material table text information

  14. User's Manual for the Object User Interface (OUI): An Environmental Resource Modeling Framework

    Science.gov (United States)

    Markstrom, Steven L.; Koczot, Kathryn M.

    2008-01-01

    The Object User Interface is a computer application that provides a framework for coupling environmental-resource models and for managing associated temporal and spatial data. The Object User Interface is designed to be easily extensible to incorporate models and data interfaces defined by the user. Additionally, the Object User Interface is highly configurable through the use of a user-modifiable, text-based control file that is written in the eXtensible Markup Language. The Object User Interface user's manual provides (1) installation instructions, (2) an overview of the graphical user interface, (3) a description of the software tools, (4) a project example, and (5) specifications for user configuration and extension.

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

    Science.gov (United States)

    1997-09-30

    changes. Cancel Button Closes the Site Information Screen, abandoning changes. APPAREL ORDER PROCESSING MODULE FIELD USER MANUAL Ordering Official...on the Ordering Official Information Screen. APPAREL ORDER PROCESSING MODULE FIELD USER MANUAL Ordering Official Information Screen (Jjj

  16. Climate change scenarios for Canada's national parks : a users manual

    International Nuclear Information System (INIS)

    Jones, B.; Wun, N.; Scott, D.; Barrow, E.

    2003-01-01

    A screening level impact assessment has shown that the implications of climate change for Canada's national parks are considerable. Climate change scenarios will be an important component in examining the potential climate change impacts and the implications of adaptation strategies. Most climate change scenarios are based on vulnerability, impact and adaptation research. This user's manual describes the development of 3 types of climate change scenarios including scenarios from global climate models (GCMs), bioclimate scenarios and daily scenarios for use by Parks Canada. The manual offers advice to first-time climate change scenario users in choosing and interpreting climate change, bioclimate and daily scenarios. It also addresses the theoretical and practical foundations of each climate scenario and shows how to access data regarding the various scenarios. Hands-on exercises are included as an interpretive aid. 20 refs., 4 tabs., 19 figs

  17. Real-Time Multiprocessor Programming Language (RTMPL) user's manual

    Science.gov (United States)

    Arpasi, D. J.

    1985-01-01

    A real-time multiprocessor programming language (RTMPL) has been developed to provide for high-order programming of real-time simulations on systems of distributed computers. RTMPL is a structured, engineering-oriented language. The RTMPL utility supports a variety of multiprocessor configurations and types by generating assembly language programs according to user-specified targeting information. Many programming functions are assumed by the utility (e.g., data transfer and scaling) to reduce the programming chore. This manual describes RTMPL from a user's viewpoint. Source generation, applications, utility operation, and utility output are detailed. An example simulation is generated to illustrate many RTMPL features.

  18. User's Manual for RESRAD-OFFSITE Version 2.

    Energy Technology Data Exchange (ETDEWEB)

    Yu, C.; Gnanapragasam, E.; Biwer, B. M.; Kamboj, S.; Cheng, J. -J.; Klett, T.; LePoire, D.; Zielen, A. J.; Chen, S. Y.; Williams, W. A.; Wallo, A.; Domotor, S.; Mo, T.; Schwartzman, A.; Environmental Science Division; DOE; NRC

    2007-09-05

    The RESRAD-OFFSITE code is an extension of the RESRAD (onsite) code, which has been widely used for calculating doses and risks from exposure to radioactively contaminated soils. The development of RESRAD-OFFSITE started more than 10 years ago, but new models and methodologies have been developed, tested, and incorporated since then. Some of the new models have been benchmarked against other independently developed (international) models. The databases used have also expanded to include all the radionuclides (more than 830) contained in the International Commission on Radiological Protection (ICRP) 38 database. This manual provides detailed information on the design and application of the RESRAD-OFFSITE code. It describes in detail the new models used in the code, such as the three-dimensional dispersion groundwater flow and radionuclide transport model, the Gaussian plume model for atmospheric dispersion, and the deposition model used to estimate the accumulation of radionuclides in offsite locations and in foods. Potential exposure pathways and exposure scenarios that can be modeled by the RESRAD-OFFSITE code are also discussed. A user's guide is included in Appendix A of this manual. The default parameter values and parameter distributions are presented in Appendix B, along with a discussion on the statistical distributions for probabilistic analysis. A detailed discussion on how to reduce run time, especially when conducting probabilistic (uncertainty) analysis, is presented in Appendix C of this manual.

  19. Activity Management System user reference manual. Revision 1

    International Nuclear Information System (INIS)

    Gates, T.A.; Burdick, M.B.

    1994-01-01

    The Activity Management System (AMS) was developed in response to the need for a simple-to-use, low-cost, user interface system for collecting and logging Hanford Waste Vitrification Plant Project (HWVP) activities. This system needed to run on user workstations and provide common user access to a database stored on a local network file server. Most important, users wanted a system that provided a management tool that supported their individual process for completing activities. Existing system treated the performer as a tool of the system. All AMS data is maintained in encrypted format. Users can feel confident that any activities they have entered into the database are private and that, as the originator, they retain sole control over who can see them. Once entered into the AMS database, the activities cannot be accessed by anyone other than the originator, the designated agent, or by authorized viewers who have been explicitly granted the right to look at specific activities by the originator. This user guide is intended to assist new AMS users in learning how to use the application and, after the initial learning process, will serve as an ongoing reference for experienced users in performing infrequently used functions. Online help screens provide reference to some of the key information in this manual. Additional help screens, encompassing all the applicable material in this manual, will be incorporated into future AMS revisions. A third, and most important, source of help is the AMS administrator(s). This guide describes the initial production version of AMS, which has been designated Revision 1.0

  20. 75 FR 52976 - Issuance of Revised Users' Manual for Commission Mediation Program for Investigations Under...

    Science.gov (United States)

    2010-08-30

    ... INTERNATIONAL TRADE COMMISSION Issuance of Revised Users' Manual for Commission Mediation Program... 65615 (Nov. 8, 2008). The Commission has determined to issue a revised Users' Manual for its program for the mediation of investigations under section 337 of the Tariff Act of 1930. The revised Users' Manual...

  1. CONPAS 1.0 (CONtainment Performance Analysis System). User's manual

    International Nuclear Information System (INIS)

    Ahn, Kwang Il; Jin, Young Ho

    1996-04-01

    CONPAS (CONtainment Performance Analysis System) is a verified computer code package to integrate the numerical, graphical, and results-operation aspects of Level 2 probabilistic safety assessments (PSA) for nuclear power plants automatically under a PC window environment. Compared with the existing DOS-based computer codes for Level 2 PSA, the most important merit of the window-based computer code is that user can easily describe and quantify the accident progression models, and manipulate the resultant outputs in a variety of ways. As a main logic for accident progression analysis, CONPAS employs a concept of the small containment phenomenological event tree (CPET) helpful to trace out visually individual accident progressions and of the large supporting event tree (LSET) for its detailed quantification. For the integrated analysis of Level 2 PSA, the code utilizes four distinct, but closely related modules; (1) ET Editor for construction of several event tree models describing the accident progressions, (2) Computer for quantification of the constructed event trees and graphical display of the resultant outputs, (3) Text Editor for preparation of input decks for quanification and utilization of calculational results, and (4) Mechanistic Code Plotter for utilization of results obtained from severe accident analysis codes. Compared with other existing computer codes for Level 2 PSA, the CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, sensitivity analysis and data interpretation, reporting aspects including tabling and graphic as well as user-friend interface. 10 refs. (Author) .new

  2. User Effect on Code Application and Qualification Needs

    International Nuclear Information System (INIS)

    D'Auria, F.; Salah, A.B.

    2008-01-01

    Experience with some code assessment case studies and also additional ISPs have shown the dominant effect of the code user on the predicted system behavior. The general findings of the user effect investigations on some of the case studies indicate, specifically, that in addition to user effects, there are other reasons which affect the results of the calculations and are hidden under the general title of user effects. The specific characteristics of experimental facilities, i.e. limitations as far as code assessment is concerned; limitations of the used thermal-hydraulic codes to simulate certain system behavior or phenomena; limitations due to interpretation of experimental data by the code user, i.e. interpretation of experimental data base. On the basis of the discussions in this paper, the following conclusions and recommendations can be made: More dialogue appears to be necessary with the experimenters in the planning of code assessment calculations, e.g. ISPs.; User guidelines are not complete for the codes and the lack of sufficient and detailed user guidelines are observed with some of the case studies; More extensive user instruction and training, improved user guidelines, or quality assurance procedures may partially reduce some of the subjective user influence on the calculated results; The discrepancies between experimental data and code predictions are due both to the intrinsic code limit and to the so called user effects. There is a worthful need to quantify the percentage of disagreement due to the poor utilization of the code and due to the code itself. This need especially arises for the uncertainty evaluation studies (e.g. [18]) which do not take into account the mentioned user effects; A much focused investigation, based on the results of comparison calculations e.g. ISPs, analyzing the experimental data and the results of the specific code in order to evaluate the user effects and the related experimental aspects should be integral part of the

  3. User Manual for SAHM package for VisTrails

    Science.gov (United States)

    Talbert, C.B.; Talbert, M.K.

    2012-01-01

    The Software for Assisted Habitat I\\•1odeling (SAHM) has been created to both expedite habitat modeling and help maintain a record of the various input data, pre-and post-processing steps and modeling options incorporated in the construction of a species distribution model. The four main advantages to using the combined VisTrail: SAHM package for species distribution modeling are: 1. formalization and tractable recording of the entire modeling process 2. easier collaboration through a common modeling framework 3. a user-friendly graphical interface to manage file input, model runs, and output 4. extensibility to incorporate future and additional modeling routines and tools. This user manual provides detailed information on each module within the SAHM package, their input, output, common connections, optional arguments, and default settings. This information can also be accessed for individual modules by right clicking on the documentation button for any module in VisTrail or by right clicking on any input or output for a module and selecting view documentation. This user manual is intended to accompany the user guide which provides detailed instructions on how to install the SAHM package within VisTrails and then presents information on the use of the package.

  4. The Montana Rivers Information System: Edit/entry program user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    1992-07-01

    The Montana Rivers Information System (MRIS) was initiated to assess the state`s fish, wildlife, and recreation value; and natural cultural and geologic features. The MRIS is now a set of data bases containing part of the information in the Natural Heritage Program natural features and threatened and endangered species data bases. The purpose of this User`s Manual is to: (1) describe to the user how to maintain the MRIS database of their choice by updating, changing, deleting, and adding records using the edit/entry programs; and (2) provide to the user all information and instructions necessary to complete data entry into the MRIS databases.

  5. APPLE-3: improvement of APPLE for neutron and gamma-ray flux, spectrum and reaction rate plotting code, and of its code manual

    International Nuclear Information System (INIS)

    Kawasaki, Hiromitu; Maki, Koichi; Seki, Yasushi.

