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Sample records for term code package

  1. Running the source term code package in Elebra MX-850

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.; Goes, A.G.A.

    1988-01-01

    The source term package (STCP) is one of the main tools applied in calculations of behavior of fission products from nuclear power plants. It is a set of computer codes to assist the calculations of the radioactive materials leaving from the metallic containment of power reactors to the environment during a severe reactor accident. The original version of STCP runs in SDC computer systems, but as it has been written in FORTRAN 77, is possible run it in others systems such as IBM, Burroughs, Elebra, etc. The Elebra MX-8500 version of STCP contains 5 codes:March 3, Trapmelt, Tcca, Vanessa and Nava. The example presented in this report has taken into consideration a small LOCA accident into a PWR type reactor. (M.I.)

  2. Verification test calculations for the Source Term Code Package

    International Nuclear Information System (INIS)

    Denning, R.S.; Wooton, R.O.; Alexander, C.A.; Curtis, L.A.; Cybulskis, P.; Gieseke, J.A.; Jordan, H.; Lee, K.W.; Nicolosi, S.L.

    1986-07-01

    The purpose of this report is to demonstrate the reasonableness of the Source Term Code Package (STCP) results. Hand calculations have been performed spanning a wide variety of phenomena within the context of a single accident sequence, a loss of all ac power with late containment failure, in the Peach Bottom (BWR) plant, and compared with STCP results. The report identifies some of the limitations of the hand calculation effort. The processes involved in a core meltdown accident are complex and coupled. Hand calculations by their nature must deal with gross simplifications of these processes. Their greatest strength is as an indicator that a computer code contains an error, for example that it doesn't satisfy basic conservation laws, rather than in showing the analysis accurately represents reality. Hand calculations are an important element of verification but they do not satisfy the need for code validation. The code validation program for the STCP is a separate effort. In general the hand calculation results show that models used in the STCP codes (e.g., MARCH, TRAP-MELT, VANESA) obey basic conservation laws and produce reasonable results. The degree of agreement and significance of the comparisons differ among the models evaluated. 20 figs., 26 tabs

  3. Use of source term code package in the ELEBRA MX-850 system

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.; Goes, A.G.A.

    1988-12-01

    The implantation of source term code package in the ELEBRA-MX850 system is presented. The source term is formed when radioactive materials generated in nuclear fuel leakage toward containment and the external environment to reactor containment. The implantated version in the ELEBRA system are composed of five codes: MARCH 3, TRAPMELT 3, THCCA, VANESA and NAVA. The original example case was used. The example consists of a small loca accident in a PWR type reactor. A sensitivity study for the TRAPMELT 3 code was carried out, modifying the 'TIME STEP' to estimate the processing time of CPU for executing the original example case. (M.C.K.) [pt

  4. Application of the source term code package to obtain a specific source term for the Laguna Verde Nuclear Power Plant

    International Nuclear Information System (INIS)

    Souto, F.J.

    1991-06-01

    The main objective of the project was to use the Source Term Code Package (STCP) to obtain a specific source term for those accident sequences deemed dominant as a result of probabilistic safety analyses (PSA) for the Laguna Verde Nuclear Power Plant (CNLV). The following programme has been carried out to meet this objective: (a) implementation of the STCP, (b) acquisition of specific data for CNLV to execute the STCP, and (c) calculations of specific source terms for accident sequences at CNLV. The STCP has been implemented and validated on CDC 170/815 and CDC 180/860 main frames as well as on a Micro VAX 3800 system. In order to get a plant-specific source term, data on the CNLV including initial core inventory, burn-up, primary containment structures, and materials used for the calculations have been obtained. Because STCP does not explicitly model containment failure, dry well failure in the form of a catastrophic rupture has been assumed. One of the most significant sequences from the point of view of possible off-site risk is the loss of off-site power with failure of the diesel generators and simultaneous loss of high pressure core spray and reactor core isolation cooling systems. The probability for that event is approximately 4.5 x 10 -6 . This sequence has been analysed in detail and the release fractions of radioisotope groups are given in the full report. 18 refs, 4 figs, 3 tabs

  5. Introduction of SCIENCE code package

    International Nuclear Information System (INIS)

    Lu Haoliang; Li Jinggang; Zhu Ya'nan; Bai Ning

    2012-01-01

    The SCIENCE code package is a set of neutronics tools based on 2D assembly calculations and 3D core calculations. It is made up of APOLLO2F, SMART and SQUALE and used to perform the nuclear design and loading pattern analysis for the reactors on operation or under construction of China Guangdong Nuclear Power Group. The purpose of paper is to briefly present the physical and numerical models used in each computation codes of the SCIENCE code pack age, including the description of the general structure of the code package, the coupling relationship of APOLLO2-F transport lattice code and SMART core nodal code, and the SQUALE code used for processing the core maps. (authors)

  6. NEACRP comparison of source term codes for the radiation protection assessment of transportation packages

    International Nuclear Information System (INIS)

    Broadhead, B.L.; Locke, H.F.; Avery, A.F.

    1994-01-01

    The results for Problems 5 and 6 of the NEACRP code comparison as submitted by six participating countries are presented in summary. These problems concentrate on the prediction of the neutron and gamma-ray sources arising in fuel after a specified irradiation, the fuel being uranium oxide for problem 5 and a mixture of uranium and plutonium oxides for problem 6. In both problems the predicted neutron sources are in good agreement for all participants. For gamma rays, however, there are differences, largely due to the omission of bremsstrahlung in some calculations

  7. The Fireball integrated code package

    Energy Technology Data Exchange (ETDEWEB)

    Dobranich, D.; Powers, D.A.; Harper, F.T.

    1997-07-01

    Many deep-space satellites contain a plutonium heat source. An explosion, during launch, of a rocket carrying such a satellite offers the potential for the release of some of the plutonium. The fireball following such an explosion exposes any released plutonium to a high-temperature chemically-reactive environment. Vaporization, condensation, and agglomeration processes can alter the distribution of plutonium-bearing particles. The Fireball code package simulates the integrated response of the physical and chemical processes occurring in a fireball and the effect these processes have on the plutonium-bearing particle distribution. This integrated treatment of multiple phenomena represents a significant improvement in the state of the art for fireball simulations. Preliminary simulations of launch-second scenarios indicate: (1) most plutonium vaporization occurs within the first second of the fireball; (2) large non-aerosol-sized particles contribute very little to plutonium vapor production; (3) vaporization and both homogeneous and heterogeneous condensation occur simultaneously; (4) homogeneous condensation transports plutonium down to the smallest-particle sizes; (5) heterogeneous condensation precludes homogeneous condensation if sufficient condensation sites are available; and (6) agglomeration produces larger-sized particles but slows rapidly as the fireball grows.

  8. The UK core performance code package

    International Nuclear Information System (INIS)

    Hutt, P.K.; Gaines, N.; McEllin, M.; White, R.J.; Halsall, M.J.

    1991-01-01

    Over the last few years work has been co-ordinated by Nuclear Electric, originally part of the Central Electricity Generating Board, with contributions from the United Kingdom Atomic Energy Authority and British Nuclear Fuels Limited, to produce a generic, easy-to-use and integrated package of core performance codes able to perform a comprehensive range of calculations for fuel cycle design, safety analysis and on-line operational support for Light Water Reactor and Advanced Gas Cooled Reactor plant. The package consists of modern rationalized generic codes for lattice physics (WIMS), whole reactor calculations (PANTHER), thermal hydraulics (VIPRE) and fuel performance (ENIGMA). These codes, written in FORTRAN77, are highly portable and new developments have followed modern quality assurance standards. These codes can all be run ''stand-alone'' but they are also being integrated within a new UNIX-based interactive system called the Reactor Physics Workbench (RPW). The RPW provides an interactive user interface and a sophisticated data management system. It offers quality assurance features to the user and has facilities for defining complex calculational sequences. The Paper reviews the current capabilities of these components, their integration within the package and outlines future developments underway. Finally, the Paper describes the development of an on-line version of this package which is now being commissioned on UK AGR stations. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-01-01

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

  10. A restructuring of CF package for MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S. H.; Kim, K. R.; Kim, D. H.; Cho, S. W.

    2004-01-01

    CF package, which evaluates user-specified 'control functions' and applies them to define or control various aspects of computation, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and modernized data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the CF package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of meaning of the variables as well as waste of memory, difficulty is more over because its data is location information of other package's data due to characteristics of CF package. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of the CF package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates data file, as well as MELCOR, which is processing a calculation. The verification has been done by comparing the results of the modified code with those from the existing code. As the trends are similar to each other, it hints that the same approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

  11. A restructuring of RN1 package for MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S. H.; Kim, D. H.; Kim, K. R.

    2003-01-01

    RN1 package, which is one of two fission product-related packages in MELCOR, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and modernized data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the RN1 package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of the RN1 package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates data file, as well as MELCOR, which is processing a calculation. The verification has been done by comparing the results of the modified code with those from the existing code. As the trends are similar to each other, it hints that the same approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

  12. A restructuring of RN2 package for MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S. H.; Kim, D. H.

    2003-01-01

    RN2 package, which is one of two fission product-related package in MELCOR, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and data structure. To do this, data transferring methods of current MELCOR code are modified and adopted into the RN2 package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of the RN2 package addressed in this paper includes module development, subroutine modification, and treats MELGEN, which generates data file, as well as MELCOR, which is processing a calculation. The validation has been done by comparing the results of the modified code with those from the existing code. As the trends are the similar to each other, it hints that the same approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

  13. Implementing and Testing the LINTAB, HEATER and PLOTTAB code package

    International Nuclear Information System (INIS)

    Cullen, D.E.; Smith, J.J.

    1987-07-01

    Enclosed is a description of the magnetic tape or floppy diskette containing the LINTAB, HEATER and PLOTTAB code package. In addition detailed information is provided on implementation and testing of these codes. These codes are documented in IAEA-NDS-84. (author)

  14. Development of the code package KASKAD for calculations of WWERs

    International Nuclear Information System (INIS)

    Bolobov, P.A.; Lazarenko, A.P.; Tomilov, M.Ju.

    2008-01-01

    The new version of software package for neutron calculation of WWER cores KASKAD 2007 consists of some calculating and service modules, which are integrated in the common framework. The package is based on the old version, which was expanded with some new functions and the new calculating modules, such as: -the BIPR-2007 code is the new one which performs calculation of power distribution in three-dimensional geometry for 2-group neutron diffusion calculation. This code is based on the BIPR-8KN model, provides all possibilities of BIPR-7A code and uses the same input data; -the PERMAK-2007 code is pin-by-pin few-group multilayer and 3-D code for neutron diffusion calculation; -graphical user interface for input data preparation of the TVS-M code. The report also includes some calculation results obtained with modified version of the KASKAD 2007 package. (Authors)

  15. A restructuring of COR package for MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S.H.; Kim, K.R.; Kim, D.H.

    2004-01-01

    The COR package, which calculates the thermal response of the core and the lower plenum internal structures and models the relocation of the core and lower plenum structural materials, has been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and a modernized data structure. To do this, the data transferring methods of the current MELCOR code are modified and adopted into the COR package. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as well as a waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which leads to an efficient memory treatment and an easy understanding of the code. Restructuring of the COR package addressed in this paper includes a module development, subroutine modification. The verification has been done by comparing the results of the modified code with those of the existing code. As the trends are similar to each other, it implies that the same approach could be extended to the entire code package. It is expected that the code restructuring will accelerated the code's domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)

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

    International Nuclear Information System (INIS)

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

    1983-04-01

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

  17. A restructuring of TF package for MIDAS computer code

    International Nuclear Information System (INIS)

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

    2002-01-01

    TF package which defines some interpolation and extrapolation condition through user defined table has been restructured in MIDAS computer code. To do this, data transferring methods of current MELCOR code are modified and adopted into TF package. The data structure of the current MELCOR code using FORTRAN77 causes a difficult grasping of the meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type, which lead to an efficient memory treatment and an easy understanding of the code. Restructuring of TF package addressed in this paper does module development and subroutine modification, and treats MELGEN which is making restart file as well as MELCOR which is processing calculation. The validation has been done by comparing the results of the modified code with those from the existing code, and it is confirmed that the results are the same. It hints that the similar approach could be extended to the entire code package. It is expected that code restructuring will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models

  18. Benchmarking of FA2D/PARCS Code Package

    International Nuclear Information System (INIS)

    Grgic, D.; Jecmenica, R.; Pevec, D.

    2006-01-01

    FA2D/PARCS code package is used at Faculty of Electrical Engineering and Computing (FER), University of Zagreb, for static and dynamic reactor core analyses. It consists of two codes: FA2D and PARCS. FA2D is a multigroup two dimensional transport theory code for burn-up calculations based on collision probability method, developed at FER. It generates homogenised cross sections both of single pins and entire fuel assemblies. PARCS is an advanced nodal code developed at Purdue University for US NRC and it is based on neutron diffusion theory for three dimensional whole core static and dynamic calculations. It is modified at FER to enable internal 3D depletion calculation and usage of neutron cross section data in a format produced by FA2D and interface codes. The FA2D/PARCS code system has been validated on NPP Krsko operational data (Cycles 1 and 21). As we intend to use this code package for development of IRIS reactor loading patterns the first logical step was to validate the FA2D/PARCS code package on a set of IRIS benchmarks, starting from simple unit fuel cell, via fuel assembly, to full core benchmark. The IRIS 17x17 fuel with erbium burnable absorber was used in last full core benchmark. The results of modelling the IRIS full core benchmark using FA2D/PARCS code package have been compared with reference data showing the adequacy of FA2D/PARCS code package model for IRIS reactor core design.(author)

  19. Abstracts of digital computer code packages. Assembled by the Radiation Shielding Information Center

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  20. SCAMPI: A code package for cross-section processing

    International Nuclear Information System (INIS)

    Parks, C.V.; Petrie, L.M.; Bowman, S.M.; Broadhead, B.L.; Greene, N.M.; White, J.E.

    1996-01-01

    The SCAMPI code package consists of a set of SCALE and AMPX modules that have been assembled to facilitate user needs for preparation of problem-specific, multigroup cross-section libraries. The function of each module contained in the SCANTI code package is discussed, along with illustrations of their use in practical analyses. Ideas are presented for future work that can enable one-step processing from a fine-group, problem-independent library to a broad-group, problem-specific library ready for a shielding analysis

  1. SCAMPI: A code package for cross-section processing

    Energy Technology Data Exchange (ETDEWEB)

    Parks, C.V.; Petrie, L.M.; Bowman, S.M.; Broadhead, B.L.; Greene, N.M.; White, J.E.

    1996-04-01

    The SCAMPI code package consists of a set of SCALE and AMPX modules that have been assembled to facilitate user needs for preparation of problem-specific, multigroup cross-section libraries. The function of each module contained in the SCANTI code package is discussed, along with illustrations of their use in practical analyses. Ideas are presented for future work that can enable one-step processing from a fine-group, problem-independent library to a broad-group, problem-specific library ready for a shielding analysis.

  2. Verification of the CONPAS (CONtainment Performance Analysis System) code package

    International Nuclear Information System (INIS)

    Kim, See Darl; Ahn, Kwang Il; Song, Yong Man; Choi, Young; Park, Soo Yong; Kim, Dong Ha; Jin, Young Ho.

    1997-09-01

    CONPAS is a computer code package to integrate the numerical, graphical, and results-oriented aspects of Level 2 probabilistic safety assessment (PSA) for nuclear power plants under a PC window environment automatically. For the integrated analysis of Level 2 PSA, the code utilizes four distinct, but closely related modules: (1) ET Editor, (2) Computer, (3) Text Editor, and (4) Mechanistic Code Plotter. Compared with other existing computer codes for Level 2 PSA, and 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-friendly interface. The computational performance of CONPAS has been verified through a Level 2 PSA to a reference plant. The results of the CONPAS code was compared with an existing level 2 PSA code (NUCAP+) and the comparison proves that CONPAS is appropriate for Level 2 PSA. (author). 9 refs., 8 tabs., 14 figs

  3. CONSUL code package application for LMFR core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Chibinyaev, A.V.; Teplov, P.S.; Frolova, M.V. [RNC ' Kurchatovskiy institute' , Kurchatov sq.1, Moscow (Russian Federation)

    2008-07-01

    CONSUL code package designed for the calculation of reactor core characteristics has been developed at the beginning of 90's. The calculation of nuclear reactor core characteristics is carried out on the basis of correlated neutron, isotope and temperature distributions. The code package has been generally used for LWR core characteristics calculations. At present CONSUL code package was adapted to calculate liquid metal fast reactors (LMFR). The comparisons with IAEA computational test 'Evaluation of benchmark calculations on a fast power reactor core with near zero sodium void effect' and BN-1800 testing calculations are presented in the paper. The IAEA benchmark core is based on the innovative core concept with sodium plenum above the core BN-800. BN-1800 core is the next development step which is foreseen for the Russian fast reactor concept. The comparison of the operational parameters has shown good agreement and confirms the possibility of CONSUL code package application for LMFR core calculation. (authors)

  4. In-core fuel management code package validation for BWRs

    International Nuclear Information System (INIS)

    1995-12-01

    The main goal of the present CRP (Coordinated Research Programme) was to develop benchmarks which are appropriate to check and improve the fuel management computer code packages and their procedures. Therefore, benchmark specifications were established which included a set of realistic data for running in-core fuel management codes. Secondly, the results of measurements and/or operating data were also provided to verify and compare with these parameters as calculated by the in-core fuel management codes or code packages. For the BWR it was established that the Mexican Laguna Verde 1 BWR would serve as the model for providing data on the benchmark specifications. It was decided to provide results for the first 2 cycles of Unit 1 of the Laguna Verde reactor. The analyses of the above benchmarks are performed in two stages. In the first stage, the lattice parameters are generated as a function of burnup at different voids and with and without control rod. These lattice parameters form the input for 3-dimensional diffusion theory codes for over-all reactor analysis. The lattice calculations were performed using different methods, such as, Monte Carlo, 2-D integral transport theory methods. Supercell Model and transport-diffusion model with proper correction for burnable absorber. Thus the variety of results should provide adequate information for any institute or organization to develop competence to analyze In-core fuel management codes. 15 refs, figs and tabs

  5. Status of reactor core design code system in COSINE code package

    International Nuclear Information System (INIS)

    Chen, Y.; Yu, H.; Liu, Z.

    2014-01-01

    For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

  6. Status of reactor core design code system in COSINE code package

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y.; Yu, H.; Liu, Z., E-mail: yuhui@snptc.com.cn [State Nuclear Power Software Development Center, SNPTC, National Energy Key Laboratory of Nuclear Power Software (NEKLS), Beijiing (China)

    2014-07-01

    For self-reliance, COre and System INtegrated Engine for design and analysis (COSINE) code package is under development in China. In this paper, recent development status of the reactor core design code system (including the lattice physics code and the core simulator) is presented. The well-established theoretical models have been implemented. The preliminary verification results are illustrated. And some special efforts, such as updated theory models and direct data access application, are also made to achieve better software product. (author)

  7. Code package to analyse behavior of the WWER fuel rods in normal operation: TOPRA's code

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.

    2001-01-01

    This paper briefly describes the code package intended for analysis of WWER fuel rod characteristics. The package includes two computer codes: TOPRA-1 and TOPRA-2 for full-scale fuel rod analyses; MRZ and MKK codes for analyzing the separate sections of fuel rods in r-z and r-j geometry. The TOPRA's codes are developed on the base of PIN-mod2 version and verified against experimental results obtained in MR, MIR and Halden research reactors (in the framework of SOFIT, FGR-2 and FUMEX experimental programs). Comparative analysis of calculation results and results from post-reactor examination of the WWER-440 and WWER-1000 fuel rod are also made as additional verification of these codes. To avoid the enlarging of uncertainties in fuel behavior prediction as a result of simplifying of the fuel geometry, MKK and MRZ codes are developed on the basis of the finite element method with use of the three nodal finite elements. Results obtained in the course of the code verification indicate the possibility for application of the method and TOPRA's code for simplified engineering calculations of WWER fuel rods thermal-physical parameters. An analysis of maximum relative errors for predicting of the fuel rod characteristics in the range of the accepted parameter values is also presented in the paper

  8. A computer code package for electron transport Monte Carlo simulation

    International Nuclear Information System (INIS)

    Popescu, Lucretiu M.

    1999-01-01

    A computer code package was developed for solving various electron transport problems by Monte Carlo simulation. It is based on condensed history Monte Carlo algorithm. In order to get reliable results over wide ranges of electron energies and target atomic numbers, specific techniques of electron transport were implemented such as: Moliere multiscatter angular distributions, Blunck-Leisegang multiscatter energy distribution, sampling of electron-electron and Bremsstrahlung individual interactions. Path-length and lateral displacement corrections algorithms and the module for computing collision, radiative and total restricted stopping powers and ranges of electrons are also included. Comparisons of simulation results with experimental measurements are finally presented. (author)

  9. NPTFit: A Code Package for Non-Poissonian Template Fitting

    International Nuclear Information System (INIS)

    Mishra-Sharma, Siddharth; Rodd, Nicholas L.; Safdi, Benjamin R.

    2017-01-01

    We present NPTFit, an open-source code package, written in Python and Cython, for performing non-Poissonian template fits (NPTFs). The NPTF is a recently developed statistical procedure for characterizing the contribution of unresolved point sources (PSs) to astrophysical data sets. The NPTF was first applied to Fermi gamma-ray data to provide evidence that the excess of ∼GeV gamma-rays observed in the inner regions of the Milky Way likely arises from a population of sub-threshold point sources, and the NPTF has since found additional applications studying sub-threshold extragalactic sources at high Galactic latitudes. The NPTF generalizes traditional astrophysical template fits to allow for the ability to search for populations of unresolved PSs that may follow a given spatial distribution. NPTFit builds upon the framework of the fluctuation analyses developed in X-ray astronomy, thus it likely has applications beyond those demonstrated with gamma-ray data. The NPTFit package utilizes novel computational methods to perform the NPTF efficiently. The code is available at http://github.com/bsafdi/NPTFit and up-to-date and extensive documentation may be found at http://nptfit.readthedocs.io.

  10. Verification of a BWR code package by gamma scan measurements

    International Nuclear Information System (INIS)

    Nakajima, Tsuyoshi; Iwamoto, Tatsuya; Kumanomido, Hironori

    1996-01-01

    High-burnup 8 x 8 fuel with a large central water rod (called step 2 fuel) has been recently introduced to the latest Japanese boiling water reactor (BWR) plants. Lanthanum-140 gamma intensity is almost directly related to nodal powers. By gamma scan measurement, the axial distribution of 140 La in the exposed fuel was measured at the end of cycle (EOC) 1 and was compared with the calculation by a BWR code package TGBLA/LOGOS. The multienrichment fuel-type core (MEC) design was adopted for the initial cycle core of the plants. The MEC design contains three different enrichment types of fuels to simulate the equilibrium cycles, achieve much higher discharge exposure, and save fuel cycle cost, and the low-enrichment fuels are loaded in periphery and in control cells. Such MEC design could be a challenge to the BWR design methods because of the large spectrum mismatch among the fuel assemblies of the different enrichments. The aforementioned comparison has shown that the accuracy of the TGBLA/LOGOS code package is satisfactory

  11. NPTFit: A Code Package for Non-Poissonian Template Fitting

    Energy Technology Data Exchange (ETDEWEB)

    Mishra-Sharma, Siddharth [Department of Physics, Princeton University, Princeton, NJ 08544 (United States); Rodd, Nicholas L.; Safdi, Benjamin R., E-mail: smsharma@princeton.edu, E-mail: nrodd@mit.edu, E-mail: bsafdi@mit.edu [Center for Theoretical Physics, Massachusetts Institute of Technology, Cambridge, MA 02139 (United States)

    2017-06-01

    We present NPTFit, an open-source code package, written in Python and Cython, for performing non-Poissonian template fits (NPTFs). The NPTF is a recently developed statistical procedure for characterizing the contribution of unresolved point sources (PSs) to astrophysical data sets. The NPTF was first applied to Fermi gamma-ray data to provide evidence that the excess of ∼GeV gamma-rays observed in the inner regions of the Milky Way likely arises from a population of sub-threshold point sources, and the NPTF has since found additional applications studying sub-threshold extragalactic sources at high Galactic latitudes. The NPTF generalizes traditional astrophysical template fits to allow for the ability to search for populations of unresolved PSs that may follow a given spatial distribution. NPTFit builds upon the framework of the fluctuation analyses developed in X-ray astronomy, thus it likely has applications beyond those demonstrated with gamma-ray data. The NPTFit package utilizes novel computational methods to perform the NPTF efficiently. The code is available at http://github.com/bsafdi/NPTFit and up-to-date and extensive documentation may be found at http://nptfit.readthedocs.io.

  12. A QR code identification technology in package auto-sorting system

    Science.gov (United States)

    di, Yi-Juan; Shi, Jian-Ping; Mao, Guo-Yong

    2017-07-01

    Traditional manual sorting operation is not suitable for the development of Chinese logistics. For better sorting packages, a QR code recognition technology is proposed to identify the QR code label on the packages in package auto-sorting system. The experimental results compared with other algorithms in literatures demonstrate that the proposed method is valid and its performance is superior to other algorithms.

  13. The development of the code package PERMAK--3D//SC--1

    International Nuclear Information System (INIS)

    Bolobov, P. A.; Oleksuk, D. A.

    2011-01-01

    Code package PERMAK-3D//SC-1 was developed for performing pin-by-pin coupled neutronic and thermal hydraulic calculation of the core fragment of seven fuel assemblies and was designed on the basis of 3D multigroup pin-by-pin code PERMAK-3D and 3D (subchannel) thermal hydraulic code SC-1 The code package predicts axial and radial pin-by-pin power distribution and coolant parameters in stimulated region (enthalpies,, velocities,, void fractions,, boiling and DNBR margins).. The report describes some new steps in code package development. Some PERMAK-3D//SC-1 outcomes of WWER calculations are presented in the report. (Authors)

  14. FUMACS-G, a Graphical User Interface for FUMACS Code Package

    International Nuclear Information System (INIS)

    Trontl, K.; Gergeta, K.; Smuc, T.

    2002-01-01

    The FUMACS (FUel MAnagement Code System) code package has been developed at Rudjer Boskovic Institute in year 1991 with the aim to enable in-core fuel management analysis of the NPP Krsko core for nominal conditions. Due to modernization and uprating of the NPP Krsko core in year 2000 and the original 1991 FUMACS inadequacy in simulating NPP Krsko core in these uprated conditions, in the year 2001 a new version of FUMACS code package has been developed - FUMACS/FEEC 2001. The code package upgrading procedure consisted of two main aspects: modifications of master files, libraries and codes necessary for proper modeling of the uprated NPP Krsko core and development of the code package structure suitable for Windows-32 environment. The latter included upgrading the source of the code from FORTRAN F77 to F90 level and development of a graphical, user-friendly interface with fully integrated electronic help system. Since the original FUMACS code package has been developed as a DOS based application, running of the code package on a Windows operating system proved to be rather inefficient and lacking in advantages of a standard Windows application. Therefore, FUMACS-G has been developed as a user friendly environment for handling off all project input and output files, as well as for easier overall project management. The design of FUMACS-G shell has been based on Microsoft application design guidelines. (author)

  15. Improvement of level-1 PSA computer code package

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Woon; Park, C. K.; Kim, K. Y.; Han, S. H.; Jung, W. D.; Chang, S. C.; Yang, J. E.; Sung, T. Y.; Kang, D. I.; Park, J. H.; Lee, Y. H.; Kim, S. H.; Hwang, M. J.; Choi, S. Y.

    1997-07-01

    This year the fifth (final) year of the phase-I of the Government-sponsored Mid- and Long-term Nuclear Power Technology Development Project. The scope of this subproject titled on `The improvement of level-1 PSA Computer Codes` is divided into two main activities : (1) improvement of level-1 PSA methodology, (2) development of applications methodology of PSA techniques to operations and maintenance of nuclear power plant. Level-1 PSA code KIRAP is converted to PC-Windows environment. For the improvement of efficiency in performing PSA, the fast cutset generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. Using about 30 foreign generic data sources, generic component reliability database (GDB) are developed considering dependency among source data. A computer program which handles dependency among data sources are also developed based on three stage bayesian updating technique. Common cause failure (CCF) analysis methods are reviewed and CCF database are established. Impact vectors can be estimated from this CCF database. A computer code, called MPRIDP, which handles CCF database are also developed. A CCF analysis reflecting plant-specific defensive strategy against CCF event is also performed. A risk monitor computer program, called Risk Monster, are being developed for the application to the operation and maintenance of nuclear power plant. The PSA application technique is applied to review the feasibility study of on-line maintenance and to the prioritization of in-service test (IST) of motor-operated valves (MOV). Finally, the root cause analysis (RCA) and reliability-centered maintenance (RCM) technologies are adopted and applied to the improvement of reliability of emergency diesel generators (EDG) of nuclear power plant. To help RCA and RCM analyses, two software programs are developed, which are EPIS and RAM Pro. (author). 129 refs., 20 tabs., 60 figs.

  16. Improvement of level-1 PSA computer code package

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Park, C. K.; Kim, K. Y.; Han, S. H.; Jung, W. D.; Chang, S. C.; Yang, J. E.; Sung, T. Y.; Kang, D. I.; Park, J. H.; Lee, Y. H.; Kim, S. H.; Hwang, M. J.; Choi, S. Y.

    1997-07-01

    This year the fifth (final) year of the phase-I of the Government-sponsored Mid- and Long-term Nuclear Power Technology Development Project. The scope of this subproject titled on 'The improvement of level-1 PSA Computer Codes' is divided into two main activities : 1) improvement of level-1 PSA methodology, 2) development of applications methodology of PSA techniques to operations and maintenance of nuclear power plant. Level-1 PSA code KIRAP is converted to PC-Windows environment. For the improvement of efficiency in performing PSA, the fast cutset generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. Using about 30 foreign generic data sources, generic component reliability database (GDB) are developed considering dependency among source data. A computer program which handles dependency among data sources are also developed based on three stage bayesian updating technique. Common cause failure (CCF) analysis methods are reviewed and CCF database are established. Impact vectors can be estimated from this CCF database. A computer code, called MPRIDP, which handles CCF database are also developed. A CCF analysis reflecting plant-specific defensive strategy against CCF event is also performed. A risk monitor computer program, called Risk Monster, are being developed for the application to the operation and maintenance of nuclear power plant. The PSA application technique is applied to review the feasibility study of on-line maintenance and to the prioritization of in-service test (IST) of motor-operated valves (MOV). Finally, the root cause analysis (RCA) and reliability-centered maintenance (RCM) technologies are adopted and applied to the improvement of reliability of emergency diesel generators (EDG) of nuclear power plant. To help RCA and RCM analyses, two software programs are developed, which are EPIS and RAM Pro. (author). 129 refs., 20 tabs., 60 figs

  17. A restructuring of the MELCOR fission product packages for the MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S.H.; Kim, K.R.; Kim, D.H.

    2004-01-01

    The RN1/RN2 packages, which are the fission product-related packages in MELCOR, have been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and a modernized data structure. To do this, the data transferring methods of the current MELCOR code are modified and adopted into the RN1/RN2 package. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as well as waste of memory. New features of FORTRAN90 make it possible to allocate the storage dynamically and to user-defined data type, which leads to an efficient memory treatment and an easy understanding of the code. Restructuring of the RN1/RN2 package addressed in this paper includes a module development, subroutine modification, and the treatment of MELGEN, which generates the data file, as well as MELCOR, which is processing the calculation. The verification has been done by comparing the results of the modified code with those of the existing code. As the trends are similar to each other, it implies that the same approach could be extended to the entire code package. It is expected that the code restructuring will accelerate the code domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)

  18. A restructuring of the CF/EDF packages for the MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S.H.; Kim, K.R.; Kim, D.H.

    2004-01-01

    The CF and EDF packages, which allow the user to define the functions of variables in a database and the usage of an external data file, have been restructured for the MIDAS computer code. MIDAS is being developed as an integrated severe accident analysis code with a user-friendly graphical user interface and a modernized data structure. To restructure the code, the data transferring methods of the current MELCOR code are modified and then partially adopted into the CF/EDF packages. The data structure of the current MELCOR code using FORTRAN77 has a difficulty in grasping the meaning of the variables as pointers are used to define their addresses. New features of FORTRAN90 make it possible to allocate the storage dynamically and to use the user-defined data type without pointers leading to an efficient memory treatment and an easy understanding of the code. Restructuring of the CF/EDF packages addressed in this paper includes a module development and subroutine modification. The verification has been done by comparing the results of the modified code with those of the existing code and the trends are almost the same to each other. Therefore the similar approach could be extended to the entire code package for code restructuring. It is expected that the code restructuring will accelerate the code's domestication thanks to a direct understanding of each variable and an easy implementation of the modified or newly developed models. (author)

  19. Code package {open_quotes}SVECHA{close_quotes}: Modeling of core degradation phenomena at severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Veshchunov, M.S.; Kisselev, A.E.; Palagin, A.V. [Nuclear Safety Institute, Moscow (Russian Federation)] [and others

    1995-09-01

    The code package SVECHA for the modeling of in-vessel core degradation (CD) phenomena in severe accidents is being developed in the Nuclear Safety Institute, Russian Academy of Science (NSI RAS). The code package presents a detailed mechanistic description of the phenomenology of severe accidents in a reactor core. The modules of the package were developed and validated on separate effect test data. These modules were then successfully implemented in the ICARE2 code and validated against a wide range of integral tests. Validation results have shown good agreement with separate effect tests data and with the integral tests CORA-W1/W2, CORA-13, PHEBUS-B9+.

  20. A restructuring of the FL package for the MIDAS computer code

    International Nuclear Information System (INIS)

    Park, S. H.; Kim, K. R.; Kim, D. H.; Cho, S. W.

    2005-01-01

    The developmental need for a localized severe accident analysis code is on the rise, and KAERI is developing a severe accident code MIDAS, based on MELCOR. The existing data saving method uses pointer variables for a fix-sized storage management, and it deteriorates the readability, maintainability and portability of the code. But new features in FORTRAN90 such as a dynamic allocation have been used for the restructuring. The restructuring of the data saving and transferring method of the existing code makes it easy to understand the code. Before an entire restructuring of the code, a restructuring template for a simple package was developed and tested. The target for the restructuring was the FL package which is responsible for modeling the thermal-hydraulic behavior of a liquid water, water vapor, and gases in MELCOR with the CVH package. The verification was done through comparing the results before and after the restructuring

  1. Code package for calculation of damage effects of medium-energy protons in metal targets

    International Nuclear Information System (INIS)

    Coulter, C.A.

    1976-12-01

    A program package was developed to calculate radiation damage effects produced in a metal target by protons in the 100-MeV to 3.5-GeV energy range. A detailed description is given of the control cards and data cards required to use the code package

  2. Improvement of Level-1 PSA computer code package -A study for nuclear safety improvement-

    International Nuclear Information System (INIS)

    Park, Chang Kyu; Kim, Tae Woon; Ha, Jae Joo; Han, Sang Hoon; Cho, Yeong Kyun; Jeong, Won Dae; Jang, Seung Cheol; Choi, Young; Seong, Tae Yong; Kang, Dae Il; Hwang, Mi Jeong; Choi, Seon Yeong; An, Kwang Il

    1994-07-01

    This year is the second year of the Government-sponsored Mid- and Long-Term Nuclear Power Technology Development Project. The scope of this subproject titled on 'The Improvement of Level-1 PSA Computer Codes' is divided into three main activities : (1) Methodology development on the under-developed fields such as risk assessment technology for plant shutdown and external events, (2) Computer code package development for Level-1 PSA, (3) Applications of new technologies to reactor safety assessment. At first, in the area of PSA methodology development, foreign PSA reports on shutdown and external events have been reviewed and various PSA methodologies have been compared. Level-1 PSA code KIRAP and CCF analysis code COCOA are converted from KOS to Windows. Human reliability database has been also established in this year. In the area of new technology applications, fuzzy set theory and entropy theory are used to estimate component life and to develop a new measure of uncertainty importance. Finally, in the field of application study of PSA technique to reactor regulation, a strategic study to develop a dynamic risk management tool PEPSI and the determination of inspection and test priority of motor operated valves based on risk importance worths have been studied. (Author)

  3. Extending R packages to support 64-bit compiled code: An illustration with spam64 and GIMMS NDVI3g data

    Science.gov (United States)

    Gerber, Florian; Mösinger, Kaspar; Furrer, Reinhard

    2017-07-01

    Software packages for spatial data often implement a hybrid approach of interpreted and compiled programming languages. The compiled parts are usually written in C, C++, or Fortran, and are efficient in terms of computational speed and memory usage. Conversely, the interpreted part serves as a convenient user-interface and calls the compiled code for computationally demanding operations. The price paid for the user friendliness of the interpreted component is-besides performance-the limited access to low level and optimized code. An example of such a restriction is the 64-bit vector support of the widely used statistical language R. On the R side, users do not need to change existing code and may not even notice the extension. On the other hand, interfacing 64-bit compiled code efficiently is challenging. Since many R packages for spatial data could benefit from 64-bit vectors, we investigate strategies to efficiently pass 64-bit vectors to compiled languages. More precisely, we show how to simply extend existing R packages using the foreign function interface to seamlessly support 64-bit vectors. This extension is shown with the sparse matrix algebra R package spam. The new capabilities are illustrated with an example of GIMMS NDVI3g data featuring a parametric modeling approach for a non-stationary covariance matrix.

  4. Qualification of the coupled RELAP5/PANTHER/COBRA code package for licensing applications

    International Nuclear Information System (INIS)

    Schneidesch, C.R.; Zhang Jinzhao

    2004-01-01

    A coupled thermal hydraulics-neutronics code package has been developed at Tractebel Engineering (TE), in which the best-estimate thermal-hydraulic system code, RELAP5/mod2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via the dynamic data exchange interface, TALINK. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by the sub-channel thermal-hydraulic analysis code COBRA-3C. The package provides the capability to accurately simulate the key physical phenomena in nuclear power plant accidents with strong asymmetric behaviours and system-core interactions. This paper presents the TE coupled code package and focuses on the methodology followed for qualifying it for licensing applications. The qualification of the coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been qualified and will be used at Tractebel Engineering (TE) for analyzing asymmetric PWR accidents with strong core-system interactions

  5. A Restructuring of the CAV and FDI Package for the MIDAS Computer Code

    International Nuclear Information System (INIS)

    Park, S. H.; Kim, D. H.; Cho, S. W.

    2006-01-01

    As one of the processes for a localized severe accident analysis code, KAERI is developing a severe accident code MIDAS. The MIDAS code is being developed based on MELCOR. The existing data saving method of MELCOR uses pointer variables for a fix-sized storage management, and it deteriorates the readability, maintainability and portability of the code. As a most important process for a localized severe accident analysis code, it is needed convenient method for data handling. So, it has been used the new features in FORTRAN90 such as a dynamic allocation for the restructuring. The restructuring of the data saving and transferring method of the existing code makes it easy to understand the code. Before an entire restructuring of the code, a restructuring template for a simple package was developed and tested. The target for the restructuring in this paper was the CAV and FDI packages. The CAV(cavity) package is responsible for modeling the attack on the basement concrete by hot core materials. The FDI(Fuel Dispersal Interactions) package is responsible for modeling both low and high pressure molten fuel ejection from the RPV into the reactor cavity, control volumes and surfaces. The verification was done through comparing the results before and after the restructuring

  6. A Restructuring of the HS Package for the MIDAS Computer Code

    International Nuclear Information System (INIS)

    Park, S. H.; Kim, K. R.; Kim, D. H.; Cho, S. W.

    2005-01-01

    As one of the processes for a localized severe accident analysis code, KAERI is developing a severe accident code MIDAS, based on MELCOR. The existing data saving method uses pointer variables for a fix-sized storage management, and it deteriorates the readability, maintainability and portability of the code. But new features in FORTRAN90 such as a dynamic allocation have been used for the restructuring. The restructuring of the data saving and transferring method of the existing code makes it easy to understand the code. Before an entire restructuring of the code, a restructuring template for a simple package was developed and tested. The target for the restructuring was the HS package which is responsible for calculation the heat conduction within an intact, solid structure and energy transfer across its boundary surfaces into control volumes. The verification was done through comparing the results before and after the restructuring

  7. Research on the improvement of nuclear safety -Improvement of level 1 PSA computer code package-

    International Nuclear Information System (INIS)

    Park, Chang Kyoo; Kim, Tae Woon; Kim, Kil Yoo; Han, Sang Hoon; Jung, Won Dae; Jang, Seung Chul; Yang, Joon Un; Choi, Yung; Sung, Tae Yong; Son, Yung Suk; Park, Won Suk; Jung, Kwang Sub; Kang Dae Il; Park, Jin Heui; Hwang, Mi Jung; Hah, Jae Joo

    1995-07-01

    This year is the third year of the Government-sponsored mid- and long-term nuclear power technology development project. The scope of this sub project titled on 'The improvement of level-1 PSA computer codes' is divided into three main activities : (1) Methodology development on the underdeveloped fields such as risk assessment technology for plant shutdown and low power situations, (2) Computer code package development for level-1 PSA, (3) Applications of new technologies to reactor safety assessment. At first, in this area of shutdown risk assessment technology development, plant outage experiences of domestic plants are reviewed and plant operating states (POS) are decided. A sample core damage frequency is estimated for over draining event in RCS low water inventory i.e. mid-loop operation. Human reliability analysis and thermal hydraulic support analysis are identified to be needed to reduce uncertainty. Two design improvement alternatives are evaluated using PSA technique for mid-loop operation situation: one is use of containment spray system as backup of shutdown cooling system and the other is installation of two independent level indication system. Procedure change is identified more preferable option to hardware modification in the core damage frequency point of view. Next, level-1 PSA code KIRAP is converted to PC-windows environment. For the improvement of efficiency in performing PSA, the fast cutest generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. 48 figs, 15 tabs, 59 refs. (Author)

  8. Research on the improvement of nuclear safety -Improvement of level 1 PSA computer code package-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Kyoo; Kim, Tae Woon; Kim, Kil Yoo; Han, Sang Hoon; Jung, Won Dae; Jang, Seung Chul; Yang, Joon Un; Choi, Yung; Sung, Tae Yong; Son, Yung Suk; Park, Won Suk; Jung, Kwang Sub; Kang Dae Il; Park, Jin Heui; Hwang, Mi Jung; Hah, Jae Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This year is the third year of the Government-sponsored mid- and long-term nuclear power technology development project. The scope of this sub project titled on `The improvement of level-1 PSA computer codes` is divided into three main activities : (1) Methodology development on the underdeveloped fields such as risk assessment technology for plant shutdown and low power situations, (2) Computer code package development for level-1 PSA, (3) Applications of new technologies to reactor safety assessment. At first, in this area of shutdown risk assessment technology development, plant outage experiences of domestic plants are reviewed and plant operating states (POS) are decided. A sample core damage frequency is estimated for over draining event in RCS low water inventory i.e. mid-loop operation. Human reliability analysis and thermal hydraulic support analysis are identified to be needed to reduce uncertainty. Two design improvement alternatives are evaluated using PSA technique for mid-loop operation situation: one is use of containment spray system as backup of shutdown cooling system and the other is installation of two independent level indication system. Procedure change is identified more preferable option to hardware modification in the core damage frequency point of view. Next, level-1 PSA code KIRAP is converted to PC-windows environment. For the improvement of efficiency in performing PSA, the fast cutest generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. 48 figs, 15 tabs, 59 refs. (Author).

  9. Development of platform to compare different wall heat transfer packages for system analysis codes

    International Nuclear Information System (INIS)

    Kim, Min-Gil; Lee, Won Woong; Lee, Jeong Ik; Shin, Sung Gil

    2016-01-01

    System thermal hydraulic (STH) analysis code is used for analyzing and evaluating the safety of a designed nuclear system. The system thermal hydraulic analysis code typically solves mass, momentum and energy conservation equations for multiple phases with sets of selected empirical constitutive equations to close the problem. Several STH codes are utilized in academia, industry and regulators, such as MARS-KS, SPACE, RELAP5, COBRA-TF, TRACE, and so on. Each system thermal hydraulic code consists of different sets of governing equations and correlations. However, the packages and sets of correlations of each code are not compared quantitatively yet. Wall heat transfer mode transition maps of SPACE and MARS-KS have a little difference for the transition from wall nucleate heat transfer mode to wall film heat transfer mode. Both codes have the same heat transfer packages and correlations in most region except for wall film heat transfer mode. Most of heat transfer coefficients calculated for the range of selected variables of SPACE are the same with those of MARS-KS. For the intervals between 500K and 540K of wall temperature, MARS-KS selects the wall film heat transfer mode and Bromley correlation but SPACE select the wall nucleate heat transfer mode and Chen correlation. This is because the transition from nucleate boiling to film boiling of MARS-KS is earlier than SPACE. More detailed analysis of the heat transfer package and flow regime package will be followed in the near future

  10. Synthesis of knowledge on the long-term behaviour of concretes. Applications to cemented waste packages

    International Nuclear Information System (INIS)

    Richet, C.; Galle, C.; Le Bescop, P.; Peycelon, H.; Bejaoui, S.; Tovena, I.; Pointeau, I.; L'Hostis, V.; Levera, P.

    2004-03-01

    As stipulated in the former law of December 91 relating to 'concrete waste package', a progress report (phenomenological reference document) was first provided in 1999. The objective was to make an assessment of the knowledge acquired on the long-term behaviour of cement-based waste packages in the context of deep disposal and/or interim storage. The present document is an updated summary report. It takes into account a new knowledge assessment, considers coupled mechanisms and should contribute to the first performance studies (operational calculations). Handling and radio-nuclides (RN) confinement are the two major functional properties requested from the concrete used for the waste packages. In unsaturated environment (interim storage/disposal prior to closing), the main problem is the generation of cracks in the material. This aspect is a key parameter from the mechanical point of view (retrievability). It can have a major impact on the disposal phase (confinement). In saturated environment (disposal post-closing phase), the main concern is the chemical degradation of the waste package concrete submitted to underground waters leaching. In this context, the major thema are: the durability of the concretes under water (chemical degradation) and in unsaturated medium (corrosion of reinforcement), matter transport, RN retention, chemistry / transport / mechanical couplings. On the other hand, laboratory data on the behaviour of concretes are used to evaluate the RN source term of waste packages in function of time (concrete waste package OPerational Model, i.e. 'Concrete MOP'). The 'MOP' provides the physico-chemical description of the RN release in relationship with the waste package degradation itself. This description is based on simplified phenomenology for which only dimensioning mechanisms are taken into account. The use of Diffu-Ca code (basic module for the MOP) on the CASTEM numerical plate-form, already allows operational predictions. (authors)

  11. Monte Carlo code criticality benchmark comparisons for waste packaging

    International Nuclear Information System (INIS)

    Alesso, H.P.; Annese, C.E.; Buck, R.M.; Pearson, J.S.; Lloyd, W.R.

    1992-07-01

    COG is a new point-wise Monte Carlo code being developed and tested at Lawrence Livermore National Laboratory (LLNL). It solves the Boltzmann equation for the transport of neutrons and photons. The objective of this paper is to report on COG results for criticality benchmark experiments both on a Cray mainframe and on a HP 9000 workstation. COG has been recently ported to workstations to improve its accessibility to a wider community of users. COG has some similarities to a number of other computer codes used in the shielding and criticality community. The recently introduced high performance reduced instruction set (RISC) UNIX workstations provide computational power that approach mainframes at a fraction of the cost. A version of COG is currently being developed for the Hewlett Packard 9000/730 computer with a UNIX operating system. Subsequent porting operations will move COG to SUN, DEC, and IBM workstations. In addition, a CAD system for preparation of the geometry input for COG is being developed. In July 1977, Babcock ampersand Wilcox Co. (B ampersand W) was awarded a contract to conduct a series of critical experiments that simulated close-packed storage of LWR-type fuel. These experiments provided data for benchmarking and validating calculational methods used in predicting K-effective of nuclear fuel storage in close-packed, neutron poisoned arrays. Low enriched UO2 fuel pins in water-moderated lattices in fuel storage represent a challenging criticality calculation for Monte Carlo codes particularly when the fuel pins extend out of the water. COG and KENO calculational results of these criticality benchmark experiments are presented

  12. Implantation, evaluation and improvement of the diffusion code package developed by the RIS0 Research Center

    International Nuclear Information System (INIS)

    Koide, M.C.M.

    1983-01-01

    The evaluation and improvement of the diffusion code package developed by the RIS0 Research Center of Denmark have been performed. The improvements made in the package consisted in the presentation of their manuals. In order to reduce the process time of the codes an analitical boundary condition capable of representing the effects of the baffle and the reflector on the flux distribution has been calculated. Such boundary condition was obtained using a one-dimensional medium in the framework of the two group diffusion theory. The results showed that the application of this boundary condition produces very accurate results and an appreciable economy of processing time. (author) [pt

  13. Verification of 3-D generation code package for neutronic calculations of WWERs

    International Nuclear Information System (INIS)

    Sidorenko, V.D.; Aleshin, S.S.; Bolobov, P.A.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Morozov, V.V.; Syslov, A.A.; Tsvetkov, V.M.

    2000-01-01

    Materials on verification of the 3 -d generation code package for WWERs neutronic calculations are presented. The package includes: - spectral code TVS-M; - 2-D fine mesh diffusion code PERMAK-A for 4- or 6-group calculation of WWER core burnup; - 3-D coarse mesh diffusion code BIPR-7A for 2-group calculations of quasi-stationary WWERs regimes. The materials include both TVS-M verification data and verification data on PERMAK-A and BIPR-7A codes using constant libraries generated with TVS-M. All materials are related to the fuel without Gd. TVS-M verification materials include results of comparison both with benchmark calculations obtained by other codes and with experiments carried out at ZR-6 critical facility. PERMAK-A verification materials contain results of comparison with TVS-M calculations and with ZR-6 experiments. BIPR-7A materials include comparison with operation data for Dukovany-2 and Loviisa-1 NPPs (WWER-440) and for Balakovo NPP Unit 4 (WWER-1000). The verification materials demonstrate rather good accuracy of calculations obtained with the use of code package of the 3 -d generation. (Authors)

  14. Validation of SCALE code package on high performance neutron shields

    International Nuclear Information System (INIS)

    Bace, M.; Jecmenica, R.; Smuc, T.

    1999-01-01

    The shielding ability and other properties of new high performance neutron shielding materials from the KRAFTON series have been recently published. A comparison of the published experimental and MCNP results for the two materials of the KRAFTON series, with our own calculations has been done. Two control modules of the SCALE-4.4 code system have been used, one of them based on one dimensional radiation transport analysis (SAS1) and other based on the three dimensional Monte Carlo method (SAS3). The comparison of the calculated neutron dose equivalent rates shows a good agreement between experimental and calculated results for the KRAFTON-N2 material.. Our results indicate that the N2-M-N2 sandwich type is approximately 10% inferior as neutron shield to the KRAFTON-N2 material. All values of neutron dose equivalent obtained by SAS1 are approximately 25% lower in comparison with the SAS3 results, which indicates proportions of discrepancies introduced by one-dimensional geometry approximation.(author)

  15. Verification of RRC Ki code package for neutronic calculations of WWER core with GD

    International Nuclear Information System (INIS)

    Aleshin, S.S.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Pavlov, V.I.; Pavlovitchev, A.M.; Sidorenko, V.D.; Tsvetkov, V.M.

    2001-01-01

    The report presented is concerned with verification results of TVS-M/PERMAK-A/BIPR-7A code package for WWERs neutronic calculation as applied to calculation of systems containing U-GD pins. The verification is based on corresponded benchmark calculations, data critical experiments and on operation data obtained WWER units with Gd. The comparison results are discussed (Authors)

  16. The development of the Nuclear Electric core performance and fault transient analysis code package in support of Sizewell B

    International Nuclear Information System (INIS)

    Hall, P.; Hutt, P.

    1994-01-01

    This paper describes Nuclear Electric's (NE) development of an integrated code package in support of all its reactors including Sizewell B, designed for the provision of fuel management design, core performance studies, operational support and fault transient analysis. The package uses the NE general purpose three-dimensional transient reactor physics code PANTHER with cross-sections derived in the PWR case from the LWRWIMS LWR lattice neutronics code. The package also includes ENIGMA a generic fuel performance code and for PWR application VIPRE-01 a subchannel thermal hydraulics code, RELAP5 the system thermal hydraulics transient code and SCORPIO an on-line surveillance system. The paper describes the capabilities and validation of the elements of this package for PWR, how they are coupled within the package and the way in which they are being applied for Sizewell B to on-line surveillance and fault transient analysis. (Author)

  17. Code Package to Analyze Parameters of the WWER Fuel Rod. TOPRA-2 Code - Verification Data

    International Nuclear Information System (INIS)

    Scheglov, A.; Proselkov, V.; Passage, G.; Stefanova, S.

    2009-01-01

    Presented are the data for computer codes to analyze WWER fuel rods, used in the WWER department of RRC 'Kurchatov Institute'. Presented is the description of TOPRA-2 code intended for the engineering analysis of thermophysical and strength parameters of the WWER fuel rod - temperature distributions along the fuel radius, gas pressures under the cladding, stresses in the cladding, etc. for the reactor operation in normal conditions. Presented are some results of the code verification against test problems and the data obtained in the experimental programs. Presented are comparison results of the calculations with TOPRA-2 and TRANSURANUS (V1M1J06) codes. Results obtained in the course of verification demonstrate possibility of application of the methodology and TOPRA-2 code for the engineering analysis of the WWER fuel rods

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

    International Nuclear Information System (INIS)

    Turula, P.

    1998-01-01

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

  19. Draft ASME code case on ductile cast iron for transport packaging

    International Nuclear Information System (INIS)

    Saegusa, T.; Arai, T.; Hirose, M.; Kobayashi, T.; Tezuka, Y.; Urabe, N.; Hueggenberg, R.

    2004-01-01

    The current Rules for Construction of ''Containment Systems for Storage and Transport Packagings of Spent Nuclear Fuel and High Level Radioactive Material and Waste'' of Division 3 in Section III of ASME Code (2001 Edition) does not include ductile cast iron in its list of materials permitted for use. The Rules specify required fracture toughness values of ferritic steel material for nominal wall thickness 5/8 to 12 inches (16 to 305 mm). New rule for ductile cast iron for transport packaging of which wall thickness is greater than 12 inches (305mm) is required

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

    International Nuclear Information System (INIS)

    Raske, D.T.; Wang, Z.

    1992-01-01

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

  1. Science version 2: the most recent capabilities of the Framatome 3-D nuclear code package

    International Nuclear Information System (INIS)

    Girieud, P.; Daudin, L.; Garat, C.; Marotte, P.; Tarle, S.

    2001-01-01

    The Framatome nuclear code package SCIENCE developed in the 1990's has been fully operational for nuclear design since 1997. Results obtained using the package demonstrate the high accuracy of its physical models. Nevertheless, since the first release of the SCIENCE package, continuous improvement work has been carried out at Framatome, which leads today to Version 2 of the package. The intensive use of the package by Framatome teams, for example, while performing reload calculations and the associated core follow, is a permanent opportunity to point out any trend or scattering in the results, even the smaller they are. Thus the main objective of improvements was to take advantage of the progress in computer performances in using more sophisticated calculation schemes conducting to more accurate results. Besides the implementation of more accurate physical models, SCIENCE Version 2 also exploits developments conducted in other fields, mainly for transient calculations using 3-D kinetics or coupling with open-channel core thermal-hydraulics and the plant simulator. These developments allow Framatome to perform accident analyses with advanced methodologies using the SCIENCE package. (author)

  2. Development of a PC code package for the analysis of research and power reactors

    International Nuclear Information System (INIS)

    Urli, N.

    1992-06-01

    Computer codes available for performing reactor physics calculations for nuclear research reactors and power reactors are normally suited for running on mainframe computers. With the fast development in speed and memory of the PCs and affordable prices it became feasible to develop PC versions of commonly used codes. The present work performed under an IAEA sponsored research contract has successfully developed a code package for running on a PC. This package includes a cross-section generating code PSU-LEOPARD and 2D and 1D spatial diffusion codes, MCRAC and MCYC 1D. For adapting PSU-LEOPARD for a PC, the binary library has been reorganized to decimal form, upgraded to FORTRAN-77 standard and arrays and subroutines reorganized to conform to PC compiler. Similarly PC version of MCRAC for FORTRAN-77 and 1D code MCYC 1D have been developed. Tests, verification and bench mark results show excellent agreement with the results obtained from mainframe calculations. The execution speeds are also very satisfactory. 12 refs, 4 figs, 3 tabs

  3. Abstracts of digital computer code packages assembled by the Radiation Shielding Information Center

    Energy Technology Data Exchange (ETDEWEB)

    Carter, B.J.; Maskewitz, B.F.

    1985-04-01

    This publication, ORNL/RSIC-13, Volumes I to III Revised, has resulted from an internal audit of the first 168 packages of computing technology in the Computer Codes Collection (CCC) of the Radiation Shielding Information Center (RSIC). It replaces the earlier three documents published as single volumes between 1966 to 1972. A significant number of the early code packages were considered to be obsolete and were removed from the collection in the audit process and the CCC numbers were not reassigned. Others not currently being used by the nuclear R and D community were retained in the collection to preserve technology not replaced by newer methods, or were considered of potential value for reference purposes. Much of the early technology, however, has improved through developer/RSIC/user interaction and continues at the forefront of the advancing state-of-the-art.

  4. Abstracts of digital computer code packages assembled by the Radiation Shielding Information Center

    International Nuclear Information System (INIS)

    Carter, B.J.; Maskewitz, B.F.

    1985-04-01

    This publication, ORNL/RSIC-13, Volumes I to III Revised, has resulted from an internal audit of the first 168 packages of computing technology in the Computer Codes Collection (CCC) of the Radiation Shielding Information Center (RSIC). It replaces the earlier three documents published as single volumes between 1966 to 1972. A significant number of the early code packages were considered to be obsolete and were removed from the collection in the audit process and the CCC numbers were not reassigned. Others not currently being used by the nuclear R and D community were retained in the collection to preserve technology not replaced by newer methods, or were considered of potential value for reference purposes. Much of the early technology, however, has improved through developer/RSIC/user interaction and continues at the forefront of the advancing state-of-the-art

  5. Core design calculations of IRIS reactor using modified CORD-2 code package

    International Nuclear Information System (INIS)

    Pevec, D.; Grgic, D.; Jecmenica, R.; Petrovic, B.

    2002-01-01

    Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)

  6. CEPXS/ONELD: A one-dimensional coupled electron-photon discrete ordinates code package

    International Nuclear Information System (INIS)

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

    1992-01-01

    CEPXS/ONELD is a discrete ordinates transport code package that can model the electron-photon cascade from 100 MeV to 1 keV. The CEPXS code generates fully-coupled multigroup-Legendre cross section data. This data is used by the general-purpose discrete ordinates code, ONELD, which is derived from the Los Alamos ONEDANT and ONBTRAN codes. Version 1.0 of CEPXS/ONELD was released in 1989 and has been primarily used to analyze the effect of radiation environments on electronics. Version 2.0 is under development and will include user-friendly features such as the automatic selection of group structure, spatial mesh structure, and S N order

  7. Windows user-friendly code package development for operation of research reactors

    International Nuclear Information System (INIS)

    Hoang Anh Tuan

    1998-01-01

    The content of the project was to developed: 1. MS Windows interface to spectral codes like THERMOS, PEACO-COLLIS, GRACE and burn-up code. 2. MS Windows C-language burn-up diffusion hexagonal lattice code. The overall scope of the project was to develop a PC-based MS Windows code package for operation of Dalat research reactor. Various problems relating to neutronic physics like thermalization, resonance treatment, fast spectral treatment, change of isotopic concentration during burn-up time as well as burn-up distribution in the reactor core are considered in parallel to application of informatics technique. The developing process is a subject of the concept of user-friendly interface between end-users and the code package. High level input features through system of icon, menu, dialog box with regard to Common User Access (CUA) convention and sophisticated graphical output in MS Windows environment was used. The user-computer interface is also enhanced by using both keyboard and mouse, which creates a very natural manner for end-user. (author)

  8. 9975 Shipping package component long-term degradation rates

    Energy Technology Data Exchange (ETDEWEB)

    Daugherty, W. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-28

    Special nuclear materials are being stored in the K-Area Complex using 3013 containers that are held within Model 9975 shipping packages. The service life for these packages in storage was recently increased from 15 to 20 years, since some of these packages have been stored for nearly 15 years. A strategy is also being developed whereby such storage might be extended beyond 20 years. This strategy is based on recent calculations that support acceptable 9975 package performance for 20 years with internal heat loads up to 19 watts, and identifies a lower heat load limit for which the package components should degrade at half the bounding rate or less, thus doubling the effective storage life for these lower wattage packages. The components of the 9975 package that are sensitive to aging under storage conditions are the fiberboard overpack and the O-ring seals, although some degradation of the lead shield and outer drum are also possible. This report summarizes degradation rates applicable to lower heat load storage conditions. In particular, the O-ring seals should provide leak-tight performance for more than 40 years in packages for which their maximum temperature is ≤135 °F. Similarly, the fiberboard should remain acceptable in performance of its required safety functions for up to 40 years in packages with a maximum fiberboard temperature ≤125 °F.

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

    International Nuclear Information System (INIS)

    Turula, P.

    1998-01-01

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

  10. DFT calculation for elastic constants of orthorhombic structure within WIEN2K code: A new package (ortho-elastic)

    International Nuclear Information System (INIS)

    Reshak, Ali H.; Jamal, Morteza

    2012-01-01

    Highlights: ► A new package for calculating elastic constants of orthorhombic structure is released. ► The package called ortho-elastic. ► It is compatible with [FP-(L)APW+lo] method implemented in WIEN2k code. ► Several orthorhombic structure compounds were used to test the new package. ► Elastic constants calculated using this package show good agreement with experiment. - Abstract: A new package for calculating the elastic constants of orthorhombic structure is released. The package called ortho-elastic. The formalism of calculating the ortho-elastic constants is described in details. The package is compatible with the highly accurate all-electron full-potential (linearized) augmented plane-wave plus local orbital [FP-(L)APW+lo] method implemented in WIEN2k code. Several orthorhombic structure compounds were used to test the new package. We found that the calculated elastic constants using the new package show better agreement with the available experimental data than the previous theoretical results used different methods. In this package the second-order derivative E ″ (ε) of polynomial fit E=E(ε) of energy vs strains at zero strain (ε=0), used to calculate the orthorhombic elastic constants.

  11. Benchmark calculation for GT-MHR using HELIOS/MASTER code package and MCNP

    International Nuclear Information System (INIS)

    Lee, Kyung Hoon; Kim, Kang Seog; Noh, Jae Man; Song, Jae Seung; Zee, Sung Quun

    2005-01-01

    The latest research associated with the very high temperature gas-cooled reactor (VHTR) is focused on the verification of a system performance and safety under operating conditions for the VHTRs. As a part of those, an international gas-cooled reactor program initiated by IAEA is going on. The key objectives of this program are the validation of analytical computer codes and the evaluation of benchmark models for the projected and actual VHTRs. New reactor physics analysis procedure for the prismatic VHTR is under development by adopting the conventional two-step procedure. In this procedure, a few group constants are generated through the transport lattice calculations using the HELIOS code, and the core physics analysis is performed by the 3-dimensional nodal diffusion code MASTER. We evaluated the performance of the HELIOS/MASTER code package through the benchmark calculations related to the GT-MHR (Gas Turbine-Modular Helium Reactor) to dispose weapon plutonium. In parallel, MCNP is employed as a reference code to verify the results of the HELIOS/MASTER procedure

  12. Package

    Directory of Open Access Journals (Sweden)

    Arsić Zoran

    2013-01-01

    Full Text Available It is duty of the seller to pack the goods in a manner which assures their safe arrival and enables their handling in transit and at the place of destination. The problem of packing is relevant in two main respects. First of all the buyer is in certain circumstances entitled to refuse acceptance of the goods if they are not properly packed. Second, the package is relevant to calculation of price and freight based on weight. In the case of export trade, the package should conform to the legislation in the country of destination. The impact of package on environment is regulated by environment protection regulation of Republic if Serbia.

  13. 'ACTIV' - a package of codes for charged particle and neutron activation analysis

    International Nuclear Information System (INIS)

    Cincu, Em.; Alexandreanu, B.; Manu, V.; Moisa, V.

    1997-01-01

    The 'ACTIV' Program is an advanced software package dedicated to applications of the thermal neutron and charged particle activation (NAA and CPA) induced reactions. The program is designed to run on personal computers compatible IBM PC-Models XT/AT, 286 or more advanced, operating under DOS version 5.0 or later, on systems with minimum 5 MB of hard disk memory. The package consists of 6 software modules and a Nuclear Data Base comprising physical, nuclear reaction and decay data for: thermal neutron, proton, deuteron and α-particle induced reactions on 15 selected metallic elements; the nuclear reaction data corresponds to the energy range (5-100) MeV. In the first version - ACTIV 1.0 - the set of input data concerns: the sample type, irradiation and measurement conditions, the γ-ray spectrum identification code, selected detection efficiency calibration curve, selected radionuclides, selected standardization method for elemental analysis, version of results. At present, the 'ACTIV' package comprises 6 soft modules for processing the experimental data, which ensure computation of the quantities: radionuclide activities, activation yield data (case of CPA) and elemental concentration by relative and absolute standardization methods. Recently, the software designed to processing complex γ-ray spectra was acquired and installed on our PC 486 (8 MB RAM, 100 MHz). The next step in developing the 'ACTIV' program envisages improving the existing computing codes, completing the data libraries, incorporating a new soft for the direct use of the 'Quantum TM MCA' data, developing modules dedicated to uncertainty computation and optimization of the activation experiments

  14. Thermal Analysis of the SAFKEG Package for Long Term Storage

    International Nuclear Information System (INIS)

    NARENDRA, GUPTA

    2005-01-01

    Interim plutonium storage for up to 10 years in the K-reactor building is currently being planned at Savannah River Site (SRS). SAFKEG package could be used to store Pu metal and oxide (PuO2) in the K-reactor complex with other packagings like 9975. The SAFKEG is designed for carrying Type-B materials across the DOE complex and meets the 10CFR71 requirements. Thermal analyses were performed to ensure that the temperatures of the SAFKEG components will not exceed their temperature limits under the K-reactor storage conditions. Thermal analyses of the SAFKEG packaging with three content configurations using BNFL 3013 outer container (Rocky Flats, SRS bagless transfer cans, and BNFL inner containers) were performed for storage of PuO2 and plutonium metal

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

    International Nuclear Information System (INIS)

    Popescu, Lucretiu M.

    2000-01-01

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

  16. VISUAL: a software package for plotting data in the RADHEAT-V4 code system

    International Nuclear Information System (INIS)

    Sasaki, Toshihiko; Yamano, Naoki

    1984-03-01

    In this report, the features, the capabilities and the constitution of the VISUAL Software Package are presented. The one of the features is that the VISUAL provides a versatile graphic display tool to plot a wide variety of data of the RADHEAT-V4 code system. And the other is to enable a user to handle easily the executing data in the Conversational Management Mode named ''CMM''. The program adopts the adjustable dimension system to increase its flexibility. VISUAL generates two-dimensional drawing, contour line map and three dimensional drawing on TSS (Time Sharing System) digital graphic equipment, NLP (Nihongo Laser Printer) or COM(Computer Output Microfilm). It is easily possible to display the calculated and experimental data in a DATA-POOL by using these functions. The purpose of this report is to describe sufficient information to enable a user to use VISUAL profitabily. (author)

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

    International Nuclear Information System (INIS)

    Miyagi, K.; Miyahara, S.

    1996-01-01

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

  18. Validation of the DRAGON/DONJON code package for MNR using the IAEA 10 MW benchmark problem

    International Nuclear Information System (INIS)

    Day, S.E.; Garland, W.J.

    2000-01-01

    The first step in developing a framework for reactor physics analysis is to establish the appropriate and proven reactor physics codes. The chosen code package is tested, by executing a benchmark problem and comparing the results to the accepted standards. The IAEA 10 MW Benchmark problem is suitable for static reactor physics calculations on plate-fueled research reactor systems and has been used previously to validate codes for the McMaster Nuclear (MNR). The flexible and advanced geometry capabilities of the DRAGON transport theory code make it a desirable tool, and the accompanying DONJON diffusion theory code also has useful features applicable to safety analysis work at MNR. This paper describes the methodology used to benchmark the DRAGON/DONJON code package against this problem and the results herein extend the domain of validation of this code package. The results are directly applicable to MNR and are relevant to a reduced-enrichment fuel program. The DRAGON transport code models, used in this study, are based on the 1-D infinite slab approximation whereas the DONJON diffusion code models are defined in 3-D Cartesian geometry. The cores under consideration are composed of HEU (93% enrichment), MEU (45% enrichment) and LEU (20% enrichment) fuel and are examined in a fresh state, as well as at beginning-of-life (BOL) and end-of-life (EOL) exposures. The required flux plots and flux-ratio plots are included, as are transport theory code k∞and diffusion theory code k eff results. In addition to this, selected isotope atom densities are charted as a function of fuel burnup. Results from this analysis are compared to and are in good agreement with previously published results. (author)

  19. BrachyTPS -Interactive point kernel code package for brachytherapy treatment planning of gynaecological cancers

    International Nuclear Information System (INIS)

    Thilagam, L.; Subbaiah, K.V.

    2008-01-01

    Brachytherapy treatment planning systems (TPS) are always recommended to account for the effect of tissue, applicator and shielding material heterogeneities exist in Intracavitary brachytherapy (ICBT) applicators. Most of the commercially available brachytherapy TPS softwares estimate the absorbed dose at a point, only taking care of the contributions of individual sources and the source distribution, neglecting the dose perturbations arising from the applicator design and construction. So the doses estimated by them are not much accurate under realistic clinical conditions. In this regard, interactive point kernel rode (BrachyTPS) has been developed to perform independent dose calculations by taking into account the effect of these heterogeneities, using two regions build up factors, proposed by Kalos. As primary input data, the code takes patients' planning data including the source specifications, dwell positions, dwell times and it computes the doses at reference points by dose point kernel formalisms, with multi-layer shield build-up factors accounting for the contributions from scattered radiation. In addition to performing dose distribution calculations, this code package is capable of displaying an isodose distribution curve into the patient anatomy images. The primary aim of this study is to validate the developed point kernel code integrated with treatment planning systems against the other tools which are available in the market. In the present work, three brachytherapy applicators commonly used in the treatment of uterine cervical carcinoma, Board of Radiation Isotope and Technology (BRIT) made low dose rate (LDR) applicator, Fletcher Green type LDR applicator and Fletcher Williamson high dose rate (HDR) applicator were studied to test the accuracy of the software

  20. Benchmarking of the computer code and the thirty foot side drop analysis for the Shippingport (RPV/NST package)

    International Nuclear Information System (INIS)

    Bumpus, S.E.; Gerhard, M.A.; Hovingh, J.; Trummer, D.J.; Witte, M.C.

    1989-01-01

    This paper presents the benchmarking of a finite element computer code and the subsequent results from the code simulating the 30 foot side drop impact of the RPV/NST transport package from the decommissioned Shippingport Nuclear Power Station. The activated reactor pressure vessel (RPV), thermal shield, and other reactor external components were encased in concrete contained by the neutron shield tank (NST) and a lifting skirt. The Shippingport RPV/NST package, a Type B Category II package, weighs approximately 900 tons and has 17.5 ft diameter and 40.7 ft. length. For transport of the activated components from Shippingport to the burial site, the Safety Analysis Report for Packaging (SARP) demonstrated that the package can withstand the hypothetical accidents of DOE Order 5480.3 including 10 CFR 71. Mathematical simulations of these accidents can substitute for actual tests if the simulated results satisfy the acceptance criteria. Any such mathematical simulation, including the modeling of the materials, must be benchmarked to experiments that duplicate the loading conditions of the tests. Additional confidence in the simulations is justified if the test specimens are configured similar to the package

  1. Color-Coded Front-of-Pack Nutrition Labels—An Option for US Packaged Foods?

    Science.gov (United States)

    Dunford, Elizabeth K.; Poti, Jennifer M.; Xavier, Dagan; Webster, Jacqui L.; Taillie, Lindsey Smith

    2017-01-01

    The implementation of a standardized front-of-pack-labelling (FoPL) scheme would likely be a useful tool for many consumers trying to improve the healthfulness of their diets. Our objective was to examine what the traffic light labelling scheme would look like if implemented in the US. Data were extracted from Label Insight’s Open Access branded food database in 2017. Nutrient levels and the proportion of products classified as “Red” (High), “Amber” (Medium) or “Green” (Low) in total fat, saturated fat, total sugar and sodium for food and beverage items were examined. The proportion of products in each category that had each possible combination of traffic light colors, and met the aggregate score for “healthy” was examined. Out of 175,198 products, >50% of all US packaged foods received a “Red” rating for total sugar and sodium. “Confectionery” had the highest mean total sugar (51.9 g/100 g) and “Meat and meat alternatives” the highest mean sodium (781 mg/100 g). The most common traffic light label combination was “Red” for total fat, saturated fat and sodium and “Green” for sugar. Only 30.1% of products were considered “healthy”. A wide variety (n = 80) of traffic light color combinations were observed. A color coded traffic light scheme appears to be an option for implementation across the US packaged food supply to support consumers in making healthier food choices. PMID:28489037

  2. GEMSFITS: Code package for optimization of geochemical model parameters and inverse modeling

    International Nuclear Information System (INIS)

    Miron, George D.; Kulik, Dmitrii A.; Dmytrieva, Svitlana V.; Wagner, Thomas

    2015-01-01

    Highlights: • Tool for generating consistent parameters against various types of experiments. • Handles a large number of experimental data and parameters (is parallelized). • Has a graphical interface and can perform statistical analysis on the parameters. • Tested on fitting the standard state Gibbs free energies of aqueous Al species. • Example on fitting interaction parameters of mixing models and thermobarometry. - Abstract: GEMSFITS is a new code package for fitting internally consistent input parameters of GEM (Gibbs Energy Minimization) geochemical–thermodynamic models against various types of experimental or geochemical data, and for performing inverse modeling tasks. It consists of the gemsfit2 (parameter optimizer) and gfshell2 (graphical user interface) programs both accessing a NoSQL database, all developed with flexibility, generality, efficiency, and user friendliness in mind. The parameter optimizer gemsfit2 includes the GEMS3K chemical speciation solver ( (http://gems.web.psi.ch/GEMS3K)), which features a comprehensive suite of non-ideal activity- and equation-of-state models of solution phases (aqueous electrolyte, gas and fluid mixtures, solid solutions, (ad)sorption. The gemsfit2 code uses the robust open-source NLopt library for parameter fitting, which provides a selection between several nonlinear optimization algorithms (global, local, gradient-based), and supports large-scale parallelization. The gemsfit2 code can also perform comprehensive statistical analysis of the fitted parameters (basic statistics, sensitivity, Monte Carlo confidence intervals), thus supporting the user with powerful tools for evaluating the quality of the fits and the physical significance of the model parameters. The gfshell2 code provides menu-driven setup of optimization options (data selection, properties to fit and their constraints, measured properties to compare with computed counterparts, and statistics). The practical utility, efficiency, and

  3. The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

    International Nuclear Information System (INIS)

    Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul

    2006-01-01

    At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

  4. The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul [Suez-Tractebel Engineering, Avenue Ariane 7, B-1200 Brussels (Belgium)

    2006-07-01

    At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

  5. Evaluation of finite element codes for demonstrating the performance of radioactive material packages in hypothetical accident drop scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Tso, C.F. [Arup (United Kingdom); Hueggenberg, R. [Gesellschaft fuer Nuklear-Behaelter mbH (Germany)

    2004-07-01

    Drop testing and analysis are the two methods for demonstrating the performance of packages in hypothetical drop accident scenarios. The exact purpose of the tests and the analyses, and the relative prominence of the two in the license application, may depend on the Competent Authority and will vary between countries. The Finite Element Method (FEM) is a powerful analysis tool. A reliable finite element (FE) code when used correctly and appropriately, will allow a package's behaviour to be simulated reliably. With improvements in computing power, and in sophistication and reliability of FE codes, it is likely that FEM calculations will increasingly be used as evidence of drop test performance when seeking Competent Authority approval. What is lacking at the moment, however, is a standardised method of assessing a FE code in order to determine whether it is sufficiently reliable or pessimistic. To this end, the project Evaluation of Codes for Analysing the Drop Test Performance of Radioactive Material Transport Containers, funded by the European Commission Directorate-General XVII (now Directorate-General for Energy and Transport) and jointly performed by Arup and Gesellschaft fuer Nuklear-Behaelter mbH, was carried out in 1998. The work consisted of three components: Survey of existing finite element software, with a view to finding codes that may be capable of analysing drop test performance of radioactive material packages, and to produce an inventory of them. Develop a set of benchmark problems to evaluate software used for analysing the drop test performance of packages. Evaluate the finite element codes by testing them against the benchmarks This paper presents a summary of this work.

  6. Evaluation of finite element codes for demonstrating the performance of radioactive material packages in hypothetical accident drop scenarios

    International Nuclear Information System (INIS)

    Tso, C.F.; Hueggenberg, R.

    2004-01-01

    Drop testing and analysis are the two methods for demonstrating the performance of packages in hypothetical drop accident scenarios. The exact purpose of the tests and the analyses, and the relative prominence of the two in the license application, may depend on the Competent Authority and will vary between countries. The Finite Element Method (FEM) is a powerful analysis tool. A reliable finite element (FE) code when used correctly and appropriately, will allow a package's behaviour to be simulated reliably. With improvements in computing power, and in sophistication and reliability of FE codes, it is likely that FEM calculations will increasingly be used as evidence of drop test performance when seeking Competent Authority approval. What is lacking at the moment, however, is a standardised method of assessing a FE code in order to determine whether it is sufficiently reliable or pessimistic. To this end, the project Evaluation of Codes for Analysing the Drop Test Performance of Radioactive Material Transport Containers, funded by the European Commission Directorate-General XVII (now Directorate-General for Energy and Transport) and jointly performed by Arup and Gesellschaft fuer Nuklear-Behaelter mbH, was carried out in 1998. The work consisted of three components: Survey of existing finite element software, with a view to finding codes that may be capable of analysing drop test performance of radioactive material packages, and to produce an inventory of them. Develop a set of benchmark problems to evaluate software used for analysing the drop test performance of packages. Evaluate the finite element codes by testing them against the benchmarks This paper presents a summary of this work

  7. The CNCSN: one, two- and three-dimensional coupled neutral and charged particle discrete ordinates code package

    International Nuclear Information System (INIS)

    Voloschenko, A.M.; Gukov, S.V.; Kryuchkov, V.P.; Dubinin, A.A.; Sumaneev, O.V.

    2005-01-01

    The CNCSN package is composed of the following codes: -) KATRIN-2.0: a three-dimensional neutral and charged particle transport code; -) KASKAD-S-2.5: a two-dimensional neutral and charged particle transport code; -) ROZ-6.6: a one-dimensional neutral and charged particle transport code; -) ARVES-2.5: a preprocessor for the working macroscopic cross-section format FMAC-M for transport calculations; -) MIXERM: a utility code for preparing mixtures on the base of multigroup cross-section libraries in ANISN format; -) CEPXS-BFP: a version of the Sandia National Lab. multigroup coupled electron-photon cross-section generating code CEPXS, adapted for solving the charged particles transport in the Boltzmann-Fokker-Planck formulation with the use of discrete ordinate method; -) SADCO-2.4: Institute for High-Energy Physics modular system for generating coupled nuclear data libraries to provide high-energy particles transport calculations by multigroup method; -) KATRIF: the post-processor for the KATRIN code; -) KASF: the post-processor for the KASKAD-S code; and ROZ6F: the post-processor for the ROZ-6 code. The coding language is Fortran-90

  8. Numerical study of the ITER divertor plasma with the B2-EIRENE code package

    Energy Technology Data Exchange (ETDEWEB)

    Kotov, V.; Reiter, D. [Forschungszentrum Juelich (DE). Inst. fuer Energieforschung (IEF), Plasmaphysik (IEF-4); Kukushkin, A.S. [ITER International Team, Cadarache (France)

    2007-11-15

    The problem of plasma-wall interaction and impurity control is one of the remaining critical issues for development of an industrial energy source based on nuclear fusion of light isotopes. In this field sophisticated integrated numerical tools are widely used both for the analysis of current experiments and for predictions guiding future device design. The present work is dedicated to the numerical modelling of the edge plasma region in divertor configurations of large-scale tokamak fusion devices. A well established software tool for this kind of modelling is the B2-EIRENE code. It was originally developed for a relatively hot (>> 10 eV) ''high recycling divertor''. It did not take into account a number of physical effects which can be potentially important for ''detached conditions'' (cold, - several eV, - high density, - {approx} 10{sup 21} m{sup -3}, - plasma) typical for large tokamak devices. This is especially critical for the modelling of the divertor plasma of ITER: an international project of an experimental tokamak fusion reactor to be built in Cadarache, France by 2016. This present work is devoted to a major upgrade of the B2-EIRENE package, which is routinely used for ITER modelling, essentially with a significantly revised version of EIRENE: the Monte-Carlo neutral transport code. The main part of the thesis address three major groups of the new physical effects which have been added to the model in frame of this work: the neutral-neutral collisions, the up-to date hydrogen molecular reaction kinetics and the line radiation transport. The impact of the each stage of the upgrade on the self-consistent (between plasma, the neutral gas and the radiation field) solution for the reference ITER case is analysed. The strongest effect is found to be due to the revised molecular collision kinetics, in particular due to hitherto neglected elastic collisions of hydrogen molecules with ions. The newly added non

  9. Numerical study of the ITER divertor plasma with the B2-EIRENE code package

    International Nuclear Information System (INIS)

    Kotov, V.; Reiter, D.; Kukushkin, A.S.

    2007-11-01

    The problem of plasma-wall interaction and impurity control is one of the remaining critical issues for development of an industrial energy source based on nuclear fusion of light isotopes. In this field sophisticated integrated numerical tools are widely used both for the analysis of current experiments and for predictions guiding future device design. The present work is dedicated to the numerical modelling of the edge plasma region in divertor configurations of large-scale tokamak fusion devices. A well established software tool for this kind of modelling is the B2-EIRENE code. It was originally developed for a relatively hot (>> 10 eV) ''high recycling divertor''. It did not take into account a number of physical effects which can be potentially important for ''detached conditions'' (cold, - several eV, - high density, - ∼ 10 21 m -3 , - plasma) typical for large tokamak devices. This is especially critical for the modelling of the divertor plasma of ITER: an international project of an experimental tokamak fusion reactor to be built in Cadarache, France by 2016. This present work is devoted to a major upgrade of the B2-EIRENE package, which is routinely used for ITER modelling, essentially with a significantly revised version of EIRENE: the Monte-Carlo neutral transport code. The main part of the thesis address three major groups of the new physical effects which have been added to the model in frame of this work: the neutral-neutral collisions, the up-to date hydrogen molecular reaction kinetics and the line radiation transport. The impact of the each stage of the upgrade on the self-consistent (between plasma, the neutral gas and the radiation field) solution for the reference ITER case is analysed. The strongest effect is found to be due to the revised molecular collision kinetics, in particular due to hitherto neglected elastic collisions of hydrogen molecules with ions. The newly added non-linear effects (neutral-neutral collisions, radiation opacity

  10. First Results for Fluid Dynamics, Neutronics and Fission Product Behaviour in HTR applying the HTR Code Package (HCP) Prototype

    International Nuclear Information System (INIS)

    Allelein, H.-J.; Kasselmann, S.; Xhonneux, A.; Lambertz, D.

    2014-01-01

    To simulate the different aspects of High Temperature Reactor (HTR) cores, a variety of specialized computer codes have been developed at Forschungszentrum Jülich (IEK-6) and Aachen University (LRST) in the last decades. In order to preserve knowledge, to overcome present limitations and to make these codes applicable to modern computer clusters, these individual programs are being integrated into a consistent code package. The so-called HTR code package (HCP) couples the related and recently applied physics models in a highly integrated manner and therefore allows to simulate phenomena with higher precision in space and time while at the same time applying state-of-the-art programming techniques and standards. This paper provides an overview of the status of the HCP and reports about first benchmark results for an HCP prototype which couples the fluid dynamics and time dependent neutronics code MGT-3D, the burn up code TNT and the fission product release code STACY. Due to the coupling of MGT-3D and TNT, a first step towards a new reactor operation and accident simulation code was made, where nuclide concentrations calculated by TNT are fed back into a new spectrum code of the HCP. Selected operation scenarios of the HTR-Module 200 concept plant and the HTTR were chosen to be simulated with the HCP prototype. The fission product release during normal operation conditions will be calculated with STACY based on a core status derived from SERPENT and MGT–3D. Comparisons will be shown against data generated by the legacy codes VSOP99/11, NAKURE and FRESCO-II. (author)

  11. Source-term model for the SYVAC3-NSURE performance assessment code

    International Nuclear Information System (INIS)

    Rowat, J.H.; Rattan, D.S.; Dolinar, G.M.

    1996-11-01

    Radionuclide contaminants in wastes emplaced in disposal facilities will not remain in those facilities indefinitely. Engineered barriers will eventually degrade, allowing radioactivity to escape from the vault. The radionuclide release rate from a low-level radioactive waste (LLRW) disposal facility, the source term, is a key component in the performance assessment of the disposal system. This report describes the source-term model that has been implemented in Ver. 1.03 of the SYVAC3-NSURE (Systems Variability Analysis Code generation 3-Near Surface Repository) code. NSURE is a performance assessment code that evaluates the impact of near-surface disposal of LLRW through the groundwater pathway. The source-term model described here was developed for the Intrusion Resistant Underground Structure (IRUS) disposal facility, which is a vault that is to be located in the unsaturated overburden at AECL's Chalk River Laboratories. The processes included in the vault model are roof and waste package performance, and diffusion, advection and sorption of radionuclides in the vault backfill. The model presented here was developed for the IRUS vault; however, it is applicable to other near-surface disposal facilities. (author). 40 refs., 6 figs

  12. Validation of the REL2005 code package on Gd-poisoned PWR type assemblies through the CAMELEON experimental program

    International Nuclear Information System (INIS)

    Blaise, Patrick; Vidal, Jean-Francois; Santamarina, Alain

    2009-01-01

    This paper details the validation of Gd-poisoned 17x17 PWR lattices, through several configurations of the CAMELEON experimental program, by using the newly qualified REL2005 French code package. After a general presentation of the CAMELEON program that took place in the EOLE critical Facility in Cadarache, one describes the new REL2005 code package relying on the deterministic transport code APOLLO2.8 based on characteristics method (MOC), and its new CEA2005 library based on the latest JEFF-3.1.1 nuclear data evaluation. For critical masses, the average Calculation-to-Experiment C/E's on the k eff are (136 ± 80) pcm and (300 ± 76) pcm for the reference 281 groups MOC and optimized 26 groups MOC schemes respectively. These values include also a drastic improvement of about 250 pcm due to the change in the library from JEF2.2 to JEFF3.1. For pin-by-pin radial power distributions, reference and REL2005 results are very close, with maximum discrepancies of the order of 2%, i.e., in the experimental uncertainty limits. The Optimized REL2005 code package allows to predict the reactivity worth of the Gd-clusters (averaged on 9 experimental configurations) to be C/E Δρ(Gd clusters) = +1.3% ± 2.3%. (author)

  13. Monitoring and inspection techniques for long term storage of higher activity waste packages

    International Nuclear Information System (INIS)

    Bolton, Gary

    2013-01-01

    In 2009, following recent changes in United Kingdom (UK) Government Policy, the Nuclear Decommissioning Authority (NDA) identified a knowledge gap in the area of long term interim storage of waste packages. A cross-industry Integrated Project Team (IPT) for Interim Storage was created with responsibility for delivering Industry Guidance on the storage of packaged Higher Activity Waste (HAW) for the current UK civil decommissioning and clean-up programmes. This included a remit to direct research and development projects via the NDA's Direct Research Portfolio (DRP) to fill the knowledge gap. The IPT for Interim Storage published Industry Guidance in 2012 which established a method to define generic package performance criteria and made recommendations on monitoring and inspection. The package performance method consists of the following steps; identification of the package safety function, identification of evolutionary processes that may affect safety function performance, determination of measurable indicators of these evolutionary processes and calibration of the indicators into package performance zones. This article provides an overview of three projects funded by the NDA's DRP that the UK National Nuclear Laboratory (NNL) have completed to address monitoring and inspection needs of waste packages in interim storage. (orig.)