    1991-03-01

    A code APPLE was produced in 1976 for calculating and plotting tritium breeding ratio and tritium production rate distributions. That code was improved as 'APPLE-2' in 1982, to calculate and plot not only tritium breeding ratio but also distributions of neutron and gamma-ray fluxes, their spectra, nuclear heating rates and other reaction rates, and dose rate distributions during operation and after shutdown in 1982. The code APPLE-2 can calculate and plot these nuclear properties derived from neutron and gamma-ray fluxes by ANISN (one dimensional transport code), DOT3.5 (two dimensional transport code) and MORSE (three dimensional Monte Carlo code). We revised the code APPLE-2 as 'APPLE-3' by adding many functions to the APPLE-2 code in accordance with users' requirements proposed in recent progress of fusion reaction nuclear design. With minor modification of APPLE-2, a number of inconsistencies have been found between the code manual and the input data in the code. In the present report, the new functions added to APPLE-2 and improved users' manual are explained. (author)

  6. ABAQUS-EPGEN: a general-purpose finite element code. Volume 3. Example problems manual

    International Nuclear Information System (INIS)

    Hibbitt, H.D.; Karlsson, B.I.; Sorensen, E.P.

    1983-03-01

    This volume is the Example and Verification Problems Manual for ABAQUS/EPGEN. Companion volumes are the User's, Theory and Systems Manuals. This volume contains two major parts. The bulk of the manual (Sections 1-8) contains worked examples that are discussed in detail, while Appendix A documents a large set of basic verification cases that provide the fundamental check of the elements in the code. The examples in Sections 1-8 illustrate and verify significant aspects of the program's capability. Most of these problems provide verification, but they have also been chosen to allow discussion of modeling and analysis techniques. Appendix A contains basic verification cases. Each of these cases verifies one element in the program's library. The verification consists of applying all possible load or flux types (including thermal loading of stress elements), and all possible foundation or film/radiation conditions, and checking the resulting force and stress solutions or flux and temperature results. This manual provides program verification. All of the problems described in the manual are run and the results checked, for each release of the program, and these verification results are made available

  7. Dairy Analytics and Nutrient Analysis (DANA) Prototype System User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Sam Alessi; Dennis Keiser

    2012-10-01

    This document is a user manual for the Dairy Analytics and Nutrient Analysis (DANA) model. DANA provides an analysis of dairy anaerobic digestion technology and allows users to calculate biogas production, co-product valuation, capital costs, expenses, revenue and financial metrics, for user customizable scenarios, dairy and digester types. The model provides results for three anaerobic digester types; Covered Lagoons, Modified Plug Flow, and Complete Mix, and three main energy production technologies; electricity generation, renewable natural gas generation, and compressed natural gas generation. Additional options include different dairy types, bedding types, backend treatment type as well as numerous production, and economic parameters. DANA’s goal is to extend the National Market Value of Anaerobic Digester Products analysis (informa economics, 2012; Innovation Center, 2011) to include a greater and more flexible set of regional digester scenarios and to provide a modular framework for creation of a tool to support farmer and investor needs. Users can set up scenarios from combinations of existing parameters or add new parameters, run the model and view a variety of reports, charts and tables that are automatically produced and delivered over the web interface. DANA is based in the INL’s analysis architecture entitled Generalized Environment for Modeling Systems (GEMS) , which offers extensive collaboration, analysis, and integration opportunities and greatly speeds the ability construct highly scalable web delivered user-oriented decision tools. DANA’s approach uses server-based data processing and web-based user interfaces, rather a client-based spreadsheet approach. This offers a number of benefits over the client-based approach. Server processing and storage can scale up to handle a very large number of scenarios, so that analysis of county, even field level, across the whole U.S., can be performed. Server based databases allow dairy and digester

  8. SCDAP/RELAP5/MOD2 code manual

    International Nuclear Information System (INIS)

    Allison, C.M.; Johnson, E.C.

    1989-09-01

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. The modeling theory and associated numerical schemes are documented in Volumes I and II to acquaint the user with the modeling base and thus aid in effective use of the code

  9. HTGR Application Economic Model Users' Manual

    Energy Technology Data Exchange (ETDEWEB)

    A.M. Gandrik

    2012-01-01

    The High Temperature Gas-Cooled Reactor (HTGR) Application Economic Model was developed at the Idaho National Laboratory for the Next Generation Nuclear Plant Project. The HTGR Application Economic Model calculates either the required selling price of power and/or heat for a given internal rate of return (IRR) or the IRR for power and/or heat being sold at the market price. The user can generate these economic results for a range of reactor outlet temperatures; with and without power cycles, including either a Brayton or Rankine cycle; for the demonstration plant, first of a kind, or nth of a kind project phases; for up to 16 reactor modules; and for module ratings of 200, 350, or 600 MWt. This users manual contains the mathematical models and operating instructions for the HTGR Application Economic Model. Instructions, screenshots, and examples are provided to guide the user through the HTGR Application Economic Model. This model was designed for users who are familiar with the HTGR design and Excel and engineering economics. Modification of the HTGR Application Economic Model should only be performed by users familiar with the HTGR and its applications, Excel, and Visual Basic.

  10. Comparative ergonomic assessment of manual wheelchairs by paraplegic users.

    Science.gov (United States)

    Gil-Agudo, Angel; Solís-Mozos, Marta; del-Ama, Antonio J; Crespo-Ruiz, Beatriz; de la Peña-González, Ana Isabel; Pérez-Nombela, Soraya

    2013-07-01

    The aim of the present study was to describe and test the reliability of a comprehensive product-centered approach to assessing functional performance and wheelchair user perceptions on device ergonomics and satisfaction of performance. A pilot study was implemented using this approach to evaluate differences among four manual wheelchairs. Six wheelchair users with complete spinal cord injury (SCI) at the thoracic level and with no previous upper limbs impairment were recruited for this study. After finishing circuit tasks, subjects were asked to complete a questionnaire about ergonomic wheelchair characteristics (manoeuvrability, stability, comfort and ease of propulsion) and satisfaction about task performance. On the other hand, objective data were recorded during user performance as the time required to complete each test, kinetic wheelchair propulsion data obtained with two SMARTWheels® and physiological parameters (heart rate and physiological index). Kuschall Champion® and Otto Bock Voyage® wheelchairs were ranked best for most ergonomic aspects specially in manoeuvrability (p importance of looking both kinds of information, user perception and user functional performance when evaluating a wheelchair or comparing across devices.

  11. KAFEPA-II program users' manual and description

    International Nuclear Information System (INIS)

    Suk, H. C.; Hwang, W.; Kim, B. G.; Sim, K. S.; Heo, Y. H.; Byun, T. S.; Park, G. S.

    1992-04-01

    KAFEPA-II is a computer program for simulating the behaviour of UO 2 fuel elements under normal operating conditions of a CANDU reactor. It computes the one-dimensional temperature distribution and thermal expansion of the fuel pellets. The amount of gas released during irradiation of the fuel is also computed. Thermal expansion and gas pressure inside the fuel element are then used to compute the strains and stresses in the sheath. This document is intended as a user's manual and description for KAFEPA-II. (Author)

  12. The development of nuclear material accountability system - software user's manual

    International Nuclear Information System (INIS)

    Byeon, Kee Hoh; Kim, Ho Dong; Song, Dae Yong; Ko, Won Il; Hong, Jong Sook; Lee, Byung Doo

    1999-07-01

    We have developed the near-real time nuclear material accountability system, named by DMAS, for DUPIC Test Facility in the basis of the survey of DUPIC process and activities for the accountability of the system, and the review of the rules and regulations related to the nuclear material accounting. Our system adopts the structure and technologies used in COREMAS which was developed by LANL. This technical report illustrates the system structure and program usage as a user manual for DMAS. (author). 56 tabs., 1 fig

  13. Synfuel program analysis. Volume 2: VENVAL users manual

    Science.gov (United States)

    Muddiman, J. B.; Whelan, J. W.

    1980-07-01

    This volume is intended for program analysts and is a users manual for the VENVAL model. It contains specific explanations as to input data requirements and programming procedures for the use of this model. VENVAL is a generalized computer program to aid in evaluation of prospective private sector production ventures. The program can project interrelated values of installed capacity, production, sales revenue, operating costs, depreciation, investment, dent, earnings, taxes, return on investment, depletion, and cash flow measures. It can also compute related public sector and other external costs and revenues if unit costs are furnished.

  14. User's manual for the two-dimensional transputer graphics toolkit

    Science.gov (United States)

    Ellis, Graham K.

    1988-01-01

    The user manual for the 2-D graphics toolkit for a transputer based parallel processor is presented. The toolkit consists of a package of 2-D display routines that can be used for the simulation visualizations. It supports multiple windows, double buffered screens for animations, and simple graphics transformations such as translation, rotation, and scaling. The display routines are written in occam to take advantage of the multiprocessing features available on transputers. The package is designed to run on a transputer separate from the graphics board.

  15. Users' guide to CACECO containment analysis code. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Peak, R.D.

    1979-06-01

    The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. The code is included in the National Energy Software Center Library at Argonne National Laboratory as Program No. 762. This users' guide describes the CACECO code and its data input requirements. The code description covers the many mathematical models used and the approximations used in their solution. The descriptions are detailed to the extent that the user can modify the code to suit his unique needs, and, indeed, the reader is urged to consider code modification acceptable.

  16. Optical code division multiple access secure communications systems with rapid reconfigurable polarization shift key user code

    Science.gov (United States)

    Gao, Kaiqiang; Wu, Chongqing; Sheng, Xinzhi; Shang, Chao; Liu, Lanlan; Wang, Jian

    2015-09-01

    An optical code division multiple access (OCDMA) secure communications system scheme with rapid reconfigurable polarization shift key (Pol-SK) bipolar user code is proposed and demonstrated. Compared to fix code OCDMA, by constantly changing the user code, the performance of anti-eavesdropping is greatly improved. The Pol-SK OCDMA experiment with a 10 Gchip/s user code and a 1.25 Gb/s user data of payload has been realized, which means this scheme has better tolerance and could be easily realized.

  17. User's manual for the reactor burnup system, REBUS

    International Nuclear Information System (INIS)

    Olson, A.P.; Regis, J.P.; Meneley, D.A.; Hoover, L.J.

    1972-01-01

    A user's manual for the REBUS System (REactor BUrnup System) is presented. Its primary purpose is to provide sufficient information about the REBUS capability to the user to ensure its efficient utilization. The current REBUS System either solves for the infinite time (equilibrium) operating conditions of a recycle system under fixed conditions, or solves for operating conditions during a single time step (non-equilibrium). The capability of studying various in-reactor fuel management and ex-reactor fuel management schemes has been included. REBUS has been operated with one- and two-dimensional diffusion theory neutronics solutions up to the present time. The model was specifically designed for extension to other neutronics models such as three-dimensional diffusion or transport theory and direct or synthesis solutions

  18. ARIES segmented gamma-ray scanner user manual

    International Nuclear Information System (INIS)

    Biddle, R.S.; Sheppard, G.A.; Schneider, C.M.