  14. First results for fluid dynamics, neutronics and fission product behavior in HTR applying the HTR code package (HCP) prototype

    Energy Technology Data Exchange (ETDEWEB)

    Allelein, H.-J., E-mail: h.j.allelein@fz-juelich.de [Forschungszentrum Jülich, 52425 Jülich (Germany); Institute for Reactor Safety and Reactor Technology, RWTH Aachen University, 52064 Aachen (Germany); Kasselmann, S.; Xhonneux, A.; Tantillo, F.; Trabadela, A.; Lambertz, D. [Forschungszentrum Jülich, 52425 Jülich (Germany)

    2016-09-15

    To simulate the different aspects of High Temperature Reactor (HTR) cores, a variety of specialized computer codes have been developed at Forschungszentrum Jülich (IEK-6) and Aachen University (LRST) in the last decades. In order to preserve knowledge, to overcome present limitations and to make these codes applicable to modern computer clusters, these individual programs are being integrated into a consistent code package. The so-called HTR code package (HCP) couples the related and recently applied physics models in a highly integrated manner and therefore allows to simulate phenomena with higher precision in space and time while at the same time applying state-of-the-art programming techniques and standards. This paper provides an overview of the status of the HCP and reports about first benchmark results for an HCP prototype which couples the fluid dynamics and time dependent neutronics code MGT-3D, the burn up code TNT and the fission product release code STACY. Due to the coupling of MGT-3D and TNT, a first step towards a new reactor operation and accident simulation code was made, where nuclide concentrations calculated by TNT lead to new cross sections, which are fed back into MGT-3D. Selected operation scenarios of the HTR-Module 200 concept plant and the HTTR were chosen to be simulated with the HCP prototype. The fission product release during normal operation conditions will be calculated with STACY based on a core status derived from SERPENT and MGT-3D. Comparisons will be shown against data generated by SERPENT and the legacy codes VSOP99/11, NAKURE and FRESCO-II.

  15. A simplified computer code based on point Kernel theory for calculating radiation dose in packages of radioactive material

    International Nuclear Information System (INIS)

    1986-03-01

    A study on radiation dose control in packages of radioactive waste from nuclear facilities, hospitals and industries, such as sources of Ra-226, Co-60, Ir-192 and Cs-137, is presented. The MAPA and MAPAM computer codes, based on point Kernel theory for calculating doses of several source-shielding type configurations, aiming to assure the safe transport conditions for these sources, was developed. The validation of the code for point sources, using the values provided by NCRP, for the thickness of lead and concrete shieldings, limiting the dose at 100 Mrem/hr for several distances from the source to the detector, was carried out. The validation for non point sources was carried out, measuring experimentally radiation dose from packages developed by Brazilian CNEN/S.P. for removing the sources. (M.C.K.) [pt

  16. An analysis of options available for developing a common laser ray tracing package for Ares and Kull code frameworks

    Energy Technology Data Exchange (ETDEWEB)

    Weeratunga, S K

    2008-11-06

    Ares and Kull are mature code frameworks that support ALE hydrodynamics for a variety of HEDP applications at LLNL, using two widely different meshing approaches. While Ares is based on a 2-D/3-D block-structured mesh data base, Kull is designed to support unstructured, arbitrary polygonal/polyhedral meshes. In addition, both frameworks are capable of running applications on large, distributed-memory parallel machines. Currently, both these frameworks separately support assorted collections of physics packages related to HEDP, including one for the energy deposition by laser/ion-beam ray tracing. This study analyzes the options available for developing a common laser/ion-beam ray tracing package that can be easily shared between these two code frameworks and concludes with a set of recommendations for its development.

  17. Long-term durability experiments with concrete-based waste packages in simulated repository conditions

    International Nuclear Information System (INIS)

    Ipatti, A.

    1993-03-01

    Two extensive experiments on long-term durability of waste packages in simulated repository conditions are described. The first one is a 'half-scale experiment' comprising radioactive waste product and half-scale concrete containers in site specific groundwater conditions. The second one is 'full-scale experiment' including simulated inactive waste product and full-scale concrete container stored in slowly flowing fresh water. The scope of the experiments is to demonstrate long-term behaviour of the designed waste packages in contact with moderately concrete aggressive groundwater, and to evaluate the possible interactions between the waste product, concrete container and ground water. As the waste packages are made of high-quality concrete, provisions have been made to continue the experiments for several years

  18. COMPASS: A source term code for investigating capillary barrier performance

    International Nuclear Information System (INIS)

    Zhou, Wei; Apted, J.J.

    1996-01-01

    A computer code COMPASS based on compartment model approach is developed to calculate the near-field source term of the High-Level-Waste repository under unsaturated conditions. COMPASS is applied to evaluate the expected performance of Richard's (capillary) barriers as backfills to divert infiltrating groundwater at Yucca Mountain. Comparing the release rates of four typical nuclides with and without the Richard's barrier, it is shown that the Richard's barrier significantly decreases the peak release rates from the Engineered-Barrier-System (EBS) into the host rock

  19. The Barrier code for predicting long-term concrete performance

    International Nuclear Information System (INIS)

    Shuman, R.; Rogers, V.C.; Shaw, R.A.

    1989-01-01

    There are numerous features incorporated into a LLW disposal facility that deal directly with critical safety objectives required by the NRC in 10 CFR 61. Engineered barriers or structures incorporating concrete are commonly being considered for waste disposal facilities. The Barrier computer code calculates the long-term degradation of concrete structures in LLW disposal facilities. It couples this degradation with water infiltration into the facility, nuclide leaching from the waste, contaminated water release from the facility, and associated doses to members of the critical population group. The concrete degradation methodology of Barrier is described

  20. Long term governance for radioactive waste management. Final report of Cowan2 - work package 4

    International Nuclear Information System (INIS)

    Schneider, Th.; Schieber, C.; Lavele, S.

    2006-12-01

    This report aims at identifying key features for the long term governance of radioactive waste. It is proposed by the COWAN2 Work Package 4 the purpose of which was to identify, discuss and analyse the institutional, ethical, economic and legal considerations raised by long term radioactive waste storage or disposal on the three interrelated issues of: responsibility and ownership of radioactive waste on the long term, continuity of local dialogue between stakeholders and monitoring of radioactive waste management facilities, and compensation and sustainable development. The aim is also to propose guidelines in order to better address long term issues in decision-making processes and start long term governance

  1. Long term governance for radioactive waste management. Final report of Cowan2 - work package 4

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, Th.; Schieber, C.; Lavele, S.

    2006-12-15

    This report aims at identifying key features for the long term governance of radioactive waste. It is proposed by the COWAN2 Work Package 4 the purpose of which was to identify, discuss and analyse the institutional, ethical, economic and legal considerations raised by long term radioactive waste storage or disposal on the three interrelated issues of: responsibility and ownership of radioactive waste on the long term, continuity of local dialogue between stakeholders and monitoring of radioactive waste management facilities, and compensation and sustainable development. The aim is also to propose guidelines in order to better address long term issues in decision-making processes and start long term governance

  2. Magnus: A New Resistive MHD Code with Heat Flow Terms

    Science.gov (United States)

    Navarro, Anamaría; Lora-Clavijo, F. D.; González, Guillermo A.

    2017-07-01

    We present a new magnetohydrodynamic (MHD) code for the simulation of wave propagation in the solar atmosphere, under the effects of electrical resistivity—but not dominant—and heat transference in a uniform 3D grid. The code is based on the finite-volume method combined with the HLLE and HLLC approximate Riemann solvers, which use different slope limiters like MINMOD, MC, and WENO5. In order to control the growth of the divergence of the magnetic field, due to numerical errors, we apply the Flux Constrained Transport method, which is described in detail to understand how the resistive terms are included in the algorithm. In our results, it is verified that this method preserves the divergence of the magnetic fields within the machine round-off error (˜ 1× {10}-12). For the validation of the accuracy and efficiency of the schemes implemented in the code, we present some numerical tests in 1D and 2D for the ideal MHD. Later, we show one test for the resistivity in a magnetic reconnection process and one for the thermal conduction, where the temperature is advected by the magnetic field lines. Moreover, we display two numerical problems associated with the MHD wave propagation. The first one corresponds to a 3D evolution of a vertical velocity pulse at the photosphere-transition-corona region, while the second one consists of a 2D simulation of a transverse velocity pulse in a coronal loop.

  3. CARP: a computer code and albedo data library for use by BREESE, the MORSE albedo package

    International Nuclear Information System (INIS)

    Emmett, M.B.; Rhoades, W.A.

    1978-10-01

    The CARP computer code was written to allow processing of DOT angular flux tapes to produce albedo data for use in the MORSE computer code. An albedo data library was produced containing several materials. 3 tables

  4. Modular Modeling System (MMS) code: a versatile power plant analysis package

    International Nuclear Information System (INIS)

    Divakaruni, S.M.; Wong, F.K.L.

    1987-01-01

    The basic version of the Modular Modeling System (MMS-01), a power plant systems analysis computer code jointly developed by the Nuclear Power and the Coal Combustion Systems Divisions of the Electric Power Research Institute (EPRI), has been released to the utility power industry in April 1983 at a code release workshop held in Charlotte, North Carolina. Since then, additional modules have been developed to analyze the Pressurized Water Reactors (PWRs) and the Boiling Water Reactors (BWRs) when the safety systems are activated. Also, a selected number of modules in the MMS-01 library have been modified to allow the code users more flexibility in constructing plant specific systems for analysis. These new PWR and BWR modules constitute the new MMS library, and it includes the modifications to the MMS-01 library. A year and half long extensive code qualification program of this new version of the MMS code at EPRI and the contractor sites, back by further code testing in an user group environment is culminating in the MMS-02 code release announcement seminar. At this seminar, the results of user group efforts and the code qualification program will be presented in a series of technical sessions. A total of forty-nine papers will be presented to describe the new code features and the code qualification efforts. For the sake of completion, an overview of the code is presented to include the history of the code development, description of the MMS code and its structure, utility engineers involvement in MMS-01 and MMS-02 validations, the enhancements made in the last 18 months to the code, and finally the perspective on the code future in the fossil and nuclear industry

  5. Verification of the 2.00 WAPPA-B [Waste Package Performance Assessment-B version] code

    International Nuclear Information System (INIS)

    Tylock, B.; Jansen, G.; Raines, G.E.

    1987-07-01

    The old version of the Waste Package Performance Assessment (WAPPA) code has been modified into a new code version, 2.00 WAPPA-B. The input files and the results for two benchmarks at repository conditions are fully documented in the appendixes of the EA reference report. The 2.00 WAPPA-B version of the code is suitable for computation of barrier failure due to uniform corrosion; however, an improved sub-version, 2.01 WAPPA-B, is recommended for general use due to minor errors found in 2.00 WAPPA-B during its verification procedures. The input files and input echoes have been modified to include behavior of both radionuclides and elements, but the 2.00 WAPPA-B version of the WAPPA code is not recommended for computation of radionuclide releases. The 2.00 WAPPA-B version computes only mass balances and the initial presence of radionuclides that can be released. Future code development in the 3.00 WAPPA-C version will include radionuclide release computations. 19 refs., 10 figs., 1 tab

  6. The DarkStars code: a publicly available dark stellar evolution package

    CERN Document Server

    Scott, Pat; Fairbairn, Malcolm

    2009-01-01

    We announce the public release of the 'dark' stellar evolution code DarkStars. The code simultaneously solves the equations of WIMP capture and annihilation in a star with those of stellar evolution assuming approximate hydrostatic equilibrium. DarkStars includes the most extensive WIMP microphysics of any dark evolution code to date. The code employs detailed treatments of the capture process from a range of WIMP velocity distributions, as well as composite WIMP distribution and conductive energy transport schemes based on the WIMP mean-free path in the star. We give a brief description of the input physics and practical usage of the code, as well as examples of its application to dark stars at the Galactic centre.

  7. 9975 Shipping Package Performance Of Alternate Materials For Long-Term Storage Application

    International Nuclear Information System (INIS)

    Skidmore, E.; Hoffman, E.; Daugherty, W.

    2010-01-01

    The Model 9975 shipping package specifies the materials of construction for its various components. With the loss of availability of material for two components (cane fiberboard overpack and Viton(reg s ign) GLT O-rings), alternate materials of construction were identified and approved for use for transport (softwood fiberboard and Viton(reg s ign) GLT-S O-rings). As these shipping packages are part of a long-term storage configuration at the Savannah River Site, additional testing is in progress to verify satisfactory long-term performance of the alternate materials under storage conditions. The test results to date can be compared to comparable results on the original materials of construction to draw preliminary conclusions on the performance of the replacement materials.

  8. Long-Term Leak Tightness Of O-Ring Seals In The 9975 Shipping Package

    International Nuclear Information System (INIS)

    Hoffman, E.; Skidmore, E.; Daugherty, W.

    2010-01-01

    O-ring seals in the 9975 shipping package containment vessels are fabricated from a Viton GLT or GLT-S compound. Long-term testing of these O-rings has been performed to support service life predictions for packages used for long-term storage. Since the only criterion for O-ring performance is to maintain a leak-tight seal, leak testing is the primary indicator of service life. Fixtures have been aging at elevated temperatures to provide data for service life predictions. Limited leak test failures have been observed at the higher temperatures. This provides the opportunity for comparison to trends based on other O-ring properties, such as compression stress relaxation. Initial data suggest that the CSR data have some predictive value for a leak-tight service life, but other factors can complicate efforts to draw definitive conclusions.

  9. The release code package REVOLS/RENONS for fission product release from a liquid sodium pool into an inert gas atmosphere

    International Nuclear Information System (INIS)

    Starflinger, J.; Scholtyssek, W.; Unger, H.

    1994-12-01

    For aerosol source term considerations in the field of nuclear safety, the investigation of the release of volatile and non-volatile species from liquid surfaces into a gas atmosphere is important. In case of a hypothetical liquid metal fast breeder reactor accident with tank failure, primary coolant sodium with suspended or solved fuel particles and fission products may be released into the containment. The computer code package REVOLS/RENONS, based on a theoretical mechanistic model with a modular structure, has been developed for the prediction of sodium release as well as volatile and non-volatile radionuclide release from a liquid pool surface into the inert gas atmosphere of the inner containment. Hereby the release of sodium and volatile fission products, like cesium and sodium iodide, is calculated using a theoretical model in a mass transfer coefficient formulation. This model has been transposed into the code version REVOLS.MOD1.1, which is discussed here. It enables parameter analysis under highly variable user-defined boundary conditions. Whereas the evaporative release of the volatile components is governed by diffusive and convective transport processes, the release of the non-volatile ones may be governed by mechanical processes which lead to droplet entrainment from the wavy pool surface under conditions of natural or forced convection into the atmosphere. The mechanistic model calculates the liquid entrainment rate of the non-volatile species, like the fission product strontium oxide and the fuel (uranium dioxide) from a liquid pool surface into a parallel gas flow. The mechanistic model has been transposed into the computer code package REVOLS/RENONS, which is discussed here. Hereby the module REVOLS (RElease of VOLatile Species) calculates the evaporative release of the volatile species, while the module RENONS (RElease of NON-Volatile Species) computes the entrainment release of the non-volatile radionuclides. (orig./HP) [de

  10. DOE progress in assessing the long term performance of waste package materials

    International Nuclear Information System (INIS)

    Berusch, A.; Gause, E.

    1987-01-01

    Under the Nuclear Waste Policy Act of 1982 (NWPA)[1], the US Dept. of Energy (DOE) is conducting activities to select and characterize candidate sites suitable for the construction and operation of a geologic repository for the disposal of high-level nuclear wastes. DOE is funding three first repository projects: Basalt Waste Isolation Project, BWIP; Nevada Nuclear Waste Isolation Project, NNWSI; and Salt Repository Project Office, SRPO. It is essential in the licensing process that DOE demonstrate to the NRC that the long-term performance of the materials and design will be in compliance with the requirements of 10 CFR 60.113 on substantially complete containment within the waste packages for 300 to 1000 years and a controlled release rate from the engineered barrier system (EBS) for 10,000 years of 1 part in 10 5 per year for radionuclides present in defined quantities 100 years after permanent closure. Obviously, the time spans involved make it impractical to base the assessment of the long term performance of waste package materials on real time, prototypical testing. The assessment of performance will be implemented by the use of models that are supported by real time field and laboratory tests, monitoring, and natural analog studies. Each of the repository projects is developing a plan for demonstrating long-term waste package material performance depending on the particular materials and the package-perturbed, time-dependent environment under which the materials must function. An overview of progress in each of these activities for each of the projects is provided in the following

  11. Characterizing and packaging BN-350 spent fuel for long-term dry storage

    International Nuclear Information System (INIS)

    Lambert, J. D. B.; Bolshinsky, I.; Haues, S.L.; Allen, K.J.; Howden, E.A.; Hill, R.N.; Planchon, H.P.; Staples, P.; Karaulov, V.N.; Blynskij, A.P.; Yakovlev, I.K.; Maev, V.; Dumchev, I. A.

    2000-01-01

    The Republic of Kazakhstan is being assisted by the U.S. Department of Energy in preparing spent fuel from the BN-350 fast reactor for long term dry storage. Argonne National Laboratory was assigned responsibility for the physical and nuclear characterization of the spent fuel, for the design and safety analysis of 6-pac and 4-pac canisters used to contain spent fuel assemblies for storage, and for the design, testing and installation of a closure station at the reactor in which the canisters of fuel are dried, filled with inert gas and welded shut. This paper briefly describes the specialized components and equipment used, the process followed, and experience gained in packaging the spent fuel. Olsen et al and Schaefer separately discuss overall safety and criticality considerations of the packaging process in parallel papers to this conference

  12. Short-Term Memory Coding in Children With Intellectual Disabilities

    OpenAIRE

    Henry, L.; Conners, F.

    2008-01-01

    To examine visual and verbal coding strategies, I asked children with intellectual disabilities and peers matched for MA and CA to perform picture memory span tasks with phonologically similar, visually similar, long, or nonsimilar named items. The CA group showed effects consistent with advanced verbal memory coding (phonological similarity and word length effects). Neither the intellectual disabilities nor MA groups showed evidence for memory coding strategies. However, children in these gr...

  13. Standard problem exercise to validate criticality codes for large arrays of packages of fissile materials

    International Nuclear Information System (INIS)

    Whitesides, G.E.; Stephens, M.E.

    1986-01-01

    A study has been conducted by an Office of Economic Cooperation and Development-Committee on the Safety of Nuclear Installations (OECD-CSNI) Working Group that examined computational methods used to compute k/sub eff/ for large greater than or equal to5 3 arrays of fissile material (in which each unit is a substantial fraction of a critical mass). Five fissile materials that might typically be transported were used in the study. The ''packages'' used for this exercise were simplified to allow studies unperturbed by the variety of structural materials which would exist in an actual package. The only material present other than the fissile material was a variation in the moderator (water) surrounding the fissile material. Consistent results were obtained from calculations using several computational methods. That is, when the bias demonstrated by each method for actual critical experiments was used to ''correct'' the results obtained for systems for which there were no experimental data, there was good agreement between the methods. Two major areas of concern were raised by this exercise. First, the lack of experimental data for arrays with size greater than 5 3 limits validation for large systems. Second, there is a distinct possibility that the comingling of two shipments of unlike units could result in a reduction of the safety margins. Additional experiments and calculations will be required to satisfactorily resolve the remaining questions regarding the safe transport of large arrays of fissile materials

  14. The Modularized Software Package ASKI - Full Waveform Inversion Based on Waveform Sensitivity Kernels Utilizing External Seismic Wave Propagation Codes

    Science.gov (United States)

    Schumacher, F.; Friederich, W.

    2015-12-01

    We present the modularized software package ASKI which is a flexible and extendable toolbox for seismic full waveform inversion (FWI) as well as sensitivity or resolution analysis operating on the sensitivity matrix. It utilizes established wave propagation codes for solving the forward problem and offers an alternative to the monolithic, unflexible and hard-to-modify codes that have typically been written for solving inverse problems. It is available under the GPL at www.rub.de/aski. The Gauss-Newton FWI method for 3D-heterogeneous elastic earth models is based on waveform sensitivity kernels and can be applied to inverse problems at various spatial scales in both Cartesian and spherical geometries. The kernels are derived in the frequency domain from Born scattering theory as the Fréchet derivatives of linearized full waveform data functionals, quantifying the influence of elastic earth model parameters on the particular waveform data values. As an important innovation, we keep two independent spatial descriptions of the earth model - one for solving the forward problem and one representing the inverted model updates. Thereby we account for the independent needs of spatial model resolution of forward and inverse problem, respectively. Due to pre-integration of the kernels over the (in general much coarser) inversion grid, storage requirements for the sensitivity kernels are dramatically reduced.ASKI can be flexibly extended to other forward codes by providing it with specific interface routines that contain knowledge about forward code-specific file formats and auxiliary information provided by the new forward code. In order to sustain flexibility, the ASKI tools must communicate via file output/input, thus large storage capacities need to be accessible in a convenient way. Storing the complete sensitivity matrix to file, however, permits the scientist full manual control over each step in a customized procedure of sensitivity/resolution analysis and full

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-11-01

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

  16. Short-Term Memory Coding in Children with Intellectual Disabilities

    Science.gov (United States)

    Henry, Lucy

    2008-01-01

    To examine visual and verbal coding strategies, I asked children with intellectual disabilities and peers matched for MA and CA to perform picture memory span tasks with phonologically similar, visually similar, long, or nonsimilar named items. The CA group showed effects consistent with advanced verbal memory coding (phonological similarity and…

  17. Short-term memory coding in children with intellectual disabilities.

    Science.gov (United States)

    Henry, Lucy

    2008-05-01

    To examine visual and verbal coding strategies, I asked children with intellectual disabilities and peers matched for MA and CA to perform picture memory span tasks with phonologically similar, visually similar, long, or nonsimilar named items. The CA group showed effects consistent with advanced verbal memory coding (phonological similarity and word length effects). Neither the intellectual disabilities nor MA groups showed evidence for memory coding strategies. However, children in these groups with MAs above 6 years showed significant visual similarity and word length effects, broadly consistent with an intermediate stage of dual visual and verbal coding. These results suggest that developmental progressions in memory coding strategies are independent of intellectual disabilities status and consistent with MA.

  18. Parylene-on-oil packaging for long-term implantable pressure sensors.

    Science.gov (United States)

    Shapero, Aubrey M; Liu, Yang; Tai, Yu-Chong

    2016-08-01

    This paper reports and analyzes the feasibility study of a parylene-on-oil encapsulation packaging method of pressure sensors targeted for long-term implantation. Commercial barometric digital-output pressure sensors are enclosed in silicone oil and then encapsulated in situ with parylene-C or -D (PA-C, PA-D) chemical vapor deposition. Experimentally, sensors encapsulated with 30,000 cSt silicone oil and 27 μm PA-D show good performance for 6 weeks in 77 °C saline with >99 % of original sensitivity, corresponding to an extrapolated lifetime of around 21 months in 37 °C saline. This work shows that, with proper designs, such a packaging method can preserve the original pressure sensor sensitivity without offset, validated throughout accelerated lifetime tests. In experiments, wires on the prototypes are used for external electronics but it is found that they contributed to early failures, which would be absent in real wireless versions, indicating a potential for even longer lifetimes. Finally, a verified model is presented to predict the pressure sensor sensitivity of parylene-on-oil packaging with and without the presence of a bubble in the oil.

  19. Containers analysis code of zero order (CACO0) - A basic design system for Type B packages

    International Nuclear Information System (INIS)

    Gaspar, C.; Benito, G.; Rey, J.C.

    1989-01-01

    Very frequently, the principal issues that have to be assessed in the design of a type B(U) package are radiation shielding and evaluation of mechanical and thermal test effects. Thermal behavior during normal transport conditions has also to be considered when the material must dissipate high thermal power. If the transported material is fissile it should be assured that it remains subcritical during transport. The containment of radioactive material must always be assured. In some cases this requires considerable effort. Usually these different design issues are very closely coupled. This coupling does not permit independent consideration. Also, some issues are competitive and generate conflicting design criteria. Given the goal of meeting pertinent transport regulations at a reasonable cost, all design-relevant issues must be balanced in order to obtain a good design. For each design-relevant issue there exists a number of methods of varying efficiency and cost, which can be used to define the key parameters of those particular issues. The overall design methodology must taken into account interactions between parameters of different issues. CACO0 is a system that integrates all design relevant issues and their interactions. The system consists of different modules, each one oriented to a different design issue. The modules are related by a control structure that enables sequentation or iteration during design in a fast and simple manner. Modules can easily be replaced or added, so the system can be updated or adapted to new design problems. The system was designed for use in factibility analysis, cost estimation, conceptual design and initial stages of basic design of type B(U) packages. To accomplish those ends, simple, fast and conservative methods are used

  20. Synthesis of knowledge on the long-term behaviour of concretes. Applications to cemented waste packages; Synthese des connaissances sur le comportement a long terme des betons. Application aux colis cimentes

    Energy Technology Data Exchange (ETDEWEB)

    Richet, C.; Galle, C.; Le Bescop, P.; Peycelon, H.; Bejaoui, S.; Tovena, I.; Pointeau, I.; L' Hostis, V.; Levera, P

    2004-03-01

    As stipulated in the former law of December 91 relating to 'concrete waste package', a progress report (phenomenological reference document) was first provided in 1999. The objective was to make an assessment of the knowledge acquired on the long-term behaviour of cement-based waste packages in the context of deep disposal and/or interim storage. The present document is an updated summary report. It takes into account a new knowledge assessment, considers coupled mechanisms and should contribute to the first performance studies (operational calculations). Handling and radio-nuclides (RN) confinement are the two major functional properties requested from the concrete used for the waste packages. In unsaturated environment (interim storage/disposal prior to closing), the main problem is the generation of cracks in the material. This aspect is a key parameter from the mechanical point of view (retrievability). It can have a major impact on the disposal phase (confinement). In saturated environment (disposal post-closing phase), the main concern is the chemical degradation of the waste package concrete submitted to underground waters leaching. In this context, the major thema are: the durability of the concretes under water (chemical degradation) and in unsaturated medium (corrosion of reinforcement), matter transport, RN retention, chemistry / transport / mechanical couplings. On the other hand, laboratory data on the behaviour of concretes are used to evaluate the RN source term of waste packages in function of time (concrete waste package OPerational Model, i.e. 'Concrete MOP'). The 'MOP' provides the physico-chemical description of the RN release in relationship with the waste package degradation itself. This description is based on simplified phenomenology for which only dimensioning mechanisms are taken into account. The use of Diffu-Ca code (basic module for the MOP) on the CASTEM numerical plate-form, already allows operational

  1. Synthesis of knowledge on the long-term behaviour of concretes. Applications to cemented waste packages; Synthese des connaissances sur le comportement a long terme des betons. Application aux colis cimentes

    Energy Technology Data Exchange (ETDEWEB)

    Richet, C; Galle, C; Le Bescop, P; Peycelon, H; Bejaoui, S; Tovena, I; Pointeau, I; L' Hostis, V; Levera, P

    2004-03-01

    As stipulated in the former law of December 91 relating to 'concrete waste package', a progress report (phenomenological reference document) was first provided in 1999. The objective was to make an assessment of the knowledge acquired on the long-term behaviour of cement-based waste packages in the context of deep disposal and/or interim storage. The present document is an updated summary report. It takes into account a new knowledge assessment, considers coupled mechanisms and should contribute to the first performance studies (operational calculations). Handling and radio-nuclides (RN) confinement are the two major functional properties requested from the concrete used for the waste packages. In unsaturated environment (interim storage/disposal prior to closing), the main problem is the generation of cracks in the material. This aspect is a key parameter from the mechanical point of view (retrievability). It can have a major impact on the disposal phase (confinement). In saturated environment (disposal post-closing phase), the main concern is the chemical degradation of the waste package concrete submitted to underground waters leaching. In this context, the major thema are: the durability of the concretes under water (chemical degradation) and in unsaturated medium (corrosion of reinforcement), matter transport, RN retention, chemistry / transport / mechanical couplings. On the other hand, laboratory data on the behaviour of concretes are used to evaluate the RN source term of waste packages in function of time (concrete waste package OPerational Model, i.e. 'Concrete MOP'). The 'MOP' provides the physico-chemical description of the RN release in relationship with the waste package degradation itself. This description is based on simplified phenomenology for which only dimensioning mechanisms are taken into account. The use of Diffu-Ca code (basic module for the MOP) on the CASTEM numerical plate-form, already allows operational predictions. (authors)

  2. NET IBK Computer code package for the needs of planning, construction and operation of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M V; Kocic, A; Marinkovic, N; Milosevic, M; Stancic, V [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1978-07-01

    Within the Nuclear Engineering Laboratory of the Boris Kidric Institute of Nuclear Sciences (NET IBK) a systematic work has been performed on collecting nuclear data for reactor calculation needs, on developing own methods and computing programs for reactor calculations, as well as on adapting and applying the foreign methods and codes. In this way a complete library of computer programs was formed for precise prediction of nuclear fuel burnup and depletion, for evaluation of the Power distribution variations with irradiation, for computing the amount of produced plutonium and its number densities etc. Programs for evaluation of location of different types of safety and economic analysis have been developed as well. The aim of this paper is to present our abilities to perform complex computations needed for planning, constructing and operating the nuclear power plants, by describing the NET IBK computer programs package. (author)

  3. An implicit steady-state initialization package for the RELAP5 computer code

    International Nuclear Information System (INIS)

    Paulsen, M.P.; Peterson, C.E.; Odar, F.

    1995-08-01

    A direct steady-state initialization (DSSI) method has been developed and implemented in the RELAP5 hydrodynamic analysis program. It provides a means for users to specify a small set of initial conditions which are then propagated through the remainder of the system. The DSSI scheme utilizes the steady-state form of the RELAP5 balance equations for nonequilibrium two-phase flow. It also employs the RELAP5 component models and constitutive model packages for wall-to-phase and interphase momentum and heat exchange. A fully implicit solution of the linearized hydrodynamic equations is implemented. An implicit coupling scheme is used to augment the standard steady-state heat conduction solution for steam generator use. It solves the primary-side tube region energy equations, heat conduction equations, wall heat flux boundary conditions, and overall energy balance equation as a coupled system of equations and improves convergence. The DSSI method for initializing RELAP5 problems to steady-state conditions has been compared with the transient solution scheme using a suite of test problems including; adiabatic single-phase liquid and vapor flow through channels with and without healing and area changes; a heated two-phase test bundle representative of BWR core conditions; and a single-loop PWR model

  4. SWAAM-LT: The long-term, sodium/water reaction analysis method computer code

    International Nuclear Information System (INIS)

    Shin, Y.W.; Chung, H.H.; Wiedermann, A.H.; Tanabe, H.

    1993-01-01

    The SWAAM-LT Code, developed for analysis of long-term effects of sodium/water reactions, is discussed. The theoretical formulation of the code is described, including the introduction of system matrices for ease of computer programming as a general system code. Also, some typical results of the code predictions for available large scale tests are presented. Test data for the steam generator design with the cover-gas feature and without the cover-gas feature are available and analyzed. The capabilities and limitations of the code are then discussed in light of the comparison between the code prediction and the test data

  5. Code of Practice on radiation protection in the mining and milling of radioactive ores (1980) - Guidelines for storage and packaging of uranium concentrates

    International Nuclear Information System (INIS)

    1986-01-01

    This Guideline is intended to provide assistance in the application of the 1980 Code of Practice on radiation protection in mining and milling of radioactive ores. Its purpose is to give advice relevant to the design, construction and operation of an uranium concentrate storage and packaging facility in which exposure to ionizing radiation from uranium-bearing concentrate will not only conform to the Code, but will also be as low as reasonably achievable. (NEA) [fr

  6. Long-term corrosion behaviour of low-/medium-level waste packages

    International Nuclear Information System (INIS)

    Jendras, M.; Bach, F.W.; Behrens, S.; Birr, Ch.; Hassel, Th.

    2009-01-01

    Full text of publication follows: Storage of low- and medium-level radioactive waste requires safe packages. This means that all materials used for the manufacturing of such packages have to show a sufficient resistance especially against corrosive attacks. Since these packages are generally made from carbon steel an additional coating for corrosion protection - mainly solvent-based polymers - is necessary. However, it is not enough to consider the selection and combination of the materials. Regarding the construction and manufacturing of corrosion-resistant drums for low- and medium-level radioactive waste there also has to be paid closer attention to the joining technologies such as welding. For lifetime prediction of low-/medium-level waste packages reliable experimental data concerning the long-term corrosion behaviour of each material as well as of the components is needed. Therefore sheet metals from carbon steel were galvanized or coated with different solvent-based and water-based corrosion protection materials (epoxy as well as silicone resins). After damaging the anti-corrosion coating of some of these sheets with predefined scratches sets of these samples were stored at higher temperatures in climatic chamber, in simulated waste or aged according to standard DIN EN ISO 9227. All corrosion damages were analyzed by means of metallography (light microscopy as well as scanning electron microscopy of micro-sections). The quantitative influence of the corrosive attacks on the mechanical properties of the materials was examined by mechanical testing according to DIN EN 10002. Besides reduction of tensile strength drastic reduction of percentage of elongation after fracture (from 30 % to 10 %) was found. Further experiments were carried out using components or scaled-down drums joined by means of innovative welding techniques such as Cold Arc or Force Arc. The relevant welding parameters (e.g. welding current, proper volume of shielding gas or wire feed) were

  7. RepoSTAR. A Code package for control and evaluation of statistical calculations with the program package RepoTREND; RepoSTAR. Ein Codepaket zur Steuerung und Auswertung statistischer Rechenlaeufe mit dem Programmpaket RepoTREND

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Dirk-Alexander

    2016-05-15

    The program package RepoTREND for the integrated long terms safety analysis of final repositories allows besides deterministic studies of defined problems also statistical or probabilistic analyses. Probabilistic uncertainty and sensitivity analyses are realized in the program package repoTREND by a specific statistic frame called RepoSTAR. The report covers the following issues: concept, sampling and data supply of single simulations, evaluation of statistical calculations with the program RepoSUN.

  8. R and D applied to the non-destructive characterization of waste packages for long term storage or deep disposal

    Energy Technology Data Exchange (ETDEWEB)

    Malvache, P.; Perot, B.; Ma, J.L.; Pettier, J.L. [CEA Cadarache, Dept. d' Etudes des Dechets, DED, 13 - Saint Paul lez Durance (France); Capdevila, J.M.; Huot, N. [CEA Saclay, Dir. du Developpement et de l' Innovation Nucleares DDIN, 91 - Gif Sur Yvette (France); Moulin, V. [CEA Grenoble, Lab. d' Electronique, de Technologie de l' Information LETI, DSYS, 38 (France)

    2001-07-01

    To ensure the quality and traceability of waste package management in the long term, knowledge on these packages is necessary so as to confirm their compliance to storage or disposal specifications. Research is focused on the management of the knowledge on these packages (fabrication means, materials contained,...) and on the acquisition, through measurement, of their characteristics. Within this context, many studies are underway at the CEA in the field of measurements so as to obtain non- destructive tools to access the parameters which allow the waste packages to be characterized. The two main R and D investigations concern: the nuclear measurement methods for the detection and quantification of radionuclides and of chemical elements considered as important for storage or disposal safety ; the measurement methods for the physical characteristics of the packages by high energy photon imaging, thus allowing pictures of the contents of large, high density and sometimes irradiating packages to be known. During the last five years, the research at the CEA focused on these two areas and resulted in a significant evolution in the non-destructive characterization means for long lived waste packages. (author)

  9. R and D applied to the non-destructive characterization of waste packages for long term storage or deep disposal

    International Nuclear Information System (INIS)

    Malvache, P.; Perot, B.; Ma, J.L.; Pettier, J.L.; Capdevila, J.M.; Huot, N.; Moulin, V.

    2001-01-01

    To ensure the quality and traceability of waste package management in the long term, knowledge on these packages is necessary so as to confirm their compliance to storage or disposal specifications. Research is focused on the management of the knowledge on these packages (fabrication means, materials contained,...) and on the acquisition, through measurement, of their characteristics. Within this context, many studies are underway at the CEA in the field of measurements so as to obtain non- destructive tools to access the parameters which allow the waste packages to be characterized. The two main R and D investigations concern: the nuclear measurement methods for the detection and quantification of radionuclides and of chemical elements considered as important for storage or disposal safety ; the measurement methods for the physical characteristics of the packages by high energy photon imaging, thus allowing pictures of the contents of large, high density and sometimes irradiating packages to be known. During the last five years, the research at the CEA focused on these two areas and resulted in a significant evolution in the non-destructive characterization means for long lived waste packages. (author)

  10. Characterization of open-cycle coal-fired MHD generators. Quarterly technical summary report No. 6, October 1--December 31, 1977. [PACKAGE code

    Energy Technology Data Exchange (ETDEWEB)

    Kolb, C.E.; Yousefian, V.; Wormhoudt, J.; Haimes, R.; Martinez-Sanchez, M.; Kerrebrock, J.L.

    1978-01-30

    Research has included theoretical modeling of important plasma chemical effects such as: conductivity reductions due to condensed slag/electron interactions; conductivity and generator efficiency reductions due to the formation of slag-related negative ion species; and the loss of alkali seed due to chemical combination with condensed slag. A summary of the major conclusions in each of these areas is presented. A major output of the modeling effort has been the development of an MHD plasma chemistry core flow model. This model has been formulated into a computer program designated the PACKAGE code (Plasma Analysis, Chemical Kinetics, And Generator Efficiency). The PACKAGE code is designed to calculate the effect of coal rank, ash percentage, ash composition, air preheat temperatures, equivalence ratio, and various generator channel parameters on the overall efficiency of open-cycle, coal-fired MHD generators. A complete description of the PACKAGE code and a preliminary version of the PACKAGE user's manual are included. A laboratory measurements program involving direct, mass spectrometric sampling of the positive and negative ions formed in a one atmosphere coal combustion plasma was also completed during the contract's initial phase. The relative ion concentrations formed in a plasma due to the methane augmented combustion of pulverized Montana Rosebud coal with potassium carbonate seed and preheated air are summarized. Positive ions measured include K/sup +/, KO/sup +/, Na/sup +/, Rb/sup +/, Cs/sup +/, and CsO/sup +/, while negative ions identified include PO/sub 3//sup -/, PO/sub 2//sup -/, BO/sub 2//sup -/, OH/sup -/, SH/sup -/, and probably HCrO/sub 3/, HMoO/sub 4//sup -/, and HWO/sub 3//sup -/. Comparison of the measurements with PACKAGE code predictions are presented. Preliminary design considerations for a mass spectrometric sampling probe capable of characterizing coal combustion plasmas from full scale combustors and flow trains are presented

  11. Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

    International Nuclear Information System (INIS)

    Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang

    2005-01-01

    Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

  12. Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

    Energy Technology Data Exchange (ETDEWEB)

    Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang [Nuclear Department, Suez-Tractebel Engineering, avenue Ariane 5, B-1200 Brussels (Belgium)

    2005-07-01

    Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

  13. A long-term, integrated impact assessment of alternative building energy code scenarios in China

    International Nuclear Information System (INIS)

    Yu, Sha; Eom, Jiyong; Evans, Meredydd; Clarke, Leon

    2014-01-01

    China is the second largest building energy user in the world, ranking first and third in residential and commercial energy consumption. Beginning in the early 1980s, the Chinese government has developed a variety of building energy codes to improve building energy efficiency and reduce total energy demand. This paper studies the impact of building energy codes on energy use and CO 2 emissions by using a detailed building energy model that represents four distinct climate zones each with three building types, nested in a long-term integrated assessment framework GCAM. An advanced building stock module, coupled with the building energy model, is developed to reflect the characteristics of future building stock and its interaction with the development of building energy codes in China. This paper also evaluates the impacts of building codes on building energy demand in the presence of economy-wide carbon policy. We find that building energy codes would reduce Chinese building energy use by 13–22% depending on building code scenarios, with a similar effect preserved even under the carbon policy. The impact of building energy codes shows regional and sectoral variation due to regionally differentiated responses of heating and cooling services to shell efficiency improvement. - Highlights: • We assessed long-term impacts of building codes and climate policy using GCAM. • Building energy codes would reduce Chinese building energy use by 13–22%. • The impacts of codes on building energy use vary by climate region and sub-sector

  14. ANITA-2000 activation code package - updating of the decay data libraries and validation on the experimental data of the 14 MeV Frascati Neutron Generator

    Directory of Open Access Journals (Sweden)

    Frisoni Manuela

    2016-01-01

    Full Text Available ANITA-2000 is a code package for the activation characterization of materials exposed to neutron irradiation released by ENEA to OECD-NEADB and ORNL-RSICC. The main component of the package is the activation code ANITA-4M that computes the radioactive inventory of a material exposed to neutron irradiation. The code requires the decay data library (file fl1 containing the quantities describing the decay properties of the unstable nuclides and the library (file fl2 containing the gamma ray spectra emitted by the radioactive nuclei. The fl1 and fl2 files of the ANITA-2000 code package, originally based on the evaluated nuclear data library FENDL/D-2.0, were recently updated on the basis of the JEFF-3.1.1 Radioactive Decay Data Library. This paper presents the results of the validation of the new fl1 decay data library through the comparison of the ANITA-4M calculated values with the measured electron and photon decay heats and activities of fusion material samples irradiated at the 14 MeV Frascati Neutron Generator (FNG of the NEA-Frascati Research Centre. Twelve material samples were considered, namely: Mo, Cu, Hf, Mg, Ni, Cd, Sn, Re, Ti, W, Ag and Al. The ratios between calculated and experimental values (C/E are shown and discussed in this paper.

  15. Methodologies for assessing long-term performance of high-level radioactive waste packages

    International Nuclear Information System (INIS)

    Stephens, K.; Boesch, L.; Crane, B.; Johnson, R.; Moler, R.; Smith, S.; Zaremba, L.

    1986-01-01

    Several methods the Nuclear Regulatory Commission (NRC) can use to independently assess Department of Energy (DOE) waste package performance were identified by The Aerospace Corporation. The report includes an overview of the necessary attributes of performance assessment, followed by discussions of DOE methods, probabilistic methods capable of predicting waste package lifetime and radionuclide releases, process modeling of waste package barriers, sufficiency of the necessary input data, and the applicability of probability density functions. It is recommended that the initial NRC performance assessment (for the basalt conceptual waste package design) should apply modular simulation, using available process models and data, to demonstrate this assessment method

  16. Lysimeter data as input to performance assessment source term codes

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Rogers, R.D.; Sullivan, T.