    1998-01-01

    The segmented gamma-ray scatter (SGS) designated as Win SGS at the Los Alamos Plutonium Facility has been installed and is intended for use in quantifying the radioisotope content of DOE-STD-3013-96 equivalent containers. The SGS features new software written in C and a new user interface that runs under Microsoft Windows trademark. The operation of the ARIES Segmented Gamma-ray Scanner is documented in this manual. It covers user instructions as well as hardware and software details. Additional information is found in the documentation for the commercially available components and modules that compose the SGS. The objective of the ARIES project is to demonstrate technology to dismantle plutonium pits from excess nuclear weapons, convert the plutonium to a metal ingot or an oxide powder, package the metal or oxide, and verify the contents of the package by nondestructive assay

  19. CALENDF-2010: user manual; Manuel d'utilisation de CALENDF - 2010

    Energy Technology Data Exchange (ETDEWEB)

    Sublet, Jean-Christophe [United Kingdom Atomic Energy Authority, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Ribon, Pierre; Coste-Delclaux, Mireille [CEA Saclay, Direction de L' energie Nucleaire, Direction Deleguee aux Activites Nucleaires de Saclay, Departement de Modelisation des Systemes et Structures, Service d' etudes des Reacteurs et de Mathematiques Appliquees, DEN/DANS/DM2S/SERMA, 91191 Gif-sur-Yvette Cedex (France)

    2011-09-15

    CALENDF-2010 represents a Fortran-95 update of the 1994, 2001 then 2005 code distribution with emphasise on programming quality and standards, physics and usage improvements. Devised to process multigroup cross-sections it relies on Gauss quadrature mathematical principle and strength. The followings processes can be handled by the code: moment probability table and effective cross-section calculation; pointwise cross section, probability table and effective cross-section regrouping; probability table condensation; probability table mix for several isotopes; probability table interpolation; effective cross section based probability table calculations; probability table calculations from effective cross-sections; cross-section comparison, complete energy pointwise cross-section processing and thickness dependent averaged transmission sample calculation. The CALENDF user manual, after having listed all principal code functions, describes sequentially each of them and gives comments on their associated output streams. Installation procedures, test cases and running time platform comparisons are given in the appendix. (authors)

  20. Calendf-2002: user manual; Calendf-2002: manuel d'utilisation

    Energy Technology Data Exchange (ETDEWEB)

    Sublet, J.Ch.; Ribon, P.; Coste-Delcaux, M. [CEA Cadarache, 13 - Saint-Paul-lez-Durance (France)]|[CEA Saclay, Dept. Modelisation de Systemes et Structures, 91 - Gif sur Yvette (France)

    2003-02-01

    CALENDF-2002 represents a Fortran-95 update of the 1994 code distribution with emphasize on programming quality and standard, physics and usage improvements. Devised to process multigroup cross-sections it relies on GAUSS quadratures mathematical principle and strength. The followings processes can be handled by the code: moment probability tables and effective cross-sections calculation; regroups pointwise cross sections, probability tables and effective cross-sections; probability table condensation; probability table mix for several isotopes; probability table interpolation; effective cross section based probability table calculations; probability table calculations from effective cross-sections; cross-section comparison, complete energies pointwise cross-section processing and thickness dependant averaged transmission sample calculation. The CALENDF user manual. after having listed all principal code functions, describes sequentially each of them and gives comments on their associated output streams. Installation procedures, test cases and running time platforms comparisons are given in the appendix. (authors)

  1. CalendF-2005: user manual; CalendF2005: manuel d'utilisation

    Energy Technology Data Exchange (ETDEWEB)

    Sublet, J.Ch. [CEA Cadarache, Dept. d' Etudes des Reacteurs, 13 - Saint Paul lez Durance (France); Ribon, P.; Coste-Delclaux, M. [CEA Saclay Dept. Modelisation de Systemes et Structures, 91 - Gif sur Yvette (France)

    2006-07-01

    CALENDF-2005 represents a Fortran-95 update of the 1994 code distribution with emphasize on programming quality and standard, physics and usage improvements. Devised to process multigroup cross-sections it relies on Gauss quadratures mathematical principle and strength. The followings processes can be handled by the code: moment probability tables and effective cross-sections calculation; regroups pointwise cross sections, probability tables and effective cross-sections; probability table condensation; probability table mix for several isotopes; probability table interpolation; effective cross section based probability table calculations; probability table calculations from effective cross-sections; cross-section comparison, complete energy pointwise cross-section processing and thickness dependant averaged transmission sample calculation, The CALENDF user manual, after having listed all principal code functions, describes sequenty each of them and gives comments on their associated output streams. Installation procedures, test cases and running time platforms comparisons are given in the appendix. (authors)

  2. Stimulation model for lenticular sands: Volume 2, Users manual

    Energy Technology Data Exchange (ETDEWEB)

    Rybicki, E.F.; Luiskutty, C.T.; Sutrick, J.S.; Palmer, I.D.; Shah, G.H.; Tomutsa, L.

    1987-07-01

    This User's Manual contains information for four fracture/proppant models. TUPROP1 contains a Geertsma and de Klerk type fracture model. The section of the program utilizing the proppant fracture geometry data from the pseudo three-dimensional highly elongated fracture model is called TUPROPC. The analogous proppant section of the program that was modified to accept fracture shape data from SA3DFRAC is called TUPROPS. TUPROPS also includes fracture closure. Finally there is the penny fracture and its proppant model, PENNPROP. In the first three chapters, the proppant sections are based on the same theory for determining the proppant distribution but have modifications to support variable height fractures and modifications to accept fracture geometry from three different fracture models. Thus, information about each proppant model in the User's Manual builds on information supplied in the previous chapter. The exception to the development of combined treatment models is the penny fracture and its proppant model. In this case, a completely new proppant model was developed. A description of how to use the combined treatment model for the penny fracture is contained in Chapter 4. 2 refs.

  3. A Procedure for Structural Weight Estimation of Single Stage to Orbit Launch Vehicles (Interim User's Manual)

    Science.gov (United States)

    Martinovic, Zoran N.; Cerro, Jeffrey A.

    2002-01-01

    This is an interim user's manual for current procedures used in the Vehicle Analysis Branch at NASA Langley Research Center, Hampton, Virginia, for launch vehicle structural subsystem weight estimation based on finite element modeling and structural analysis. The process is intended to complement traditional methods of conceptual and early preliminary structural design such as the application of empirical weight estimation or application of classical engineering design equations and criteria on one dimensional "line" models. Functions of two commercially available software codes are coupled together. Vehicle modeling and analysis are done using SDRC/I-DEAS, and structural sizing is performed with the Collier Research Corp. HyperSizer program.

  4. AutoBayes Program Synthesis System Users Manual

    Science.gov (United States)

    Schumann, Johann; Jafari, Hamed; Pressburger, Tom; Denney, Ewen; Buntine, Wray; Fischer, Bernd

    2008-01-01

    Program synthesis is the systematic, automatic construction of efficient executable code from high-level declarative specifications. AutoBayes is a fully automatic program synthesis system for the statistical data analysis domain; in particular, it solves parameter estimation problems. It has seen many successful applications at NASA and is currently being used, for example, to analyze simulation results for Orion. The input to AutoBayes is a concise description of a data analysis problem composed of a parameterized statistical model and a goal that is a probability term involving parameters and input data. The output is optimized and fully documented C/C++ code computing the values for those parameters that maximize the probability term. AutoBayes can solve many subproblems symbolically rather than having to rely on numeric approximation algorithms, thus yielding effective, efficient, and compact code. Statistical analysis is faster and more reliable, because effort can be focused on model development and validation rather than manual development of solution algorithms and code.

  5. DeCART v1.2 User's Manual

    International Nuclear Information System (INIS)

    Cho, J. Y.; Kim, K. S.; Kim, H. Y.; Lee, C. C.; Zee, S. Q; Joo, H. G.

    2007-07-01

    DeCART (Deterministic Core Analysis based on Ray Tracing) is a whole core neutron transport code capable of direct subpin level flux calculation at power generating conditions. It does not require a priori homogenization nor group condensation needed in conventional reactor physics calculations. The depletion and transient calculation capabilities are also available. This manual serves as a self-sufficient guide to use the code. First of all, the various features of the code are explained which encompass various modeling options as well as the basic calculation functionalities. The instructions for running the code are also given with a description of the output files generated. Next, the underlying concepts and principles of preparing a DeCART model for a problem under consideration are presented. Each part of the input needed to specify the geometry, material composition, thermal operating condition, program execution control parameters are explained with examples. The descriptions of all the input cards are then followed. Finally, various sample model inputs ranging from a simple 2D pin cell to a realistic 3D core problem, steady-state to transient problems, and from rectangular to hexagonal core problems are presented

  6. DeCART v1.1 user's manual

    International Nuclear Information System (INIS)

    Cho, J. Y.; Kim, K. S.; Kim, H. Y.; Lee, C. C.; Zee, S. Q.; Joo, H. G.

    2005-03-01

    DeCART (Deterministic Core Analysis based on Ray Tracing) is a whole core neutron transport code capable of direct subpin level flux calculation at power generating conditions. It does not require a priori homogenization nor group condensation needed in conventional reactor physics calculations. The depletion and transient calculation capabilities are also available. This manual serves as a self-sufficient guide to use the code. First of all, the various features of the code are explained which encompass various modeling options as well as the basic calculation functionalities. The instructions for running the code are also given with a description of the output files generated. Next, the underlying concepts and principles of preparing a DeCART model for a problem under consideration are presented. Each part of the input needed to specify the geometry, material composition, thermal operating condition, program execution control parameters are explained with examples. The descriptions of all the input cards are then followed. Finally, various sample model inputs ranging from a simple 2D pin cell to a realistic 3D core problem, steady-state to transient problems, are presented

  7. URSULA2 computer program. Volume 3. User's manual. Final report

    International Nuclear Information System (INIS)

    Singhal, A.K.

    1980-01-01

    This report is intended to provide documentation for the users of the URSULA2 code so that they can appreciate its important features such as: code structure, flow chart, grid notations, coding style, usage of secondary storage and its interconnection with the input preparation program (Reference H3201/4). Subroutines and subprograms have been divided into four functional groups. The functions of all subroutines have been explained with particular emphasis on the control subroutine (MAIN) and the data input subroutine (BLOCK DATA). Computations for the flow situations similar to the reference case can be performed simply by making alterations in BLOCK DATA. Separate guides for the preparation of input data and for the interpretation of program output have been provided. Furthermore, two appendices; one for the URSULA2 listing and the second for the glossary of FORTRAN variables, are included to make this report self-sufficient

  8. User's manual and guide to SALT3 and SALT4: two-dimensional computer codes for analysis of test-scale underground excavations for the disposal of radioactive waste in bedded salt deposits

    International Nuclear Information System (INIS)

    Lindner, E.N.; St John, C.M.; Hart, R.D.