    1992-01-01

    The Field Lysimeter Investigation: Low-Level Waste Data Base Development Program is obtaining information on the performance of radioactive waste in a disposal environment. Waste forms fabricated using ion-exchange resins from EPICOR-II c prefilters employed in the cleanup of the Three Mile Island (TMI) Nuclear Power Station are being tested to develop a low-level waste data base and to obtain information on survivability of waste forms in a disposal environment. In this paper, radionuclide releases from waste forms in the first seven years of sampling are presented and discussed. Application of lysimeter data to be used in performance assessment source term models is presented. Initial results from use of data in two models are discussed

  17. Phonological, visual, and semantic coding strategies and children's short-term picture memory span.

    Science.gov (United States)

    Henry, Lucy A; Messer, David; Luger-Klein, Scarlett; Crane, Laura

    2012-01-01

    Three experiments addressed controversies in the previous literature on the development of phonological and other forms of short-term memory coding in children, using assessments of picture memory span that ruled out potentially confounding effects of verbal input and output. Picture materials were varied in terms of phonological similarity, visual similarity, semantic similarity, and word length. Older children (6/8-year-olds), but not younger children (4/5-year-olds), demonstrated robust and consistent phonological similarity and word length effects, indicating that they were using phonological coding strategies. This confirmed findings initially reported by Conrad (1971), but subsequently questioned by other authors. However, in contrast to some previous research, little evidence was found for a distinct visual coding stage at 4 years, casting doubt on assumptions that this is a developmental stage that consistently precedes phonological coding. There was some evidence for a dual visual and phonological coding stage prior to exclusive use of phonological coding at around 5-6 years. Evidence for semantic similarity effects was limited, suggesting that semantic coding is not a key method by which young children recall lists of pictures.

  18. Development of Coupled Interface System between the FADAS Code and a Source-term Evaluation Code XSOR for CANDU Reactors

    International Nuclear Information System (INIS)

    Son, Han Seong; Song, Deok Yong; Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon

    2006-01-01

    An accident prevention system is essential to the industrial security of nuclear industry. Thus, the more effective accident prevention system will be helpful to promote safety culture as well as to acquire public acceptance for nuclear power industry. The FADAS(Following Accident Dose Assessment System) which is a part of the Computerized Advisory System for a Radiological Emergency (CARE) system in KINS is used for the prevention against nuclear accident. In order to enhance the FADAS system more effective for CANDU reactors, it is necessary to develop the various accident scenarios and reliable database of source terms. This study introduces the construction of the coupled interface system between the FADAS and the source-term evaluation code aimed to improve the applicability of the CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors

  19. ANITA-IEAF activation code package - updating of the decay and cross section data libraries and validation on the experimental data from the Karlsruhe Isochronous Cyclotron

    Science.gov (United States)

    Frisoni, Manuela

    2017-09-01

    ANITA-IEAF is an activation package (code and libraries) developed in the past in ENEA-Bologna in order to assess the activation of materials exposed to neutrons with energies greater than 20 MeV. An updated version of the ANITA-IEAF activation code package has been developed. It is suitable to be applied to the study of the irradiation effects on materials in facilities like the International Fusion Materials Irradiation Facility (IFMIF) and the DEMO Oriented Neutron Source (DONES), in which a considerable amount of neutrons with energies above 20 MeV is produced. The present paper summarizes the main characteristics of the updated version of ANITA-IEAF, able to use decay and cross section data based on more recent evaluated nuclear data libraries, i.e. the JEFF-3.1.1 Radioactive Decay Data Library and the EAF-2010 neutron activation cross section library. In this paper the validation effort related to the comparison between the code predictions and the activity measurements obtained from the Karlsruhe Isochronous Cyclotron is presented. In this integral experiment samples of two different steels, SS-316 and F82H, pure vanadium and a vanadium alloy, structural materials of interest in fusion technology, were activated in a neutron spectrum similar to the IFMIF neutron field.

  20. Phonological, visual, and semantic coding strategies and children's short-term picture memory span

    OpenAIRE

    Henry, L.; Messer, D. J.; Luger-Klein, S.; Crane, L.

    2012-01-01

    Three experiments addressed controversies in the previous literature on the development of phonological and other forms of short-term memory coding in children, using assessments of picture memory span that ruled out potentially confounding effects of verbal input and output. Picture materials were varied in terms of phonological similarity, visual similarity, semantic similarity, and word length. Older children (6/8-year-olds), but not younger children (4/5-year-olds), demonstrated robust an...

  1. Code-switched English pronunciation modeling for Swahili spoken term detection

    CSIR Research Space (South Africa)

    Kleynhans, N

    2016-05-01

    Full Text Available Computer Science 81 ( 2016 ) 128 – 135 5th Workshop on Spoken Language Technology for Under-resourced Languages, SLTU 2016, 9-12 May 2016, Yogyakarta, Indonesia Code-switched English Pronunciation Modeling for Swahili Spoken Term Detection Neil...

  2. pulver: an R package for parallel ultra-rapid p-value computation for linear regression interaction terms.

    Science.gov (United States)

    Molnos, Sophie; Baumbach, Clemens; Wahl, Simone; Müller-Nurasyid, Martina; Strauch, Konstantin; Wang-Sattler, Rui; Waldenberger, Melanie; Meitinger, Thomas; Adamski, Jerzy; Kastenmüller, Gabi; Suhre, Karsten; Peters, Annette; Grallert, Harald; Theis, Fabian J; Gieger, Christian

    2017-09-29

    Genome-wide association studies allow us to understand the genetics of complex diseases. Human metabolism provides information about the disease-causing mechanisms, so it is usual to investigate the associations between genetic variants and metabolite levels. However, only considering genetic variants and their effects on one trait ignores the possible interplay between different "omics" layers. Existing tools only consider single-nucleotide polymorphism (SNP)-SNP interactions, and no practical tool is available for large-scale investigations of the interactions between pairs of arbitrary quantitative variables. We developed an R package called pulver to compute p-values for the interaction term in a very large number of linear regression models. Comparisons based on simulated data showed that pulver is much faster than the existing tools. This is achieved by using the correlation coefficient to test the null-hypothesis, which avoids the costly computation of inversions. Additional tricks are a rearrangement of the order, when iterating through the different "omics" layers, and implementing this algorithm in the fast programming language C++. Furthermore, we applied our algorithm to data from the German KORA study to investigate a real-world problem involving the interplay among DNA methylation, genetic variants, and metabolite levels. The pulver package is a convenient and rapid tool for screening huge numbers of linear regression models for significant interaction terms in arbitrary pairs of quantitative variables. pulver is written in R and C++, and can be downloaded freely from CRAN at https://cran.r-project.org/web/packages/pulver/ .

  3. LISA package user guide. Part II: LISA (Long Term Isolation Safety Assessment) program description and user guide

    International Nuclear Information System (INIS)

    Prado, P.; Saltelli, A.; Homma, T.

    1992-01-01

    This manual is subdivided into three parts. In this second part, this document describes the LISA (Long term Isolation Safety Assessment) Code and its submodels. LISA is a tool for analysis of the safety of an underground disposal of nuclear waste. It has the capability to handle nuclide chain of arbitrary length and to evaluate the migration of nuclide through a geosphere medium composed of an arbitrary number of segments. LISA makes use of Monte Carlo methodology to evaluate the uncertainty in the quantity being assessed (eg dose) arising from the uncertainty in the model input parameters. In the present version LISA is equipped with a very simple source term submodel, a relatively complex geosphere and a simplified biosphere. The code is closely associated with its statistical pre-processor code (PREP), which generates the input Monte Carlo sample from the assigned parameter probability density functions and with its post-processor code (SPOP) which provides useful statistics on the output sample (uncertainty and sensitivity analysis). This report describes the general structure of LISA, its subroutines and submodels, the code input ant output files. It is intended to provide the user with enough information to know and run the code as well as the capacity to incorporate different submodels. 15 refs., 6 figs

  4. Knowledge synthesis on the long time behavior of packages MOP-CLTBE 2004; Synthese des connaissances sur le comportement a long terme des colis MOP-CLTBE 2004

    Energy Technology Data Exchange (ETDEWEB)

    Peycelon, H.; Bejaoui, S.; Richet, C.; Bary, B

    2004-10-15

    This document is realized in the framework of the law of 1991 on the radioactive wastes management. It presents the main evolution factors of the concrete packages for non saturated medium storage and saturated disposal. It takes stock on the operational model describing the package alteration and the consequences in terms of radionuclide releases. (A.L.B.)

  5. Long-Term Stability of NIST Chip-Scale Atomic Clock Physics Packages

    Science.gov (United States)

    2007-01-01

    vacuum packaging), as has been demonstrated by Lutwak et al. [3]. Nevertheless, we tried to investigate the causes for the frequency shifts of...stability,” Optics Express, 13, 1249-1253. [3] R. Lutwak , J. Deng, W. Riley, M. Varghese, J. Leblanc, G. Tepolt, M. Mescher, D. K. Serkland, K. M. Geib...the 1st Annual Multiconference on Electronics and Photonics, 7-11 November 2006, Guanajuato, Mexico, in press. [6] R. Lutwak , P. Vlitas, M

  6. Influence of interconnection on the long-term reliability of UV LED packages

    Science.gov (United States)

    Nieland, S.; Mitrenga, D.; Karolewski, D.; Brodersen, O.; Ortlepp, T.

    2017-02-01

    High power LEDs have conquered the mass market in recent years. Besides the main development focus to achieve higher productivity in the field of visible semiconductor LED processing, the wavelength range is further enhanced by active research and development in the direction of UVA / UVB / UVC. UVB and UVC LEDs are new and promising due to their numerous advantages. UV LEDs emit in a near range of one single emission peak with a width (FWHM) below 15 nm compared to conventional mercury discharge lamps and xenon sources, which show broad spectrums with many emission peaks over a wide range of wavelengths. Furthermore, the UV LED size is in the range of a few hundred microns and offers a high potential of significant system miniaturization. Of course, LED efficiency, lifetime and output power have to be increased [1]. Lifetime limiting issues of UVB/UVC-LED are the very high thermal stress in the chip resulting from the higher forward voltages (6-10 V @ 350 mA), the lower external quantum efficiency, below 10 % (most of the power disappears as heat), and the thermal resistance Rth of conventional LED packages being not able to dissipate these large amounts of heat for spreading. Beside the circuit boards and submounts which should have maximum thermal conductivity, the dimension of contacts as well as the interconnection of UV LED to the submount/package determinates the resolvable amount of heat [2]. In the paper different innovative interconnection techniques for UVC-LED systems will be discussed focused on the optimization of thermal conductivity in consideration of the assembly costs. Results on thermal simulation for the optimal contact dimensions and interconnections will be given. In addition, these theoretical results will be compared with results on electrical characterization as well as IR investigations on real UV LED packages in order to give recommendations for optimal UV LED assembly.

  7. Uncertainties in source term calculations generated by the ORIGEN2 computer code for Hanford Production Reactors

    International Nuclear Information System (INIS)

    Heeb, C.M.

    1991-03-01

    The ORIGEN2 computer code is the primary calculational tool for computing isotopic source terms for the Hanford Environmental Dose Reconstruction (HEDR) Project. The ORIGEN2 code computes the amounts of radionuclides that are created or remain in spent nuclear fuel after neutron irradiation and radioactive decay have occurred as a result of nuclear reactor operation. ORIGEN2 was chosen as the primary code for these calculations because it is widely used and accepted by the nuclear industry, both in the United States and the rest of the world. Its comprehensive library of over 1,600 nuclides includes any possible isotope of interest to the HEDR Project. It is important to evaluate the uncertainties expected from use of ORIGEN2 in the HEDR Project because these uncertainties may have a pivotal impact on the final accuracy and credibility of the results of the project. There are three primary sources of uncertainty in an ORIGEN2 calculation: basic nuclear data uncertainty in neutron cross sections, radioactive decay constants, energy per fission, and fission product yields; calculational uncertainty due to input data; and code uncertainties (i.e., numerical approximations, and neutron spectrum-averaged cross-section values from the code library). 15 refs., 5 figs., 5 tabs

  8. Neural coding of basic reward terms of animal learning theory, game theory, microeconomics and behavioural ecology.

    Science.gov (United States)

    Schultz, Wolfram

    2004-04-01

    Neurons in a small number of brain structures detect rewards and reward-predicting stimuli and are active during the expectation of predictable food and liquid rewards. These neurons code the reward information according to basic terms of various behavioural theories that seek to explain reward-directed learning, approach behaviour and decision-making. The involved brain structures include groups of dopamine neurons, the striatum including the nucleus accumbens, the orbitofrontal cortex and the amygdala. The reward information is fed to brain structures involved in decision-making and organisation of behaviour, such as the dorsolateral prefrontal cortex and possibly the parietal cortex. The neural coding of basic reward terms derived from formal theories puts the neurophysiological investigation of reward mechanisms on firm conceptual grounds and provides neural correlates for the function of rewards in learning, approach behaviour and decision-making.

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

    International Nuclear Information System (INIS)

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

    1992-03-01

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

  10. Short-term retention of pictures and words: evidence for dual coding systems.

    Science.gov (United States)

    Pellegrino, J W; Siegel, A W; Dhawan, M

    1975-03-01

    The recall of picture and word triads was examined in three experiments that manipulated the type of distraction in a Brown-Peterson short-term retention task. In all three experiments recall of pictures was superior to words under auditory distraction conditions. Visual distraction produced high performance levels with both types of stimuli, whereas combined auditory and visual distraction significantly reduced picture recall without further affecting word recall. The results were interpreted in terms of the dual coding hypothesis and indicated that pictures are encoded into separate visual and acoustic processing systems while words are primarily acoustically encoded.

  11. Dissipation terms in one-dimensional hydrodynamic code for ICF target

    International Nuclear Information System (INIS)

    Tamba, Moritake; Niu, Keishiro.

    1985-01-01

    The one-dimensional hydrodynamic code so far used for the simulation of the target improsion in ICF, the artificial viscosity has been employed as the dissipation terms. This artificial viscosity depends on the mesh width of the space using in the simulation and is much large in comparison with the real viscosity. In this paper, it is shown that this artificial viscosity leads to the unreasonable fusion parameters depending on the used mesh width of the space. Several methods to modify the dissipation term are given in this paper. (author)

  12. Nuclear GUI: a Graphical User Interface for 3D discrete ordinates neutral particle transport codes in the doors and BOT3P packages

    International Nuclear Information System (INIS)

    Saintagne, P.W.; Azmy, Y.Y.

    2005-01-01

    A GUI (Graphical User Interface) provides a graphical, interactive and intuitive link between the user and the software. It translates the user'actions into information, e.g; input data that is interpretable by the software. In order to develop an efficient GUI, it is important to master the target computational code. An initial version of a complete GUI for the DOORS and BOT3P packages for solving neutral particle transport problems in 3-dimensional geometry has been completed. This GUI is made of 4 components. The first component GipGui aims at handling cross-sections by mixing microscopic cross-sections from different libraries. The second component TORT-GUI provides the user a simple way to create or modify input files for the TORT codes that is a general purpose neutral transport code able to solve large problems with complex configurations. The third component GGTM-GUI prepares the data describing the problem configuration like the geometrical data, material assignment or key flux positions. The fourth component DTM3-GUI helps the user to visualize TORT results by providing data for a graphics post-processor

  13. The European source term code ESTER - basic ideas and tools for coupling of ATHLET and ESTER

    International Nuclear Information System (INIS)

    Schmidt, F.; Schuch, A.; Hinkelmann, M.

    1993-04-01

    The French software house CISI and IKE of the University of Stuttgart have developed during 1990 and 1991 in the frame of the Shared Cost Action Reactor Safety the informatic structure of the European Source TERm Evaluation System (ESTER). Due to this work tools became available which allow to unify on an European basis both code development and code application in the area of severe core accident research. The behaviour of reactor cores is determined by thermal hydraulic conditions. Therefore for the development of ESTER it was important to investigate how to integrate thermal hydraulic code systems with ESTER applications. This report describes the basic ideas of ESTER and improvements of ESTER tools in view of a possible coupling of the thermal hydraulic code system ATHLET and ESTER. Due to the work performed during this project the ESTER tools became the most modern informatic tools presently available in the area of severe accident research. A sample application is given which demonstrates the use of the new tools. (orig.) [de

  14. Sensitivity analysis and benchmarking of the BLT low-level waste source term code

    International Nuclear Information System (INIS)

    Suen, C.J.; Sullivan, T.M.

    1993-07-01

    To evaluate the source term for low-level waste disposal, a comprehensive model had been developed and incorporated into a computer code, called BLT (Breach-Leach-Transport) Since the release of the original version, many new features and improvements had also been added to the Leach model of the code. This report consists of two different studies based on the new version of the BLT code: (1) a series of verification/sensitivity tests; and (2) benchmarking of the BLT code using field data. Based on the results of the verification/sensitivity tests, the authors concluded that the new version represents a significant improvement and it is capable of providing more realistic simulations of the leaching process. Benchmarking work was carried out to provide a reasonable level of confidence in the model predictions. In this study, the experimentally measured release curves for nitrate, technetium-99 and tritium from the saltstone lysimeters operated by Savannah River Laboratory were used. The model results are observed to be in general agreement with the experimental data, within the acceptable limits of uncertainty

  15. Waste package performance assessment

    International Nuclear Information System (INIS)

    Lester, D.H.

    1981-01-01

    This paper describes work undertaken to assess the life-expectancy and post-failure nuclide release behavior of high-level and waste packages in a geologic repository. The work involved integrating models of individual phenomena (such as heat transfer, corrosion, package deformation, and nuclide transport) and using existing data to make estimates of post-emplacement behavior of waste packages. A package performance assessment code was developed to predict time to package failure in a flooded repository and subsequent transport of nuclides out of the leaking package. The model has been used to evaluate preliminary package designs. The results indicate, that within the limitation of model assumptions and data base, packages lasting a few hundreds of years could be developed. Very long lived packages may be possible but more comprehensive data are needed to confirm this

  16. BOT3P: a mesh generation software package for the transport analysis codes Dort, Tort, Twodant, Threedant and MCNP

    International Nuclear Information System (INIS)

    Orsi, R.

    2003-01-01

    Bot3p consists of a set of standard Fortran 77 language programs that gives the users of the deterministic transport codes Dort and Tort some useful diagnostic tools to prepare and check the geometry of their input data files for both Cartesian and cylindrical geometries including graphical display modules. Bot3p produces at the same time the geometrical and material distribution data for the deterministic transport codes Twodant and Threedant and, only in three-dimensional (3D) Cartesian geometry, for the Monte Carlo Transport Code MCNP. This makes it possible to compare directly for the same geometry the effects stemming from the use of different data libraries and solution approaches on transport analysis results. Through the use of Bot3p, radiation transport problems with complex 3D geometrical structures can be modelled easily, as a relatively small amount of engineer-time is required and refinement is achieved by changing few parameters. This tool is useful for solving very large challenging problems. (author)

  17. The long-term management of information and records on radioactive waste packages in the United Kingdom

    International Nuclear Information System (INIS)

    Upshall, I.R.; Wisbey, S.J.

    2002-01-01

    It is generally accepted that information associated with the creation, conditioning and packaging of radioactive waste must be maintained for a considerable period of time. During the retention period not only must the storage media be preserved, but the data must remain accessible and in a form that can be interpreted with minimum 'processing'. This will ensure, as far as practicable, that the information accurately represents the nature of the waste and its packaging, that the media employed is suitable for long-term storage and that the storage conditions provide a stable environment. Careful consideration must be given to the type and form of the retained information and the threats to its continued integrity. United Kingdom Nirex Limited (Nirex), in association with experts in records media and management, has undertaken a programme of work to consider the range of media currently available, the threats to media integrity and the implications of a general move towards 'electronic' records. The results of this study are being used to develop an information management system strategy, capable of retaining data for all future phases of radioactive waste management. (author)

  18. Harmonic Enhancement in Low Bitrate Audio Coding Using an Efficient Long-Term Predictor

    Directory of Open Access Journals (Sweden)

    Song Jeongook

    2010-01-01

    Full Text Available This paper proposes audio coding using an efficient long-term prediction method to enhance the perceptual quality of audio codecs to speech input signals at low bit-rates. The MPEG-4 AAC-LTP exploited a similar concept, but its improvement was not significant because of small prediction gain due to long prediction lags and aliased components caused by the transformation with a time-domain aliasing cancelation (TDAC technique. The proposed algorithm increases the prediction gain by employing a deharmonizing predictor and a long-term compensation filter. The look-back memory elements are first constructed by applying the de-harmonizing predictor to the input signal, then the prediction residual is encoded and decoded by transform audio coding. Finally, the long-term compensation filter is applied to the updated look-back memory of the decoded prediction residual to obtain synthesized signals. Experimental results show that the proposed algorithm has much lower spectral distortion and higher perceptual quality than conventional approaches especially for harmonic signals, such as voiced speech.

  19. Code Disentanglement: Initial Plan

    Energy Technology Data Exchange (ETDEWEB)

    Wohlbier, John Greaton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kelley, Timothy M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rockefeller, Gabriel M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Calef, Matthew Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-01-27

    The first step to making more ambitious changes in the EAP code base is to disentangle the code into a set of independent, levelized packages. We define a package as a collection of code, most often across a set of files, that provides a defined set of functionality; a package a) can be built and tested as an entity and b) fits within an overall levelization design. Each package contributes one or more libraries, or an application that uses the other libraries. A package set is levelized if the relationships between packages form a directed, acyclic graph and each package uses only packages at lower levels of the diagram (in Fortran this relationship is often describable by the use relationship between modules). Independent packages permit independent- and therefore parallel|development. The packages form separable units for the purposes of development and testing. This is a proven path for enabling finer-grained changes to a complex code.

  20. The European source-term evaluation code ASTEC: status and applications, including CANDU plant applications

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.P.; Giordano, P.; Kissane, M.P.; Montanelli, T.; Schwinges, B.; Ganju, S.; Dickson, L.

    2004-01-01

    Research on light-water reactor severe accidents (SA) is still required in a limited number of areas in order to confirm accident-management plans. Thus, 49 European organizations have linked their SA research in a durable way through SARNET (Severe Accident Research and management NETwork), part of the European 6th Framework Programme. One goal of SARNET is to consolidate the integral code ASTEC (Accident Source Term Evaluation Code, developed by IRSN and GRS) as the European reference tool for safety studies; SARNET efforts include extending the application scope to reactor types other than PWR (including VVER) such as BWR and CANDU. ASTEC is used in IRSN's Probabilistic Safety Analysis level 2 of 900 MWe French PWRs. An earlier version of ASTEC's SOPHAEROS module, including improvements by AECL, is being validated as the Canadian Industry Standard Toolset code for FP-transport analysis in the CANDU Heat Transport System. Work with ASTEC has also been performed by Bhabha Atomic Research Centre, Mumbai, on IPHWR containment thermal hydraulics. (author)

  1. A Long-term Ring Current Measure Created by Using the VMO MANGO Service Package

    Science.gov (United States)

    Bargatze, L. F.; King, T. A.

    2008-12-01

    A set of computational routines called MANGO (Magnetogram Analysis for the Network of Geomagnetic Observatories) is utilized to calculate a new measure of magnetic storm activity for the years 1932 to the near present. The MANGO routines are part of an effort to enhance data services available to users of the Heliophysics VxOs, specifically for the Virtual Magnetospheric Observatory (VMO). The community can utilize MANGO to derive value-added data products and images suitable for publication via the VMO web site. MANGO routines will be demonstrated through their application to study magnetic storms, a field of research that began in 1828 when von Humboldt launched an investigation of observations taken simultaneously from magnetic field stations spread around the Earth. The defining signature of magnetic storms is a worldwide decrease of the horizontal component of the magnetic field caused by fluctuations in the strength of the ring current. In the 1940's, Bartel pushed for deriving an index to measure the strength of magnetic storms. Progress intensified during the International Geophysical Year leading to the definition of the Dst index. The definitive Dst index is calculated at WDC-C2 for Geomagnetism in Kyoto by using a derivation scheme certified by Division V of IAGA. The Dst index time series spans the years 1957 to present with a cadence equal to 1-hr. The new data set we will present is a magnetic storm measure that is similar to the Dst index though it is calculated by using MANGO and a method that differs slightly from the official scheme. The MANGO data service package is based on a set of IDL routines that decompose ground magnetic field observations to isolate secular, diurnal, and disturbance variations of the magnetic field station-by-station. Each MANGO subroutine has been written in modular fashion to allow "plug and play"- style flexibility and each has been designed to account for failure modes and noisy data so that the programs will run to

  2. 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.

  3. 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

  4. New Source Term Model for the RESRAD-OFFSITE Code Version 3

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Charley [Argonne National Lab. (ANL), Argonne, IL (United States); Gnanapragasam, Emmanuel [Argonne National Lab. (ANL), Argonne, IL (United States); Cheng, Jing-Jy [Argonne National Lab. (ANL), Argonne, IL (United States); Kamboj, Sunita [Argonne National Lab. (ANL), Argonne, IL (United States); Chen, Shih-Yew [Argonne National Lab. (ANL), Argonne, IL (United States)

    2013-06-01

    This report documents the new source term model developed and implemented in Version 3 of the RESRAD-OFFSITE code. This new source term model includes: (1) "first order release with transport" option, in which the release of the radionuclide is proportional to the inventory in the primary contamination and the user-specified leach rate is the proportionality constant, (2) "equilibrium desorption release" option, in which the user specifies the distribution coefficient which quantifies the partitioning of the radionuclide between the solid and aqueous phases, and (3) "uniform release" option, in which the radionuclides are released from a constant fraction of the initially contaminated material during each time interval and the user specifies the duration over which the radionuclides are released.

  5. Instrument packages to study long-term sediment transport processes in a shallow bay

    Science.gov (United States)

    Strahle, William J.; Martini, Marinna A.; Davis, Ray E.

    1994-01-01

    Pressure and near-surface and near-bottom measurements of current, temperature, salinity and light transmission were required in Mobile Bay, a 3 m deep estuary on the Gulf of Mexico. This environment presented several obstacles to obtaining long term observations. Boat traffic, soft estuary bottom, heavy biofouling, rapid sample rates and large data storage were overcome by using instrumentation techniques that are applicable to other estuary systems. Nearly two years of continuous data was collected.

  6. Improvement of level-1 PSA computer code package - Modeling and analysis for dynamic reliability of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang Hoon; Baek, Sang Yeup; Shin, In Sup; Moon, Shin Myung; Moon, Jae Phil; Koo, Hoon Young; Kim, Ju Shin [Seoul National University, Seoul (Korea, Republic of); Hong, Jung Sik [Seoul National Polytechnology University, Seoul (Korea, Republic of); Lim, Tae Jin [Soongsil University, Seoul (Korea, Republic of)

    1996-08-01

    The objective of this project is to develop a methodology of the dynamic reliability analysis for NPP. The first year`s research was focused on developing a procedure for analyzing failure data of running components and a simulator for estimating the reliability of series-parallel structures. The second year`s research was concentrated on estimating the lifetime distribution and PM effect of a component from its failure data in various cases, and the lifetime distribution of a system with a particular structure. Computer codes for performing these jobs were also developed. The objectives of the third year`s research is to develop models for analyzing special failure types (CCFs, Standby redundant structure) that were nor considered in the first two years, and to complete a methodology of the dynamic reliability analysis for nuclear power plants. The analysis of failure data of components and related researches for supporting the simulator must be preceded for providing proper input to the simulator. Thus this research is divided into three major parts. 1. Analysis of the time dependent life distribution and the PM effect. 2. Development of a simulator for system reliability analysis. 3. Related researches for supporting the simulator : accelerated simulation analytic approach using PH-type distribution, analysis for dynamic repair effects. 154 refs., 5 tabs., 87 figs. (author)

  7. A Source Term Calculation for the APR1400 NSSS Auxiliary System Components Using the Modified SHIELD Code

    International Nuclear Information System (INIS)

    Park, Hong Sik; Kim, Min; Park, Seong Chan; Seo, Jong Tae; Kim, Eun Kee

    2005-01-01

    The SHIELD code has been used to calculate the source terms of NSSS Auxiliary System (comprising CVCS, SIS, and SCS) components of the OPR1000. Because the code had been developed based upon the SYSTEM80 design and the APR1400 NSSS Auxiliary System design is considerably changed from that of SYSTEM80 or OPR1000, the SHIELD code cannot be used directly for APR1400 radiation design. Thus the hand-calculation is needed for the portion of design changes using the results of the SHIELD code calculation. In this study, the SHIELD code is modified to incorporate the APR1400 design changes and the source term calculation is performed for the APR1400 NSSS Auxiliary System components

  8. London 2012 packaging guidelines

    OpenAIRE

    2013-01-01

    These guidelines are intended to provide supplemental advice to suppliers and licensees regarding the provisions of the LOCOG Sustainable Sourcing Code that relate to packaging design and materials selection.

  9. Trends in Food Packaging.

    Science.gov (United States)

    Ott, Dana B.

    1988-01-01

    This article discusses developments in food packaging, processing, and preservation techniques in terms of packaging materials, technologies, consumer benefits, and current and potential food product applications. Covers implications due to consumer life-style changes, cost-effectiveness of packaging materials, and the ecological impact of…

  10. The coding theorem for a class of quantum channels with long-term memory

    International Nuclear Information System (INIS)

    Datta, Nilanjana; Dorlas, Tony C

    2007-01-01

    In this paper, we consider the transmission of classical information through a class of quantum channels with long-term memory, which are convex combinations of memoryless channels. Hence, the memory of such channels can be considered to be given by a Markov chain which is aperiodic but not irreducible. We prove the coding theorem and weak converse for this class of channels. The main techniques that we employ are a quantum version of Feinstein's fundamental lemma (Feinstein A 1954 IRE Trans. PGIT 4 2-22, Khinchin A I 1957 Mathematical Foundations of Information Theory: II. On the Fundamental Theorems of Information Theory (New York: Dover) chapter IV) and a generalization of Helstrom's theorem (Helstrom C W 1976 Quantum detection and estimation theory Mathematics in Science and Engineering vol 123 (London: Academic))

  11. The ENSDF Java Package

    International Nuclear Information System (INIS)

    Sonzogni, A.A.

    2005-01-01

    A package of computer codes has been developed to process and display nuclear structure and decay data stored in the ENSDF (Evaluated Nuclear Structure Data File) library. The codes were written in an object-oriented fashion using the java language. This allows for an easy implementation across multiple platforms as well as deployment on web pages. The structure of the different java classes that make up the package is discussed as well as several different implementations

  12. Spatial coding of ordinal information in short- and long-term memory

    Directory of Open Access Journals (Sweden)

    Veronique eGinsburg

    2015-01-01

    Full Text Available The processing of numerical information induces a spatial response bias: Faster responses to small numbers with the left hand and faster responses to large numbers with the right hand. Most theories agree that long-term representations underlie this so called SNARC effect (Spatial Numerical Association of Response Codes; Dehaene, Bossini, & Giraux, 1993. However, a spatial response bias was also observed with the activation of temporary position-space associations in working memory (ordinal position effect; van Dijck & Fias, 2011. Items belonging to the beginning of a memorized sequence are responded to faster with the left hand side while items at the end of the sequence are responded to faster with the right hand side. The theoretical possibility was put forward that the SNARC effect is an instance of the ordinal position effect, with the empirical consequence that the SNARC effect and the ordinal position effect cannot be observed simultaneously. In two experiments we falsify this claim by demonstrating that the SNARC effect and the ordinal position effect are not mutually exclusive. Consequently, this suggests that the SNARC effect and the ordinal position effect result from the activation of different representations. We conclude that spatial response biases can result from the activation of both pre-existing positions in long-term memory and from temporary space associations in working memory at the same time.

  13. Spatial coding of ordinal information in short- and long-term memory.

    Science.gov (United States)

    Ginsburg, Véronique; Gevers, Wim

    2015-01-01

    The processing of numerical information induces a spatial response bias: Faster responses to small numbers with the left hand and faster responses to large numbers with the right hand. Most theories agree that long-term representations underlie this so called SNARC effect (Spatial Numerical Association of Response Codes; Dehaene et al., 1993). However, a spatial response bias was also observed with the activation of temporary position-space associations in working memory (ordinal position effect; van Dijck and Fias, 2011). Items belonging to the beginning of a memorized sequence are responded to faster with the left hand side while items at the end of the sequence are responded to faster with the right hand side. The theoretical possibility was put forward that the SNARC effect is an instance of the ordinal position effect, with the empirical consequence that the SNARC effect and the ordinal position effect cannot be observed simultaneously. In two experiments we falsify this claim by demonstrating that the SNARC effect and the ordinal position effect are not mutually exclusive. Consequently, this suggests that the SNARC effect and the ordinal position effect result from the activation of different representations. We conclude that spatial response biases can result from the activation of both pre-existing positions in long-term memory and from temporary space associations in working memory at the same time.

  14. Spatial coding of ordinal information in short- and long-term memory

    Science.gov (United States)

    Ginsburg, Véronique; Gevers, Wim

    2015-01-01

    The processing of numerical information induces a spatial response bias: Faster responses to small numbers with the left hand and faster responses to large numbers with the right hand. Most theories agree that long-term representations underlie this so called SNARC effect (Spatial Numerical Association of Response Codes; Dehaene et al., 1993). However, a spatial response bias was also observed with the activation of temporary position-space associations in working memory (ordinal position effect; van Dijck and Fias, 2011). Items belonging to the beginning of a memorized sequence are responded to faster with the left hand side while items at the end of the sequence are responded to faster with the right hand side. The theoretical possibility was put forward that the SNARC effect is an instance of the ordinal position effect, with the empirical consequence that the SNARC effect and the ordinal position effect cannot be observed simultaneously. In two experiments we falsify this claim by demonstrating that the SNARC effect and the ordinal position effect are not mutually exclusive. Consequently, this suggests that the SNARC effect and the ordinal position effect result from the activation of different representations. We conclude that spatial response biases can result from the activation of both pre-existing positions in long-term memory and from temporary space associations in working memory at the same time. PMID:25688199

  15. Genetic algorithm to optimize two-echelon inventory control system for perishable goods in terms of active packaging

    Directory of Open Access Journals (Sweden)

    Babak Yousefi Yegane

    2012-01-01

    Full Text Available This paper considers an inventory control policy for a two-echelon inventory control system with one supplier-one buyer. We consider the case of deteriorating items which lead to shortage in supply chain. Therefore, it is necessary to decrease the deterioration rate by adding some specification to the packaging of these items that is known as active packaging. Although this packaging can reduce the deteriorating rate of products, but may be increases the cost of both supplier and buyer. Because of the complexity of the mathematical model, a genetic algorithm has been developed to determine the best policy of this inventory control system.

  16. Progress in waste package and engineered barrier system performance assessment and design

    International Nuclear Information System (INIS)

    Van Luik, A.; Stahl, D.; Harrison, D.

    1993-01-01

    As part of the U.S. Department of Energy's evaluation of site suitability for a potential high-level radioactive waste repository, long-term interactions between the engineered barrier system and the site must be determined. This requires a waste-package/engineered-system design, a description of the environment around the emplacement zone, and models that simulate operative processes describing these engineered/natural systems interactions. Candidate designs are being evaluated, including a more robust, multi-barrier waste package, and a drift emplacement mode. Tools for evaluating designs, and emplacement mode are the currently available waste-package/engineered-system performance assessment codes development for the project. For assessments that support site suitability, environmental impact, or licensing decisions, more capable codes are needed. Code capability requirements are being written, and existing codes are to be evaluated against those requirements. Recommendations are being made to focus waste-packaging/engineered-system code-development

  17. Uncertainty analysis methods for quantification of source terms using a large computer code

    International Nuclear Information System (INIS)

    Han, Seok Jung

    1997-02-01

    Quantification of uncertainties in the source term estimations by a large computer code, such as MELCOR and MAAP, is an essential process of the current probabilistic safety assessments (PSAs). The main objectives of the present study are (1) to investigate the applicability of a combined procedure of the response surface method (RSM) based on input determined from a statistical design and the Latin hypercube sampling (LHS) technique for the uncertainty analysis of CsI release fractions under a hypothetical severe accident sequence of a station blackout at Young-Gwang nuclear power plant using MAAP3.0B code as a benchmark problem; and (2) to propose a new measure of uncertainty importance based on the distributional sensitivity analysis. On the basis of the results obtained in the present work, the RSM is recommended to be used as a principal tool for an overall uncertainty analysis in source term quantifications, while using the LHS in the calculations of standardized regression coefficients (SRC) and standardized rank regression coefficients (SRRC) to determine the subset of the most important input parameters in the final screening step and to check the cumulative distribution functions (cdfs) obtained by RSM. Verification of the response surface model for its sufficient accuracy is a prerequisite for the reliability of the final results obtained by the combined procedure proposed in the present work. In the present study a new measure has been developed to utilize the metric distance obtained from cumulative distribution functions (cdfs). The measure has been evaluated for three different cases of distributions in order to assess the characteristics of the measure: The first case and the second are when the distribution is known as analytical distributions and the other case is when the distribution is unknown. The first case is given by symmetry analytical distributions. The second case consists of two asymmetry distributions of which the skewness is non zero

  18. Does packaging with a calendar feature improve adherence to self-administered medication for long-term use? A systematic review.

    Science.gov (United States)

    Zedler, Barbara K; Kakad, Priyanka; Colilla, Susan; Murrelle, Lenn; Shah, Nirav R

    2011-01-01

    The therapeutic benefit of self-administered medications for long-term use is limited by an average 50% nonadherence rate. Patient forgetfulness is a common factor in unintentional nonadherence. Unit-of-use packaging that incorporates a simple day-and-date feature (calendar packaging) is designed to improve adherence by prompting patients to maintain the prescribed dosing schedule. To review systematically, in accordance with the Preferred Reporting Items for Systematic Reviews and Meta-Analyses (PRISMA) statement, randomized controlled trial evidence of the adherence benefits and harms of calendar blister packaging (CBP) and calendar pill organizers (CPO) for self-administered, long-term medication use. Data sources included the MEDLINE and Web of Science and Cochrane Library databases from their inception to September 2010 and communication with researchers in the field. Key search terms included blister-calendar pack, blister pack, drug packaging, medication adherence, medication compliance, medication compliance devices, medication containers, medication organizers, multicompartment compliance aid, persistence, pill-box organizers, prescription refill, randomized controlled trials, and refill compliance. Selected studies had an English-language title; a randomized controlled design; medication packaged in CBP or CPO; a requirement of solid, oral medication self-administered daily for longer than 1 month in community-dwelling adults; and at least 1 quantitative outcome measure of adherence. Two reviewers extracted data independently on study design, sample size, type of intervention and control, and outcomes. Ten trials with a total of 1045 subjects met the inclusion criteria, and 9 also examined clinical outcomes (seizures, blood pressure, psychiatric symptoms) or health care resource utilization. Substantial heterogeneity among trials precluded meta-analysis. In 3 studies, calendar packaging was part of a multicomponent adherence intervention. Six of 10 trials

  19. Long term behaviour of low and intermediate level waste packages under repository conditions. Results of a co-ordinated research project 1997-2002

    International Nuclear Information System (INIS)

    2004-06-01

    The development and application of approaches and technologies that provide long term safety is an essential issue in the disposal of radioactive waste. For low and intermediate level radioactive waste, engineered barriers play an important role in the overall safety and performance of near surface repositories. Thus, developing a strong technical basis for understanding the behaviour and performance of engineered barriers is an important consideration in the development and establishment of near surface repositories for radioactive waste. In 1993, a Co-ordinated Research Project (CRP) on Performance of Engineered Barrier Materials in Near Surface Disposal Facilities for Radioactive Waste was initiated by the IAEA with the twin goals of addressing some of the gaps in the database on radionuclide isolation and long term performance of a wide variety of materials and components that constitute the engineered barriers system (IAEA-TECDOC-1255 (2001)). However, during the course of the CRP, it was realized that that the scope of the CRP did not include studies of the behaviour of waste packages over time. Given that a waste package represents an important component of the overall near surface disposal system and the fact that many Member States have active R and D programmes related to waste package testing and evaluation, a new CRP was launched, in 1997, on Long Term Behaviour of Low and Intermediate Level Waste Packages Under Repository Conditions. The CRP was intended to promote research activities on the subject area in Member States, share information on the topic among the participating countries, and contribute to advancing technologies for near surface disposal of radioactive waste. Thus, this CRP complements the afore mentioned CRP on studies of engineered barriers. With the active participation and valuable contributions from twenty scientists and engineers from Argentina, Canada, Czech Republic, Egypt, Finland, India, Republic of Korea, Norway, Romania

  20. Utilizing an Adaptive Grey Model for Short-Term Time Series Forecasting: A Case Study of Wafer-Level Packaging

    Directory of Open Access Journals (Sweden)

    Che-Jung Chang

    2013-01-01

    Full Text Available The wafer-level packaging process is an important technology used in semiconductor manufacturing, and how to effectively control this manufacturing system is thus an important issue for packaging firms. One way to aid in this process is to use a forecasting tool. However, the number of observations collected in the early stages of this process is usually too few to use with traditional forecasting techniques, and thus inaccurate results are obtained. One potential solution to this problem is the use of grey system theory, with its feature of small dataset modeling. This study thus uses the AGM(1,1 grey model to solve the problem of forecasting in the pilot run stage of the packaging process. The experimental results show that the grey approach is an appropriate and effective forecasting tool for use with small datasets and that it can be applied to improve the wafer-level packaging process.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-15

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

  2. Design analysis of various transportation package options for BN-350 SNF in terms of nuclear radiation safety in planning for long-terms dry storage

    International Nuclear Information System (INIS)

    Aisabekov, A.Z.; Mukenova, S.A.; Tur, E.S.; Tsyngaev, V.M.

    2004-01-01

    Full text: This effort is performed under the BN-350 reactor facility decommissioning project. One of the project tasks - spent nuclear fuel handling - includes the following: fuel packaging into sealed canisters, transportation of the canisters in multi-seat metallo-concrete containers and placement of the containers for a long-term dry storage. The goal of this effort is to computationally validate nuclear and radiation safety of the SNF containers placed for storage both under normal storage conditions and probable accident situations. The basic unit structure and design configurations are presented: assemblies, canisters, transportation containers. The major factors influencing nuclear and radiation safety are presented: fuel burn-up, enrichment, fabrication tolerance, types of fuel assemblies, configuration of assemblies in the canister and canisters in the container, background of assemblies placed in the reactor and cooling pool. Conditions under which the SNF containers will be stored are described and probable accident situations are listed. Proceeding from the conservatism principle, selection of the assemblies posing the greatest nuclear hazard is validated. A neutron effective multiplication factor is calculated for the SNF containers under the normal storage conditions and for the case of emergency. The effective multiplication factor is shown to be within a standard value of 0.95 in any situation. Based on the experimental data on assembly and canister dose rates, canisters posing the highest radiation threat are selected. Activities of sources and gamma-radiation spectral composition are calculated. Distribution of the dose rate outside the containers both under the normal storage conditions and accident situations are calculated. The results obtained are analyzed

  3. Long-Term Waste Package Degradation Studies at the Yucca Mountain Potential High-Level Nuclear Waste Repository

    International Nuclear Information System (INIS)

    Mon, K. G.; Bullard, B. E.; Longsine, D. E.; Mehta, S.; Lee, J. H.; Monib, A. M.