    1984-02-01

    SALT3 and SALT4 are two-dimensional analytical/displacement-discontinuity codes designed to evaluate temperatures, deformation, and stresses associated with underground disposal of radioactive waste in bedded salt. These codes were developed by the University of Minnesota for the Office of Nuclear Waste Isolation in 1979. The present documentation describes the mathematical equations of the physical system being modeled, the numerical techniques utilized, and the organization of these computer codes. The SALT3 and SALT4 codes can simulate: (a) viscoelastic behavior in pillars adjacent to excavations; (b) transversely isotropic elastic moduli such as those exhibited by bedded or stratified rock; and (c) excavation sequence. Major advantages of these codes are: (a) computational efficiency; (b) the small amount of input data required; and (c) a creep law based on laboratory experimental data for salt. The main disadvantage is that some of the assumptions in the formulation of the codes, i.e., the homogeneous elastic half-space and temperature-independent material properties, render it unsuitable for canister-scale analysis or analysis of lateral deformation of the pillars. The SALT3 and SALT4 codes can be used for parameter sensitivity analyses of two-dimensional, repository-scale, thermomechanical response in bedded salt during the excavation, operational, and post-closure phases. It is especially useful in evaluating alternative patterns and sequences of excavation or waste canister placement. SALT3 is a refinement of an earlier code, SALT, and includes a fully anelastic creep model and thermal stress routine. SALT4 is a later version, and incorporates a revised creep model which is strain-hardening

  9. User manual for SPLASH (Single Panel Lamp and Shroud Helper).

    Energy Technology Data Exchange (ETDEWEB)

    Larsen, Marvin Elwood

    2006-02-01

    The radiant heat test facility develops test sets providing well-characterized thermal environments, often representing fires. Many of the components and procedures have become standardized to such an extent that the development of a specialized design tool to determine optimal configurations for radiant heat experiments was appropriate. SPLASH (Single Panel Lamp and Shroud Helper) is that tool. SPLASH is implemented as a user-friendly, Windows-based program that allows a designer to describe a test setup in terms of parameters such as number of lamps, power, position, and separation distance. This document is a user manual for that software. Any incidental descriptions of theory are only for the purpose of defining the model inputs. The theory for the underlying model is described in SAND2005-2947 (Ref. [1]). SPLASH provides a graphical user interface to define lamp panel and shroud designs parametrically, solves the resulting radiation enclosure problem for up to 2500 surfaces, and provides post-processing to facilitate understanding and documentation of analyzed designs.

  10. Tree Cover Mapping Tool—Documentation and user manual

    Science.gov (United States)

    Cotillon, Suzanne E.; Mathis, Melissa L.

    2016-06-02

    The Tree Cover Mapping (TCM) tool was developed by scientists at the U.S. Geological Survey Earth Resources Observation and Science Center to allow a user to quickly map tree cover density over large areas using visual interpretation of high resolution imagery within a geographic information system interface. The TCM tool uses a systematic sample grid to produce maps of tree cover. The TCM tool allows the user to define sampling parameters to estimate tree cover within each sample unit. This mapping method generated the first on-farm tree cover maps of vast regions of Niger and Burkina Faso. The approach contributes to implementing integrated landscape management to scale up re-greening and restore degraded land in the drylands of Africa. The TCM tool is easy to operate, practical, and can be adapted to many other applications such as crop mapping, settlements mapping, or other features. This user manual provides step-by-step instructions for installing and using the tool, and creating tree cover maps. Familiarity with ArcMap tools and concepts is helpful for using the tool.

  11. User's manual for the HYPGEN hyperbolic grid generator and the HGUI graphical user interface

    Science.gov (United States)

    Chan, William M.; Chiu, Ing-Tsau; Buning, Pieter G.

    1993-01-01

    The HYPGEN program is used to generate a 3-D volume grid over a user-supplied single-block surface grid. This is accomplished by solving the 3-D hyperbolic grid generation equations consisting of two orthogonality relations and one cell volume constraint. In this user manual, the required input files and parameters and output files are described. Guidelines on how to select the input parameters are given. Illustrated examples are provided showing a variety of topologies and geometries that can be treated. HYPGEN can be used in stand-alone mode as a batch program or it can be called from within a graphical user interface HGUI that runs on Silicon Graphics workstations. This user manual provides a description of the menus, buttons, sliders, and typein fields in HGUI for users to enter the parameters needed to run HYPGEN. Instructions are given on how to configure the interface to allow HYPGEN to run either locally or on a faster remote machine through the use of shell scripts on UNIX operating systems. The volume grid generated is copied back to the local machine for visualization using a built-in hook to PLOT3D.

  12. GENII (Generation II): The Hanford Environmental Radiation Dosimetry Software System: Volume 3, Code maintenance manual: Hanford Environmental Dosimetry Upgrade Project

    Energy Technology Data Exchange (ETDEWEB)

    Napier, B.A.; Peloquin, R.A.; Strenge, D.L.; Ramsdell, J.V.

    1988-09-01

    The Hanford Environmental Dosimetry Upgrade Project was undertaken to incorporate the internal dosimetry models recommended by the International Commission on Radiological Protection (ICRP) in updated versions of the environmental pathway analysis models used at Hanford. The resulting second generation of Hanford environmental dosimetry computer codes is compiled in the Hanford Environmental Dosimetry System (Generation II, or GENII). This coupled system of computer codes is intended for analysis of environmental contamination resulting from acute or chronic releases to, or initial contamination of, air, water, or soil, on through the calculation of radiation doses to individuals or populations. GENII is described in three volumes of documentation. This volume is a Code Maintenance Manual for the serious user, including code logic diagrams, global dictionary, worksheets to assist with hand calculations, and listings of the code and its associated data libraries. The first volume describes the theoretical considerations of the system. The second volume is a Users' Manual, providing code structure, users' instructions, required system configurations, and QA-related topics. 7 figs., 5 tabs.

  13. GENII [Generation II]: The Hanford Environmental Radiation Dosimetry Software System: Volume 3, Code maintenance manual: Hanford Environmental Dosimetry Upgrade Project

    International Nuclear Information System (INIS)

    Napier, B.A.; Peloquin, R.A.; Strenge, D.L.; Ramsdell, J.V.

    1988-09-01

    The Hanford Environmental Dosimetry Upgrade Project was undertaken to incorporate the internal dosimetry models recommended by the International Commission on Radiological Protection (ICRP) in updated versions of the environmental pathway analysis models used at Hanford. The resulting second generation of Hanford environmental dosimetry computer codes is compiled in the Hanford Environmental Dosimetry System (Generation II, or GENII). This coupled system of computer codes is intended for analysis of environmental contamination resulting from acute or chronic releases to, or initial contamination of, air, water, or soil, on through the calculation of radiation doses to individuals or populations. GENII is described in three volumes of documentation. This volume is a Code Maintenance Manual for the serious user, including code logic diagrams, global dictionary, worksheets to assist with hand calculations, and listings of the code and its associated data libraries. The first volume describes the theoretical considerations of the system. The second volume is a Users' Manual, providing code structure, users' instructions, required system configurations, and QA-related topics. 7 figs., 5 tabs

  14. 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.

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

    International Nuclear Information System (INIS)

    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.

    2002-01-01

    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. SCDAP/RELAP5/MOD2 code manual

    International Nuclear Information System (INIS)

    Allison, C.M.; Johnson, E.C.

    1989-09-01

    The SCDAP/RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, and the fission products and aerosols in the system during a severe accident transient as well as large and small break loss-of-coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. The modeling theory and associated numerical schemes are documented in Volumes I and in this document, Volume II, to acquaint the user with the modeling base and thus aid in effective use of the code. 135 refs., 48 figs., 8 tabs

  17. User's manual of JT-60 experimental data analysis system

    International Nuclear Information System (INIS)

    Hirayama, Takashi; Morishima, Soichi; Yoshioka, Yuji

    2010-02-01

    In the Japan Atomic Energy Agency Naka Fusion Institute, a lot of experiments have been conducted by using the large tokamak device JT-60 aiming to realize fusion power plant. In order to optimize the JT-60 experiment and to investigate complex characteristics of plasma, JT-60 experimental data analysis system was developed and used for collecting, referring and analyzing the JT-60 experimental data. Main components of the system are a data analysis server and a database server for the analyses and accumulation of the experimental data respectively. Other peripheral devices of the system are magnetic disk units, NAS (Network Attached Storage) device, and a backup tape drive. This is a user's manual of the JT-60 experimental data analysis system. (author)

  18. Image enhancement software for underwater recovery operations: User's manual

    Science.gov (United States)

    Partridge, William J.; Therrien, Charles W.

    1989-06-01

    This report describes software for performing image enhancement on live or recorded video images. The software was developed for operational use during underwater recovery operations at the Naval Undersea Warfare Engineering Station. The image processing is performed on an IBM-PC/AT compatible computer equipped with hardware to digitize and display video images. The software provides the capability to provide contrast enhancement and other similar functions in real time through hardware lookup tables, to automatically perform histogram equalization, to capture one or more frames and average them or apply one of several different processing algorithms to a captured frame. The report is in the form of a user manual for the software and includes guided tutorial and reference sections. A Digital Image Processing Primer in the appendix serves to explain the principle concepts that are used in the image processing.

  19. Water Security Toolkit User Manual: Version 1.3 | Science ...

    Science.gov (United States)

    User manual: Data Product/Software The Water Security Toolkit (WST) is a suite of tools that help provide the information necessary to make good decisions resulting in the minimization of further human exposure to contaminants, and the maximization of the effectiveness of intervention strategies. WST assists in the evaluation of multiple response actions in order to select the most beneficial consequence management strategy. It includes hydraulic and water quality modeling software and optimization methodologies 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 to confirm contamination or cleanup and (6) valves to close in order to isolate contaminated areas of the network.

  20. User Manual for the Allpix$^2$ Simulation Framework

    CERN Document Server

    AUTHOR|(SzGeCERN)818092; Spannagel, Simon; Hynds, Daniel

    2017-01-01

    Several simulation tools exist for the detailed study of position sensitive silicon detectors, covering aspects ranging from the electrical properties of sensors to the behaviour of charged particles traversing a given detector setup. Each of these toolkits performs a very specialised task, and for the complete description of a silicon detector several such software packages must typically be used. Allpix$^2$ builds upon this work by providing a complete and easy-to-use C++ software package for simulating detector performance, from the interaction of particles to the digitisation of propagated carriers by the front-end electronics. A modular framework is used to flexibly add or remove modules from the simulation chain, each performing specific tasks such as interfacing to Geant4 to deposit energy in the detector and provide an accurate description of material effects, or the propagation of deposited charges through the sensor bulk. This document presents the user manual of the software as of release version 1...

  1. EPANET Multi-Species Extension Software and User's Manual ...

    Science.gov (United States)

    Software and User's Manual EPANET is used in homeland security research to model contamination threats to water systems. Historically, EPANET has been limited to tracking the dynamics of a single chemical transported through a network of pipes and storage tanks, such as a fluoride used in a tracer study or free chlorine used in a disinfection decay study. Recently, the NHSRC released a new extension to EPANET called EPANET-MSX (Multi-Species eXtension) that allows for the consideration of multiple interacting species in the bulk flow and on the pipe walls. This capability has been incorporated into both a stand-alone executable program as well as a toolkit library of functions that programmers can use to build customized applications.