    2002-01-01

    The Site Recommendation (SR) process for the potential repository for spent nuclear fuel (SNF) and high-level nuclear waste (HLW) at Yucca Mountain, Nevada is underway. Fulfillment of the requirements for substantially complete containment of the radioactive waste emplaced in the potential repository and subsequent slow release of radionuclides from the Engineered Barrier System (EBS) into the geosphere will rely on a robust waste container design, among other EBS components. Part of the SR process involves sensitivity studies aimed at elucidating which model parameters contribute most to the drip shield and waste package degradation characteristics. The model parameters identified included (a) general corrosion rate model parameters (temperature-dependence and uncertainty treatment), and (b) stress corrosion cracking (SCC) model parameters (uncertainty treatment of stress and stress intensity factor profiles in the Alloy 22 waste package outer barrier closure weld regions, the SCC initiation stress threshold, and the fraction of manufacturing flaws oriented favorably for through-wall penetration by SCC). These model parameters were reevaluated and new distributions were generated. Also, early waste package failures due to improper heat treatment were added to the waste package degradation model. The results of these investigations indicate that the waste package failure profiles are governed by the manufacturing flaw orientation model parameters and models used

  4. 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

  5. Use of WIMS-E lattice code for prediction of the transuranic source term for spent fuel dose estimation

    International Nuclear Information System (INIS)

    Schwinkendorf, K.N.

    1996-01-01

    A recent source term analysis has shown a discrepancy between ORIGEN2 transuranic isotopic production estimates and those produced with the WIMS-E lattice physics code. Excellent agreement between relevant experimental measurements and WIMS-E was shown, thus exposing an error in the cross section library used by ORIGEN2

  6. Software package as an information center product

    International Nuclear Information System (INIS)

    Butler, M.K.

    1977-01-01

    The Argonne Code Center serves as a software exchange and information center for the U.S. Energy Research and Development Administration and the Nuclear Regulatory Commission. The goal of the Center's program is to provide a means for sharing of software among agency offices and contractors, and for transferring computing applications and technology, developed within the agencies, to the information-processing community. A major activity of the Code Center is the acquisition, review, testing, and maintenance of a collection of software--computer systems, applications programs, subroutines, modules, and data compilations--prepared by agency offices and contractors to meet programmatic needs. A brief review of the history of computer program libraries and software sharing is presented to place the Code Center activity in perspective. The state-of-the-art discussion starts off with an appropriate definition of the term software package, together with descriptions of recommended package contents and the Carter's package evaluation activity. An effort is made to identify the various users of the product, to enumerate their individual needs, to document the Center's efforts to meet these needs and the ongoing interaction with the user community. Desirable staff qualifications are considered, and packaging problems, reviewed. The paper closes with a brief look at recent developments and a forecast of things to come. 2 tables

  7. Packaging fluency

    DEFF Research Database (Denmark)

    Mocanu, Ana; Chrysochou, Polymeros; Bogomolova, Svetlana

    2011-01-01

    Research on packaging stresses the need for packaging design to read easily, presuming fast and accurate processing of product-related information. In this paper we define this property of packaging as “packaging fluency”. Based on the existing marketing and cognitive psychology literature on pac...

  8. 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)

  9. 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)

  10. EMOS: Program package for the long-term safety analysis of a repository for radioactive waste, version 4

    International Nuclear Information System (INIS)

    Storck, R.; Buhmann, D.; Hirsekorn, R.P.; Nies, A.; Rausch, H.

    1990-09-01

    Model calculations were performed with the EMOS/REPOS code describing the release of radionuclides from a salt formation, and radionuclide transport in the cap rock was computed separately for one-dimensional or multidimensional dispersion using the SWIFT code. The EMOS code, however, is basically designed so as to allow both the calculation of radionuclide release from a mine opening in a salt formation with the REPOS code, and the subsequent disperson processes in the cap rock and the resulting radiation exposure in the biosphere, with the help of several other modules. These modules have been developed and implemented. Furthermore, the EMOS code was provided with a Monte-Carlo frame, in order to take into account the range of variation of input data, referring to the mine opening, the cap rock, and the biosphere. (orig.) [de

  11. Verbal Short-Term Memory Span in Speech-Disordered Children: Implications for Articulatory Coding in Short-Term Memory.

    Science.gov (United States)

    Raine, Adrian; And Others

    1991-01-01

    Children with speech disorders had lower short-term memory capacity and smaller word length effect than control children. Children with speech disorders also had reduced speech-motor activity during rehearsal. Results suggest that speech rate may be a causal determinant of verbal short-term memory capacity. (BC)

  12. Microelectronic packaging

    CERN Document Server

    Datta, M; Schultze, J Walter

    2004-01-01

    Microelectronic Packaging analyzes the massive impact of electrochemical technologies on various levels of microelectronic packaging. Traditionally, interconnections within a chip were considered outside the realm of packaging technologies, but this book emphasizes the importance of chip wiring as a key aspect of microelectronic packaging, and focuses on electrochemical processing as an enabler of advanced chip metallization.Divided into five parts, the book begins by outlining the basics of electrochemical processing, defining the microelectronic packaging hierarchy, and emphasizing the impac

  13. Analysis of source term aspects in the experiment Phebus FPT1 with the MELCOR and CFX codes

    Energy Technology Data Exchange (ETDEWEB)

    Martin-Fuertes, F. [Universidad Politecnica de Madrid, UPM, Nuclear Engineering Department, Jose Gutierrez Abascal 2, 28006 Madrid (Spain)]. E-mail: francisco.martinfuertes@upm.es; Barbero, R. [Universidad Politecnica de Madrid, UPM, Nuclear Engineering Department, Jose Gutierrez Abascal 2, 28006 Madrid (Spain); Martin-Valdepenas, J.M. [Universidad Politecnica de Madrid, UPM, Nuclear Engineering Department, Jose Gutierrez Abascal 2, 28006 Madrid (Spain); Jimenez, M.A. [Universidad Politecnica de Madrid, UPM, Nuclear Engineering Department, Jose Gutierrez Abascal 2, 28006 Madrid (Spain)

    2007-03-15

    Several aspects related to the source term in the Phebus FPT1 experiment have been analyzed with the help of MELCOR 1.8.5 and CFX 5.7 codes. Integral aspects covering circuit thermalhydraulics, fission product and structural material release, vapours and aerosol retention in the circuit and containment were studied with MELCOR, and the strong and weak points after comparison to experimental results are stated. Then, sensitivity calculations dealing with chemical speciation upon release, vertical line aerosol deposition and steam generator aerosol deposition were performed. Finally, detailed calculations concerning aerosol deposition in the steam generator tube are presented. They were obtained by means of an in-house code application, named COCOA, as well as with CFX computational fluid dynamics code, in which several models for aerosol deposition were implemented and tested, while the models themselves are discussed.

  14. BLT [Breach, Leach, and Transport]: A source term computer code for low-level waste shallow land burial

    International Nuclear Information System (INIS)

    Suen, C.J.; Sullivan, T.M.

    1990-01-01

    This paper discusses the development of a source term model for low-level waste shallow land burial facilities and separates the problem into four individual compartments. These are water flow, corrosion and subsequent breaching of containers, leaching of the waste forms, and solute transport. For the first and the last compartments, we adopted the existing codes, FEMWATER and FEMWASTE, respectively. We wrote two new modules for the other two compartments in the form of two separate Fortran subroutines -- BREACH and LEACH. They were incorporated into a modified version of the transport code FEMWASTE. The resultant code, which contains all three modules of container breaching, waste form leaching, and solute transport, was renamed BLT (for Breach, Leach, and Transport). This paper summarizes the overall program structure and logistics, and presents two examples from the results of verification and sensitivity tests. 6 refs., 7 figs., 1 tab

  15. Glossary of terms recommended for use in radiation control legislation and associated codes of practice

    International Nuclear Information System (INIS)

    Swindon, T.; Kennedy, K.N.; Elliott, G.

    1983-06-01

    A glossary of terms associated with radiological health is presented. It includes basic physics terms, quantities and units, radiological physics terms and radiation protection terms. The National Health and Medical Research Council has recommended that the glossary be used by organizations producing official documents and by appropriate regulatory authorities

  16. Code bench-marking for long-term tracking and adaptive algorithms

    OpenAIRE

    Schmidt, Frank; Alexahin, Yuri; Amundson, James; Bartosik, Hannes; Franchetti, Giuliano; Holmes, Jeffrey; Huschauer, Alexander; Kapin, Valery; Oeftiger, Adrian; Stern, Eric; Titze, Malte

    2016-01-01

    At CERN we have ramped up a program to investigate space charge effects in the LHC pre-injectors with high brightness beams and long storage times. This in view of the LIU upgrade project for these accelerators. These studies require massive simulation over large number of turns. To this end we have been looking at all available codes and started collaborations on code development with several laboratories: pyORBIT from SNS, SYNERGIA from Fermilab, MICROMAP from GSI and our in-house MAD-X cod...

  17. A modified carrier-to-code leveling method for retrieving ionospheric observables and detecting short-term temporal variability of receiver differential code biases

    Science.gov (United States)

    Zhang, Baocheng; Teunissen, Peter J. G.; Yuan, Yunbin; Zhang, Xiao; Li, Min

    2018-03-01

    Sensing the ionosphere with the global positioning system involves two sequential tasks, namely the ionospheric observable retrieval and the ionospheric parameter estimation. A prominent source of error has long been identified as short-term variability in receiver differential code bias (rDCB). We modify the carrier-to-code leveling (CCL), a method commonly used to accomplish the first task, through assuming rDCB to be unlinked in time. Aside from the ionospheric observables, which are affected by, among others, the rDCB at one reference epoch, the Modified CCL (MCCL) can also provide the rDCB offsets with respect to the reference epoch as by-products. Two consequences arise. First, MCCL is capable of excluding the effects of time-varying rDCB from the ionospheric observables, which, in turn, improves the quality of ionospheric parameters of interest. Second, MCCL has significant potential as a means to detect between-epoch fluctuations experienced by rDCB of a single receiver.

  18. Study of the source term of radiation of the CDTN GE-PET trace 8 cyclotron with the MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Benavente C, J. A.; Lacerda, M. A. S.; Fonseca, T. C. F.; Da Silva, T. A. [Centro de Desenvolvimento da Tecnologia Nuclear / CNEN, Av. Pte. Antonio Carlos 6627, 31270-901 Belo Horizonte, Minas Gerais (Brazil); Vega C, H. R., E-mail: jhonnybenavente@gmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

    2015-10-15

    Full text: The knowledge of the neutron spectra in a PET cyclotron is important for the optimization of radiation protection of the workers and individuals of the public. The main objective of this work is to study the source term of radiation of the GE-PET trace 8 cyclotron of the Development Center of Nuclear Technology (CDTN/CNEN) using computer simulation by the Monte Carlo method. The MCNPX version 2.7 code was used to calculate the flux of neutrons produced from the interaction of the primary proton beam with the target body and other cyclotron components, during 18F production. The estimate of the source term and the corresponding radiation field was performed from the bombardment of a H{sub 2}{sup 18}O target with protons of 75 μA current and 16.5 MeV of energy. The values of the simulated fluxes were compared with those reported by the accelerator manufacturer (GE Health care Company). Results showed that the fluxes estimated with the MCNPX codes were about 70% lower than the reported by the manufacturer. The mean energies of the neutrons were also different of that reported by GE Health Care. It is recommended to investigate other cross sections data and the use of physical models of the code itself for a complete characterization of the source term of radiation. (Author)

  19. Opening the gift: social inclusion, professional codes and gift-giving in long-term mental healthcare.

    Science.gov (United States)

    Ootes, S T C; Pols, A J; Tonkens, E H; Willems, D L

    2013-03-01

    Deinstitutionalisation has not only made the social inclusion of clients a key objective in long-term mental healthcare, it may also affect the role of the care professional. This article investigates whether the social inclusion objective clashes with other long-standing professional values, specifically when clients give gifts to care professionals. In making a typology of gifts, we compare the literature on gift-giving with professional codes for gifts and relate both to the objective of social inclusion of clients. Our typology draws on an analysis of ethnographic fieldwork carried out in 2007/2008 at a Dutch mental healthcare centre. We identify four types of gifts for professionals in long-term mental healthcare, each relating individually to professional codes and the objective of social inclusion of clients. Only the 'personal gift' directly supports social inclusion, by fostering personal relationships between professionals and clients. Acceptance of this type of gift is advocated only for long-term care professionals. We suggest that professional codes need to consider this typology of gifts, and we advocate promoting reflexivity as a means of accounting for professional behaviour in deinstitutionalised care settings.

  20. WASTE PACKAGE TRANSPORTER DESIGN

    International Nuclear Information System (INIS)

    Weddle, D.C.; Novotny, R.; Cron, J.

    1998-01-01

    The purpose of this Design Analysis is to develop preliminary design of the waste package transporter used for waste package (WP) transport and related functions in the subsurface repository. This analysis refines the conceptual design that was started in Phase I of the Viability Assessment. This analysis supports the development of a reliable emplacement concept and a retrieval concept for license application design. The scope of this analysis includes the following activities: (1) Assess features of the transporter design and evaluate alternative design solutions for mechanical components. (2) Develop mechanical equipment details for the transporter. (3) Prepare a preliminary structural evaluation for the transporter. (4) Identify and recommend the equipment design for waste package transport and related functions. (5) Investigate transport equipment interface tolerances. This analysis supports the development of the waste package transporter for the transport, emplacement, and retrieval of packaged radioactive waste forms in the subsurface repository. Once the waste containers are closed and accepted, the packaged radioactive waste forms are termed waste packages (WP). This terminology was finalized as this analysis neared completion; therefore, the term disposal container is used in several references (i.e., the System Description Document (SDD)) (Ref. 5.6). In this analysis and the applicable reference documents, the term ''disposal container'' is synonymous with ''waste package''

  1. WASTE PACKAGE TRANSPORTER DESIGN

    Energy Technology Data Exchange (ETDEWEB)

    D.C. Weddle; R. Novotny; J. Cron

    1998-09-23

    The purpose of this Design Analysis is to develop preliminary design of the waste package transporter used for waste package (WP) transport and related functions in the subsurface repository. This analysis refines the conceptual design that was started in Phase I of the Viability Assessment. This analysis supports the development of a reliable emplacement concept and a retrieval concept for license application design. The scope of this analysis includes the following activities: (1) Assess features of the transporter design and evaluate alternative design solutions for mechanical components. (2) Develop mechanical equipment details for the transporter. (3) Prepare a preliminary structural evaluation for the transporter. (4) Identify and recommend the equipment design for waste package transport and related functions. (5) Investigate transport equipment interface tolerances. This analysis supports the development of the waste package transporter for the transport, emplacement, and retrieval of packaged radioactive waste forms in the subsurface repository. Once the waste containers are closed and accepted, the packaged radioactive waste forms are termed waste packages (WP). This terminology was finalized as this analysis neared completion; therefore, the term disposal container is used in several references (i.e., the System Description Document (SDD)) (Ref. 5.6). In this analysis and the applicable reference documents, the term ''disposal container'' is synonymous with ''waste package''.

  2. An Object-Oriented Serial DSMC Simulation Package

    Science.gov (United States)

    Liu, Hongli; Cai, Chunpei

    2011-05-01

    A newly developed three-dimensional direct simulation Monte Carlo (DSMC) simulation package, named GRASP ("Generalized Rarefied gAs Simulation Package"), is reported in this paper. This package utilizes the concept of simulation engine, many C++ features and software design patterns. The package has an open architecture which can benefit further development and maintenance of the code. In order to reduce the engineering time for three-dimensional models, a hybrid grid scheme, combined with a flexible data structure compiled by C++ language, are implemented in this package. This scheme utilizes a local data structure based on the computational cell to achieve high performance on workstation processors. This data structure allows the DSMC algorithm to be very efficiently parallelized with domain decomposition and it provides much flexibility in terms of grid types. This package can utilize traditional structured, unstructured or hybrid grids within the framework of a single code to model arbitrarily complex geometries and to simulate rarefied gas flows. Benchmark test cases indicate that this package has satisfactory accuracy for complex rarefied gas flows.

  3. PAINeT: An object-oriented software package for simulations of flow-field, transport coefficients and flux terms in non-equilibrium gas mixture flows

    Science.gov (United States)

    Istomin, V. A.

    2018-05-01

    The software package Planet Atmosphere Investigator of Non-equilibrium Thermodynamics (PAINeT) has been devel-oped for studying the non-equilibrium effects associated with electronic excitation, chemical reactions and ionization. These studies are necessary for modeling process in shock tubes, in high enthalpy flows, in nozzles or jet engines, in combustion and explosion processes, in modern plasma-chemical and laser technologies. The advantages and possibilities of the package implementation are stated. Within the framework of the package implementation, based on kinetic theory approximations (one-temperature and state-to-state approaches), calculations are carried out, and the limits of applicability of a simplified description of shock-heated air flows and any other mixtures chosen by the user are given. Using kinetic theory algorithms, a numerical calculation of the heat fluxes and relaxation terms can be performed, which is necessary for further comparison of engineering simulation with experi-mental data. The influence of state-to-state distributions over electronic energy levels on the coefficients of thermal conductivity, diffusion, heat fluxes and diffusion velocities of the components of various gas mixtures behind shock waves is studied. Using the software package the accuracy of different approximations of the kinetic theory of gases is estimated. As an example state-resolved atomic ionized mixture of N/N+/O/O+/e- is considered. It is shown that state-resolved diffusion coefficients of neutral and ionized species vary from level to level. Comparing results of engineering applications with those given by PAINeT, recommendations for adequate models selection are proposed.

  4. MMAPDNG: A new, fast code backed by a memory-mapped database for simulating delayed γ-ray emission with MCNPX package

    Science.gov (United States)

    Lou, Tak Pui; Ludewigt, Bernhard

    2015-09-01

    The simulation of the emission of beta-delayed gamma rays following nuclear fission and the calculation of time-dependent energy spectra is a computational challenge. The widely used radiation transport code MCNPX includes a delayed gamma-ray routine that is inefficient and not suitable for simulating complex problems. This paper describes the code "MMAPDNG" (Memory-Mapped Delayed Neutron and Gamma), an optimized delayed gamma module written in C, discusses usage and merits of the code, and presents results. The approach is based on storing required Fission Product Yield (FPY) data, decay data, and delayed particle data in a memory-mapped file. When compared to the original delayed gamma-ray code in MCNPX, memory utilization is reduced by two orders of magnitude and the ray sampling is sped up by three orders of magnitude. Other delayed particles such as neutrons and electrons can be implemented in future versions of MMAPDNG code using its existing framework.

  5. R2SM: a package for the analytic computation of the R2 Rational terms in the Standard Model of the Electroweak interactions

    International Nuclear Information System (INIS)

    Garzelli, M.V.

    2011-01-01

    The analytical package written in FORM presented in this paper allows the computation of the complete set of Feynman Rules producing the Rational terms of kind R 2 contributing to the virtual part of NLO corrections in the Standard Model of the Electroweak interactions. Building block topologies filled by means of generic scalars, vectors and fermions, allowing to build these Feynman Rules in terms of specific elementary particles, are explicitly given in the R ξ gauge class, together with the automatic dressing procedure to obtain the Feynman Rules from them. The results in more specific gauges, like the 't Hooft Feynman one, follow as particular cases, in both the HV and the FDH dimensional regularization schemes. As a check on our formulas, the gauge independence of the total Rational contribution (R 1 +R 2 ) to renormalized S-matrix elements is verified by considering the specific example of the H →γγ decay process at 1-loop. This package can be of interest for people aiming at a better understanding of the nature of the Rational terms. It is organized in a modular way, allowing a further use of some its files even in different contexts. Furthermore, it can be considered as a first seed in the effort towards a complete automation of the process of the analytical calculation of the R 2 effective vertices, given the Lagrangian of a generic gauge theory of particle interactions. (orig.)

  6. MEMS packaging

    CERN Document Server

    Hsu , Tai-Ran

    2004-01-01

    MEMS Packaging discusses the prevalent practices and enabling techniques in assembly, packaging and testing of microelectromechanical systems (MEMS). The entire spectrum of assembly, packaging and testing of MEMS and microsystems, from essential enabling technologies to applications in key industries of life sciences, telecommunications and aerospace engineering is covered. Other topics included are bonding and sealing of microcomponents, process flow of MEMS and microsystems packaging, automated microassembly, and testing and design for testing.The Institution of Engineering and Technology is

  7. Long-term behaviour of concrete: development of operational model to predict the evolution of its containment performance. Application to cemented waste packages

    International Nuclear Information System (INIS)

    Peycelon, H.; Le Bescop, P.; Richet, C.; Adenot, F.

    2001-01-01

    In order to describe the main phenomena during different stages of cement waste packages life-time and to predict the long-term behaviour (containment performance) of concrete, coupled experiments and modelling studies are achieved. With respect to logical methodology, improvement of these studies is accomplished. Degradation of concrete in low mineralized, carbonated and sulfated water lead to an evolution of chemical characteristics (dissolution/precipitation of solid phases) and of transport properties which must be included or coupled in retention/transport modelling of radio nuclides to predict containment performance. (author)

  8. 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)

  9. Current status and future direction of the MONK software package

    International Nuclear Information System (INIS)

    Smith, Nigel; Armishaw, Malcolm; Cooper, Andrew

    2003-01-01

    The current status of the MONK criticality software package is summarized in terms of recent and current developments and envisaged directions for the future. The areas of the discussion are physics modeling, geometry modeling, source modeling, nuclear data, validation, supporting tools and customer services. In future development plan, MONK continues to be focused on meeting the short and long-term needs of the code user community. (J.P.N.)

  10. Application of modified version of SPPS-1 - HEXAB-2DB computer code package for operational analyses of fuel behaviour in WWER-440 reactors at Kozloduy NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kharalampieva, Ts; Stoyanova, I; Antonov, A; Simeonov, T [Kombinat Atomna Energetika, Kozloduj (Bulgaria); Petkov, P [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

    1994-12-31

    The modified version of SPPS-1 code called SPPS-1-HEXAB-2DB was applied for the purposes of the operational analysis and power peaking factors and reactor core critical parameters predictions of WWER-440s. The results of the calculations performed by the use of SPPS-1-HEXAB-2DB code and the corresponding parameters obtained from experiments at Kozloduy NPP WWER-440s as well as the results of fuel rod power distribution are presented. The method of operation simulation of reactor core with 349 assemblies (Unit 4) and with 313 fuel assemblies and 36 dummy fuel assemblies (Unit 1) is outlined. The modified code calculates not only fuel burnup and Pm-149 and Sm-149 concentrations distributions but also the space distribution of I-135 and Xe-135 concentrations. In this way it makes possible to perform the reactor operation simulation during the immediate periods after the reactor start-up or shut-down and to predict the critical reactor core parameters during transients. The results obtained show that SPPS-1-HEXAB-2DB code describes adequately the reactor core status. The new SPPS-1 code algorithm for estimation of assembly-wise power peaking factors distribution in reactor core is also described. The new code provides an option for checking the correctness of reactor core symmetry. The experience from the use of the modified SPPS-1-HEXAB-2DB code system confirms the provision of improved availability of operational analysis, prediction of Kozloduy NPP WWER-440s safe operations and fuel behaviour estimation. 14 tabs., 4 figs., 5 refs.

  11. A prescription and fast code for the long-term evolution of star clusters

    Science.gov (United States)

    Alexander, Poul E. R.; Gieles, Mark

    2012-06-01

    We introduce the star cluster evolution code Evolve Me A Cluster of StarS (EMACSS), a simple yet physically motivated computational model that describes the evolution of some fundamental properties of star clusters in static tidal fields. We base our prescription upon the flow of energy within the cluster, which is a constant fraction of the total energy per half-mass relaxation time. According to Hénon's predictions, this flow is independent of the precise mechanisms for energy production within the core, and therefore does not require a complete description of the many-body interactions therein. For a cluster of equal-mass stars, we use dynamical theory and analytic descriptions of escape mechanisms to construct a series of coupled differential equations expressing the evolution of cluster mass and radius. These equations are numerically solved using a fourth-order Runge-Kutta integration kernel, and the results benchmarked against a data base of direct N-body simulations. We use simulations containing a modest initial number of stars (1024 ≤N≤ 65 536) and point-mass tidal fields of various strengths. Our prescription is publicly available and reproduces the N-body results to within ˜10 per cent accuracy for the entire post-collapse evolution of star clusters.

  12. Packaging microservices

    DEFF Research Database (Denmark)

    Montesi, Fabrizio; Thrane, Dan Sebastian

    2017-01-01

    We describe a first proposal for a new packaging system for microservices based on the Jolie programming language, called the Jolie Package Manager (JPM). Its main features revolve around service interfaces, which make the functionalities that a service provides and depends on explicit. For the f......We describe a first proposal for a new packaging system for microservices based on the Jolie programming language, called the Jolie Package Manager (JPM). Its main features revolve around service interfaces, which make the functionalities that a service provides and depends on explicit...

  13. PROBCON-HDW: A probability and consequence system of codes for long-term analysis of Hanford defense wastes

    International Nuclear Information System (INIS)

    Piepho, M.G.; Nguyen, T.H.

    1988-12-01

    The PROBCON-HDW (PROBability and CONsequence analysis for Hanford defense waste) computer code system calculates the long-term cumulative releases of radionuclides from the Hanford defense wastes (HDW) to the accessible environment and compares the releases to environmental release limits as defined in 40 CFR 191. PROBCON-HDW takes into account the variability of input parameter values used in models to calculate HDW release and transport in the vadose zone to the accessible environment (taken here as groundwater). A human intrusion scenario, which consists of drilling boreholes into the waste beneath the waste sites and bringing waste to the surface, is also included in PROBCON-HDW. PROBCON-HDW also includes the capability to combine the cumulative releases according to various long-term (10,000 year) scenarios into a composite risk curve or complementary cumulative distribution function (CCDF). The system structure of the PROBCON-HDW codes, the mathematical models in PROBCON-HDW, the input files, the input formats, the command files, and the graphical output results of several HDW release scenarios are described in the report. 3 refs., 7 figs., 9 tabs

  14. Current and proposed revisions, changes, and modifications to American codes and standards to address packaging, handling, and transportation of radioactive materials and how they relate to comparable international regulations

    International Nuclear Information System (INIS)

    Borter, W.H.; Froehlich, C.H.

    2004-01-01

    This paper addresses current and proposed revisions, additions, and modifications to American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) (i.e., ''ASMEthe Code'') Section III, Division 3 and American National Standards Institute (ANSI)/ASME N14.6. It provides insight into the ongoing processes of the associated committees and highlights important revisions, changes, and modifications to this Code and Standard. The ASME Code has developed and issued Division 3 to address items associated with the transportation and storage of radioactive materials. It currently only addresses ''General Requirements'' in Subsections WA and ''Class TP (Type B) Containments'' (Transportation Packages) in Subsection WB, but is in the process of adding a new Subsection WC to address ''Class SC'' (Storage Containments). ANSI/ASME Standard N14.6 which interacts with components constructed to Division 3 by addressinges special lifting devices for radioactive material shipping containers. This Standard is in the process of a complete re-write. This Code and Standard can be classified as ''dynamic'' in that their committees meet at least four times a year to evaluate proposed modifications and additions that reflect current safety practices in the nuclear industry. These evaluations include the possible addition of new materials, fabrication processes, examination methods, and testing requirements. An overview of this ongoing process is presented in this paper along with highlights of the more important proposed revisions, changes, and modifications and how they relate to United States (US) and international regulations and guidance like International Atomic Energy Agency (IAEA) Requirement No. TS-R-1

  15. 9 CFR 381.144 - Packaging materials.

    Science.gov (United States)

    2010-01-01

    ... to health. All packaging materials must be safe for the intended use within the meaning of section..., from the packaging supplier under whose brand name and firm name the material is marketed to the... distinguishing brand name or code designation appearing on the packaging material shipping container; must...

  16. Fruit-related terms and images on food packages and advertisements affect children's perceptions of foods' fruit content.

    Science.gov (United States)

    Heller, Rebecca; Martin-Biggers, Jennifer; Berhaupt-Glickstein, Amanda; Quick, Virginia; Byrd-Bredbenner, Carol

    2015-10-01

    To determine whether food label information and advertisements for foods containing no fruit cause children to have a false impression of the foods' fruit content. In the food label condition, a trained researcher showed each child sixteen different food label photographs depicting front-of-food label packages that varied with regard to fruit content (i.e. real fruit v. sham fruit) and label elements. In the food advertisement condition, children viewed sixteen, 30 s television food advertisements with similar fruit content and label elements as in the food label condition. After viewing each food label and advertisement, children responded to the question 'Did they use fruit to make this?' with responses of yes, no or don't know. Schools, day-care centres, after-school programmes and other community groups. Children aged 4-7 years. In the food label condition, χ 2 analysis of within fruit content variation differences indicated children (n 58; mean age 4·2 years) were significantly more accurate in identifying real fruit foods as the label's informational load increased and were least accurate when neither a fruit name nor an image was on the label. Children (n 49; mean age 5·4 years) in the food advertisement condition were more likely to identify real fruit foods when advertisements had fruit images compared with when no image was included, while fruit images in advertisements for sham fruit foods significantly reduced accuracy of responses. Findings suggest that labels and advertisements for sham fruit foods mislead children with regard to the food's real fruit content.

  17. Levels of processing and the coding of position cues in motor short-term memory.

    Science.gov (United States)

    Ho, L; Shea, J B

    1978-06-01

    The present study investigated the appropriateness of the levels-of-processing framework of memory for explaining retention of information in motor short-term memory. Subjects were given labels descriptive of the positions to be remembered by the experimenter (EL), were given no labels (NL), or provided their own labels (SL). A control group (CONT) was required to count backwards during the presentation of the criterion positions. The inclusion of a 30-sec filled retention interval as well as 0-sec and 30-sec unfilled retention intervals tested a prediction by Craik and Lockhart (1972), when attention is diverted from an item, information will be lost at a rate appropriate to its level of processing - that is, slower rates for deeper levels. Groups EL and SL had greater accuracy at recall for all three retention intervals than groups CONT and NL. In addition, there was no significant increase in error between 30-sec unfilled and 30-sec filled intervals for groups EL and SL, while there was a significant increase in error for groups CONT and NL. The data were interpreted in terms of Craik and Lockhart's (1972) levels-of-processing approach to memory.

  18. Department of Transportation -- Exemption for using the Transuranic Package Transporter-I (TRUPACT-I) at the Idaho National Engineering Laboratory (Code of Federal Regulations, Title 49, Part 107, Subpart B -- Exemptions, 107-103 Application for Exemption)

    International Nuclear Information System (INIS)

    Tyacke, M.J.; Macdonald, R.J.

    1992-08-01

    Exemption from specific regulations is being sought for the Transuranic Package Transporter Model I (TRUPACT-I) container. The design has successfully undergone extensive testing of a quarter-scale model and a full-scale prototype of the container. Results from the analysis and testing are in the TRUPACT-1 Safely Analysis Report for Packaging (SARP), GA-Al8695/SAND 87-7104 (TTC0735), April 1987 (see Attachment 1). The container was never certified or used because of questions raised during the certification process. Two features of the container design failed to satisfy the regulations for Type B packaging. First, the design utilizes a venting system to control internal and external pressures; this venting system is not allowed by the Code of Federal Regulations, Title 10, Parts 71(h) and 71.51(b) [10 CFR 71.(h) and 71.51(b)]. Second, the maximum quantity fissile material proposed to be hauled in TRUPACT-I exceeded the limits in 10 CFR 71.63(b) for a single-containment container. To correct these design deficiencies, the vents would be plugged during transport, and the maximum quantity of fissile material would be limited to the allowables for a single-containment container. An engineering analysis showed that the container could safely transport radioactive material within the boundaries of the Idaho National Engineering Laboratory (INEL) with the vent system plugged (see Attachment 2). However, some of the requirements for determining pressure on a container need to be changed (i.e., exempted) to reflect conditions unique to the INEL. The following are the requirements needing to be changed for INEL conditions, variances being sought, and justifications for the variances

  19. Functional interplay of top-down attention with affective codes during visual short-term memory maintenance.

    Science.gov (United States)

    Kuo, Bo-Cheng; Lin, Szu-Hung; Yeh, Yei-Yu

    2018-06-01

    Visual short-term memory (VSTM) allows individuals to briefly maintain information over time for guiding behaviours. Because the contents of VSTM can be neutral or emotional, top-down influence in VSTM may vary with the affective codes of maintained representations. Here we investigated the neural mechanisms underlying the functional interplay of top-down attention with affective codes in VSTM using functional magnetic resonance imaging. Participants were instructed to remember both threatening and neutral objects in a cued VSTM task. Retrospective cues (retro-cues) were presented to direct attention to the hemifield of a threatening object (i.e., cue-to-threat) or a neutral object (i.e., cue-to-neutral) during VSTM maintenance. We showed stronger activity in the ventral occipitotemporal cortex and amygdala for attending threatening relative to neutral representations. Using multivoxel pattern analysis, we found better classification performance for cue-to-threat versus cue-to-neutral objects in early visual areas and in the amygdala. Importantly, retro-cues modulated the strength of functional connectivity between the frontoparietal and early visual areas. Activity in the frontoparietal areas became strongly correlated with the activity in V3a-V4 coding the threatening representations instructed to be relevant for the task. Together, these findings provide the first demonstration of top-down modulation of activation patterns in early visual areas and functional connectivity between the frontoparietal network and early visual areas for regulating threatening representations during VSTM maintenance. Copyright © 2018 Elsevier Ltd. All rights reserved.

  20. Long term energy plan of Mongolian with nuclear power plant using Message code

    Energy Technology Data Exchange (ETDEWEB)

    Tsolmonbaatar, Batmunkh; Roh, Myung Sub [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-10-15

    In this paper, an overview of the current situation of Mongolia's energy sector and its role and contribution in the country's economy and environment, and a comprehensive assessment of the sector, are provided. Most importantly, the Model for Energy Supply Strategy Alternatives and their General Environmental Impacts (MESSAGE) model used to forecast the future energy supply and demand and to build and compare possible scenarios that could sustain economic development, environmental sustainability, and energy security in the country. In this paper, two scenarios for long-term energy development in Mongolia by 2040 were built using the MESSAGE model, and 2010 was set as the base year. The forecasting of the energy demand and supply was shown as a build the first model of Mongolia energy system to forecast the potentials of using NPP technology in the energy mix of Mongolia energy system. The industry and mining project sectors are expected to remain as the main energy consumers in the next decades. The demand of the sector will be three times of the total energy demand by 2040. The development of big mining projects and the increasing number of vehicles in Mongolia are the main factors that will contribute to the increased demand. Coal still accounts for the largest share in Mongolia's total primary energy demand. MESSAGE result shown in fig.2, which shows an exponential increase in coal resources demand in the future. It will be big issue to environmental impact (e.g., CO{sub 2} emission). The estimated CO{sub 2} emissions from fuel combustion are 13.0 million tones in Mongolia, this should be mitigated significantly, to meet the CO{sub 2} reduction target.

  1. Contribution and limits of geochemical calculation codes to evaluate the long term behavior of nuclear waste glasses

    International Nuclear Information System (INIS)

    Fritz, B.; Crovisier, J.L.

    1997-01-01

    Geochemical models have been intensively developed by researchers since more than twenty five years in order to be able to better understand and/or predict the long term stability/instability of water-rock systems. These geochemical codes were ail built first on a thermodynamic approach deriving from the application of Mass Action Law. The resulting first generation of models allowed to detect or predict the possible mass transfers (thermodynamic models) between aqueous and mineral phases including irreversible dissolutions of primary minerals and/or precipitation near equilibrium of secondary mineral phases. The recent development of models based on combined thermodynamics and kinetics opens the field of Lime dependent reactions prediction. This is crucial if one thinks to combine geochemical and hydrological studies in the so-called coupled models for transport and reaction calculations. All these models are progressively applied to the prediction of long term behavior of mineral phases, and more specifically glasses. In order to succeed in chat specific extension of the models, but also the data bases, there is a great need for additional new data from experimental approaches and from natural analogues. The modelling approach appears than also very useful in order to interpret the results of experimental data and to relate them to long term data extracted from natural analogues. Specific functions for modelling solid solution phases mat' also be used for describing the products of glasses alterations. (authors)

  2. Retroviral packaging cells encapsulated in TheraCyte immunoisolation devices enable long-term in vivo gene delivery.

    Science.gov (United States)

    Krupetsky, Anna; Parveen, Zahida; Marusich, Elena; Goodrich, Adrienne; Dornburg, Ralph

    2003-05-01

    The method of delivering a therapeutic gene into a patient is still one of the major obstacles towards successful human gene therapy. Here we describe a novel gene delivery approach using TheraCyte immunoisolation devices. Retroviral vector producing cells, derived from the avian retrovirus spleen necrosis virus, SNV, were encapsulated in TheraCyte devices and tested for the release of retroviral vectors. In vitro experiments show that such devices release infectious retroviral vectors into the tissue culture medium for up to 4 months. When such devices were implanted subcutaneously in SCID mice, infectious virus was released into the blood stream. There, the vectors were transported to and infected tumors, which had been induced by subcutaneous injection of tissue culture cells. Thus, this novel concept of a continuous, long-term gene delivery may constitute an attractive approach for future in vivo human gene therapy.

  3. Full Core Criticality Modeling of Gas-Cooled Fast Reactor Using the SCALE6.0 and MCNP5 Code Packages

    International Nuclear Information System (INIS)

    Matijevic, M.; Jecmenica, R.; Pevec, D.; Trontl, K.

    2012-01-01

    The Gas-Cooled Fast Reactor (GFR) is one of the reactor concepts selected by the Generation IV International Forum (GIF) for the next generation of innovative nuclear energy systems. It was selected among a group of more than 100 prototypes and his commercial availability is expected by 2030. GFR has common goals of the rest GIF advanced reactor types: economy, safety, proliferation resistance, availability and sustainability. Several GFR fuel design concepts such as plates, rod pins and pebbles are currently being investigated in order to meet the high temperature constraints characteristic for a GFR working enviroment. In the previous study we have compared the fuel depletion results for heterogeneous GFR fuel assembly (FA), obtained with TRITON6 sequence of SCALE6.0 code system, with the MCNPX-CINDER90 and TRIPOLI-4-D codes. Present work is a continuation of neutronic criticality analysis of heterogeneous FA and full core configurations of a GFR concept using 3-D Monte Carlo codes KENO-VI/SCALE6.0 and MCNP5. The FA is based on a hexagonal mesh of fuel rods (uranium and plutonium carbide fuel, silicon carbide clad, helium gas coolant) with axial reflector thickness being varied for the purpose of optimization. Three reflector materials were analysed: zirconium carbide (ZrC), silicon carbide (SiC) and natural uranium. ZrC has been selected as a reflector material, having the best contribution to the neutron economy and to the reactivity of the core. The core safety parameters were also analysed: a negative temperature coefficient of reactivity was verified for the heavy metal fuel and coolant density loss. Criticality calculations of different FA active heights were performed and the reflector thickness was also adjusted. Finally, GFR full core criticality calculations using different active fuel rod heights and fixed ZrC reflector height were done to find the optimal height of the core. The Shannon entropy of the GFR core fission distribution was proved to be

  4. Nuclear Criticality Safety Assessment Using the SCALE Computer Code Package. A demonstration based on an independent review of a real application

    International Nuclear Information System (INIS)

    Mennerdahl, Dennis

    1998-06-01

    The purpose of this project was to instruct a young scientist from the Lithuanian Energy Institute (LEI) on how to carry out an independent review of a safety report. In particular, emphasis, was to be put on how to use the personal computer version of the calculation system SCALE 4.3 in this process. Nuclear criticality safety together with radiation shielding from gamma and neutron sources were areas of interest. This report concentrates on nuclear criticality safety aspects while a separate report covers radiation shielding. The application was a proposed storage cask for irradiated fuel assemblies from the Ignalina RBMK reactors in Lithuania. The safety report contained various documents involving many design and safety considerations. A few other documents describing the Ignalina reactors and their operation were available. The time for the project was limited to approximately one month, starting 'clean' with a SCALE 4.3 CD-ROM, a thick safety report and a fast personal computer. The results should be of general interest to Swedish authorities, in particular related to shielding where experience in using advanced computer codes like those available in SCALE is limited. It has been known for many years that criticality safety is very complicated, and that independent reviews are absolutely necessary to reduce the risk from quite common errors in the safety assessments. Several important results were obtained during the project. Concerning use of SCALE 4.3, it was confirmed that a young scientist, without extensive previous experience in the code system, can learn to use essentially all options. During the project, it was obvious that familiarity with personal computers, operating systems (including network system) and office software (word processing, spreadsheet and Internet browser software) saved a lot of time. Some of the Monte Carlo calculations took several hours. Experience is valuable in quickly picking out input or source document errors. Understanding

  5. Review of the status of validation of the computer codes used in the severe accident source term reassessment study (BMI-2104)

    International Nuclear Information System (INIS)

    Kress, T.S.

    1985-04-01

    The determination of severe accident source terms must, by necessity it seems, rely heavily on the use of complex computer codes. Source term acceptability, therefore, rests on the assessed validity of such codes. Consequently, one element of NRC's recent efforts to reassess LWR severe accident source terms is to provide a review of the status of validation of the computer codes used in the reassessment. The results of this review is the subject of this document. The separate review documents compiled in this report were used as a resource along with the results of the BMI-2104 study by BCL and the QUEST study by SNL to arrive at a more-or-less independent appraisal of the status of source term modeling at this time

  6. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2005-01-01

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: ''each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.'' They further state: ''each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SARP charges the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.

  7. Coding early naturalists' accounts into long-term fish community changes in the Adriatic Sea (1800-2000.

    Directory of Open Access Journals (Sweden)

    Tomaso Fortibuoni

    Full Text Available The understanding of fish communities' changes over the past centuries has important implications for conservation policy and marine resource management. However, reconstructing these changes is difficult because information on marine communities before the second half of the 20(th century is, in most cases, anecdotal and merely qualitative. Therefore, historical qualitative records and modern quantitative data are not directly comparable, and their integration for long-term analyses is not straightforward. We developed a methodology that allows the coding of qualitative information provided by early naturalists into semi-quantitative information through an intercalibration with landing proportions. This approach allowed us to reconstruct and quantitatively analyze a 200-year-long time series of fish community structure indicators in the Northern Adriatic Sea (Mediterranean Sea. Our analysis provides evidence of long-term changes in fish community structure, including the decline of Chondrichthyes, large-sized and late-maturing species. This work highlights the importance of broadening the time-frame through which we look at marine ecosystem changes and provides a methodology to exploit, in a quantitative framework, historical qualitative sources. To the purpose, naturalists' eyewitness accounts proved to be useful for extending the analysis on fish community back in the past, well before the onset of field-based monitoring programs.

  8. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2008-01-01

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the pplication.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

  9. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2009-01-01

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

  10. RH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2006-01-01

    The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' It further states: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M and O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 Code of Federal Regulations (CFR) 1.8, 'Deliberate Misconduct.' Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. In accordance with 10 CFR Part 71, 'Packaging and Transportation of Radioactive Material,' certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, 'Reporting of Defects and Noncompliance,' regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these

  11. RH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2008-01-01

    The purpose of this program guidance document is to provide the technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package (also known as the 'RH-TRU 72-B cask') and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the C of C shall govern. The C of C states: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' It further states: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M and O) Contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8, 'Deliberate Misconduct.' Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, 'Packaging and Transportation of Radioactive Material,' certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21, 'Reporting of Defects and Noncompliance,' regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous

  12. Software requirements specification document for the AREST code development

    International Nuclear Information System (INIS)

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

    1993-11-01

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

  13. Safety analysis report on Model UC-609 shipping package

    International Nuclear Information System (INIS)

    Sandberg, R.R.