  2. CONPAS 1.0 (CONtainment Performance Analysis System). User`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Kwang Il; Jin, Young Ho [Korea Atomic Energy Research Institute, Daeduk (Korea, Republic of)

    1996-04-01

    CONPAS (CONtainment Performance Analysis System) is a verified computer code package to integrate the numerical, graphical, and results-operation aspects of Level 2 probabilistic safety assessments (PSA) for nuclear power plants automatically under a PC window environment. Compared with the existing DOS-based computer codes for Level 2 PSA, the most important merit of the window-based computer code is that user can easily describe and quantify the accident progression models, and manipulate the resultant outputs in a variety of ways. As a main logic for accident progression analysis, CONPAS employs a concept of the small containment phenomenological event tree (CPET) helpful to trace out visually individual accident progressions and of the large supporting event tree (LSET) for its detailed quantification. For the integrated analysis of Level 2 PSA, the code utilizes four distinct, but closely related modules; (1) ET Editor for construction of several event tree models describing the accident progressions, (2) Computer for quantification of the constructed event trees and graphical display of the resultant outputs, (3) Text Editor for preparation of input decks for quanification and utilization of calculational results, and (4) Mechanistic Code Plotter for utilization of results obtained from severe accident analysis codes. Compared with other existing computer codes for Level 2 PSA, the CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, sensitivity analysis and data interpretation, reporting aspects including tabling and graphic as well as user-friend interface. 10 refs. (Author) .new.

  3. User's manual for the Composite HTGR Analysis Program (CHAP-1)

    International Nuclear Information System (INIS)

    Gilbert, J.S.; Secker, P.A. Jr.; Vigil, J.C.; Wecksung, M.J.; Willcutt, G.J.E. Jr.

    1977-03-01

    CHAP-1 is the first release version of an HTGR overall plant simulation program with both steady-state and transient solution capabilities. It consists of a model-independent systems analysis program and a collection of linked modules, each representing one or more components of the HTGR plant. Detailed instructions on the operation of the code and detailed descriptions of the HTGR model are provided. Information is also provided to allow the user to easily incorporate additional component modules, to modify or replace existing modules, or to incorporate a completely new simulation model into the CHAP systems analysis framework

  4. User's guide to the biosphere code BIOMOD

    International Nuclear Information System (INIS)

    Kane, P.

    1983-05-01

    BIOMOD has been designed to interface with SYVAC, the function of which is to perform generic risk assessments on hypothetical repository-geosphere-biosphere combinations. The user's guide contains the detailed specifications for the models used, a description of the interim user-interface, a specification for required input and definition of output. Sources of error are indicated and reference is made to the database description and other documents issued relating to BIOMOD. (author)

  5. Performance of automated and manual coding systems for occupational data: a case study of historical records.

    Science.gov (United States)

    Patel, Mehul D; Rose, Kathryn M; Owens, Cindy R; Bang, Heejung; Kaufman, Jay S

    2012-03-01

    Occupational data are a common source of workplace exposure and socioeconomic information in epidemiologic research. We compared the performance of two occupation coding methods, an automated software and a manual coder, using occupation and industry titles from U.S. historical records. We collected parental occupational data from 1920-40s birth certificates, Census records, and city directories on 3,135 deceased individuals in the Atherosclerosis Risk in Communities (ARIC) study. Unique occupation-industry narratives were assigned codes by a manual coder and the Standardized Occupation and Industry Coding software program. We calculated agreement between coding methods of classification into major Census occupational groups. Automated coding software assigned codes to 71% of occupations and 76% of industries. Of this subset coded by software, 73% of occupation codes and 69% of industry codes matched between automated and manual coding. For major occupational groups, agreement improved to 89% (kappa = 0.86). Automated occupational coding is a cost-efficient alternative to manual coding. However, some manual coding is required to code incomplete information. We found substantial variability between coders in the assignment of occupations although not as large for major groups.

  6. Extension of an Object-Oriented Optimization Tool: User's Reference Manual

    Science.gov (United States)

    Pak, Chan-Gi; Truong, Samson S.

    2015-01-01

    The National Aeronautics and Space Administration Armstrong Flight Research Center has developed a cost-effective and flexible object-oriented optimization (O (sup 3)) tool that leverages existing tools and practices and allows easy integration and adoption of new state-of-the-art software. This object-oriented framework can integrate the analysis codes for multiple disciplines, as opposed to relying on one code to perform analysis for all disciplines. Optimization can thus take place within each discipline module, or in a loop between the O (sup 3) tool and the discipline modules, or both. Six different sample mathematical problems are presented to demonstrate the performance of the O (sup 3) tool. Instructions for preparing input data for the O (sup 3) tool are detailed in this user's manual.

  7. Tokamak plasma power balance calculation code (TPC code) outline and operation manual

    International Nuclear Information System (INIS)

    Fujieda, Hirobumi; Murakami, Yoshiki; Sugihara, Masayoshi.

    1992-11-01

    This report is a detailed description on the TPC code, that calculates the power balance of a tokamak plasma according to the ITER guidelines. The TPC code works on a personal computer (Macintosh or J-3100/ IBM-PC). Using input data such as the plasma shape, toroidal magnetic field, plasma current, electron temperature, electron density, impurities and heating power, TPC code can determine the operation point of the fusion reactor (Ion temperature is assumed to be equal to the electron temperature). Supplied flux (Volt · sec) and burn time are also estimated by coil design parameters. Calculated energy confinement time is compared with various L-mode scaling laws and the confinement enhancement factor (H-factor) is evaluated. Divertor heat load is predicted by using simple scaling models (constant-χ, Bohm-type-χ and JT-60U empirical scaling models). Frequently used data can be stored in a 'device file' and used as the default values. TPC code can generate 2-D mesh data and the POPCON plot is drawn by a contour line plotting program (CONPLT). The operation manual about CONPLT code is also described. (author)

  8. LLCEDATA and LLCECALC for Windows version 1.0, Volume 1: User`s manual

    Energy Technology Data Exchange (ETDEWEB)

    McFadden, J.G.

    1998-09-04

    LLCEDATA and LLCECALC for Windows are user-friendly computer software programs that work together to determine the proper waste designation, handling, and disposition requirements for Long Length Contaminated Equipment (LLCE). LLCEDATA reads from a variety of data bases to produce an equipment data file (EDF) that represents a snapshot of both the LLCE and the tank it originates from. LLCECALC reads the EDF and a gamma assay (AV2) file that is produced by the Flexible Receiver Gamma Energy Analysis System. LLCECALC performs corrections to the AV2 file as it is being read and characterizes the LLCE. Both programs produce a variety of reports, including a characterization report and a status report. The status report documents each action taken by the user, LLCEDATA, and LLCECALC. Documentation for LLCEDATA and LLCECALC for Windows is available in three volumes. Volume 1 is a user`s manual, which is intended as a quick reference for both LLCEDATA and LLCECALC. Volume 2 is a technical manual, and Volume 3 is a software verification and validation document.

  9. GENII: The Hanford Environmental Radiation Dosimetry Software System: Volume 2, Users' manual: Hanford Environmental Dosimetry Upgrade Project

    International Nuclear Information System (INIS)

    Napier, B.A.; Peloquin, R.A.; Strenge, D.L.; Ramsdell, J.V.

    1988-11-01

    The Hanford Environmental Dosimetry Upgrade Project was undertaken to incorporate the internal dosimetry models recommended by the International Commission on Radiological Protection (ICRP) in updated versions of the environmental pathway analysis models used at Hanford. The resulting second generation of Hanford environmental dosimetry computer codes is compiled in the Hanford Environmental Dosimetry System (Generation II, or GENII). The purpose of this coupled system of computer codes is to analyze environmental contamination of, air, water, or soil. This is accomplished by calculating radiation doses to individuals or populations. GENII is described in three volumes of documentation. This second volume is a Users' Manual, providing code structure, users' instructions, required system configurations, and QA-related topics. The first volume describes the theoretical considerations of the system. The third volume is a Code Maintenance Manual for the user who requires knowledge of code detail. It includes logic diagrams, global dictionary, worksheets, example hand calculations, and listings of the code and its associated data libraries. 27 refs., 17 figs., 23 tabs

  10. Multi-Sector Sustainability Browser (MSSB) User Manual: A ...

    Science.gov (United States)

    EPA’s Sustainable and Healthy Communities (SHC) Research Program is developing methodologies, resources, and tools to assist community members and local decision makers in implementing policy choices that facilitate sustainable approaches in managing their resources affecting the built environment, natural environment, and human health. In order to assist communities and decision makers in implementing sustainable practices, EPA is developing computer-based systems including models, databases, web tools, and web browsers to help communities decide upon approaches that support their desired outcomes. Communities need access to resources that will allow them to achieve their sustainability objectives through intelligent decisions in four key sustainability areas: • Land Use • Buildings and Infrastructure • Transportation • Materials Management (i.e., Municipal Solid Waste [MSW] processing and disposal) The Multi-Sector Sustainability Browser (MSSB) is designed to support sustainable decision-making for communities, local and regional planners, and policy and decision makers. Document is an EPA Technical Report, which is the user manual for the Multi-Sector Sustainability Browser (MSSB) tool. The purpose of the document is to provide basic guidance on use of the tool for users

  11. Computer Program for Calculation of Complex Chemical Equilibrium Compositions and Applications II. Users Manual and Program Description. 2; Users Manual and Program Description

    Science.gov (United States)

    McBride, Bonnie J.; Gordon, Sanford

    1996-01-01

    This users manual is the second part of a two-part report describing the NASA Lewis CEA (Chemical Equilibrium with Applications) program. The program obtains chemical equilibrium compositions of complex mixtures with applications to several types of problems. The topics presented in this manual are: (1) details for preparing input data sets; (2) a description of output tables for various types of problems; (3) the overall modular organization of the program with information on how to make modifications; (4) a description of the function of each subroutine; (5) error messages and their significance; and (6) a number of examples that illustrate various types of problems handled by CEA and that cover many of the options available in both input and output. Seven appendixes give information on the thermodynamic and thermal transport data used in CEA; some information on common variables used in or generated by the equilibrium module; and output tables for 14 example problems. The CEA program was written in ANSI standard FORTRAN 77. CEA should work on any system with sufficient storage. There are about 6300 lines in the source code, which uses about 225 kilobytes of memory. The compiled program takes about 975 kilobytes.

  12. User effects on the transient system code calculations. Final report

    International Nuclear Information System (INIS)

    Aksan, S.N.; D'Auria, F.

    1995-01-01

    Large thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants to optimize operational procedures and the plant design itself. Evaluation of the capabilities of these codes are accomplished by comparing the code predictions with the measured experimental data obtained from various types of separate effects and integral test facilities. In recent years, some attempts have been made to establish methodologies to evaluate the accuracy and the uncertainty of the code predictions and consequently judgement on the acceptability of the codes. In none of the methodologies has the influence of the code user on the calculated results been directly addressed. In this paper, the results of the investigations on the user effects for the thermal-hydraulic transient system codes is presented and discussed on the basis of some case studies. The general findings of the investigations show that in addition to user effects, there are other reasons that affect the results of the calculations and which are hidden under user effects. Both the hidden factors and the direct user effects are discussed in detail and general recommendations and conclusions are presented to control and limit them

  13. Graphical user interface development for the MARS code

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  14. High School and Beyond Transcripts Survey (1982). Data File User's Manual. Contractor Report.