    1977-08-01

    This Safety Analysis Report for Packaging demonstrates that model UC-609 shipping package can safely transport tritium in any of its forms. The package and its contents are described. The package when subjected to the transport conditions specified in the Code of Federal Regulations, Title 10, Part 71 is evaluated. Finally, compliance with these regulations is discussed

  14. RASPLAV package

    International Nuclear Information System (INIS)

    1990-01-01

    The RASPLAV package for investigation of post-accident mass transport and heat transfer processes is presented. The package performs three dimensional thermal conduction calculations in space nonuniform and temperature dependent conductivities and variable heat sources, taking into account phase transformations. The processes of free-moving bulk material, mixing of melting fuel due to advection and dissolution, and also evaporation/adsorption are modelled. Two-dimensional hydrodynamics with self-consistent heat transfer are also performed. The paper briefly traces the ways the solution procedures are carried out in the program package and outlines the major results of the simulation of reactor vessel melting after a core meltdown. The theoretical analysis and the calculations in this case were carried out in order to define the possibility of localization of the zone reminders. The interactions between the reminders and the concrete are simulated and evaluation of the interaction parameters is carried out. 4 refs. (R.Ts)

  15. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2006-01-01

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant| (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations(CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

  16. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    2007-01-01

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and U.S. Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: 'each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.' They further state: 'each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.' Chapter 9.0 of the SARP charges the U.S. Department of Energy (DOE) or the Waste Isolation Pilot Plant (WIPP) management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with Title 10 Code of Federal Regulations (CFR) 71.8. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. The CBFO will evaluate the issue and notify the NRC if required.In accordance with 10 CFR Part 71, certificate holders, packaging users, and contractors or subcontractors who use, design, fabricate, test, maintain, or modify the packaging shall post copies of (1) 10 CFR Part 21 regulations, (2) Section 206 of the Energy Reorganization Act of 1974, and (3) NRC Form 3, Notice to Employees. These documents must be posted in a conspicuous location where the activities subject to these regulations

  17. Evaluation of an organic package of practice towards integrated management of Solanum tuberosum and its comparison with conventional farming in terms of yield, quality, energy efficiency and economics

    Directory of Open Access Journals (Sweden)

    Antara Seal

    2017-09-01

    Full Text Available A study was taken up during 2014-16 for evaluating the potential of an organic package of practice towards integrated crop production (green farming in comparison to conventional farmers’ practice in West Bengal, India. Under green farming, compost was integrated with chemical fertilizer for soil management while organic plant/ pest management was undertaken utilizing Inhana Rational Farming (IRF Technology. The study indicated higher yield (9.7 %, higher nutrient use efficiency and economic sustainability under green farming irrespective of study area or potato variety. Higher qualitative expression in terms of starch content, pulp pH, vitamin C etc. under green farming might be due to the organic plant management aimed at energization of plant biochemical functions. Soil quality development as noted under green farming might have been influenced by the on-farm produced compost containing rich self- generated micro flora (in order of 1016 per colony forming unit.. The study indicated that green farming may serve as an efficient substitute of conventional farming towards yield sustenance, abatement of food toxicity and quality end product; through higher use of renewable energy and activation of plant physiological functions.

  18. 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

  19. Development of waste packages for the long-term confinement of C-14 in TRU waste disposal. 2. Confinement container with titanium alloy

    International Nuclear Information System (INIS)

    Nakamura, Ario; Owada, Hitoshi; Asano, Hidekazu; Jintoku, Takashi; Nakayama, Gen

    2008-01-01

    The long-term integrity of TRU waste package, with a titanium alloy for the outer corrosion resistance layer and carbon steel for the inner structural vessel, has been evaluated. The target confinement period is settled at 60,000 years in this study (i.e., 10 times of half-life). So the outer corrosion resistance layer with titanium (Ti-Pd alloy) is developed through focus on the high corrosion resistance of Ti alloy as a technology that has long-term confinement. In investigation about integrity of its passive film, the thickness of corrosion layer of 60,000 years is calculated by building an empirical formula between temperature and corrosion current density, considering the results of constant voltage examination in the TRU waste repository assumed environment. About crevice corrosion, its occurrence conditions is investigated in the TRU waste repository assumed environment, then, Ti.Gr-17 is selected as candidate material of the corrosion resistance layer. In investigation about SCC in Ti alloy, using the models of growth of hydride-layer, the thickness of hydride-layer after 60,000 years is estimated by the results of constant currents tests. Then, the hydride-layer of this thickness is confirmed not to generate cracks, in consideration of destructive critical hydride cracking thickness and the models of crack propagation. The knowledge that the process of generation of hydrogenated layers changes with differences in acceleration conditions (i.e., current density) is obtained. So we must confirm the adequacy of this model by the examination with natural condition. (author)

  20. Calculation of source terms for NUREG-1150

    International Nuclear Information System (INIS)

    Breeding, R.J.; Williams, D.C.; Murfin, W.B.; Amos, C.N.; Helton, J.C.

    1987-10-01

    The source terms estimated for NUREG-1150 are generally based on the Source Term Code Package (STCP), but the actual source term calculations used in computing risk are performed by much smaller codes which are specific to each plant. This was done because the method of estimating the uncertainty in risk for NUREG-1150 requires hundreds of source term calculations for each accident sequence. This is clearly impossible with a large, detailed code like the STCP. The small plant-specific codes are based on simple algorithms and utilize adjustable parameters. The values of the parameters appearing in these codes are derived from the available STCP results. To determine the uncertainty in the estimation of the source terms, these parameters were varied as specified by an expert review group. This method was used to account for the uncertainties in the STCP results and the uncertainties in phenomena not considered by the STCP

  1. Watermarking spot colors in packaging

    Science.gov (United States)

    Reed, Alastair; Filler, TomáÅ.¡; Falkenstern, Kristyn; Bai, Yang

    2015-03-01

    In January 2014, Digimarc announced Digimarc® Barcode for the packaging industry to improve the check-out efficiency and customer experience for retailers. Digimarc Barcode is a machine readable code that carries the same information as a traditional Universal Product Code (UPC) and is introduced by adding a robust digital watermark to the package design. It is imperceptible to the human eye but can be read by a modern barcode scanner at the Point of Sale (POS) station. Compared to a traditional linear barcode, Digimarc Barcode covers the whole package with minimal impact on the graphic design. This significantly improves the Items per Minute (IPM) metric, which retailers use to track the checkout efficiency since it closely relates to their profitability. Increasing IPM by a few percent could lead to potential savings of millions of dollars for retailers, giving them a strong incentive to add the Digimarc Barcode to their packages. Testing performed by Digimarc showed increases in IPM of at least 33% using the Digimarc Barcode, compared to using a traditional barcode. A method of watermarking print ready image data used in the commercial packaging industry is described. A significant proportion of packages are printed using spot colors, therefore spot colors needs to be supported by an embedder for Digimarc Barcode. Digimarc Barcode supports the PANTONE spot color system, which is commonly used in the packaging industry. The Digimarc Barcode embedder allows a user to insert the UPC code in an image while minimizing perceptibility to the Human Visual System (HVS). The Digimarc Barcode is inserted in the printing ink domain, using an Adobe Photoshop plug-in as the last step before printing. Since Photoshop is an industry standard widely used by pre-press shops in the packaging industry, a Digimarc Barcode can be easily inserted and proofed.

  2. Modelling of blackout sequence at Atucha-1 using the MARCH3 code

    International Nuclear Information System (INIS)

    Baron, J.; Bastianelli, B.

    1997-01-01

    This paper presents the modelling of a complete blackout at the Atucha-1 NPP as preliminary phase for a Level II safety probabilistic analysis. The MARCH3 code of the STCP (Source Term Code Package) is used, based on a plant model made in accordance with particularities of the plant design. The analysis covers all the severe accident phases. The results allow to view the time sequence of the events, and provide the basis for source term studies. (author). 6 refs., 2 figs

  3. Waste package characterisation

    Energy Technology Data Exchange (ETDEWEB)

    Sannen, L.; Bruggeman, M.; Wannijn, J.P

    1998-09-01

    Radioactive wastes originating from the hot labs of the Belgian Nuclear Research Centre SCK-CEN contain a wide variety of radiotoxic substances. The accurate characterisation of the short- and long-term radiotoxic components is extremely difficult but required in view of geological disposal. This paper describes the methodology which was developed and adopted to characterise the high- and medium-level waste packages at the SCK-CEN hot laboratories. The proposed method is based on the estimation of the fuel inventory evacuated in a particular waste package; a calculation of the relative fission product contribution on the fuel fabrication and irradiation footing; a comparison of the calculated, as expected, dose rate and the real measured dose rate of the waste package. To cope with the daily practice an appropriate fuel inventory estimation route, a user friendly computer programme for fission product and corresponding dose rate calculation, and a simple dose rate measurement method have been developed and implemented.

  4. Waste package characterisation

    International Nuclear Information System (INIS)

    Sannen, L.; Bruggeman, M.; Wannijn, J.P.

    1998-09-01

    Radioactive wastes originating from the hot labs of the Belgian Nuclear Research Centre SCK-CEN contain a wide variety of radiotoxic substances. The accurate characterisation of the short- and long-term radiotoxic components is extremely difficult but required in view of geological disposal. This paper describes the methodology which was developed and adopted to characterise the high- and medium-level waste packages at the SCK-CEN hot laboratories. The proposed method is based on the estimation of the fuel inventory evacuated in a particular waste package; a calculation of the relative fission product contribution on the fuel fabrication and irradiation footing; a comparison of the calculated, as expected, dose rate and the real measured dose rate of the waste package. To cope with the daily practice an appropriate fuel inventory estimation route, a user friendly computer programme for fission product and corresponding dose rate calculation, and a simple dose rate measurement method have been developed and implemented

  5. Codes and curves

    CERN Document Server

    Walker, Judy L

    2000-01-01

    When information is transmitted, errors are likely to occur. Coding theory examines efficient ways of packaging data so that these errors can be detected, or even corrected. The traditional tools of coding theory have come from combinatorics and group theory. Lately, however, coding theorists have added techniques from algebraic geometry to their toolboxes. In particular, by re-interpreting the Reed-Solomon codes, one can see how to define new codes based on divisors on algebraic curves. For instance, using modular curves over finite fields, Tsfasman, Vladut, and Zink showed that one can define a sequence of codes with asymptotically better parameters than any previously known codes. This monograph is based on a series of lectures the author gave as part of the IAS/PCMI program on arithmetic algebraic geometry. Here, the reader is introduced to the exciting field of algebraic geometric coding theory. Presenting the material in the same conversational tone of the lectures, the author covers linear codes, inclu...

  6. A study on Prediction of Radioactive Source-term from the Decommissioning of Domestic NPPs by using CRUDTRAN Code

    Energy Technology Data Exchange (ETDEWEB)

    Song, Jong Soon; Lee, Sang Heon; Cho, Hoon Jo [Department of Nuclear Engineering Chosun University, Gwangju (Korea, Republic of)

    2016-10-15

    For the study, the behavior mechanism of corrosion products in the primary system of the Kori no.1 was analyzed, and the volume of activated corrosion products in the primary system was assessed based on domestic plant data with the CRUDTRAN code used to predict the volume. It is expected that the study would be utilized in predicting radiation exposure of workers performing maintenance and repairs in high radiation areas and in selecting the process of decontaminations and decommissioning in the primary system. It is also expected that in the future it would be used as the baseline data to estimate the volume of radioactive wastes when decommissioning a nuclear plant in the future, which would be an important criterion in setting the level of radioactive wastes used to compute the quantity of radioactive wastes. The results of prediction of the radioactive nuclide inventory in the primary system performed in this study would be used as baseline data for the estimation of the volume of radioactive wastes when decommissioning NPPs in the future. It is also expected that the data would be important criteria used to classify the level of radioactive wastes to calculate the volume. In addition, it is expected that the data would be utilized in reducing radiation exposure of workers in charge of system maintenance and repairing in high radiation zones and also predicting the selection of decontaminations and decommissioning processes in the primary systems. In future researches, it is planned to conduct the source term assessment against other NPP types such as CANDU and OPR-1000, in addition to the Westinghouse type nuclear plants.

  7. Short-term impact of oral hygiene training package to Anganwadi workers on improving oral hygiene of preschool children in North Indian City

    Science.gov (United States)

    2013-01-01

    Background Globally, dental caries is categorized in the list of public health problems in preschool children. In India, lack of availability and affordability of oral health enhances the cost of treatment and care. Empowering community workers like anganwadi workers (AWWs) in oral health, and providing basic oral health awareness to the mothers through them can be feasible model. So, the present study was conducted to evaluate the short-term impact of Oral Hygiene Training Package (OHTP) to AWWs on improving oral hygiene of preschool children. Methods This before and after comparison field trial was done in Anganwadi centres (AWCs) of Chandigarh city, India. 534 children aged 36-72 months attending 21 AWCs were examined before and after imparting trainings to AWWs. OHTP was administered to AWWs, which consisted of power-point presentation and demonstrated the skills like proper brushing technique, plaque disclosure, flossing technique, gum massaging etc. The AWWs later imparted training to mothers in their respective AWCs. Post intervention data was collected after three months. Outcome measures were improvement in oral health status (plaque, debris, gingival health), oral habits (brushing, rinsing) and decrease in caries activity (Snyder test). Results Prevalence of dental caries was found to be 48.3%. Only 4.1% of the population reported brushing twice which increased significantly to 9.9% post-intervention (p = 0.000). There was a significant decrease in debris (78.3% to 54.1%), and stage-1 plaque (75.5 to 66.5%) in the oral cavity. Caries activity by Snyder’s test decreased from 48.2% to 31.2% (p = 0.01) post-intervention. Conclusions Controlled trials of using AWWs to improve oral hygiene appear to be justified. Trial registration CTRI/2012/07/002786 PMID:24279468

  8. Spent fuel packaging and its safety analysis

    International Nuclear Information System (INIS)

    Takada, Kimitaka; Nakaoki, Kozo; Tamamura, Tadao; Matsuda, Fumio; Fukudome, Kazuyuki

    1983-01-01

    An all stainless steel B(U) type packaging is proposed to transport spent fuels discharged from research reactors and other radioactive materials. The package is used dry and provided with surface fins to absorb drop shock and to dissipate decay heat. Safety was analyzed for structural, thermal, containment shielding and criticality factors, and the integrity of the package was confirmed with the MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, and KENO computer codes. (author)

  9. The WECHSL-Mod2 code: A computer program for the interaction of a core melt with concrete including the long term behavior

    International Nuclear Information System (INIS)

    Reimann, M.; Stiefel, S.

    1989-06-01

    The WECHSL-Mod2 code is a mechanistic computer code developed for the analysis of the thermal and chemical interaction of initially molten LWR reactor materials with concrete in a two-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. The code assumes that the metallic phases of the melt pool form a layer at the bottom overlayed by the oxide melt atop. 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 WECHSL code in its present version was validated by the BETA experiments. The test samples include a typical BETA post test calculation and a WECHSL application to a reactor accident. (orig.) [de

  10. Spin wave Feynman diagram vertex computation package

    Science.gov (United States)

    Price, Alexander; Javernick, Philip; Datta, Trinanjan

    Spin wave theory is a well-established theoretical technique that can correctly predict the physical behavior of ordered magnetic states. However, computing the effects of an interacting spin wave theory incorporating magnons involve a laborious by hand derivation of Feynman diagram vertices. The process is tedious and time consuming. Hence, to improve productivity and have another means to check the analytical calculations, we have devised a Feynman Diagram Vertex Computation package. In this talk, we will describe our research group's effort to implement a Mathematica based symbolic Feynman diagram vertex computation package that computes spin wave vertices. Utilizing the non-commutative algebra package NCAlgebra as an add-on to Mathematica, symbolic expressions for the Feynman diagram vertices of a Heisenberg quantum antiferromagnet are obtained. Our existing code reproduces the well-known expressions of a nearest neighbor square lattice Heisenberg model. We also discuss the case of a triangular lattice Heisenberg model where non collinear terms contribute to the vertex interactions.

  11. Speech coding

    Energy Technology Data Exchange (ETDEWEB)

    Ravishankar, C., Hughes Network Systems, Germantown, MD

    1998-05-08

    Speech is the predominant means of communication between human beings and since the invention of the telephone by Alexander Graham Bell in 1876, speech services have remained to be the core service in almost all telecommunication systems. Original analog methods of telephony had the disadvantage of speech signal getting corrupted by noise, cross-talk and distortion Long haul transmissions which use repeaters to compensate for the loss in signal strength on transmission links also increase the associated noise and distortion. On the other hand digital transmission is relatively immune to noise, cross-talk and distortion primarily because of the capability to faithfully regenerate digital signal at each repeater purely based on a binary decision. Hence end-to-end performance of the digital link essentially becomes independent of the length and operating frequency bands of the link Hence from a transmission point of view digital transmission has been the preferred approach due to its higher immunity to noise. The need to carry digital speech became extremely important from a service provision point of view as well. Modem requirements have introduced the need for robust, flexible and secure services that can carry a multitude of signal types (such as voice, data and video) without a fundamental change in infrastructure. Such a requirement could not have been easily met without the advent of digital transmission systems, thereby requiring speech to be coded digitally. The term Speech Coding is often referred to techniques that represent or code speech signals either directly as a waveform or as a set of parameters by analyzing the speech signal. In either case, the codes are transmitted to the distant end where speech is reconstructed or synthesized using the received set of codes. A more generic term that is applicable to these techniques that is often interchangeably used with speech coding is the term voice coding. This term is more generic in the sense that the

  12. Package Formats for Preserved Digital Material

    DEFF Research Database (Denmark)

    Zierau, Eld

    2012-01-01

    This paper presents an investigation of the best suitable package formats for long term digital preservation. The choice of a package format for preservation is crucial for future access, thus a thorough analysis of choice is important. The investigation presented here covers setting up requireme......This paper presents an investigation of the best suitable package formats for long term digital preservation. The choice of a package format for preservation is crucial for future access, thus a thorough analysis of choice is important. The investigation presented here covers setting up...... requirements for package formats used for long term preserved digital material, and using these requirements as the basis for analysing a range of package formats. The result of the concrete investigation is that the WARC format is the package format best suited for the listed requirements. Fulfilling...

  13. Lively package

    International Nuclear Information System (INIS)

    Jaremko, G.

    1997-01-01

    Progress on the Lloydminster Heavy Oil Interpretive Centre, sponsored by the Lloydminster Oilfield Technical Society and expected to open in late 1998, was discussed. Some $150,000 of the $750,000 budget is already in the bank, and another $150,000 is in the pipeline. The Centre will be added to an existing and well-established visitor's site. It is reported to contain a lively and imaginatively-designed exhibit package, and promises to become a combination of educational tool and tourist attraction for the town of Lloydminster, Saskatchewan, in the heart of heavy oil country

  14. Reflective Packaging

    Science.gov (United States)

    1994-01-01

    The aluminized polymer film used in spacecraft as a radiation barrier to protect both astronauts and delicate instruments has led to a number of spinoff applications. Among them are aluminized shipping bags, food cart covers and medical bags. Radiant Technologies purchases component materials and assembles a barrier made of layers of aluminized foil. The packaging reflects outside heat away from the product inside the container. The company is developing new aluminized lines, express mailers, large shipping bags, gel packs and insulated panels for the building industry.

  15. Documentation of the seawater intrusion (SWI2) package for MODFLOW

    Science.gov (United States)

    Bakker, Mark; Schaars, Frans; Hughes, Joseph D.; Langevin, Christian D.; Dausman, Alyssa M.

    2013-01-01

    The SWI2 Package is the latest release of the Seawater Intrusion (SWI) Package for MODFLOW. The SWI2 Package allows three-dimensional vertically integrated variable-density groundwater flow and seawater intrusion in coastal multiaquifer systems to be simulated using MODFLOW-2005. Vertically integrated variable-density groundwater flow is based on the Dupuit approximation in which an aquifer is vertically discretized into zones of differing densities, separated from each other by defined surfaces representing interfaces or density isosurfaces. The numerical approach used in the SWI2 Package does not account for diffusion and dispersion and should not be used where these processes are important. The resulting differential equations are equivalent in form to the groundwater flow equation for uniform-density flow. The approach implemented in the SWI2 Package allows density effects to be incorporated into MODFLOW-2005 through the addition of pseudo-source terms to the groundwater flow equation without the need to solve a separate advective-dispersive transport equation. Vertical and horizontal movement of defined density surfaces is calculated separately using a combination of fluxes calculated through solution of the groundwater flow equation and a simple tip and toe tracking algorithm. Use of the SWI2 Package in MODFLOW-2005 only requires the addition of a single additional input file and modification of boundary heads to freshwater heads referenced to the top of the aquifer. Fluid density within model layers can be represented using zones of constant density (stratified flow) or continuously varying density (piecewise linear in the vertical direction) in the SWI2 Package. The main advantage of using the SWI2 Package instead of variable-density groundwater flow and dispersive solute transport codes, such as SEAWAT and SUTRA, is that fewer model cells are required for simulations using the SWI2 Package because every aquifer can be represented by a single layer of cells

  16. Remote-Handled Transuranic Content Codes

    International Nuclear Information System (INIS)

    2001-01-01

    The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document represents the development of a uniform content code system for RH-TRU waste to be transported in the 72-Bcask. It will be used to convert existing waste form numbers, content codes, and site-specific identification codes into a system that is uniform across the U.S. Department of Energy (DOE) sites.The existing waste codes at the sites can be grouped under uniform content codes without any lossof waste characterization information. The RH-TRUCON document provides an all-encompassing description for each content code and compiles this information for all DOE sites. Compliance with waste generation, processing, and certification procedures at the sites (outlined in this document foreach content code) ensures that prohibited waste forms are not present in the waste. The content code gives an overall description of the RH-TRU waste material in terms of processes and packaging, as well as the generation location. This helps to provide cradle-to-grave traceability of the waste material so that the various actions required to assess its qualification as payload for the 72-B cask can be performed. The content codes also impose restrictions and requirements on the manner in which a payload can be assembled. The RH-TRU Waste Authorized Methods for Payload Control (RH-TRAMPAC), Appendix 1.3.7 of the 72-B Cask Safety Analysis Report (SAR), describes the current governing procedures applicable for the qualification of waste as payload for the 72-B cask. The logic for this classification is presented in the 72-B Cask SAR. Together, these documents (RH-TRUCON, RH-TRAMPAC, and relevant sections of the 72-B Cask SAR) present the foundation and justification for classifying RH-TRU waste into content codes. Only content codes described in thisdocument can be considered for transport in the 72-B cask. Revisions to this document will be madeas additional waste qualifies for transport. Each content code uniquely

  17. Safety analysis of spent fuel packaging

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

    1987-01-01

    Many types of spent fuel packagings have been manufactured and been used for transport of spent fuels discharged from nuclear power plant. These spent fuel packagings need to be assesed thoroughly about safety transportation because spent fuels loaded into the packaging have high radioactivity and generation of heat. This paper explains the outline of safety analysis of a packaging, Safety analysis is performed for structural, thermal, containment, shielding and criticality factors, and MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, KENO, etc computer codes are used for such analysis. (author)

  18. Decoding Codes on Graphs

    Indian Academy of Sciences (India)

    Shannon limit of the channel. Among the earliest discovered codes that approach the. Shannon limit were the low density parity check (LDPC) codes. The term low density arises from the property of the parity check matrix defining the code. We will now define this matrix and the role that it plays in decoding. 2. Linear Codes.

  19. Group Self-Definitions and Religious Collectivities in the 16th Book of the Theodosian Code. Notes on the Term coetus (with a Preliminary Analysis of populus

    Directory of Open Access Journals (Sweden)

    Alessandro Saggioro

    2015-06-01

    Full Text Available This essay starts from the well-known definition of populus as well as it appears in some Cicero’s works and aims at analyze the religious significance of coetus, a term strictly connected with the first, at least focusing on the 16th book of the Theodosian Code. This article intends to cast light some transformations and representations of religious groups between Constantinian and Post-Constantinian ages.

  20. Alternate Materials In Design Of Radioactive Material Packages

    International Nuclear Information System (INIS)

    Blanton, P.; Eberl, K.

    2010-01-01

    This paper presents a summary of design and testing of material and composites for use in radioactive material packages. These materials provide thermal protection and provide structural integrity and energy absorption to the package during normal and hypothetical accident condition events as required by Title 10 Part 71 of the Code of Federal Regulations. Testing of packages comprising these materials is summarized.

  1. 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

  2. Short-term hydro generation scheduling of Xiluodu and Xiangjiaba cascade hydropower stations using improved binary-real coded bee colony optimization algorithm

    International Nuclear Information System (INIS)

    Lu, Peng; Zhou, Jianzhong; Wang, Chao; Qiao, Qi; Mo, Li

    2015-01-01

    Highlights: • STHGS problem is decomposed into two parallel sub-problems of UC and ELD. • Binary coded BCO is used to solve UC sub-problem with 0–1 discrete variables. • Real coded BCO is used to solve ELD sub-problem with continuous variables. • Some heuristic repairing strategies are designed to handle various constraints. • The STHGS of Xiluodu and Xiangjiaba cascade stations is solved by IB-RBCO. - Abstract: Short-term hydro generation scheduling (STHGS) of cascade hydropower stations is a typical nonlinear mixed integer optimization problem to minimize the total water consumption while simultaneously meeting the grid requirements and other hydraulic and electrical constraints. In this paper, STHGS problem is decomposed into two parallel sub-problems of unit commitment (UC) and economic load dispatch (ELD), and the methodology of improved binary-real coded bee colony optimization (IB-RBCO) algorithm is proposed to solve them. Firstly, the improved binary coded BCO is used to solve the UC sub-problem with 0–1 discrete variables, and the heuristic repairing strategy for unit state constrains is applied to generate the feasible unit commitment schedule. Then, the improved real coded BCO is used to solve the ELD sub-problem with continuous variables, and an effective method is introduced to handle various unit operation constraints. Especially, the new updating strategy of DE/best/2/bin method with dynamic parameter control mechanism is applied to real coded BCO to improve the search ability of IB-RBCO. Finally, to verify the feasibility and effectiveness of the proposed IB-RBCO method, it is applied to solve the STHGS problem of Xiluodu and Xiangjiaba cascaded hydropower stations, and the simulating results are compared with other intelligence algorithms. The simulation results demonstrate that the proposed IB-RBCO method can get higher-quality solutions with less water consumption and shorter calculating time when facing the complex STHGS problem

  3. The CORSYS neutronics code system

    International Nuclear Information System (INIS)

    Caner, M.; Krumbein, A.D.; Saphier, D.; Shapira, M.

    1994-01-01

    The purpose of this work is to assemble a code package for LWR core physics including coupled neutronics, burnup and thermal hydraulics. The CORSYS system is built around the cell code WIMS (for group microscopic cross section calculations) and 3-dimension diffusion code CITATION (for burnup and fuel management). We are implementing such a system on an IBM RS-6000 workstation. The code was rested with a simplified model of the Zion Unit 2 PWR. (authors). 6 refs., 8 figs., 1 tabs

  4. Recycling source terms for edge plasma fluid models and impact on convergence behaviour of the BRAAMS 'B2' code

    International Nuclear Information System (INIS)

    Maddison, G.P.; Reiter, D.

    1994-02-01

    Predictive simulations of tokamak edge plasmas require the most authentic description of neutral particle recycling sources, not merely the most expedient numerically. Employing a prototypical ITER divertor arrangement under conditions of high recycling, trial calculations with the 'B2' steady-state edge plasma transport code, plus varying approximations or recycling, reveal marked sensitivity of both results and its convergence behaviour to details of sources incorporated. Comprehensive EIRENE Monte Carlo resolution of recycling is implemented by full and so-called 'shot' intermediate cycles between the plasma fluid and statistical neutral particle models. As generally for coupled differencing and stochastic procedures, though, overall convergence properties become more difficult to assess. A pragmatic criterion for the 'B2'/EIRENE code system is proposed to determine its success, proceeding from a stricter condition previously identified for one particular analytic approximation of recycling in 'B2'. Certain procedures are also inferred potentially to improve their convergence further. (orig.)

  5. Safety Analysis Report - Packages, 9965, 9968, 9972-9975 Packages

    International Nuclear Information System (INIS)

    Van Alstine, M.N.

    1999-01-01

    This Safety Analysis Report for Packaging (SARP) documents the performance of the 9965 B, 9968 B, 9972 B(U), 9973 B(U), 9974 B(U), and 9975 B(U) packages in satisfying the regulatory safety requirements of the Code of Federal Regulations (CFR) 711 and the International Atomic Energy Agency (IAEA) Safety Series No. 6, Regulations for the Safe Transport of Radioactive Material, 1985 edition2. Results of the analysis and testing performed on the 9965 B, 9968 B, 9972 B(U), 9973 B(U), 9974 B(U), and 9975 B(U) packages are presented in this SARP, which was prepared in accordance with U.S. Department of energy (DOE) Order 5480.33 and in the format specified in the Nuclear Regulatory Commission (NRC) Regulatory Guides 7.94 and 7.10.5

  6. Safety analysis report - packages 9965, 9968, 9972-9975 packages

    International Nuclear Information System (INIS)

    Van Alstine, M.N.

    1997-10-01

    This Safety Analysis Report for Packaging (SARP) documents the performance of the 9965 B( ), 9968 B( ), 9972 B(U), 9973 B(U), 9974 B(U), and 9975 B(U) packages in satisfying the regulatory safety requirements of the Code of Federal Regulations (CFR) 10 CFR 71 and the International Atomic Energy Agency (IAEA) Safety Series No. 6, Regulations for the Safe Transport of Radioactive Material, 1985 edition. Results of the analysis and testing performed on the 9965 B(), 9968 B(), 9972 B(U), 9973 B(U), and 9975 B(U) packages are presented in this SARP, which was prepared in accordance with U.S. Department of Energy (DOE) Order 5480.3 and in the format specified in the Nuclear Regulatory Commission (NRC) Regulatory Guides 7.9 and 7.10

  7. The environmental effects of taxes on packages

    International Nuclear Information System (INIS)

    Schroten, A.; Nelissen, D.; Bergsma, G.C.; Blom, M.J.

    2010-08-01

    The results of an evaluation of the environmental impacts of taxes for packages are presented, differentiated for greenhouse gas emissions. The evaluation used a qualitative analysis of information from eighteen depth-interviews with experts in the packaging market, foreign experiences, relevant price elasticities and 'expert guesses'. It appears that tax package so far had a limited effect on the packaging market. For the longer term (ten years) larger, but probably also limited, effects are expected. The environmental impact of packaging tax can be increased if the taxes are substantially increased. [nl

  8. Vertical Drop of the Naval SNF Long Waste Package On Unyielding Surface

    International Nuclear Information System (INIS)

    S. Mastilovic

    2006-01-01

    The purpose of this calculation is to determine the structural response of a Naval SNF (Spent Nuclear Fuel) Long Waste Package (WP) subjected to 2 m-vertical drop on unyielding surface (US). The scope of this document is limited to reporting the calculation results in terms of maximum stress intensities. This calculation is associated with the waste package design; calculation is performed by the Waste Package Design group. AP-3.12Q, Revision 0, ICN 0, Calculations, is used to perform the calculation and develop the document. The finite element calculation is performed by using the commercially available ANSYS Version (V) 5.4 finite element code. The result of this calculation is provided in terms of maximum stress intensities

  9. Science packages

    Science.gov (United States)

    1997-01-01

    Primary science teachers in Scotland have a new updating method at their disposal with the launch of a package of CDi (Compact Discs Interactive) materials developed by the BBC and the Scottish Office. These were a response to the claim that many primary teachers felt they had been inadequately trained in science and lacked the confidence to teach it properly. Consequently they felt the need for more in-service training to equip them with the personal understanding required. The pack contains five disks and a printed user's guide divided up as follows: disk 1 Investigations; disk 2 Developing understanding; disks 3,4,5 Primary Science staff development videos. It was produced by the Scottish Interactive Technology Centre (Moray House Institute) and is available from BBC Education at £149.99 including VAT. Free Internet distribution of science education materials has also begun as part of the Global Schoolhouse (GSH) scheme. The US National Science Teachers' Association (NSTA) and Microsoft Corporation are making available field-tested comprehensive curriculum material including 'Micro-units' on more than 80 topics in biology, chemistry, earth and space science and physics. The latter are the work of the Scope, Sequence and Coordination of High School Science project, which can be found at http://www.gsh.org/NSTA_SSandC/. More information on NSTA can be obtained from its Web site at http://www.nsta.org.

  10. Naval Waste Package Design Sensitivity

    International Nuclear Information System (INIS)

    T. Schmitt

    2006-01-01

    The purpose of this calculation is to determine the sensitivity of the structural response of the Naval waste packages to varying inner cavity dimensions when subjected to a comer drop and tip-over from elevated surface. This calculation will also determine the sensitivity of the structural response of the Naval waste packages to the upper bound of the naval canister masses. The scope of this document is limited to reporting the calculation results in terms of through-wall stress intensities in the outer corrosion barrier. This calculation is intended for use in support of the preliminary design activities for the license application design of the Naval waste package. It examines the effects of small changes between the naval canister and the inner vessel, and in these dimensions, the Naval Long waste package and Naval Short waste package are similar. Therefore, only the Naval Long waste package is used in this calculation and is based on the proposed potential designs presented by the drawings and sketches in References 2.1.10 to 2.1.17 and 2.1.20. All conclusions are valid for both the Naval Long and Naval Short waste packages

  11. 21 CFR 106.90 - Coding.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 2 2010-04-01 2010-04-01 false Coding. 106.90 Section 106.90 Food and Drugs FOOD... of Infant Formulas § 106.90 Coding. The manufacturer shall code all infant formulas in conformity with the coding requirements that are applicable to thermally processed low-acid foods packaged in...

  12. Thermal analysis of transportation packaging for nuclear spent fuel

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki

    1989-01-01

    Safety analysis of transportation packaging for nuclear spent fuel comprises structural, thermal, containment, shielding and criticality factors, and the safety of a packaging is verified by these analyses. In thermal analysis, the temperature of each part of the packaging is calculated under normal and accident test conditions. As an example of thermal analysis, the temperature distribution of a packaging being subjected to a normal test was calculated by the TRUMP code and compared with measured data. (author)

  13. A study on the application of CRUDTRAN code in primary systems of domestic pressurized heavy-water reactors for prediction of radiation source term

    Energy Technology Data Exchange (ETDEWEB)

    Song, Jong Soon; Cho, Hoon Jo; Jung, Min Young; Lee, Sang Heon [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2017-04-15

    The importance of developing a source-term assessment technology has been emphasized owing to the decommissioning of Kori nuclear power plant (NPP) Unit 1 and the increase of deteriorated NPPs. We analyzed the behavioral mechanism of corrosion products in the primary system of a pressurized heavy-water reactor-type NPP. In addition, to check the possibility of applying the CRUDTRAN code to a Canadian Deuterium Uranium Reactor (CANDU)-type NPP, the type was assessed using collected domestic onsite data. With the assessment results, it was possible to predict trends according to operating cycles. Values estimated using the code were similar to the measured values. The results of this study are expected to be used to manage the radiation exposures of operators in high-radiation areas and to predict decommissioning processes in the primary system.

  14. An improvement of estimation method of source term to the environment for interfacing system LOCA for typical PWR using MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Han, Seok Jung; Kim, Tae Woon; Ahn, Kwang Il [Risk and Environmental Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    Interfacing-system loss-of-coolant-accident (ISLOCA) has been identified as the most hazardous accident scenario in the typical PWR plants. The present study as an effort to improve the knowledge of the source term to the environment during ISLOCA focuses on an improvement of the estimation method. The improvement was performed to take into account an effect of broken pipeline and auxiliary building structures relevant to ISLOCA. An estimation of the source term to the environment was for the OPR-1000 plants by MELOCR code version 1.8.6. The key features of the source term showed that the massive amount of fission products departed from the beginning of core degradation to the vessel breach. The release amount of fission products may be affected by the broken pipeline and the auxiliary building structure associated with release pathway.

  15. Reassessment of the technical bases for estimating source terms. Final report

    International Nuclear Information System (INIS)

    Silberberg, M.; Mitchell, J.A.; Meyer, R.O.; Ryder, C.P.

    1986-07-01

    This document describes a major advance in the technology for calculating source terms from postulated accidents at US light-water reactors. The improved technology consists of (1) an extensive data base from severe accident research programs initiated following the TMI accident, (2) a set of coupled and integrated computer codes (the Source Term Code Package), which models key aspects of fission product behavior under severe accident conditions, and (3) a number of detailed mechanistic codes that bridge the gap between the data base and the Source Term Code Package. The improved understanding of severe accident phenonmena has also allowed an identification of significant sources of uncertainty, which should be considered in estimating source terms. These sources of uncertainty are also described in this document. The current technology provides a significant improvement in evaluating source terms over that available at the time of the Reactor Safety Study (WASH-1400) and, because of this significance, the Nuclear Regulatory Commission staff is recommending its use

  16. ADAM adaptation and mitigation strategies: supporting European climate policy. Deliverable D3 of work package M1 (code D-M1.3). ADAM 2-degree scenario for Europe - policies and impacts

    Energy Technology Data Exchange (ETDEWEB)

    Schade, Wolfgang; Jochem, Eberhard; Barker, Terry [and others

    2009-07-31

    ADAM research identifies and appraises existing and new policy options that can contribute to different combinations of adaptation and mitigation strategies. These options address the demands a changing climate will place on protecting citizens and valuable ecosystems - i.e., adaptation - as well as addressing the necessity to restrain/control humankind's perturbation to global climate to a desirable level - i.e., mitigation. The work package Mitigation 1 (Ml) has the core objective to simulate mitigation options and their related costs for Europe until 2050 and 2100 respectively. The focus of this deliverable is on the period 2005 to 2050. The long-term period until 2100 is covered in the previous deliverable D2, applying the POLES model for this time horizon. The analysis constitutes basically a techno-economic analysis. Depending on the sector analyzed it is either directly combined with a policy analysis (e.g. in the transport sector, renewables sector) or the policy analysis is performed qualitatively as a subsequent and independent step after the techno-economic analysis is completed (e.g. in the residential and service sectors). The book includes the following chapters: scenarios and macroeconomic assumptions; methodological issues analyzing mitigation options; the integrated global energy model POLES and its projections for the reference and 2 deg C scenarios; forest and basic materials sector; residential sector in Europe; the service (tertiary) and the primary sectors in Europe; basic products and other manufacturing industry sectors; transport sectors in Europe; renewable sector in Europe; conversion sector in Europe; syntheses and sectoral analysis in Europe; macroeconomic impacts of climate policy in the EU; the effects of the financial crisis on baseline simulations with implications for climate policy modeling: an analysis using the global model E3MG 2008-2012; conclusions and policy recommendations.

  17. ADAM adaptation and mitigation strategies: supporting European climate policy. Deliverable D3 of work package M1 (code D-M1.3). ADAM 2-degree scenario for Europe - policies and impacts

    Energy Technology Data Exchange (ETDEWEB)

    Schade, Wolfgang; Jochem, Eberhard; Barker, Terry (and others)

    2009-07-31

    ADAM research identifies and appraises existing and new policy options that can contribute to different combinations of adaptation and mitigation strategies. These options address the demands a changing climate will place on protecting citizens and valuable ecosystems - i.e., adaptation - as well as addressing the necessity to restrain/control humankind's perturbation to global climate to a desirable level - i.e., mitigation. The work package Mitigation 1 (Ml) has the core objective to simulate mitigation options and their related costs for Europe until 2050 and 2100 respectively. The focus of this deliverable is on the period 2005 to 2050. The long-term period until 2100 is covered in the previous deliverable D2, applying the POLES model for this time horizon. The analysis constitutes basically a techno-economic analysis. Depending on the sector analyzed it is either directly combined with a policy analysis (e.g. in the transport sector, renewables sector) or the policy analysis is performed qualitatively as a subsequent and independent step after the techno-economic analysis is completed (e.g. in the residential and service sectors). The book includes the following chapters: scenarios and macroeconomic assumptions; methodological issues analyzing mitigation options; the integrated global energy model POLES and its projections for the reference and 2 deg C scenarios; forest and basic materials sector; residential sector in Europe; the service (tertiary) and the primary sectors in Europe; basic products and other manufacturing industry sectors; transport sectors in Europe; renewable sector in Europe; conversion sector in Europe; syntheses and sectoral analysis in Europe; macroeconomic impacts of climate policy in the EU; the effects of the financial crisis on baseline simulations with implications for climate policy modeling: an analysis using the global model E3MG 2008-2012; conclusions and policy recommendations.

  18. PERCEVAL v4.0: a new PC code for gamma radiation studies based upon most recent development about point kernel

    Energy Technology Data Exchange (ETDEWEB)

    Bindel, Laurent; Clouet, Laurent; Castanier, Eric; Bonnet, Jerome; Fleury, Guillaume; Vermuse, Manuel; Gamess, Andre; Lejeune, Eric [Societe Generale pour les techniques Nouvelles, Saint Quentin en Yvelines (France)

    2000-03-01

    The present paper presents the capabilities of a new code named PERCEVAL v4.0 and based on the new point kernel described in an attached issue. Two linked codes named SPECTRE{sub G} and GRAPH{sub 3}D are part of the code package in order to establish the energetic source term and visualize the tri-dimensional scene respectively. (author)

  19. Packaging supplier inspection guide

    International Nuclear Information System (INIS)

    Stromberg, H.M.; Gregg, R.E.; Kido, C.; Boyle, C.D.

    1991-05-01

    This is document is a guide for conducting quality assurance inspections of transportations packaging suppliers, where suppliers are defined as designers, fabricators, distributors, users, or owners of transportation packaging. This document can be used during an inspection to determine regulatory compliance within the requirements of 10 Code of Federal Regulations, Part 71, Subpart H (10 CFR 71.101--71.135). The guidance described in this document provides a framework for an inspection. It provides the inspector with the flexibility to adapt the methods and concepts presented here to meet the needs of the particular facility being inspected. The guide was developed to ensure a structured and consistent approach for inspections. The method treats each activity at a supplier facility as a separate entity (or functional element), and combines the activities within the framework of an ''inspection tree.'' The method separates each functional element into several areas of performance and then identifies guidelines, based on regulatory requirements, to be used to qualitatively rate each area. This document was developed to serve as a field manual to facilitate the work of inspectors. 1 ref., 1 fig., 5 tabs

  20. (02.2) Scoping experiments; (02.3) long-term corrosion testing and properties evaluation of candidate waste package basket material

    International Nuclear Information System (INIS)

    VanKonynenburg, R. A.

    1996-01-01

    The work described in this activity plan addresses Information Need 2.7.3 of the Yucca Mountain Site Characterization Plan (l), which reads Determination that the design criteria in lOCFR60.130 through 60.133 and any appropriate additional design objectives pertaining to criticality control have been met. This work falls under section WBS 1.2.2.5 2 (Basket Materials) of WBS 1.2.2.5 (Waste Package Materials) in the Work Breakdown Structure of the Yucca Mountain Site Characterization Project

  1. Hazardous materials package performance regulations

    International Nuclear Information System (INIS)

    Russell, N.A.; Glass, R.E.; McClure, J.D.; Finley, N.C.

    1993-01-01

    Two regulatory philosophies, one based on 'specification' packaging standards and the other based on 'performance' packaging standards, currently define the hazmat packaging certification process. A main concern when setting performance standards is determining the appropriate standards necessary to assure adequate public protection. This paper discusses a Hazmat Packaging Performance Evaluation (HPPE) project being conducted at Sandia National Laboratories for the U.S. Department of Transportation Research and Special Programs Administration. In this project, the current bulk packagings (larger than 2000 gallons) for transporting Materials Extremely Toxic By Inhalation (METBI) are being evaluated and performance standards will be recommended. A computer software system, HazCon, has been developed which can calculate the dispersion of dense, neutral, and buoyant gases. HazCon also has a database of thermodynamic and toxicity data for the METBI materials, a user-friendly menu-driven format for creating input data sets for calculating dispersion of the METBI in the event of an accidental release, and a link between the METBI database and the dense gas dispersion code (which requires thermodynamic properties). The primary output of HazCon is a listing of mass concentrations of the released material at distances downwind from the release point. (J.P.N.)