    Science.gov (United States)

    Jones, Calvin; And Others

    This data file user's manual documents the procedures used to collect and process high school transcripts for a large sample of the younger cohort (1980 sophomores) in the High School and Beyond survey. The manual provides the user with the technical assistance needed to use the computer file and also discusses the following: (1) sample design for…

  15. Users' Manual for Research: Translating Head Start Findings Into Action (Expanded Notebook Version).

    Science.gov (United States)

    Grotberg, Edith H.; Fowler, Austine

    This users' manual, intended for use with a Project Head Start teacher training notebook, describes the purpose, development and field testing of the training materials and suggests procedures for using the notebook as a resource in teacher training sessions. The training notebook to which the users' manual refers is based on 11 questions in the…

  16. Energy Code Enforcement Training Manual : Covering the Washington State Energy Code and the Ventilation and Indoor Air Quality Code.

    Energy Technology Data Exchange (ETDEWEB)

    Washington State Energy Code Program

    1992-05-01

    This manual is designed to provide building department personnel with specific inspection and plan review skills and information on provisions of the 1991 edition of the Washington State Energy Code (WSEC). It also provides information on provisions of the new stand-alone Ventilation and Indoor Air Quality (VIAQ) Code.The intent of the WSEC is to reduce the amount of energy used by requiring energy-efficient construction. Such conservation reduces energy requirements, and, as a result, reduces the use of finite resources, such as gas or oil. Lowering energy demand helps everyone by keeping electricity costs down. (It is less expensive to use existing electrical capacity efficiently than it is to develop new and additional capacity needed to heat or cool inefficient buildings.) The new VIAQ Code (effective July, 1991) is a natural companion to the energy code. Whether energy-efficient or not, an homes have potential indoor air quality problems. Studies have shown that indoor air is often more polluted than outdoor air. The VIAQ Code provides a means of exchanging stale air for fresh, without compromising energy savings, by setting standards for a controlled ventilation system. It also offers requirements meant to prevent indoor air pollution from building products or radon.

  17. SAFEPAQ-II. User manual[Nuclear data

    Energy Technology Data Exchange (ETDEWEB)

    Forrest, R.A

    2001-03-01

    SAFEPAQ-II is the new software tool that has been developed to enable efficient production of the EAF nuclear data libraries that are required as input to the FISPACT activation code. It forms part of the European Activation System (EASY), and replaces SAFEPAQ and SYMPAL that were used previously. It enables all the nuclear data to be stored in relational databases (Access) and by using an interactive user interface allows the data to be viewed, modified, validated and then produced in the required EAF format as text files. It is written in Visual Basic and runs under the Windows NT4 and 98 operating systems. The Windows operating system has the great advantage of portability and SAFEPAQ-II has been successfully installed at two external sites for use by UKAEA's international collaborators. It has been used in the production of the EAF-2001 data libraries. (author)

  18. The International Nuclear Event Scale (INES) user's manual. 2001 edition

    International Nuclear Information System (INIS)

    2001-12-01

    The International Nuclear Event Scale (INES) was introduced in March 1990 jointly by the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA). Its primary purpose is to facilitate communication and understanding between the nuclear community, the media and the public on the safety significance of events occurring at nuclear installations. The scale was refined in 1992 in the light of experience gained and extended to be applicable to any event associated with radioactive material and/or radiation, including the transport of radioactive materials.This edition of the INES User's Manual incorporates experience gained from applying the 1992 version of the scale and the document entitled 'Clarification of Issues Raised'. As such, it replaces those earlier publications. It does not amend the technical basis of the INES rating procedure but is expected to facilitate the task of those who are required to rate the safety significance of events using the INES scale. The INES communication network currently receives and disseminates event information to the INES National Officers of 60 Member States on special Event Rating Forms which represent official information on the events, including the rating. The INES communication process has led each participating country to set up an internal network which ensures that all events are promptly communicated and rated whenever they have to be reported outside or inside the country. The IAEA provides training services on the use of INES on request

  19. INES - The International Nuclear Event Scale. User's manual

    International Nuclear Information System (INIS)

    2005-01-01

    The International Nuclear Event Scale (INES) was introduced in March 1990 jointly by the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA). Its primary purpose is to facilitate communication and understanding between the nuclear community, the media and the public on the safety significance of events occurring at nuclear installations. The scale was refined in 1992 in the light of experience gained and extended to be applicable to any event associated with radioactive material and/or radiation, including the transport of radioactive materials. This edition of the INES User's Manual incorporates experience gained from applying the 1992 version of the scale and the document entitled 'Clarification of Issues Raised'. As such, it replaces those earlier publications. It does not amend the technical basis of the INES rating procedure but is expected to facilitate the task of those who are required to rate the safety significance of events using the INES scale. The INES communication network currently receives and disseminates event information to the INES National Officers of 60 Member States on special Event Rating Forms which represent official information on the events, including the rating. The INES communication process has led each participating country to set up an internal network which ensures that all events are promptly communicated and rated whenever they have to be reported outside or inside the country. The IAEA provides training services on the use of INES on request

  20. The International Nuclear Event Scale (INES) user's manual. 2001 edition

    International Nuclear Information System (INIS)

    2001-02-01

    The International Nuclear Event Scale (INES) was introduced in March 1990 jointly by the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA). Its primary purpose is to facilitate communication and understanding between the nuclear community, the media and the public on the safety significance of events occurring at nuclear installations. The scale was refined in 1992 in the light of experience gained and extended to be applicable to any event associated with radioactive material and/or radiation, including the transport of radioactive materials.This edition of the INES User's Manual incorporates experience gained from applying the 1992 version of the scale and the document entitled ''Clarification of Issues Raised''. As such, it replaces those earlier publications. It does not amend the technical basis of the INES rating procedure but is expected to facilitate the task of those who are required to rate the safety significance of events using the INES scale. The INES communication network currently receives and disseminates event information to the INES National Officers of 60 Member States on special Event Rating Forms which represent official information on the events, including the rating. The INES communication process has led each participating country to set up an internal network which ensures that all events are promptly communicated and rated whenever they have to be reported outside or inside the country. The IAEA provides training services on the use of INES on request

  1. User manual of nuclide dispersion in phreatic aquifers model

    International Nuclear Information System (INIS)

    Rives, D.E.

    1999-01-01

    The Nuclide Dispersion in Phreatic Aquifers (DRAF) model was developed in the 'Division Estudios Ambientales' of the 'Gerencia de Seguridad Radiologica y Nuclear, Comision Nacional de Energia Atomica' (1991), for the Safety Assessment of Near Surface Radioactive Waste Disposal Facilities. Afterwards, it was modified in several opportunities, adapting it to a number of application conditions. The 'Manual del usuario del codigo DRAF' here presented is a reference document for the use of the last three versions of the code developed for the 'Autoridad Regulatoria Nuclear' between 1995 and 1996. The DRAF model solves the three dimension's solute transport equation for porous media by the finite differences method. It takes into account the advection, dispersion, radioactive decay, and retention in the solid matrix processes, and has multiple possibilities for the source term. There are three versions of the model, two of them for the saturated zone and one for the unsaturated zone. All the versions have been verified in different conditions, and have been applied in exercises of the International Atomic Energy Agency and also in real cases. (author)

  2. Micromechanical combined stress analysis: MICSTRAN, a user manual

    Science.gov (United States)

    Naik, R. A.

    1992-01-01

    Composite materials are currently being used in aerospace and other applications. The ability to tailor the composite properties by the appropriate selection of its constituents, the fiber and matrix, is a major advantage of composite materials. The Micromechanical Combined Stress Analysis (MICSTRAN) code provides the materials engineer with a user-friendly personal computer (PC) based tool to calculate overall composite properties given the constituent fiber and matrix properties. To assess the ability of the composite to carry structural loads, the materials engineer also needs to calculate the internal stresses in the composite material. MICSTRAN is a simple tool to calculate such internal stresses with a composite ply under combined thermomechanical loading. It assumes that the fibers have a circular cross-section and are arranged either in a repeating square or diamond array pattern within a ply. It uses a classical elasticity solution technique that has been demonstrated to calculate accurate stress results. Input to the program consists of transversely isotropic fiber properties and isotropic matrix properties such as moduli, Poisson's ratios, coefficients of thermal expansion, and volume fraction. Output consists of overall thermoelastic constants and stresses. Stresses can be computed under the combined action of thermal, transverse, longitudinal, transverse shear, and longitudinal shear loadings. Stress output can be requested along the fiber-matrix interface, the model boundaries, circular arcs, or at user-specified points located anywhere in the model. The MICSTRAN program is Windows compatible and takes advantage of the Microsoft Windows graphical user interface which facilitates multitasking and extends memory access far beyond the limits imposed by the DOS operating system.

  3. Radiological Safety Analysis Computer (RSAC) Program Version 7.0 Users Manual

    International Nuclear Information System (INIS)

    Schrader, Bradley J.

    2009-01-01

    The Radiological Safety Analysis Computer (RSAC) Program Version 7.0 (RSAC-7) is the newest version of the RSAC legacy code. It calculates the consequences of a release of radionuclides to the atmosphere. A user can generate a fission product inventory from either reactor operating history or a nuclear criticality event. RSAC-7 models the effects of high-efficiency particulate air filters or other cleanup systems and calculates the decay and ingrowth during transport through processes, facilities, and the environment. Doses are calculated for inhalation, air immersion, ground surface, ingestion, and cloud gamma pathways. RSAC-7 can be used as a tool to evaluate accident conditions in emergency response scenarios, radiological sabotage events and to evaluate safety basis accident consequences. This users manual contains the mathematical models and operating instructions for RSAC-7. Instructions, screens, and examples are provided to guide the user through the functions provided by RSAC-7. This program was designed for users who are familiar with radiological dose assessment methods

  4. Investigation of Advanced Counterrotation Blade Configuration Concepts for High Speed Turboprop Systems. Task 8: Cooling Flow/heat Transfer Analysis User's Manual

    Science.gov (United States)

    Hall, Edward J.; Topp, David A.; Heidegger, Nathan J.; Delaney, Robert A.

    1994-01-01

    The focus of this task was to validate the ADPAC code for heat transfer calculations. To accomplish this goal, the ADPAC code was modified to allow for a Cartesian coordinate system capability and to add boundary conditions to handle spanwise periodicity and transpiration boundaries. This user's manual describes how to use the ADPAC code as developed in Task 5, NAS3-25270, including the modifications made to date in Tasks 7 and 8, NAS3-25270.

  5. ATHENA code manual. Volume 1. Code structure, system models, and solution methods

    International Nuclear Information System (INIS)

    Carlson, K.E.; Roth, P.A.; Ransom, V.H.

    1986-09-01

    The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code has been developed to perform transient simulation of the thermal hydraulic systems which may be found in fusion reactors, space reactors, and other advanced systems. A generic modeling approach is utilized which permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of a complete facility. Several working fluids are available to be used in one or more interacting loops. Different loops may have different fluids with thermal connections between loops. The modeling theory and associated numerical schemes are documented in Volume I in order to acquaint the user with the modeling base and thus aid effective use of the code. The second volume contains detailed instructions for input data preparation

  6. 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

  7. DECADES Tools. User's manual for version 1.0

    International Nuclear Information System (INIS)

    2000-01-01

    databases for analysis of costs and environmental burdens at the power plant, full energy chain and electric system levels, and a decision analysis tool. The DECADES Tools were developed jointly by the International Atomic Energy Agency (IAEA), Argonne National Laboratory (ANL) under contract with the IAEA funded by extrabudgetary contribution from the US Government to the IAEA, and external consultants, working under contract with the IAEA. The DECADES Tools User's Manual is published as part of a series of technical reports and documents prepared in the framework of the DECADES project. The Joint Steering Committee for the project hopes that these software tools will contribute to the process of strengthening and improving the capabilities for the design and implementation of sustainable strategies in the power sector, in particular in developing countries and countries in transition to market economies

  8. ELIST v.8.1 : User's Manual.; TOPICAL

    International Nuclear Information System (INIS)

    Van Groningen, Blachowicz D.; Duffy Braun, M.; Clemmons, M. A.; Simunich, K. L.; Timmerman, D.; VanderZee, H.; Widing, M. A.

    2002-01-01

    This user's manual documents the capabilities and functions of the Enhanced Logistics Intratheater Support Tool (ELIST) software application. Steps for using the Expanded Time Phase Force Deployment Data (ETPFDD) Editor (ETEdit), which is included in ELIST but is also a stand-alone software application, are contained in a separate document. ELIST is a discrete event simulation tool developed for use by military planners in both the continental United States (CONUS) and outside the continental United States (OCONUS). It simulates the reception, staging, onward movement, and integration (RSOI) of military personnel and equipment from all services within, between, or among countries. ELIST not only runs a simulation, but it also provides the capability to edit asset sets, networks, and scenarios. These capabilities show how various changes can affect the outcome of a simulation. Further, ELIST incorporates topographic maps on which the network is displayed. The system also allows planners to simulate scenarios at the vehicle level. Prior to the implementation of ELIST, planners were able to simulate military deployment from the point of departure (origin) to the point of arrival in the theater (the port of debarkation). Since the development and implementation of ELIST, however, planners can simulate military deployment from the point of departure (airport or seaport), through the staging area, through the theater-staging base, to the final destination. A typical scenario might be set up to transport personnel and cargo to a location by aircraft or ship. Upon arrival at the airport or seaport, the cargo would be sent to a staging area where it would be set up and transferred to a vehicle, or in the case of petroleum, oil, and lubricants (POL), a pipeline. The vehicle then would transport the cargo to the theater-staging base where it would ''marry up'' with the main body of personnel. From this point, the cargo and the main body would be transported to the final

  9. The user's manual of 'Manyo Library' data reduction software framework at MLF, J-PARC

    International Nuclear Information System (INIS)

    Inamura, Yasuhiro; Nakatani, Takeshi; Ito, Takayoshi; Suzuki, Jiro

    2016-06-01

    Manyo Library is a software framework for developing analysis software of neutron scattering data produced at MLF, J-PARC. This software framework is required to work on many instruments in MLF and to include base functions applied to various scientific purposes at beam lines. This framework mainly consists of data containers, which enable to store 1, 2 and 3 dimensional axes data for neutron scattering. Data containers have many functions to calculate four arithmetic operations with errors distribution between containers, to store the meta-data about measurements and to read or write text file. The analysis codes are constructed using various analysis operators defined in Manyo Library, which executes functions with given data containers and output the results. On the other hands, the main interface for instrument scientists and users must be easy and interactive to treat data containers and functions or to develop new analysis codes. Therefore we chose Python as user interface. Since Manyo Library is built in C++ language, we've introduced the technology to call C++ function from Python environment into the framework. As a result, we have already developed a lot of software for data reduction, analysis and visualization, which are utilized widely in beam lines at MLF. This document is the manual for the beginner to touch this framework. (author)

  10. Treatment for Abused and Neglected Children. The User Manual Series.

    Science.gov (United States)

    Martin, Harold P.

    This manual brings current knowledge about the needs of abused and neglected children together with practical ideas about what can be done to meet those needs. Designed primarily for use by child protective service (CPS) workers, the manual may also be used by other professionals dealing with child abuse and neglect. Chapter 1 briefly discusses…

  11. PRISMA single crystal cold neutron spectrometer and diffractometer. User manual

    International Nuclear Information System (INIS)

    Harris, M.; Bull, M.

    1998-01-01

    PRISMA has undergone a complete rebuild since the First Edition of this manual, resulting in a rather different instrument to that of January 1995. Now that the development of the instrument has reached a quiescent stage, this new edition of the manual describes the operation of PRISMA for the foreseeable future

  12. Spent fuel management fee methodology and computer code user's manual

    International Nuclear Information System (INIS)

    Engel, R.L.; White, M.K.

    1982-01-01

    The methodology and computer model described here were developed to analyze the cash flows for the federal government taking title to and managing spent nuclear fuel. The methodology has been used by the US Department of Energy (DOE) to estimate the spent fuel disposal fee that will provide full cost recovery. Although the methodology was designed to analyze interim storage followed by spent fuel disposal, it could be used to calculate a fee for reprocessing spent fuel and disposing of the waste. The methodology consists of two phases. The first phase estimates government expenditures for spent fuel management. The second phase determines the fees that will result in revenues such that the government attains full cost recovery assuming various revenue collection philosophies. These two phases are discussed in detail in subsequent sections of this report. Each of the two phases constitute a computer module, called SPADE (SPent fuel Analysis and Disposal Economics) and FEAN (FEe ANalysis), respectively

  13. AIDA Asia. Artificial Insemination Database Application. User manual. 1

    International Nuclear Information System (INIS)

    Garcia Podesta, Mario

    2002-01-01

    Artificial Insemination Database Application (AIDA-Asia) is a computer application to store and analyze information from AI Services (farms, females, inseminated, semen, estrus characteristics, inseminator and pregnancy diagnosis data). The need for such an application arose during a consultancy undertaken by the author for the International Atomic Energy Agency (IAEA, Vienna) under the framework of its Regional Co-operative Agreement for Asia and the Pacific (RCA) which is implementing a project on 'Improving Animal Productivity and Reproductive Efficiency' (RAS/5/035). The detailed specifications for the application were determined through a Task Force Meeting of National Consultants from five RCA Member States, organized by the IAEA and held in Sri Lanka in April 2001. The application has been developed in MS Access 2000 and Visual Basic for Applications (VBA) 6.0. However, it can run as a stand-alone application through its own executable files. It is based on screen forms for data entry or editing of information and command buttons. The structure of the data, the design of the application and VBA codes cannot be seen and cannot be modified by users. However, the designated administrator of AIDA-Asia in each country can customize it

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

    International Nuclear Information System (INIS)

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

    1987-10-01

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

  15. User's guide to the repository intrusion risk evaluation code INTRUDE

    International Nuclear Information System (INIS)

    Nancarrow, D.J.; Thorne, M.C.

    1986-05-01

    The report, commissioned by the Department of the Environment as part of its radioactive waste management research programme, constitutes the user's guide to the repository intrusion risk evaluation code INTRUDE. It provides an explanation of the mathematical basis of the code, the database used and the operation of the code. INTRUDE is designed to facilitate the estimation of individual risks arising from the possibility of intrusion into shallow land burial facilities for radioactive wastes. It considers a comprehensive inventory of up to 65 long-lived radionuclides and produces risk estimates for up to 20 modes of intrusion and up to 50 times of evaluation. (author)

  16. User's guide to the radionuclide inventory and hazard code INVENT

    International Nuclear Information System (INIS)

    Nancarrow, D.J.; Thorne, M.C.

    1986-05-01

    This report constitutes the user's guide to the radionuclide inventory and hazard index code INVENT and provides an explanation of the mathematical basis of the code, the database used and operation of the code. INVENT was designed to facilitate the post-closure radiological assessment of land-based repositories for low and intermediate-level wastes. For those radionuclides identified to be of potential significance, it allows the calculation of time-dependent radionuclide activities, hazard indices for both inhalation and ingestion of activity, and photon spectra. (author)

  17. NetMOD Version 2.0 User?s Manual.

    Energy Technology Data Exchange (ETDEWEB)

    Merchant, Bion J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-10-01

    NetMOD ( Net work M onitoring for O ptimal D etection) is a Java-based software package for conducting simulation of seismic, hydracoustic, and infrasonic networks. Specifically, NetMOD simulates the detection capabilities of monitoring networks. Network simulations have long been used to study network resilience to station outages and to determine where additional stations are needed to reduce monitoring thresholds. NetMOD makes use of geophysical models to determine the source characteristics, signal attenuation along the path between the source and station, and the performance and noise properties of the station. These geophysical models are combined to simulate the relative amplitudes of signal and noise that are observed at each of the stations. From these signal-to-noise ratios (SNR), the probability of detection can be computed given a detection threshold. This manual describes how to configure and operate NetMOD to perform detection simulations. In addition, NetMOD is distributed with simulation datasets for the Comprehensive Nuclear-Test-Ban Treaty Organization (CTBTO) International Monitoring System (IMS) seismic, hydroacoustic, and infrasonic networks for the purpose of demonstrating NetMOD's capabilities and providing user training. The tutorial sections of this manual use this dataset when describing how to perform the steps involved when running a simulation. ACKNOWLEDGEMENTS We would like to thank the reviewers of this document for their contributions.

  18. 18 CFR 410.1 - Basin regulations-Water Code and Administrative Manual-Part III Water Quality Regulations.

    Science.gov (United States)

    2010-04-01

    ... Code and Administrative Manual-Part III Water Quality Regulations. 410.1 Section 410.1 Conservation of... CODE AND ADMINISTRATIVE MANUAL-PART III WATER QUALITY REGULATIONS § 410.1 Basin regulations—Water Code and Administrative Manual—Part III Water Quality Regulations. (a) The Water Code of the Delaware River...

  19. Ensuring that User Defined Code does not See Uninitialized Fields

    DEFF Research Database (Denmark)

    Nielsen, Anders Bach

    2007-01-01

    Initialization of objects is commonly handled by user code, often in special routines known as constructors. This applies even in a virtual machine with multiple concurrent execution engines that all share the same heap. But for a language where run-time values play a role in the type system...

  20. Prototype road weather performance management (RWPM) tool installation instructions & user manual.

    Science.gov (United States)

    2016-07-20

    This document is the Installation Instructions and User Manual for the Road Weather Performance Management (RW-PM) Tool developed for the project on Development and Demonstration of a Prototype Road Weather Performance Management Application that Use...

  1. National Solar Radiation Database 1991-2010 Update: User's Manual

    Energy Technology Data Exchange (ETDEWEB)

    Wilcox, S. M.