  2. Integrated computer codes for nuclear power plant severe accident analysis

    International Nuclear Information System (INIS)

    Jordanov, I.; Khristov, Y.

    1995-01-01

    This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs

  3. Integrated computer codes for nuclear power plant severe accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Jordanov, I; Khristov, Y [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika

    1996-12-31

    This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs.

  4. Opening the gift: social inclusion, professional codes and gift-giving in long-term mental healthcare

    NARCIS (Netherlands)

    Ootes, S. T. C.; Pols, A. J.; Tonkens, E. H.; Willems, D. L.

    2013-01-01

    Deinstitutionalisation has not only made the social inclusion of clients a key objective in long-term mental healthcare, it may also affect the role of the care professional. This article investigates whether the social inclusion objective clashes with other long-standing professional values,

  5. actuar: An R Package for Actuarial Science

    OpenAIRE

    Dutang, Christophe; Goulet, Vincent; Pigeon, Mathieu

    2008-01-01

    actuar is a package providing additional Actuarial Science functionality to the R statistical system. The project was launched in 2005 and the package is available on the Comprehensive R Archive Network since February 2006. The current version of the package contains functions for use in the fields of loss distributions modeling, risk theory (including ruin theory), simulation of compound hierarchical models and credibility theory. This paper presents in detail but with few technical terms t...

  6. Code Cactus; Code Cactus

    Energy Technology Data Exchange (ETDEWEB)

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

    1966-09-01

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

  7. Research and Development Program for transportation packagings at Sandia National Laboratories

    International Nuclear Information System (INIS)

    Hohnstreiter, G.F.; Sorenson, K.B.

    1995-01-01

    This document contains information about the research and development programs dealing with waste transport at Sandia National Laboratories. This paper discusses topics such as: Why new packaging is needed; analytical methodologies and design codes;evaluation of packaging components; materials characterization; creative packaging concepts; packaging engineering and analysis; testing; and certification support

  8. Safety analysis report for packages: packaging of fissile and other radioactive materials. Final report

    International Nuclear Information System (INIS)

    Chalfant, G.G.

    1984-01-01

    The 9965, 9966, 9967, and 9968 packages are designed for surface shipment of fissile and other radioactive materials where a high degree of containment (either single or double) is required. Provisions are made to add shielding material to the packaging as required. The package was physically tested to demonstrate that it meets the criteria specified in USDOE Order No. 5480.1, chapter III, dated 5/1/81, which invokes Title 10, Code of Federal Regulations, Part 71 (10 CFR 71), Packing and Transportation of Radioactive Material, and Title 49, Code of Federal Regulations, Part 100-179, Transportation. By restricting the maximum normal operating pressure of the packages to less than 7 kg/cm 2 (gauge) (99 to 54 psig), the packages will comply with Type B(U) regulations of the International Atomic Energy Agency (IAEA) in its Regulations for the Safe Transport of Radioactive Materials, Safety Series No. 6, 1973 Revised Edition, and may be used for export and import shipments. These packages have been assessed for transport of up to 14.5 kilograms of uranium, excluding uranium-233, or 4.4 kilograms of plutonium metal, oxides, or scrap having a maximum radioactive decay energy of 30 watts. Specific maximum package contents are given. This quantity and the configuration of uranium or plutonium metal cannot be made critical by any combination of hydrogeneous reflection and moderation regardless of the condition of the package. For a uranium-233 shipment, a separate criticality evaluation for the specific package is required

  9. A waste package strategy for regulatory compliance

    International Nuclear Information System (INIS)

    Stahl, D.; Cloninger, M.O.

    1990-01-01

    This paper summarizes the strategy given in the Site Characterization Plan for demonstrating compliance with the post closure performance objectives for the waste package and the Engineered Barrier System contained in the Code of Federal Regulations. The strategy consists of the development of a conservative waste package design that will meet the regulatory requirements with sufficient margin for uncertainty using a multi-barrier approach that takes advantage of the unsaturated nature of the Yucca Mountain site. 7 refs., 1 fig

  10. Provision of transport packaging for radioactive materials

    International Nuclear Information System (INIS)

    1981-04-01

    The safe transport of radioactive materials is governed by various regulations based on International Atomic Energy Agency Regulations. This code of practice is a supplement to the regulations, its objects being (a) to advise designers of packaging on the technical features necessary to conform to the regulations, and (b) to outline the requirements for obtaining approval of package designs from the competent authority. (U.K.)

  11. AutoDipole - Automated generation of dipole subtraction terms

    International Nuclear Information System (INIS)

    Hasegawa, K.; Uwer, P.

    2009-11-01

    We present an automated generation of the subtraction terms for next-to-leading order QCD calculations in the Catani-Seymour dipole formalism. For a given scattering process with n external particles our Mathematica package generates all dipole terms, allowing for bothmassless and massive dipoles. The numerical evaluation of the subtraction terms proceeds with MadGraph, which provides Fortran code for the necessary scattering amplitudes. Checks of the numerical stability are discussed. (orig.)

  12. AutoDipole - Automated generation of dipole subtraction terms

    Energy Technology Data Exchange (ETDEWEB)

    Hasegawa, K.; Uwer, P. [Humboldt-Universitaet, Berlin (Germany). Inst. fuer Physik; Moch, S. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany)

    2009-11-15

    We present an automated generation of the subtraction terms for next-to-leading order QCD calculations in the Catani-Seymour dipole formalism. For a given scattering process with n external particles our Mathematica package generates all dipole terms, allowing for bothmassless and massive dipoles. The numerical evaluation of the subtraction terms proceeds with MadGraph, which provides Fortran code for the necessary scattering amplitudes. Checks of the numerical stability are discussed. (orig.)

  13. Numerical modelling of the long-term evolution of EDZ. Development of material models, implementation in finite-element codes, and validation

    International Nuclear Information System (INIS)

    Pudewills, A.

    2005-11-01

    Construction of deep underground structures disturbs the initial stress field in the surrounding rock. This effect can generate microcracks and alter the hydromechanical properties of the rock salt around the excavations. For the long-term performance of an underground repository in rock salt, the evolution of the 'Excavation Disturbed Zone' (EDZ) and the hydromechanical behaviour of this zone represent important issues with respect to the integrity of the geological and technical barriers. Within the framework of the NF-PRO project, WP 4.4, attention focuses on the mathematical modelling of the development and evolution of the EDZ in the rock near a disposal drift due to its relevance on the integrity of the geological and technical barriers. To perform this task, finite-element codes containing a set of time- and temperature-dependent constitutive models have been improved. A new viscoplastic constitutive model for rock salt that can describe the damage of the rock has been implemented in the finite-element codes available. The model parameters were evaluated based on experimental results. Additionally, the long-term evolution of the EDZ around a gallery in a salt mine at about 700 m below the surface was analysed and the numerical results were compared with in-situ measurements. The calculated room closure, stress distribution and the increase of rock permeability in the EDZ were compared with in-situ data, thus providing confidence in the model used. (orig.)

  14. TRUPACT-II Content Codes (TRUCON), Revision 8 and list of chemicals and materials in TRUCON (chemical list), Revision 7

    International Nuclear Information System (INIS)

    1996-03-01

    The Transuranic Package Transporter (TRUPACT-II) Content Codes document (TRUCON) represents the development of a new content code system for shipping contact handled transuranic (CH-TRU) waste in TRUPACT-II. It will be used to convert existing waste forms, content codes, and any other identification codes into a system that is uniform throughout for all the Department of Energy (DOE) sites. These various codes can be grouped under the newly formed shipping content codes without any loss of waste characterization information. The TRUCON document provides a parametric description for each content code for waste generated and compiles this information for all ten DOE sites. Compliance with waste generation, processing and certification procedures at the sites (outlined in the TRUCON document for each content code) ensures that prohibited waste forms are not present in the waste. The content code essentially gives a description of the CH-TRU waste material in terms of processes and packaging, and the generation location. This helps to provide cradle-to-grave traceability of the waste material so that the various actions required to assess its qualification as payload for the TRUPACT-II package can be performed

  15. INTRA graphical package - INTRA-Graph 1.0

    International Nuclear Information System (INIS)

    Hofman, D.; Edlund, O.

    2001-04-01

    INTRA-Graph 1.0 has been developed at Studsvik Eco and Safety AB in the frame of the European Fusion Technology Programme for application in the safety analysis using the INTRA code. INTRA-Graph 1.0 is a graphical package producing 2-dimensional plots of results generated by the INTRA code. INTRA-Graph 1.0 has been developed by extending the Grace package source code, distributed under the terms of GNU General Public License. The changes in the Grace source files are limited to provide easy updates of the INTRA-Graph when a new version of Grace will be released. The INTRA-related functionality has been implemented in new source files. The present report describes and gives complete listing of these files. The changes in the Grace source files are also described and the listing of the changed parts of the files is presented. The report gives detailed explanations and examples of files required for installation and configuration of INTRA-Graph on the different types of Unix workstations

  16. SCALE criticality safety verification and validation package

    International Nuclear Information System (INIS)

    Bowman, S.M.; Emmett, M.B.; Jordan, W.C.

    1998-01-01

    Verification and validation (V and V) are essential elements of software quality assurance (QA) for computer codes that are used for performing scientific calculations. V and V provides a means to ensure the reliability and accuracy of such software. As part of the SCALE QA and V and V plans, a general V and V package for the SCALE criticality safety codes has been assembled, tested and documented. The SCALE criticality safety V and V package is being made available to SCALE users through the Radiation Safety Information Computational Center (RSICC) to assist them in performing adequate V and V for their SCALE applications

  17. Use of computer codes for system reliability analysis

    International Nuclear Information System (INIS)

    Sabek, M.; Gaafar, M.; Poucet, A.

    1988-01-01

    This paper gives a collective summary of the studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRANTIC, FTAP, computer code package RALLY, and BOUNDS codes. Two reference study cases were executed by each code. The results obtained logic/probabilistic analysis as well as computation time are compared

  18. Modeling of Radiotherapy Linac Source Terms Using ARCHER Monte Carlo Code: Performance Comparison for GPU and MIC Parallel Computing Devices

    Science.gov (United States)

    Lin, Hui; Liu, Tianyu; Su, Lin; Bednarz, Bryan; Caracappa, Peter; Xu, X. George

    2017-09-01

    Monte Carlo (MC) simulation is well recognized as the most accurate method for radiation dose calculations. For radiotherapy applications, accurate modelling of the source term, i.e. the clinical linear accelerator is critical to the simulation. The purpose of this paper is to perform source modelling and examine the accuracy and performance of the models on Intel Many Integrated Core coprocessors (aka Xeon Phi) and Nvidia GPU using ARCHER and explore the potential optimization methods. Phase Space-based source modelling for has been implemented. Good agreements were found in a tomotherapy prostate patient case and a TrueBeam breast case. From the aspect of performance, the whole simulation for prostate plan and breast plan cost about 173s and 73s with 1% statistical error.

  19. Modeling of Radiotherapy Linac Source Terms Using ARCHER Monte Carlo Code: Performance Comparison for GPU and MIC Parallel Computing Devices

    Directory of Open Access Journals (Sweden)

    Lin Hui

    2017-01-01

    Full Text Available Monte Carlo (MC simulation is well recognized as the most accurate method for radiation dose calculations. For radiotherapy applications, accurate modelling of the source term, i.e. the clinical linear accelerator is critical to the simulation. The purpose of this paper is to perform source modelling and examine the accuracy and performance of the models on Intel Many Integrated Core coprocessors (aka Xeon Phi and Nvidia GPU using ARCHER and explore the potential optimization methods. Phase Space-based source modelling for has been implemented. Good agreements were found in a tomotherapy prostate patient case and a TrueBeam breast case. From the aspect of performance, the whole simulation for prostate plan and breast plan cost about 173s and 73s with 1% statistical error.

  20. Data Packages for the Hanford Immobilized Low-Activity Tank Waste Performance Assessment: 2001 Version

    International Nuclear Information System (INIS)

    MANN, F.M.

    2000-01-01

    Data package supporting the 2001 Immobilized Low-Activity Waste Performance Analysis. Geology, hydrology, geochemistry, facility, waste form, and dosimetry data based on recent investigation are provided. Verification and benchmarking packages for selected software codes are provided

  1. actuar: An R Package for Actuarial Science

    Directory of Open Access Journals (Sweden)

    Christophe Dutang

    2008-02-01

    Full Text Available actuar is a package providing additional Actuarial Science functionality to the R statistical system. The project was launched in 2005 and the package is available on the Comprehensive R Archive Network since February 2006. The current version of the package contains functions for use in the fields of loss distributions modeling, risk theory (including ruin theory, simulation of compound hierarchical models and credibility theory. This paper presents in detail but with few technical terms the most recent version of the package.

  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. Long-term integrity of waste package final closure for HLW geological disposal, (2). Applicability of TIG welding method to overpack final closure

    International Nuclear Information System (INIS)

    Asano, Hidekazu; Sawa, Shuusuke; Aritomi, Masanori

    2005-01-01

    Overpack, a high-level radioactive waste package for geological disposal, seals vitrified waste and in line with Japan's waste management program is required to isolate it from contact with groundwater for 1,000 years. In this study, TIG (Tungsten Inert Gas) welding method, a typical arc welding method and widely used in various industries, was examined for its applicability to seal a carbon steel overpack lid with a thickness of 190 mm. Welding conditions and welding parameters were examined for multi-layer welding in a narrow gap for four different groove depths. Weld joint tests were conducted and weld flaws, macro- and microstructure, and mechanical properties were assessed within tentatively applied criteria for weld joints. Measurement and numerical calculation for residual stress were also conducted and the tendency of residual stress distribution was discussed. These test results were compared with the basic requirements of the welding method for overpack which were pointed out in our first report. It is assessed that the TIG welding method has the potential to provide the necessary requirements to complete the final closure of overpack with a maximum thickness of 190 mm. (author)

  4. ENDF utility codes version 6.8

    International Nuclear Information System (INIS)

    McLaughlin, P.K.

    1992-01-01

    Description and operating instructions are given for a package of utility codes operating on evaluated nuclear data files in the formats ENDF-5 and ENDF-6. Included are the data checking codes CHECKER, FIZCON, PSYCHE; the code INTER for retrieving thermal cross-sections and some other data; graphical plotting codes PLOTEF, GRALIB, graphic device interface subroutine library INTLIB; and the file maintenance and retrieval codes LISTEF, SETMDC, GETMAT, STANEF. This program package which is designed for CDC, IBM, DEC and PC computers, can be obtained on magnetic tape or floppy diskette, free of charge, from the IAEA Nuclear Data Section. (author)

  5. Design Aspects of the Rayleigh Convection Code

    Science.gov (United States)

    Featherstone, N. A.

    2017-12-01

    Understanding the long-term generation of planetary or stellar magnetic field requires complementary knowledge of the large-scale fluid dynamics pervading large fractions of the object's interior. Such large-scale motions are sensitive to the system's geometry which, in planets and stars, is spherical to a good approximation. As a result, computational models designed to study such systems often solve the MHD equations in spherical geometry, frequently employing a spectral approach involving spherical harmonics. We present computational and user-interface design aspects of one such modeling tool, the Rayleigh convection code, which is suitable for deployment on desktop and petascale-hpc architectures alike. In this poster, we will present an overview of this code's parallel design and its built-in diagnostics-output package. Rayleigh has been developed with NSF support through the Computational Infrastructure for Geodynamics and is expected to be released as open-source software in winter 2017/2018.

  6. Packaging for Food Service

    Science.gov (United States)

    Stilwell, E. J.

    1985-01-01

    Most of the key areas of concern in packaging the three principle food forms for the space station were covered. It can be generally concluded that there are no significant voids in packaging materials availability or in current packaging technology. However, it must also be concluded that the process by which packaging decisions are made for the space station feeding program will be very synergistic. Packaging selection will depend heavily on the preparation mechanics, the preferred presentation and the achievable disposal systems. It will be important that packaging be considered as an integral part of each decision as these systems are developed.

  7. Waste Package Lifting Calculation

    International Nuclear Information System (INIS)

    H. Marr

    2000-01-01

    The objective of this calculation is to evaluate the structural response of the waste package during the horizontal and vertical lifting operations in order to support the waste package lifting feature design. The scope of this calculation includes the evaluation of the 21 PWR UCF (pressurized water reactor uncanistered fuel) waste package, naval waste package, 5 DHLW/DOE SNF (defense high-level waste/Department of Energy spent nuclear fuel)--short waste package, and 44 BWR (boiling water reactor) UCF waste package. Procedure AP-3.12Q, Revision 0, ICN 0, calculations, is used to develop and document this calculation

  8. Development of Level-2 PSA Technology: A Development of the Database of the Parametric Source Term for Kori Unit 1 Using the MAAP4 Code

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Chang Soon; Mun, Ju Hyun; Yun, Jeong Ick; Cho, Young Hoo; Kim, Chong Uk [Seoul National University, Seoul (Korea, Republic of)

    1997-07-15

    To quantify the severe accident source term of the parametric model method, the uncertainty of the parameters should be analyzed. Generally, to analyze the uncertainties, the cumulative distribution functions(CDF`S) of the parameters are derived. This report introduces a method of derivation of the CDF`s of the basic parameters, FCOR, FVES and FDCH. The calculation tool of the source term is the MAAP version 4.0. In the MAAP code, there are model parameters to consider an uncertain physical and/or chemical phenomenon. In general, the parameters have not a point value but a range. In this paper, considering this point, the input values of model parameters influencing each parameter are sampled using LHS. Then, the calculation results are shown in the cumulative distribution form. For a case study, the CDF`s of FCOR, FVES and FDCH of KORI unit 1 are derived. The target scenarios for the calculation are the ones whose initial events are large LOCA, small LOCA and transient, respectively. It is found that the distributions of this study are consistent to those of NUREG-1150 and are proven to be adequate in assessing the uncertainties in the severe accident source term of KORI Unit 1. 15 refs., 27 tabs., 4 figs. (author)

  9. CONTAIN code calculations of the effects on the source term of CsI to I/sub 2/ conversion due to severe hydrogen burns

    International Nuclear Information System (INIS)

    Valdez, G.D.; Williams, D.C.

    1986-01-01

    In experiments conducted at Sandia National Laboratories large amounts of elemental iodine were produced when CsI-Al 2 O 3 aerosol was exposed to hydrogen/air combustion. To evaluate some of the implications of the iodide conversion (observed to occur with up to 75% efficiency) for the severe accident source term, computational simulations of representative accident sequences were conducted with the CONTAIN code. The following conclusions can be drawn from this preliminary source term assessment: (1) If the containment sprays are inoperative during the accident, or failed by the hydrogen burn, the late-time source term is almost tripled when the iodide is converted to I 2 . (2) With the sprays active, the amount released without conversion of the CsI aerosol is 63% higher than for the case when conversion occurs. (3) For the case where CsI is converted to I 2 continued operation of the sprays reduces the release by a factor of 40, relative to the case in which the sprays fail at the time of the hydrogen burn. When there is no conversion, the reduction factor for continued spray operation is about a factor of 9, relative to the failed spray case

  10. 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

  11. Short-term calculations to supplement the RS 16 B PWR experiments with internals (PWR1 to PWR5), using the LECK 4 computer code

    International Nuclear Information System (INIS)

    Hughes, G.; Mueller, R.

    1980-03-01

    Within the framework of research project RS 16 B sponsored by the German BMFT a series of a blowdown experiments, DWR1 to DWR5, were performed using a vessel with dummy internals under conditions similar to those in a PWR. The prime objective of these experiments was the investigation of the highly transient blowdown phenomena in the discharge nozzle and the determination of the induced loads on the internals. As a partner in the project, KWU carried out both pre-test predictions and post-test analyses of these experiments using, among others, the computer code LECK 4. For the most severe blowdown test DWR5, the influence of the most important model parameters on the blowdown analysis was investigated in detail. These investigations suggest that, similar to the long-term analyses, calculations using the homogeneous critical flow model would improve agreement between calculation and experiment. (orig./RW) [de

  12. Merganser Download Package

    Data.gov (United States)

    U.S. Environmental Protection Agency — This data download package contains an Esri 10.0 MXD, file geodatabase and copy of this FGDC metadata record. The data in this package are used in support of the...

  13. Creative Thinking Package

    Science.gov (United States)

    Jones, Clive

    1972-01-01

    A look at the latest package from a British managment training organization, which explains and demonstrates creative thinking techniques, including brainstorming. The package, designed for groups of twelve or more, consists of tapes, visuals, and associated exercises. (Editor/JB)

  14. Bituminous reference document: synthesis of knowledge on the long time behavior of bituminous packages; Dossier de reference bitume: synthese des connaissances sur le comportement a long terme des colis bitumes

    Energy Technology Data Exchange (ETDEWEB)

    Sercombe, J.; Adenot, F.; Vistoli, P.P.; Parraud, S.; Riglet-Martial, C.; Gwinner, B.; Felines, I.; Tiffreau, C.; Libert, M

    2004-07-01

    This document is a synthesis of the knowledge acquired at the CEA on the behavior of bituminous packages. In this framework, the CEA studied bituminous packages in generic conditions of a package lifetime. The main factors, the evolution mechanisms and influential parameters have been determined and quantitative s simulation have been developed. After a description of the main initial bituminous packages characterizations, the evolutions in saturated and un-saturated environment are exposed. (A.L.B.)

  15. Packaging Printing Today

    OpenAIRE

    Stanislav Bolanča; Igor Majnarić; Kristijan Golubović

    2015-01-01

    Printing packaging covers today about 50% of all the printing products. Among the printing products there are printing on labels, printing on flexible packaging, printing on folding boxes, printing on the boxes of corrugated board, printing on glass packaging, synthetic and metal ones. The mentioned packaging are printed in flexo printing technique, offset printing technique, intaglio halftone process, silk – screen printing, ink ball printing, digital printing and hybrid printing process. T...

  16. Genome packaging in viruses

    OpenAIRE

    Sun, Siyang; Rao, Venigalla B.; Rossmann, Michael G.

    2010-01-01

    Genome packaging is a fundamental process in a viral life cycle. Many viruses assemble preformed capsids into which the genomic material is subsequently packaged. These viruses use a packaging motor protein that is driven by the hydrolysis of ATP to condense the nucleic acids into a confined space. How these motor proteins package viral genomes had been poorly understood until recently, when a few X-ray crystal structures and cryo-electron microscopy structures became available. Here we discu...

  17. Statistical package for improved analysis of hillslope monitoring data collected as part of the Board of Forestry's long-term monitoring program

    Science.gov (United States)

    Jack Lewis; Jim Baldwin

    1997-01-01

    The State of California has embarked upon a Long-Term Monitoring Program whose primary goal is to assess the effectiveness of the Forest Practice Rules and Review Process in protecting the beneficial uses of waters from the impacts of timber operations on private timberlands. The Board of Forestry's Monitoring Study Group concluded that hillslope monitoring should...

  18. Work package 1B.2 under the European Commission: Integrated wind turbine design (UPWIND): Specification of long-term load measurement technique

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt Paulsen, U.; Cutululis, N.; Soerensen, Poul

    2007-02-15

    The present report is the 12 month intermediate report of the UPWIND WP1B2 transmission and conversion. It describes the developed measurement technique for long-term load measurement technique, presents the hardware details, type of sensors and location, data storage and data analysis technique to verify design load assumptions. (au)

  19. PRIDE Surveillance Projects Data Packaging Project, Information Package Specification Version 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Kelleher, D.M.; Shipp, R. L.; Mason, J. D.

    2009-09-28

    This document contains a specification for a standard XML document format called an information package that can be used to store information and the context required to understand and use that information in information management systems and other types of information archives. An information package consists of packaged information, a set of information metadata that describes the packaged information, and an XML signature that protects the packaged information. The information package described in this specification was designed to be used to store Department of Energy (DOE) and National Nuclear Security Administration (NNSA) information and includes the metadata required for that information: a unique package identifier, information marking that conforms to DOE and NNSA requirements, and access control metadata. Information package metadata can also include information search terms, package history, and notes. Packaged information can be text content, binary content, and the contents of files and other containers. A single information package can contain multiple types of information. All content not in a text form compatible with XML must be in a text encoding such as base64. Package information is protected by a digital XML signature that can be used to determine whether the information has changed since it was signed and to identify the source of the information. This specification has been tested but has not been used to create production information packages. The authors expect that gaps and unclear requirements in this specification will be identified as this specification is used to create information packages and as information stored in information packages is used. The authors expect to issue revised versions of this specification as needed to address these issues.

  20. The drift-flux correlation package MDS

    International Nuclear Information System (INIS)

    Hoeld, A.

    2001-01-01

    Based on the SONNENBURG drift-flux correlation, developed at GRS/Garching (Germany), a comprehensive drift-flux correlation package (MDS) has been established. Its aim is to support thermal-hydraulic mixture-fluid models, models being used for the simulation of the steady state and transient behaviour of characteristic thermal-hydraulic parameters of single- or two-phase fluids flowing along coolant channels of different types (being, e.g., parts of NPP-s, steam generators etc.). The characteristic properties of this package with respect to the behaviour at co- and counter-current flow, its inverse solutions needed for steady state simulations, its behaviour when approaching the lower or upper boundary of a two-phase region, its verification and behaviour with respect to other correlations will be discussed. An adequate driver code, MDSDRI, has been established too, allowing to test the package very thoroughly out of the complex thermal-hydraulic codes. (author)

  1. The drift-flux correlation package MDS

    Energy Technology Data Exchange (ETDEWEB)

    Hoeld, A. [Bernaysstr. 16A, Munich, F.R. (Germany)

    2001-07-01

    Based on the SONNENBURG drift-flux correlation, developed at GRS/Garching (Germany), a comprehensive drift-flux correlation package (MDS) has been established. Its aim is to support thermal-hydraulic mixture-fluid models, models being used for the simulation of the steady state and transient behaviour of characteristic thermal-hydraulic parameters of single- or two-phase fluids flowing along coolant channels of different types (being, e.g., parts of NPP-s, steam generators etc.). The characteristic properties of this package with respect to the behaviour at co- and counter-current flow, its inverse solutions needed for steady state simulations, its behaviour when approaching the lower or upper boundary of a two-phase region, its verification and behaviour with respect to other correlations will be discussed. An adequate driver code, MDSDRI, has been established too, allowing to test the package very thoroughly out of the complex thermal-hydraulic codes. (author)

  2. Relevance of biotic pathways to the long-term regulation of nuclear waste disposal. Estimation of radiation dose to man resulting from biotic transport: the BIOPORT/MAXI1 software package. Volume 5

    Energy Technology Data Exchange (ETDEWEB)

    McKenzie, D.H.; Cadwell, L.L.; Gano, K.A.; Kennedy, W.E. Jr.; Napier, B.A.; Peloquin, R.A.; Prohammer, L.A.; Simmons, M.A.

    1985-10-01

    BIOPORT/MAXI1 is a collection of five computer codes designed to estimate the potential magnitude of the radiation dose to man resulting from biotic transport processes. Dose to man is calculated for ingestion of agricultural crops grown in contaminated soil, inhalation of resuspended radionuclides, and direct exposure to penetrating radiation resulting from the radionuclide concentrations established in the available soil surface by the biotic transport model. This document is designed as both an instructional and reference document for the BIOPORT/MAXI1 computer software package and has been written for two major audiences. The first audience includes persons concerned with the mathematical models of biological transport of commercial low-level radioactive wastes and the computer algorithms used to implement those models. The second audience includes persons concerned with exercising the computer program and exposure scenarios to obtain results for specific applications. The report contains sections describing the mathematical models, user operation of the computer programs, and program structure. Input and output for five sample problems are included. In addition, listings of the computer programs, data libraries, and dose conversion factors are provided in appendices.

  3. Relevance of biotic pathways to the long-term regulation of nuclear waste disposal. Estimation of radiation dose to man resulting from biotic transport: the BIOPORT/MAXI1 software package. Volume 5

    International Nuclear Information System (INIS)

    McKenzie, D.H.; Cadwell, L.L.; Gano, K.A.; Kennedy, W.E. Jr.; Napier, B.A.; Peloquin, R.A.; Prohammer, L.A.; Simmons, M.A.

    1985-10-01

    BIOPORT/MAXI1 is a collection of five computer codes designed to estimate the potential magnitude of the radiation dose to man resulting from biotic transport processes. Dose to man is calculated for ingestion of agricultural crops grown in contaminated soil, inhalation of resuspended radionuclides, and direct exposure to penetrating radiation resulting from the radionuclide concentrations established in the available soil surface by the biotic transport model. This document is designed as both an instructional and reference document for the BIOPORT/MAXI1 computer software package and has been written for two major audiences. The first audience includes persons concerned with the mathematical models of biological transport of commercial low-level radioactive wastes and the computer algorithms used to implement those models. The second audience includes persons concerned with exercising the computer program and exposure scenarios to obtain results for specific applications. The report contains sections describing the mathematical models, user operation of the computer programs, and program structure. Input and output for five sample problems are included. In addition, listings of the computer programs, data libraries, and dose conversion factors are provided in appendices

  4. “All Inclusive” Package Travel

    Directory of Open Access Journals (Sweden)

    Nicola Soldati

    2011-01-01

    Full Text Available Travel and tourism contracts are regulated under Articles from 82 to 100 of the Italian Consumer Code. Those articles are specifically dedicated to the regulation of tourist services, providing an accurate and precise definition of "tourist packages". This paper presents a study of travel and tourism under Italian law.

  5. Computer Package for Graphite Total Cross-Section Calculations

    International Nuclear Information System (INIS)

    Adib, M.; Fathalla, M.

    2008-01-01

    An additive formula is given which allows calculating the contribution of the total neut.>neutron transmission through crystalline graphite. The formula takes into account the graphite form of poly or pyrolytic crystals and its parameters. Computer package Graphite has been designed in order to provide the required calculations in the neutron energy range from 0.1 MeV to 10 eV. The package includes three codes: PCG (Polycrystalline Graphite), PG (Pyrolytic Graphite) and HOPG (Highly Oriented Pyrolytic Graphite) for calculating neutron transmission through fine graphite powder (polycrystalline), neutron transmission and removal coefficient of PG crystal in terms of its mosaic spread for neutrons incident along its c-axis and the transmission of neutrons incident on HOPG crystal at different angles, respectively. For comparison of the experimental neutron transmission data with the calculated values, the program takes into consideration the effect of both wavelength and neutron beam divergence in either 2 constant wavelength spread mode (δλ=constant) or constant wavelength resolution mode (δλ/λ=constant). In order to check the validity for application of computer package Graphite in cross-section calculations, a comparison between calculated values with the available experimental data were carried out. An overall agreement is indicated with an accuracy sufficient for determine the neutron transmission characteristics

  6. Development of waste packages for tuff

    International Nuclear Information System (INIS)

    Rothman, A.J.

    1982-01-01

    The objective of this program is to develop nuclear waste packages that meet the Nuclear Regulatory Commission's requirements for a licensed repository in tuff at the Nevada Test Site. Selected accomplishments for FY82 are: (1) Selection, collection of rock, and characterization of suitable outcrops (for lab experiments); (2) Rock-water interactions (Bullfrog Tuff); (3) Corrosion tests of ferrous metals; (4) Thermal modeling of waste package in host rock; (5) Preliminary fabrication tests of alternate backfills (crushed tuff); (6) Reviewed Westinghouse conceptual waste package designs for tuff and began modification for unsaturated zone; and (7) Waste Package Codes (BARIER and WAPPA) now running on our computer. Brief discussions are presented for rock-water interactions, corrosion tests of ferrous metals, and thermal and radionuclide migration modelling

  7. Active Packaging Coatings

    Directory of Open Access Journals (Sweden)

    Luis J. Bastarrachea

    2015-11-01

    Full Text Available Active food packaging involves the packaging of foods with materials that provide an enhanced functionality, such as antimicrobial, antioxidant or biocatalytic functions. This can be achieved through the incorporation of active compounds into the matrix of the commonly used packaging materials, or by the application of coatings with the corresponding functionality through surface modification. The latter option offers the advantage of preserving the packaging materials’ bulk properties nearly intact. Herein, different coating technologies like embedding for controlled release, immobilization, layer-by-layer deposition, and photografting are explained and their potential application for active food packaging is explored and discussed.

  8. Contribution and limits of geochemical calculation codes to evaluate the long term behavior of nuclear waste glasses; Apports et limites des modeles geochimiques pour l'evaluation du comportement a long terme des verres de confinement des dechets radioactifs

    Energy Technology Data Exchange (ETDEWEB)

    Fritz, B; Crovisier, J L [Universite Louis Pasteur, Centre de Geochimie de la Surface, CNRS ULP, Ecole et Observatoire des Sciences de la Terre, 67 - Strasbourg (France)

    1997-07-01

    Geochemical models have been intensively developed by researchers since more than twenty five years in order to be able to better understand and/or predict the long term stability/instability of water-rock systems. These geochemical codes were ail built first on a thermodynamic approach deriving from the application of Mass Action Law. The resulting first generation of models allowed to detect or predict the possible mass transfers (thermodynamic models) between aqueous and mineral phases including irreversible dissolutions of primary minerals and/or precipitation near equilibrium of secondary mineral phases. The recent development of models based on combined thermodynamics and kinetics opens the field of Lime dependent reactions prediction. This is crucial if one thinks to combine geochemical and hydrological studies in the so-called coupled models for transport and reaction calculations. All these models are progressively applied to the prediction of long term behavior of mineral phases, and more specifically glasses. In order to succeed in chat specific extension of the models, but also the data bases, there is a great need for additional new data from experimental approaches and from natural analogues. The modelling approach appears than also very useful in order to interpret the results of experimental data and to relate them to long term data extracted from natural analogues. Specific functions for modelling solid solution phases mat' also be used for describing the products of glasses alterations. (authors)

  9. Advancements in meat packaging.

    Science.gov (United States)

    McMillin, Kenneth W

    2017-10-01

    Packaging of meat provides the same or similar benefits for raw chilled and processed meats as other types of food packaging. Although air-permeable packaging is most prevalent for raw chilled red meat, vacuum and modified atmosphere packaging offer longer shelf life. The major advancements in meat packaging have been in the widely used plastic polymers while biobased materials and their integration into composite packaging are receiving much attention for functionality and sustainability. At this time, active and intelligent packaging are not widely used for antioxidant, antimicrobial, and other functions to stabilize and enhance meat properties although many options are being developed and investigated. The advances being made in nanotechnology will be incorporated into food packaging and presumably into meat packaging when appropriate and useful. Intelligent packaging using sensors for transmission of desired information and prompting of subsequent changes in packaging materials, environments or the products to maintain safety and quality are still in developmental stages. Copyright © 2017 Elsevier Ltd. All rights reserved.

  10. Edible packaging materials.

    Science.gov (United States)

    Janjarasskul, Theeranun; Krochta, John M

    2010-01-01

    Research groups and the food and pharmaceutical industries recognize edible packaging as a useful alternative or addition to conventional packaging to reduce waste and to create novel applications for improving product stability, quality, safety, variety, and convenience for consumers. Recent studies have explored the ability of biopolymer-based food packaging materials to carry and control-release active compounds. As diverse edible packaging materials derived from various by-products or waste from food industry are being developed, the dry thermoplastic process is advancing rapidly as a feasible commercial edible packaging manufacturing process. The employment of nanocomposite concepts to edible packaging materials promises to improve barrier and mechanical properties and facilitate effective incorporation of bioactive ingredients and other designed functions. In addition to the need for a more fundamental understanding to enable design to desired specifications, edible packaging has to overcome challenges such as regulatory requirements, consumer acceptance, and scaling-up research concepts to commercial applications.

  11. Waste Form Release Data Package for the 2001 Immobilized Low-Activity Waste Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    McGrail, B. Peter; Icenhower, Jonathan P.; Martin, Paul F.; Schaef, Herbert T.; O' Hara, Matthew J.; Rodriguez, Eugenio; Steele, Jackie L.

    2001-02-01

    This data package documents the experimentally derived input data on the representative waste glasses LAWABP1 and HLP-31 that will be used for simulations of the immobilized lowactivity waste disposal system with the Subsurface Transport Over Reactive Multiphases (STORM) code. The STORM code will be used to provide the near-field radionuclide release source term for a performance assessment to be issued in March of 2001. Documented in this data package are data related to 1) kinetic rate law parameters for glass dissolution, 2) alkali-H ion exchange rate, 3) chemical reaction network of secondary phases that form in accelerated weathering tests, and 4) thermodynamic equilibrium constants assigned to these secondary phases. The kinetic rate law and Na+-H+ ion exchange rate were determined from single-pass flow-through experiments. Pressurized unsaturated flow and vapor hydration experiments were used for accelerated weathering or aging of the glasses. The majority of the thermodynamic data were extracted from the thermodynamic database package shipped with the geochemical code EQ3/6. However, several secondary reaction products identified from laboratory tests with prototypical LAW glasses were not included in this database, nor are the thermodynamic data available in the open literature. One of these phases, herschelite, was determined to have a potentially significant impact on the release calculations and so a solubility product was estimated using a polymer structure model developed for zeolites. Although this data package is relatively complete, final selection of ILAW glass compositions has not been done by the waste treatment plant contractor. Consequently, revisions to this data package to address new ILAW glass formulations are to be regularly expected.

  12. NHRIC (National Health Related Items Code)

    Data.gov (United States)

    U.S. Department of Health & Human Services — The National Health Related Items Code (NHRIC) is a system for identification and numbering of marketed device packages that is compatible with other numbering...

  13. NORTICA - a new code for cyclotron analysis

    International Nuclear Information System (INIS)

    Gorelov, D.; Johnson, D.; Marti, F.

    2001-01-01

    The new package NORTICA (Numerical ORbit Tracking In Cyclotrons with Analysis) of computer codes for beam dynamics simulations is under development at NSCL. The package was started as a replacement for the code MONSTER developed in the laboratory in the past. The new codes are capable of beam dynamics simulations in both CCF (Coupled Cyclotron Facility) accelerators, the K500 and K1200 superconducting cyclotrons. The general purpose of this package is assisting in setting and tuning the cyclotrons taking into account the main field and extraction channel imperfections. The computer platform for the package is Alpha Station with UNIX operating system and X-Windows graphic interface. A multiple programming language approach was used in order to combine the reliability of the numerical algorithms developed over the long period of time in the laboratory and the friendliness of modern style user interface. This paper describes the capability and features of the codes in the present state

  14. Software infrastructure progress in the RAVEN code

    Energy Technology Data Exchange (ETDEWEB)

    Cogliati, Joshua J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Permann, Cody J. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    The milestones have been achieved. RAVEN has been migrated to Gitlab which adds new abilities for code review and management. Standalone RAVEN framework packages have been created for OSX and two Linux distributions.

  15. Vegetation cover and long-term conservation of radioactive waste packages: the case study of the CSM waste disposal facility (Manche District, France).

    Science.gov (United States)

    Petit-Berghem, Yves; Lemperiere, Guy

    2012-03-01

    The CSM is the first French waste disposal facility for radioactive waste. Waste material is buried several meters deep and protected by a multi-layer cover, and equipped with a drainage system. On the surface, the plant cover is a grassland vegetation type. A scientific assessment has been carried out by the Géophen laboratory, University of Caen, in order to better characterize the plant cover (ecological groups and associated soils) and to observe its medium and long term evolution. Field assessments made on 10 plots were complemented by laboratory analyses carried out over a period of 1 year. The results indicate scenarios and alternative solutions which could arise, in order to passively ensure the long-term safety of the waste disposal system. Several proposals for a blanket solution are currently being studied and discussed, under the auspices of international research institutions in order to determine the most appropriate materials for the storage conditions. One proposal is an increased thickness of these materials associated with a geotechnical barrier since it is well adapted to the forest plants which are likely to colonize the site. The current experiments that are carried out will allow to select the best option and could provide feedback for other waste disposal facility sites already being operated in France (CSFMA waste disposal facility, Aube district) or in other countries.

  16. STRUCTURAL CALCULATION OF AN EMPLACEMENT PALLET STATICALLY LOADED BY A WASTE PACKAGE

    International Nuclear Information System (INIS)

    S. Mastilovic

    2000-01-01

    The purpose of this calculation is to determine the structural response of the emplacement pallet (EP) subjected to static load from the mounted waste package (WP). The scope of this document is limited to reporting the calculation results in terms of stress intensity magnitudes. This calculation is associated with the waste emplacement systems design; calculations are performed by the Waste Package Design group. AP-3.12Q, Revision 0, ICN 0, Calculations, is used to perform the calculation and develop the document. The finite element solutions are performed by using the commercially available ANSYS Version (V) 5.4 finite element code. The results of these calculations are provided in terms of maximum stress intensity magnitudes

  17. Radioactive material package seal tests

    International Nuclear Information System (INIS)

    Madsen, M.M.; Humphreys, D.L.; Edwards, K.R.

    1990-01-01

    General design or test performance requirements for radioactive materials (RAM) packages are specified in Title 10 of the US Code of Federal Regulations Part 71 (US Nuclear Regulatory Commission, 1983). The requirements for Type B packages provide a broad range of environments under which the system must contain the RAM without posing a threat to health or property. Seals that provide the containment system interface between the packaging body and the closure must function in both high- and low-temperature environments under dynamic and static conditions. A seal technology program, jointly funded by the US Department of Energy Office of Environmental Restoration and Waste Management (EM) and the Office of Civilian Radioactive Waste Management (OCRWM), was initiated at Sandia National Laboratories. Experiments were performed in this program to characterize the behavior of several static seal materials at low temperatures. Helium leak tests on face seals were used to compare the materials. Materials tested include butyl, neoprene, ethylene propylene, fluorosilicone, silicone, Eypel, Kalrez, Teflon, fluorocarbon, and Teflon/silicone composites. Because most elastomer O-ring applications are for hydraulic systems, manufacturer low-temperature ratings are based on methods that simulate this use. The seal materials tested in this program with a fixture similar to a RAM cask closure, with the exception of silicone S613-60, are not leak tight (1.0 x 10 -7 std cm 3 /s) at manufacturer low-temperature ratings. 8 refs., 3 figs., 1 tab

  18. Hazardous Material Packaging and Transportation

    Energy Technology Data Exchange (ETDEWEB)

    Hypes, Philip A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-02-04

    This is a student training course. Some course objectives are to: recognize and use standard international and US customary units to describe activities and exposure rates associated with radioactive material; determine whether a quantity of a single radionuclide meets the definition of a class 7 (radioactive) material; determine, for a given single radionuclide, the shipping quantity activity limits per 49 Code of Federal Regulations (CFR) 173.435; determine the appropriate radioactive material hazard class proper shipping name for a given material; determine when a single radionuclide meets the DOT definition of a hazardous substance; determine the appropriate packaging required for a given radioactive material; identify the markings to be placed on a package of radioactive material; determine the label(s) to apply to a given radioactive material package; identify the entry requirements for radioactive material labels; determine the proper placement for radioactive material label(s); identify the shipping paper entry requirements for radioactive material; select the appropriate placards for a given radioactive material shipment or vehicle load; and identify allowable transport limits and unacceptable transport conditions for radioactive material.

  19. Work Package 1B.2 under the European Commission, Integrated Wind Turbine Design (UPWIND): Verification of long-term load measurement technique

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt Paulsen, U.