    2012-08-01

    This user's manual provides information on the updated 1991-2010 National Solar Radiation Database. Included are data format descriptions, data sources, production processes, and information about data uncertainty.

  2. Computer-based guidelines for concrete pavements : HIPERPAV III : user manual

    Science.gov (United States)

    2009-10-01

    This user manual provides guidance on how to use the new High PERformance PAVing (HIPERPAV) III software program for the analysis of early-age Portland cement concrete pavement (PCCP) behavior. HIPERPAV III includes several improvements over prev...

  3. A Manual for Coding Teacher's Enacted Interpersonal Understanding.

    Science.gov (United States)

    DeVries, Rheta; And Others

    This manual was developed in order to study the sociomoral atmospheres of three kindergarten classrooms. Previous research by Robert Selman et al. conceptualized developmental levels of interpersonal understanding in terms of two types of experiences: negotiation, where the developmental goal is identity separate from others; and shared…

  4. The Intelligent Monitoring System: Generic Database Interface (GDI). User Manual. Revision

    Science.gov (United States)

    1994-01-03

    Summary of Lo=catos Nan* Decufptin Directory Location User Manual FrameMaker ’ source organized inlo, a book UBSW~ftbendb~doclim/user-manual named gdibk A...functions. LNSRCf1bgenrdb/srC I. Framemaker is a docment publishing tool fium Fame Technology Cororation Baseline: 21.1 3-1 anoAW ftnua ?bewd uw on 3.2

  5. Users manual for an expert system (HSPEXP) for calibration of the hydrological simulation program; Fortran

    Science.gov (United States)

    Lumb, A.M.; McCammon, R.B.; Kittle, J.L.

    1994-01-01

    Expert system software was developed to assist less experienced modelers with calibration of a watershed model and to facilitate the interaction between the modeler and the modeling process not provided by mathematical optimization. A prototype was developed with artificial intelligence software tools, a knowledge engineer, and two domain experts. The manual procedures used by the domain experts were identified and the prototype was then coded by the knowledge engineer. The expert system consists of a set of hierarchical rules designed to guide the calibration of the model through a systematic evaluation of model parameters. When the prototype was completed and tested, it was rewritten for portability and operational use and was named HSPEXP. The watershed model Hydrological Simulation Program--Fortran (HSPF) is used in the expert system. This report is the users manual for HSPEXP and contains a discussion of the concepts and detailed steps and examples for using the software. The system has been tested on watersheds in the States of Washington and Maryland, and the system correctly identified the model parameters to be adjusted and the adjustments led to improved calibration.

  6. Vault Safety and Inventory System users manual, PRIME 2350. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Downey, N.J.

    1994-12-14

    This revision is issued to request review of the attached document: VSIS User Manual, PRIME 2350, which provides user information for the operation of the VSIS (Vault Safety and Inventory System). It describes operational aspects of Prime 2350 minicomputer and vault data acquisition equipment. It also describes the User`s Main Menu and menu functions, including REPORTS. Also, system procedures for the Prime 2350 minicomputer are covered.

  7. VIPRE-01: a thermal-hydraulic code for reactor cores. Volume 3: programmer's manual (Revision 2)

    International Nuclear Information System (INIS)

    Stewart, C.W.; Koontz, A.S.; Cuta, J.M.; Montgomery, S.D.

    1985-07-01

    The VIPRE thermal-hydraulic computer code for PWR and BWR core analysis has undergone a detailed design review by a committee of experts. A new version of the code, incorporating the committee's recommendations, has been submitted for NRC review and issuance of a safety evaluation report. The changes in the programmers's manual are given

  8. Reference manual for the KfK code PCROSS

    International Nuclear Information System (INIS)

    Ravndal, S.; Oblozinsky, P.; Kelzenberg, S.; Cierjacks, S.

    1991-12-01

    The PCROSS code calculates the so-called 'pseudo' cross sections for sequential (x,n) reactions and merges them together with 'effective' cross section for neutron induced reactions into one file of 'collapsed' cross sections. The file is tailored to provide an input for the FISPACT inventory code that calculates the activation and related radiological quantities of material irradiated in given neutron fields. The report summarizes calculational procedure and provides the reader with essential technical details of the code PCROSS (version 1.0) such as description of parameters, common blocks and subroutines. (orig.) [de

  9. Integrated Tiger Series of electron/photon Monte Carlo transport codes: a user's guide for use on IBM mainframes

    International Nuclear Information System (INIS)

    Kirk, B.L.

    1985-12-01

    The ITS (Integrated Tiger Series) Monte Carlo code package developed at Sandia National Laboratories and distributed as CCC-467/ITS by the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory (ORNL) consists of eight codes - the standard codes, TIGER, CYLTRAN, ACCEPT; the P-codes, TIGERP, CYLTRANP, ACCEPTP; and the M-codes ACCEPTM, CYLTRANM. The codes have been adapted to run on the IBM 3081, VAX 11/780, CDC-7600, and Cray 1 with the use of the update emulator UPEML. This manual should serve as a guide to a user running the codes on IBM computers having 370 architecture. The cases listed were tested on the IBM 3033, under the MVS operating system using the VS Fortran Level 1.3.1 compiler

  10. Data acquisition for X ray microprobe. User's manual

    International Nuclear Information System (INIS)

    2002-01-01

    A modified data acquisition software for X ray microprobe was developed by the Physics Group, Instrumentation Unit, IAEA Laboratories at Seibersdorf, with assistance from M. Bogovac, Croatia. The software consists of data acquisition (scanning and calibration), automatic positioning and micro-movement of sample, data reduction and evaluation. The acquisition software was designed in order to support different measurement set-ups which are applied in low-energy nuclear physics. The modification was done in 1999-2000 under the projects Nuclear Spectrometry and Utilization of Particle Accelerators. The manual supersedes the first version entitled Microanalysis Data Acquisition and Control Program published under Computer Manual Series, No. 9 in 1996. The software described in this manual is freely available from the IAEA upon request

  11. User manual of the multicompenent variably - saturated flow and transport model HP1

    International Nuclear Information System (INIS)

    Jacques, D.; Simunek, J.

    2005-06-01

    (iii) calculations obtained from an independent geochemical transport model (CRUNCH) for several relatively complex problems. Nine verification examples of increasing complexity are described in this report. This report serves as both a user manual and reference document. Detailed instructions for input data preparation and interpretation of output data are given in the manuals of the original HYDRUS-1D and PHREEQC codes. The graphical user interfaces of both HYDRUS-1D and PHREEQC can be used for easy input data preparation and output display in the MS Windows environment. (author)

  12. 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.

  13. Sampling large landscapes with small-scale stratification-User's Manual

    Science.gov (United States)

    Bart, Jonathan

    2011-01-01

    This manual explains procedures for partitioning a large landscape into plots, assigning the plots to strata, and selecting plots in each stratum to be surveyed. These steps are referred to as the "sampling large landscapes (SLL) process." We assume that users of the manual have a moderate knowledge of ArcGIS and Microsoft ® Excel. The manual is written for a single user but in many cases, some steps will be carried out by a biologist designing the survey and some steps will be carried out by a quantitative assistant. Thus, the manual essentially may be passed back and forth between these users. The SLL process primarily has been used to survey birds, and we refer to birds as subjects of the counts. The process, however, could be used to count any objects. ®

  14. 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

  15. RELAP5/MOD3 code manual: Summaries and reviews of independent code assessment reports. Volume 7, Revision 1

    International Nuclear Information System (INIS)

    Moore, R.L.; Sloan, S.M.; Schultz, R.R.; Wilson, G.E.

    1996-10-01

    Summaries of RELAP5/MOD3 code assessments, a listing of the assessment matrix, and a chronology of the various versions of the code are given. Results from these code assessments have been used to formulate a compilation of some of the strengths and weaknesses of the code. These results are documented in the report. Volume 7 was designed to be updated periodically and to include the results of the latest code assessments as they become available. Consequently, users of Volume 7 should ensure that they have the latest revision available

  16. User's guide for FRMOD, a zero dimensional FRM burn code

    International Nuclear Information System (INIS)

    Driemeryer, D.; Miley, G.H.

    1979-01-01

    The zero-dimensional FRM plasma burn code, FRMOD is written in the FORTRAN language and is currently available on the Control Data Corporation (CDC) 7600 computer at the Magnetic Fusion Energy Computer Center (MFECC), sponsored by the US Department of Energy, in Livermore, CA. This guide assumes that the user is familiar with the system architecture and some of the utility programs available on the MFE-7600 machine, since online documentation is available for system routines through the use of the DOCUMENT utility. Users may therefore refer to it for answers to system related questions

  17. No Code Required Giving Users Tools to Transform the Web

    CERN Document Server

    Cypher, Allen; Lau, Tessa; Nichols, Jeffrey

    2010-01-01

    Revolutionary tools are emerging from research labs that enable all computer users to customize and automate their use of the Web without learning how to program. No Code Required takes cutting edge material from academic and industry leaders - the people creating these tools -- and presents the research, development, application, and impact of a variety of new and emerging systems. *The first book since Web 2.0 that covers the latest research, development, and systems emerging from HCI research labs on end user programming tools *Featuring contributions from the creators of Adobe's Zoet

  18. pix2code: Generating Code from a Graphical User Interface Screenshot

    OpenAIRE

    Beltramelli, Tony

    2017-01-01

    Transforming a graphical user interface screenshot created by a designer into computer code is a typical task conducted by a developer in order to build customized software, websites, and mobile applications. In this paper, we show that deep learning methods can be leveraged to train a model end-to-end to automatically generate code from a single input image with over 77% of accuracy for three different platforms (i.e. iOS, Android and web-based technologies).

  19. C.A.S.H. - a transient integrated plant model for a HTR-module power plant. User manual

    International Nuclear Information System (INIS)

    Biesenbach, R.; Lauer, A.; Struth, S.

    1997-07-01

    The computer code C.A.S.H. has been developed as an integrated plant model for the HTR-Module reactor, in order to treat safety related questions about this type of power plant which require a detailed numeric simulation of the transient behaviour of the integrated plant. The present report contains the user manual for this plant model. It consists of three parts: In the first part, the code structure and functions, the course of the simulation calculations, and important code parts are described. The second part is devoted to the practical application and explains extensively the handling of the complex code system with several sample calculations. These computing cases comprise load-follow transients and the shutdown procedure of the HTR-Module and are presented and discussed with the full input data, job patterns, and numerous computer graphics. The third part contains the input manual of C.A.S.H. and is rather extensive as it includes the complete inputs of several reactor component computer codes along with the control program of the integrated plant model. (orig./DG) [de

  20. Simplified Procedures for Eutrophication Assessment and Prediction: User Manual

    Science.gov (United States)

    1996-09-01

    1975), for use in the Lake Erie Wastewater Management Study and is described by Verhoff, Yaksich, and Melfi (1980) and Westerdahl et al. (1981). This...manual," Technical Re- port E-81-9, U.S. Army Engineer Waterways Experiment Station, Vicksburg, MS. Westerdahl , H. E., Ford, W. B., Harris, J., and