    2011-02-15

    The present report is the final effort of tasks carried out under UPWIND WP1B2 transmission and conversion, which describes: 1) results and recommendations developed in the course of developing the long-term load measurement technique. 2) the hardware details, type of sensors and location, data storage and. 3) data analysis technique to verify design load assumptions. The work is carried out under Contract no 019945 (SES6) 'UPWIND' within the European Commission The interaction between the mechanical and electrical generator subsystems is described rudimentarily, based primarily on HAWC2 simulations below stall of the mechanical system with simple generator and gearbox systems. The electrical system simulations were not carried out as intended in DOW[2], but indications of the conditions for establishing the interaction have been described by measurements and by argument, that this might have an effect as indicated. The hypothesis stating, that the power signal can be utilized as a basic signal for extended analysis of mechanical as well as electrical power signal with static and dynamic features, has been demonstrated on performance and dynamic bandwidth capability. It is however from present analysis obvious that improved signal conditions could be achieved with other mechanical joint solutions than with the present torque signal as measured with the cardan joint. For the reasons mentioned, the comparison with a signal showing the mechanical properties could be improved, with a likely gain on the accuracy as a result. (Author)

  20. Estimation and analysis of the short-term variations of multi-GNSS receiver differential code biases using global ionosphere maps

    Science.gov (United States)

    Li, Min; Yuan, Yunbin; Wang, Ningbo; Liu, Teng; Chen, Yongchang

    2017-12-01

    Care should be taken to minimize the adverse impact of differential code biases (DCBs) on global navigation satellite systems (GNSS)-derived ionospheric information determinations. For the sake of convenience, satellite and receiver DCB products provided by the International GNSS Service (IGS) are treated as constants over a period of 24 h (Li et al. (2014)). However, if DCB estimates show remarkable intra-day variability, the DCBs estimated as constants over 1-day period will partially account for ionospheric modeling error; in this case DCBs will be required to be estimated over shorter time period. Therefore, it is important to further gain insight into the short-term variation characteristics of receiver DCBs. In this contribution, the IGS combined global ionospheric maps and the German Aerospace Center (DLR)-provided satellite DCBs are used in the improved method to determine the multi-GNSS receiver DCBs with an hourly time resolution. The intra-day stability of the receiver DCBs is thereby analyzed in detail. Based on 1 month of data collected within the multi-GNSS experiment of the IGS, a good agreement within the receiver DCBs is found between the resulting receiver DCB estimates and multi-GNSS DCB products from the DLR at a level of 0.24 ns for GPS, 0.28 ns for GLONASS, 0.28 ns for BDS, and 0.30 ns for Galileo. Although most of the receiver DCBs are relatively stable over a 1-day period, large fluctuations (more than 9 ns between two consecutive hours) within the receiver DCBs can be found. We also demonstrate the impact of the significant short-term variations in receiver DCBs on the extraction of ionospheric total electron content (TEC), at a level of 12.96 TECu (TEC unit). Compared to daily receiver DCB estimates, the hourly DCB estimates obtained from this study can reflect the short-term variations of the DCB estimates more dedicatedly. The main conclusion is that preliminary analysis of characteristics of receiver DCB variations over short-term

  1. Grandfathering of competent authority approved packages

    International Nuclear Information System (INIS)

    Osgood, N.L.

    2004-01-01

    International Atomic Energy Agency transportation regulations are reviewed and revised on a periodic basis as new technical and scientific information becomes available. The 1996 Edition of the Regulations for the Safe Transport of Radioactive Materials in TS-R-1 includes provisions for the use of package designs approved to previous editions of the regulations. This assures that there is no disruption of transport when the regulations are updated and revised. The revision of package design standards may make certain designs obsolete, though not necessarily unsafe. The U.S. Nuclear Regulatory Commission is the agency in the United States that certifies transportation packages for Type B and fissile materials. NRC regulations include grandfathering provisions that are comparable to and compatible with the IAEA standards. NRC staff is promoting a new system that would eliminate the need to grandfather package designs. Under the new method, any new or revised provision of the regulations that affects package standards would include its own transitional arrangements. In this way, each change would be evaluated for its safety importance. Changes in the package standards that are important to safety would be implemented immediately upon the regulations coming into force. Other changes, that do not significantly affect safety, would have longer implementation periods. In this way, all packages in use would be compatible with the regulations in force, and no specific grandfathering provisions for older designs would be needed. NRC staff has concluded that the package design standards are mature and have been shown to be protective over the past 40 years of shipping experience. We predict that future changes in package design standards will not be substantive in terms of resulting in significant changes in physical performance of a package in transport, including actual transportation accidents. The benefits of the new system would be a more predictable regulatory structure

  2. Inspection, testing, and operating requiremens for the packaging and shipping of uranium trioxide in 55-gallon Department of Transportation (DOT) Specification 6M shipping packagings

    International Nuclear Information System (INIS)

    Toomer, D.V.

    1991-06-01

    This document identifies the inspection, testing and operating requirements for the packaging, loading, and shipping of uranium trioxide (UO 3 ) in 55-gallon DOT Specification 6M shipping packagings from the Idaho Chemical Processing Plant (ICPP). Compliance with this document assures established controls for the purchasing, packaging, loading, and shipping of DOT Specification 6M shipping packagings are maintained in strict accordance with applicable Code of Federal Regulations (CFRs) and Department of Energy (DOE) Orders. 7 refs., 3 figs., 1 tab

  3. Functional anthology of intrinsic disorder. 2. Cellular components, domains, technical terms, developmental processes, and coding sequence diversities correlated with long disordered regions.

    Science.gov (United States)

    Vucetic, Slobodan; Xie, Hongbo; Iakoucheva, Lilia M; Oldfield, Christopher J; Dunker, A Keith; Obradovic, Zoran; Uversky, Vladimir N

    2007-05-01

    Biologically active proteins without stable ordered structure (i.e., intrinsically disordered proteins) are attracting increased attention. Functional repertoires of ordered and disordered proteins are very different, and the ability to differentiate whether a given function is associated with intrinsic disorder or with a well-folded protein is crucial for modern protein science. However, there is a large gap between the number of proteins experimentally confirmed to be disordered and their actual number in nature. As a result, studies of functional properties of confirmed disordered proteins, while helpful in revealing the functional diversity of protein disorder, provide only a limited view. To overcome this problem, a bioinformatics approach for comprehensive study of functional roles of protein disorder was proposed in the first paper of this series (Xie, H.; Vucetic, S.; Iakoucheva, L. M.; Oldfield, C. J.; Dunker, A. K.; Obradovic, Z.; Uversky, V. N. Functional anthology of intrinsic disorder. 1. Biological processes and functions of proteins with long disordered regions. J. Proteome Res. 2007, 5, 1882-1898). Applying this novel approach to Swiss-Prot sequences and functional keywords, we found over 238 and 302 keywords to be strongly positively or negatively correlated, respectively, with long intrinsically disordered regions. This paper describes approximately 90 Swiss-Prot keywords attributed to the cellular components, domains, technical terms, developmental processes, and coding sequence diversities possessing strong positive and negative correlation with long disordered regions.

  4. Functional Anthology of Intrinsic Disorder. II. Cellular Components, Domains, Technical Terms, Developmental Processes and Coding Sequence Diversities Correlated with Long Disordered Regions

    Science.gov (United States)

    Vucetic, Slobodan; Xie, Hongbo; Iakoucheva, Lilia M.; Oldfield, Christopher J.; Dunker, A. Keith; Obradovic, Zoran; Uversky, Vladimir N.

    2008-01-01

    Biologically active proteins without stable ordered structure (i.e., intrinsically disordered proteins) are attracting increased attention. Functional repertoires of ordered and disordered proteins are very different, and the ability to differentiate whether a given function is associated with intrinsic disorder or with a well-folded protein is crucial for modern protein science. However, there is a large gap between the number of proteins experimentally confirmed to be disordered and their actual number in nature. As a result, studies of functional properties of confirmed disordered proteins, while helpful in revealing the functional diversity of protein disorder, provide only a limited view. To overcome this problem, a bioinformatics approach for comprehensive study of functional roles of protein disorder was proposed in the first paper of this series (Xie H., Vucetic S., Iakoucheva L.M., Oldfield C.J., Dunker A.K., Obradovic Z., Uversky V.N. (2006) Functional anthology of intrinsic disorder. I. Biological processes and functions of proteins with long disordered regions. J. Proteome Res.). Applying this novel approach to Swiss-Prot sequences and functional keywords, we found over 238 and 302 keywords to be strongly positively or negatively correlated, respectively, with long intrinsically disordered regions. This paper describes ~90 Swiss-Prot keywords attributed to the cellular components, domains, technical terms, developmental processes and coding sequence diversities possessing strong positive and negative correlation with long disordered regions. PMID:17391015

  5. Lithium niobate packaging challenges

    International Nuclear Information System (INIS)

    Murphy, E.J.; Holmes, R.J.; Jander, R.B.; Schelling, A.W.

    1988-01-01

    The use of lithium niobate integrated optic devices outside of the research laboratory is predicated on the development of a sound packaging method. The authors present a discussion of the many issues that face the development of a viable, robust packaging technology. The authors emphasize the interaction of lithium niobate's physical properties with available packaging materials and technologies. The broad range of properties (i.e. electro-optic, piezo-electric, pyro-electric, photorefractive...) that make lithium niobate an interesting material in many device applications also make it a packaging challenge. The package design, materials and packaging technologies must isolate the device from the environment so that lithium niobate's properties do not adversely affect the device performance

  6. Packaged die heater

    Science.gov (United States)

    Spielberger, Richard; Ohme, Bruce Walker; Jensen, Ronald J.

    2011-06-21

    A heater for heating packaged die for burn-in and heat testing is described. The heater may be a ceramic-type heater with a metal filament. The heater may be incorporated into the integrated circuit package as an additional ceramic layer of the package, or may be an external heater placed in contact with the package to heat the die. Many different types of integrated circuit packages may be accommodated. The method provides increased energy efficiency for heating the die while reducing temperature stresses on testing equipment. The method allows the use of multiple heaters to heat die to different temperatures. Faulty die may be heated to weaken die attach material to facilitate removal of the die. The heater filament or a separate temperature thermistor located in the package may be used to accurately measure die temperature.

  7. Packaging for Sustainability

    CERN Document Server

    Lewis, Helen; Fitzpatrick, Leanne

    2012-01-01

    The packaging industry is under pressure from regulators, customers and other stakeholders to improve packaging’s sustainability by reducing its environmental and societal impacts. This is a considerable challenge because of the complex interactions between products and their packaging, and the many roles that packaging plays in the supply chain. Packaging for Sustainability is a concise and readable handbook for practitioners who are trying to implement sustainability strategies for packaging. Industry case studies are used throughout the book to illustrate possible applications and scenarios. Packaging for Sustainability draws on the expertise of researchers and industry practitioners to provide information on business benefits, environmental issues and priorities, environmental evaluation tools, design for environment, marketing strategies, and challenges for the future.

  8. Cigarette package design: opportunities for disease prevention.

    Science.gov (United States)

    Difranza, J R; Clark, D M; Pollay, R W

    2002-06-15

    To learn how cigarette packages are designed and to determine to what extent cigarette packages are designed to target children. A computer search was made of all Internet websites that post tobacco industry documents using the search terms: packaging, package design, package study, box design, logo, trademark and design study. All documents were retrieved electronically and analyzed by the first author for recurrent themes. Cigarette manufacturers devote a great deal of attention and expense to package design because it is central to their efforts to create brand images. Colors, graphic elements, proportioning, texture, materials and typography are tested and used in various combinations to create the desired product and user images. Designs help to create the perceived product attributes and project a personality image of the user with the intent of fulfilling the psychological needs of the targeted type of smoker. The communication of these images and attributes is conducted through conscious and subliminal processes. Extensive testing is conducted using a variety of qualitative and quantitative research techniques. The promotion of tobacco products through appealing imagery cannot be stopped without regulating the package design. The same marketing research techniques used by the tobacco companies can be used to design generic packaging and more effective warning labels targeted at specific consumers.

  9. Cigarette package design: opportunities for disease prevention

    Directory of Open Access Journals (Sweden)

    Pollay RW

    2003-01-01

    Full Text Available Abstract Objective To learn how cigarette packages are designed and to determine to what extent cigarette packages are designed to target children. Methods A computer search was made of all Internet websites that post tobacco industry documents using the search terms: packaging, package design, package study, box design, logo, trademark and design study. All documents were retrieved electronically and analyzed by the first author for recurrent themes. Data Synthesis Cigarette manufacturers devote a great deal of attention and expense to package design because it is central to their efforts to create brand images. Colors, graphic elements, proportioning, texture, materials and typography are tested and used in various combinations to create the desired product and user images. Designs help to create the perceived product attributes and project a personality image of the user with the intent of fulfilling the psychological needs of the targeted type of smoker. The communication of these images and attributes is conducted through conscious and subliminal processes. Extensive testing is conducted using a variety of qualitative and quantitative research techniques. Conclusion The promotion of tobacco products through appealing imagery cannot be stopped without regulating the package design. The same marketing research techniques used by the tobacco companies can be used to design generic packaging and more effective warning labels targeted at specific consumers.

  10. MARS software package status

    International Nuclear Information System (INIS)

    Azhgirej, I.L.; Talanov, V.V.

    2000-01-01

    The MARS software package is intended for simulating the nuclear-electromagnetic cascades and the secondary neutrons and muons transport in the heterogeneous medium of arbitrary complexity in the magnetic fields presence. The inclusive approach to describing the particle production in the nuclear and electromagnetic interactions and by the unstable particles decay is realized in the package. The MARS software package was actively applied for solving various radiation physical problems [ru

  11. User friendly packaging

    DEFF Research Database (Denmark)

    Geert Jensen, Birgitte

    2010-01-01

    Most consumers have experienced occasional problems with opening packaging. Tomato sauce from the tinned mackerel splattered all over the kitchen counter, the unrelenting pickle jar lid, and the package of sliced ham that cannot be opened without a knife or a pair of scissors. The research project...... “User-friendly Packaging” aims to create a platform for developing more user-friendly packaging. One intended outcome of the project is a guideline that industry can use in development efforts. The project also points the way for more extended collaboration between companies and design researchers. How...... can design research help industry in packaging innovation?...

  12. The ZOOM minimization package

    International Nuclear Information System (INIS)

    Fischler, Mark S.; Sachs, D.

    2004-01-01

    A new object-oriented Minimization package is available for distribution in the same manner as CLHEP. This package, designed for use in HEP applications, has all the capabilities of Minuit, but is a re-write from scratch, adhering to modern C++ design principles. A primary goal of this package is extensibility in several directions, so that its capabilities can be kept fresh with as little maintenance effort as possible. This package is distinguished by the priority that was assigned to C++ design issues, and the focus on producing an extensible system that will resist becoming obsolete

  13. Plasma physics plotting package

    International Nuclear Information System (INIS)

    Hyman, D.H.

    1981-02-01

    We describe a package of plotting routines that do up to six two- or three-dimensional plots on a frame with minimal loss of resolution. The package now runs on a PDP-10 with PLOT-10 TCS primitives and on a Control Data Corporation-7600 and a Cray-1 with TV80LIB primitives on the National Magnetic Fusion Energy Computer Center network. The package is portable to other graphics systems because only the primitive plot calls are used from the underlying system's graphics package

  14. Active food packaging technologies.

    Science.gov (United States)

    Ozdemir, Murat; Floros, John D

    2004-01-01

    Active packaging technologies offer new opportunities for the food industry, in the preservation of foods. Important active packaging systems currently known to date, including oxygen scavengers, carbon dioxide emitters/absorbers, moisture absorbers, ethylene absorbers, ethanol emitters, flavor releasing/absorbing systems, time-temperature indicators, and antimicrobial containing films, are reviewed. The principle of operation of each active system is briefly explained. Recent technological advances in active packaging are discussed, and food related applications are presented. The effects of active packaging systems on food quality and safety are cited.

  15. Supervised Convolutional Sparse Coding

    KAUST Repository

    Affara, Lama Ahmed; Ghanem, Bernard; Wonka, Peter

    2018-01-01

    coding, which aims at learning discriminative dictionaries instead of purely reconstructive ones. We incorporate a supervised regularization term into the traditional unsupervised CSC objective to encourage the final dictionary elements

  16. MODLIB, library of Fortran modules for nuclear reaction codes

    International Nuclear Information System (INIS)

    Talou, Patrick

    2006-01-01

    1 - Description of program or function: ModLib is a library of Fortran (90-compatible) modules to be used in existing and future nuclear reaction codes. The development of the library is an international effort being undertaken under the auspices of the long-term Subgroup A of the OECD/NEA Working Party on Evaluation and Cooperation. The aim is to constitute a library of well-tested and well-documented pieces of codes that can be used with confidence in all our coding efforts. This effort will undoubtedly help avoid the duplication of work, and most certainly facilitate the very important inter-comparisons between existing codes. 2 - Methods: - Width f luctuations [Talou, Chadwick]: calculates width fluctuation correction factors (output) for a set of transmission coefficients (input). Three methods are available: HRTW, Moldauer, and Verbaarschot (also called GOE approach). So far, no distinction is made according to the type of the coefficients channels (particle emission, gamma-ray emission, fission). - Gamma s trength [Herman]: calculates gamma-ray transmission coefficients using a Giant Resonance formalism. - Level d ensity [Koning]: computes the Gilbert-Cameron-Ignatyuk formalism for the continuum nuclear level density. - CHECKR, FIZCON, INTER, PSYCHE, STANEF [Dunford]: these modules are used in the MODLIB project but are not included in this package. They are available from the NEA Data Bank Computer Program Service under Package Ids: CHECKR (USCD1208), FIZCON (USCD1209), INTER (USCD1212), PSYCHE (USCD1216), STANEF (USCD1218)

  17. The application package DeCA for calculating cyclic accelerators

    International Nuclear Information System (INIS)

    Gladkikh, P.I.; Zelinsky, A.Yu.; Strelkov, M.A.

    1993-01-01

    The application Package DeCA (Design Cyclic Accelerator) is offered to solve a set of problem which arise on designing electron storage rings. The package is based on the block principle. This makes it extremely flexible in designing storage rings and investigating beam dynamics in them. The package is intended for a user not familiar with programming languages, it is arranged so that the user familiar with FORTRAN-77 can easily extend the package functions. This is of particular interest, when the input data are the storage ring or electron bunch parameters. The code allows operation in both the batch and interactive modes. The programming language is FORTRAN-77. The capacity of the total package is 40,000 code lines. The necessary main storage capacity for the total version is 4 Mbytes

  18. Applications guide to the MORSE Monte Carlo code

    International Nuclear Information System (INIS)

    Cramer, S.N.

    1985-08-01

    A practical guide for the implementation of the MORESE-CG Monte Carlo radiation transport computer code system is presented. The various versions of the MORSE code are compared and contrasted, and the many references dealing explicitly with the MORSE-CG code are reviewed. The treatment of angular scattering is discussed, and procedures for obtaining increased differentiality of results in terms of reaction types and nuclides from a multigroup Monte Carlo code are explained in terms of cross-section and geometry data manipulation. Examples of standard cross-section data input and output are shown. Many other features of the code system are also reviewed, including (1) the concept of primary and secondary particles, (2) fission neutron generation, (3) albedo data capability, (4) DOMINO coupling, (5) history file use for post-processing of results, (6) adjoint mode operation, (7) variance reduction, and (8) input/output. In addition, examples of the combinatorial geometry are given, and the new array of arrays geometry feature (MARS) and its three-dimensional plotting code (JUNEBUG) are presented. Realistic examples of user routines for source, estimation, path-length stretching, and cross-section data manipulation are given. A deatiled explanation of the coupling between the random walk and estimation procedure is given in terms of both code parameters and physical analogies. The operation of the code in the adjoint mode is covered extensively. The basic concepts of adjoint theory and dimensionality are discussed and examples of adjoint source and estimator user routines are given for all common situations. Adjoint source normalization is explained, a few sample problems are given, and the concept of obtaining forward differential results from adjoint calculations is covered. Finally, the documentation of the standard MORSE-CG sample problem package is reviewed and on-going and future work is discussed

  19. Data Packages for the Hanford Immobilized Low Activity Tank Waste Performance Assessment 2001 Version [SEC 1 THRU 5

    Energy Technology Data Exchange (ETDEWEB)

    MANN, F.M.

    2000-03-02

    Data package supporting the 2001 Immobilized Low-Activity Waste Performance Analysis. Geology, hydrology, geochemistry, facility, waste form, and dosimetry data based on recent investigation are provided. Verification and benchmarking packages for selected software codes are provided.

  20. Aspects of safety assessments for package with additional equipment components

    International Nuclear Information System (INIS)

    Reiche, I.; Boerst, F.M.; Krietsch, T.

    2004-01-01

    Many paragraphs in TS-R-1 contain the terms ''package'' or ''packaging''. These terms are defined in TS-R-1 paras 230 and 231 and explained in TS-G-1.1 paras 230.1 - 230.6. The importance of a consistent understanding of these definitions has been shown by recent discussions during the assessment of applications for package design approval. There was disagreement, if equipment components attached to the container body during transport, e.g. a transport frame, should be considered part of the package and taken into account in the safety assessment for the package. Discussions were also caused by the way inner design components are treated in the safety assessment of the package. This paper summarises the regulatory requirements to such additional equipment components and presents the way of their inclusion into the package design approval process in Germany

  1. Neutron and photon measurements through concrete from a 15 GeV electron beam on a target-comparison with models and calculations. [Intermediate energy source term, Monte Carlo code

    Energy Technology Data Exchange (ETDEWEB)

    Jenkins, T M [Stanford Linear Accelerator Center, CA (USA)

    1979-02-15

    Measurements of neutron and photon dose equivalents from a 15 GeV electron beam striking an iron target inside a scale model of a PEP IR hall are described, and compared with analytic-empirical calculations and with the Monte Carlo code, MORSE. The MORSE code is able to predict both absolute neutron and photon dose equivalents for geometries where the shield is relatively thin, but fails as the shield thickness is increased. An intermediate energy source term is postulated for analytic-empirical neutron shielding calculations to go along with the giant resonance and high energy terms, and a new source term due to neutron capture is postulated for analytic-empirical photon shielding calculations. The source strengths for each energy source term, and each type, are given from analysis of the measurements.

  2. Coding Partitions

    Directory of Open Access Journals (Sweden)

    Fabio Burderi

    2007-05-01

    Full Text Available Motivated by the study of decipherability conditions for codes weaker than Unique Decipherability (UD, we introduce the notion of coding partition. Such a notion generalizes that of UD code and, for codes that are not UD, allows to recover the ``unique decipherability" at the level of the classes of the partition. By tacking into account the natural order between the partitions, we define the characteristic partition of a code X as the finest coding partition of X. This leads to introduce the canonical decomposition of a code in at most one unambiguouscomponent and other (if any totally ambiguouscomponents. In the case the code is finite, we give an algorithm for computing its canonical partition. This, in particular, allows to decide whether a given partition of a finite code X is a coding partition. This last problem is then approached in the case the code is a rational set. We prove its decidability under the hypothesis that the partition contains a finite number of classes and each class is a rational set. Moreover we conjecture that the canonical partition satisfies such a hypothesis. Finally we consider also some relationships between coding partitions and varieties of codes.

  3. The generation of recombinant influenza A viruses expressing a PB2 fusion protein requires the conservation of a packaging signal overlapping the coding and noncoding regions at the 5' end of the PB2 segment

    International Nuclear Information System (INIS)

    Dos Santos Afonso, Emmanuel; Escriou, Nicolas; Leclercq, India; Werf, Sylvie van der; Naffakh, Nadia

    2005-01-01

    We generated recombinant A/WSN/33 influenza A viruses expressing a PB2 protein fused to a Flag epitope at the N- (Flag-PB2) or C-terminus (PB2-Flag), which replicated efficiently and proved to be stable upon serial passage in vitro on MDCK cells. Rescue of PB2-Flag viruses required that the 5' end of the PB2 segment was kept identical to the wild-type beyond the 34 noncoding terminal nucleotides. This feature was achieved by a duplication of the 109 last nucleotides encoding PB2 between the Flag sequence and the 5'NCR. In PB2 minigenomes rescue experiments, both the 5' and 3' coding ends of the PB2 segment were found to promote the incorporation of minigenomes into virions. However, the presence of the Flag sequence at the junction between the 3'NCR and the coding sequence did not prevent the rescue of Flag-PB2 viruses. Our observations define requirements that may be useful for the purpose of engineering influenza RNAs

  4. CYPROS - Cybernetic Program Packages

    Directory of Open Access Journals (Sweden)

    Arne Tyssø

    1980-10-01

    Full Text Available CYPROS is an interactive program system consisting of a number of special purpose packages for simulation, identification, parameter estimation and control system design. The programming language is standard FORTRAN IV and the system is implemented on a medium size computer system (Nord-10. The system is interactive and program control is obtained by the use of numeric terminals. Output is rapidly examined by extensive use of video colour graphics. The subroutines included in the packages are designed and documented according to standardization rules given by the SCL (Scandinavian Control Library organization. This simplifies the exchange of subroutines throughout the SCL system. Also, this makes the packages attractive for implementation by industrial users. In the simulation package, different integration methods are available and it can be easily used for off-line, as well as real time, simulation problems. The identification package consists of programs for single-input/single-output and multivariablc problems. Both transfer function models and state space models can be handled. Optimal test signals can be designed. The control package consists of programs based on multivariable time domain and frequency domain methods for analysis and design. In addition, there is a package for matrix and time series manipulation. CYPROS has been applied successfully to industrial problems of various kinds, and parts of the system have already been implemented on different computers in industry. This paper will, in some detail, describe the use and the contents of the packages and some examples of application will be discussed.

  5. User friendly packaging

    DEFF Research Database (Denmark)

    Geert Jensen, Birgitte

    2010-01-01

    Most consumers have experienced occasional problems with opening packaging. Tomato sauce from the tinned mackerel splattered all over the kitchen counter, the unrelenting pickle jar lid, and the package of sliced ham that cannot be opened without a knife or a pair of scissors. The research project...

  6. NRF TRIGA packaging

    International Nuclear Information System (INIS)

    Clements, M.D.

    1995-11-01

    Training Reactor Isotopes, General Atomics (TRIGA reg-sign) Reactors are in use at four US Department of Energy (DOE) complex facilities and at least 23 university, commercial, or government facilities. The development of the Neutron Radiography Facility (NRF) TRIGA packaging system began in October 1993. The Hanford Site NRF is being shut down and requires an operationally user-friendly transportation and storage packaging system for removal of the TRIGA fuel elements. The NRF TRIGA packaging system is designed to remotely remove the fuel from the reactor and transport the fuel to interim storage (up to 50 years) on the Hanford Site. The packaging system consists of a cask and an overpack. The overpack is used only for transport and is not necessary for storage. Based upon the cask's small size and light weight, small TRIGA reactors will find it versatile for numerous refueling and fuel storage needs. The NRF TRIGA packaging design also provides the basis for developing a certifiable and economical packaging system for other TRIGA reactor facilities. The small size of the NRF TRIGA cask also accommodates placing the cask into a larger certified packaging for offsite transport. The Westinghouse Hanford Company NRF TRIGA packaging, as described herein can serve other DOE sites for their onsite use, and the design can be adapted to serve university reactor facilities, handling a variety of fuel payloads

  7. The BINSYN Program Package

    Directory of Open Access Journals (Sweden)

    Albert P. Linnell

    2012-06-01

    Full Text Available The BINSYN program package, recently expanded to calculate synthetic spectra of cataclysmic variables, is being further extended to include synthetic photometry of ordinary binary stars in addition to binary stars with optically thick accretion disks. The package includes a capability for differentials correction optimization of eclipsing binary systems using synthetic photometry.

  8. The LCDROOT Analysis Package

    International Nuclear Information System (INIS)

    Abe, Toshinori

    2001-01-01

    The North American Linear Collider Detector group has developed simulation and analysis program packages. LCDROOT is one of the packages, and is based on ROOT and the C++ programing language to maximally benefit from object oriented programming techniques. LCDROOT is constantly improved and now has a new topological vertex finder, ZVTOP3. In this proceeding, the features of the LCDROOT simulation are briefly described

  9. Grooming. Learning Activity Package.

    Science.gov (United States)

    Stark, Pamela

    This learning activity package on grooming for health workers is one of a series of 12 titles developed for use in health occupations education programs. Materials in the package include objectives, a list of materials needed, information sheets, reviews (self evaluations) of portions of the content, and answers to reviews. These topics are…

  10. The particle tracking package Kassiopeia

    Energy Technology Data Exchange (ETDEWEB)

    Groh, Stefan [Karlsruhe Institute of Technology (Germany); Collaboration: KATRIN-Collaboration

    2016-07-01

    The Kassiopeia particle tracking framework is an object-oriented software package utilizing modern C++ techniques, written originally to meet the needs of the Katrin collaboration. Kassiopeia's target consists of simulating particle trajectories governed by arbitrarily complex differential equations of motion, continuous physics processes that may in part be modeled as terms perturbing that equation of motion, stochastic processes that occur in flight such as bulk scattering and decay, and potentially stochastic surface processes occurring at interfaces, including transmission and reflection effects. This entire set of computations takes place against the backdrop of a fully-featured geometry package which serves a variety of roles, including initialization of electromagnetic field simulations, gas flow simulations, and the support of state-dependent algorithm-swapping and behavioral changes. Kassiopeia has been well validated and widely used within the Katrin collaboration, playing a primary role in many theses and refereed publications.

  11. Crisis, Stimulus Package and Migration in China

    NARCIS (Netherlands)

    Csanádi, Maria; Nie, Zihan; Li, Shi

    2015-01-01

    This paper analyzes the short-term and long-term effects that the global economic crisis and the investment priorities of the Chinese Government's stimulus package had on Chinese migrant flows between 2008 and 2014. Combining micro-level household survey data and macro-level statistics, the

  12. Majorana fermion codes

    International Nuclear Information System (INIS)

    Bravyi, Sergey; Terhal, Barbara M; Leemhuis, Bernhard

    2010-01-01

    We initiate the study of Majorana fermion codes (MFCs). These codes can be viewed as extensions of Kitaev's one-dimensional (1D) model of unpaired Majorana fermions in quantum wires to higher spatial dimensions and interacting fermions. The purpose of MFCs is to protect quantum information against low-weight fermionic errors, that is, operators acting on sufficiently small subsets of fermionic modes. We examine to what extent MFCs can surpass qubit stabilizer codes in terms of their stability properties. A general construction of 2D MFCs is proposed that combines topological protection based on a macroscopic code distance with protection based on fermionic parity conservation. Finally, we use MFCs to show how to transform any qubit stabilizer code to a weakly self-dual CSS code.

  13. Theory of epigenetic coding.

    Science.gov (United States)

    Elder, D

    1984-06-07

    The logic of genetic control of development may be based on a binary epigenetic code. This paper revises the author's previous scheme dealing with the numerology of annelid metamerism in these terms. Certain features of the code had been deduced to be combinatorial, others not. This paradoxical contrast is resolved here by the interpretation that these features relate to different operations of the code; the combinatiorial to coding identity of units, the non-combinatorial to coding production of units. Consideration of a second paradox in the theory of epigenetic coding leads to a new solution which further provides a basis for epimorphic regeneration, and may in particular throw light on the "regeneration-duplication" phenomenon. A possible test of the model is also put forward.

  14. Waste package performance analysis

    International Nuclear Information System (INIS)

    Lester, D.H.; Stula, R.T.; Kirstein, B.E.

    1982-01-01

    A performance assessment model for multiple barrier packages containing unreprocessed spent fuel has been applied to several package designs. The resulting preliminary assessments were intended for use in making decisions about package development programs. A computer model called BARIER estimates the package life and subsequent rate of release of selected nuclides. The model accounts for temperature, pressure (and resulting stresses), bulk and localized corrosion, and nuclide retardation by the backfill after water intrusion into the waste form. The assessment model assumes a post-closure, flooded, geologic repository. Calculations indicated that, within the bounds of model assumptions, packages could last for several hundred years. Intact backfills of appropriate design may be capable of nuclide release delay times on the order of 10 7 yr for uranium, plutonium, and americium. 8 references, 6 figures, 9 tables

  15. RH Packaging Operations Manual

    International Nuclear Information System (INIS)

    Washington TRU Solutions LLC

    2003-01-01

    This procedure provides operating instructions for the RH-TRU 72-B Road Cask, Waste Shipping Package. In this document, ''Packaging'' refers to the assembly of components necessary to ensure compliance with the packaging requirements (not loaded with a payload). ''Package'' refers to a Type B packaging that, with its radioactive contents, is designed to retain the integrity of its containment and shielding when subject to the normal conditions of transport and hypothetical accident test conditions set forth in 10 CFR Part 71. Loading of the RH 72-B cask can be done two ways, on the RH cask trailer in the vertical position or by removing the cask from the trailer and loading it in a facility designed for remote-handling (RH). Before loading the 72-B cask, loading procedures and changes to the loading procedures for the 72-B cask must be sent to CBFO at sitedocuments at wipp.ws for approval

  16. Advanced flip chip packaging

    CERN Document Server

    Lai, Yi-Shao; Wong, CP

    2013-01-01

    Advanced Flip Chip Packaging presents past, present and future advances and trends in areas such as substrate technology, material development, and assembly processes. Flip chip packaging is now in widespread use in computing, communications, consumer and automotive electronics, and the demand for flip chip technology is continuing to grow in order to meet the need for products that offer better performance, are smaller, and are environmentally sustainable. This book also: Offers broad-ranging chapters with a focus on IC-package-system integration Provides viewpoints from leading industry executives and experts Details state-of-the-art achievements in process technologies and scientific research Presents a clear development history and touches on trends in the industry while also discussing up-to-date technology information Advanced Flip Chip Packaging is an ideal book for engineers, researchers, and graduate students interested in the field of flip chip packaging.

  17. Short-term changes in quitting-related cognitions and behaviours after the implementation of plain packaging with larger health warnings: findings from a national cohort study with Australian adult smokers.

    Science.gov (United States)

    Durkin, Sarah; Brennan, Emily; Coomber, Kerri; Zacher, Meghan; Scollo, Michelle; Wakefield, Melanie

    2015-04-01

    Plain packaging (PP) with larger graphic health warnings (GHWs) was implemented in Australia in late 2012. This study examined effects of these packaging changes on short-term changes in quitting-related cognitions and behaviours. We used a series of cohorts of Australian adult cigarette smokers originally sourced from a nationally representative cross-sectional tracking survey, followed up approximately 1 month after their baseline interview (n(weighted)=5441). Logistic regression analyses compared changes in seven quitting-related outcomes over this 1-month follow-up period for the cohorts surveyed before PP, over the period of transition to PP, and during the first year of PP, adjusting for baseline levels of the outcome and covariates. Compared to the referent group of smokers who completed their follow-up survey pre-PP, those who were followed-up in the early transition period showed significantly greater increases in rates of stopping themselves from smoking (OR=1.51, 95% CI (1.08 to 2.10)) and higher quit attempt rates (OR=1.43, 95% CI (1.00 to 2.03)), those followed-up in the late transition period showed greater increases in intentions to quit (OR=1.42, 95% CI (1.06 to 1.92)) and pack concealment (OR=1.55, 95% CI (1.05 to 2.31)), and those followed-up in the first year of PP showed higher levels of pack concealment (OR=1.65, 95% CI (1.01 to 2.72)), more premature stubbing out of cigarettes (OR=1.55, 95% CI (1.01 to 2.36)), and higher quit attempt rates (OR=1.52, 95% CI (1.01 to 2.30)). These findings provide some of the strongest evidence to date that implementation of PP with larger GHWs was associated with increased rates of quitting cognitions, microindicators of concern and quit attempts among adult cigarette smokers. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://group.bmj.com/group/rights-licensing/permissions.

  18. Large transport packages for decommissioning waste

    International Nuclear Information System (INIS)

    Price, M.S.T.

    1988-03-01

    The main tasks performed during the period related to the influence of manufacture, transport and disposal on the design of such packages. It is deduced that decommissioning wastes will be transported under the IAEA Transport Regulations under either the Type B or Low Specific Activity (LSA) categories. If the LSA packages are self-shielded, reinforced concrete is the preferred material of construction. But the high cost of disposal implies that there is a strong reason to investigate the use of returnable shields for LSA packages and in such cases they are likely to be made of ferrous metal. Economic considerations favour the use of spheroidal graphite cast iron for this purpose. Transport operating hazards have been investigated using a mixture of desk studies, routes surveys and operations data from the railway organisations. Reference routes were chosen in the Federal Republic of Germany, France and the United Kingdom. This work has led to a description of ten accident scenarios and an evaluation of the associated accident probabilities. The effect of disposal on design of packages has been assessed in terms of the radiological impact of decommissioning wastes, an in addition corrosion and gas evolution have been examined. The inventory of radionuclides in a decommissioning waste package has low environmental impact. If metal clad reinforced concrete packages are to be used, the amount of gas evolution is such that a vent would need to be included in the design. Similar unclad packages would be sufficiently permeable to gases to prevent a pressure build-up. (author)

  19. Packaging Printing Today

    Directory of Open Access Journals (Sweden)

    Stanislav Bolanča

    2015-12-01

    Full Text Available Printing packaging covers today about 50% of all the printing products. Among the printing products there are printing on labels, printing on flexible packaging, printing on folding boxes, printing on the boxes of corrugated board, printing on glass packaging, synthetic and metal ones. The mentioned packaging are printed in flexo printing technique, offset printing technique, intaglio halftone process, silk – screen printing, ink ball printing, digital printing and hybrid printing process. The possibilities of particular printing techniques for optimal production of the determined packaging were studied in the paper. The problem was viewed from the technological and economical aspect. The possible printing quality and the time necessary for the printing realization were taken as key parameters. An important segment of the production and the way of life is alocation value and it had also found its place in this paper. The events in the field of packaging printing in the whole world were analyzed. The trends of technique developments and the printing technology for packaging printing in near future were also discussed.

  20. Consumer response to packaging design

    NARCIS (Netherlands)

    Steenis, Nigel D.; Herpen, van Erica; Lans, van der Ivo A.; Ligthart, Tom N.; Trijp, van Hans C.M.

    2017-01-01

    Building on theories of cue utilization, this paper investigates whether and how packaging sustainability influences consumer perceptions, inferences and attitudes towards packaged products. A framework is tested in an empirical study among 249 students using soup products varying in packaging

  1. Development of a Code for the Long Term Radiological Safety Assessment of Radioactive Wastes from Advanced Nuclear Fuel Cycle Facilities in Republic of Korea

    International Nuclear Information System (INIS)

    Hwang, Yong Soo

    2010-01-01

    For the purpose of evaluating annual individual doses from a potential repository disposing of radioactive wastes from the operation of the prospective advanced nuclear fuel cycle facilities in Korea, the new safety assessment code based on the Goldsim has been developed. It was designed to compare the environmental impacts from many fuel cycle options such as direct disposal, wet and dry recycling. The code based on the compartment theory can be applied to assess both normal and what if scenarios

  2. Packaging Concerns/Techniques for Large Devices

    Science.gov (United States)

    Sampson, Michael J.

    2009-01-01

    This slide presentation reviews packaging challenges and options for electronic parts. The presentation includes information about non-hermetic packages, space challenges for packaging and complex package variations.

  3. Hermeticity of electronic packages

    CERN Document Server

    Greenhouse, Hal; Romenesco, Bruce

    2011-01-01

    This is a book about the integrity of sealed packages to resist foreign gases and liquids penetrating the seal or an opening (crack) in the packageùespecially critical to the reliability and longevity of electronics. The author explains how to predict the reliability and the longevity of the packages based on leak rate measurements and the assumptions of impurities. Non-specialists in particular will benefit from the author's long involvement in the technology. Hermeticity is a subject that demands practical experience, and solving one problem does not necessarily give one the background to so

  4. Hermeticity of electronic packages

    CERN Document Server

    Greenhouse, Hal

    2000-01-01

    This is a book about the integrity of sealed packages to resist foreign gases and liquids penetrating the seal or an opening (crack) in the package-especially critical to the reliability and longevity of electronics. The author explains how to predict the reliability and the longevity of the packages based on leak rate measurements and the assumptions of impurities. Non-specialists in particular will benefit from the author's long involvement in the technology. Hermeticity is a subject that demands practical experience, and solving one problem does not necessarily give one the background to so

  5. AMPX-77 Phase 1 certification package

    Energy Technology Data Exchange (ETDEWEB)

    Niemer, K.A.

    1994-03-01

    The AMPX-77 Phase 1 modules have been certified. AMPX-77 is a modular code system for generating coupled multigroup neutron-gamma cross section libraries from Evaluated Nuclear Data Files (ENDF/B). All basic cross-section data are input from the formats used by the ENDF/B, and output can be obtained from a variety of formats, included in its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-ray data. The AMPX-77 code system will be used at SRS to perform critical calculations related to nuclear criticality safety. The AMPX-77 modular codes system contains forty-seven separate modules. For the certification process, the 47 modules have been divided into three groups or phases. This Certification Package is for the Phase 1 modules: BONAMI, LAPHNGAS, MALOCS, NITAWL, ROLAIDS, SMUG, and XSDRNPM.

  6. Development of polymer packaging for power cable

    Directory of Open Access Journals (Sweden)

    S. Sremac

    2014-10-01

    Full Text Available This paper discusses the issues of product design and the procedure of developing polymer packaging as one of the most important engineering tasks. For the purpose of packing power cables a polymer packaging has been designed in the form of drum. Packaging and many other consumer products are largely produced using polymeric materials due to many positive features. High Density Polyethylene is the type of polyethylene proposed for packaging purposes due to its low degree of branching and strong intermolecular forces. Transport and storage processes were automated based on the radio-frequency identification technology. The proposed system is flexible in terms of its possibility of accepting and processing different types of cables and other products.

  7. 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

  8. MIFT: GIFT Combinatorial Geometry Input to VCS Code

    Science.gov (United States)

    1977-03-01

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

  9. Safety Analysis Report - Packages, 9965, 9968, 9972-9975 Packages

    International Nuclear Information System (INIS)

    Blanton, P.

    2000-01-01

    This Safety Analysis Report for Packaging (SARP) documents the analysis and testing performed on four type B Packages: the 9972, 9973, 9974, and 9975 packages. Because all four packages have similar designs with very similar performance characteristics, all of them are presented in a single SARP. The performance evaluation presented in this SARP documents the compliance of the 9975 package with the regulatory safety requirements. Evaluations of the 9972, 9973, and 9974 packages support that of the 9975. To avoid confusion arising from the inclusion of four packages in a single document, the text segregates the data for each package in such a way that the reader interested in only one package can progress from Chapter 1 through Chapter 9. The directory at the beginning of each chapter identifies each section that should be read for a given package. Sections marked ''all'' are generic to all packages

  10. Glass packages in interim storage

    International Nuclear Information System (INIS)

    Jacquet-Francillon, N.

    1994-10-01

    This report summarize the current state of knowledge concerning the behavior of type C waste packages consisting of vitrified high-level solutions produced by reprocessing spent fuel. The composition and the physical and chemical properties of the feed solutions are reviewed, and the vitrification process is described. Sodium alumino-borosilicate glass compositions are generally employed - the glass used at la Hague for LWR fuel solutions, for example, contains 45 % SiO 2 . The major physical, chemical, mechanical and thermal properties of the glass are reviewed. In order to allow their thermal power to diminish, the 3630 glass packages produced (as of January 1993) in the vitrification facilities at Marcoule and La Hague are placed in interim storage for several decades. The actual interim storage period has not been defined, as it is closely related to the concept and organization selected for the final destination of the packages: a geological repository. The glass behavior under irradiation is described. Considerable basic and applied research has been conducted to assess the aqueous leaching behavior of nuclear containment glass. The effects of various repository parameters (temperature, flow rate, nature of the environmental materials) have been investigated. The experimental findings have been used to specify a model describing the kinetics of aqueous corrosion of the glass. More generally all the ''source term'' models developed in France by the CEA or by ANDRA are summarized. (author). 152 refs., 33 figs

  11. Waste Package Design Methodology Report

    Energy Technology Data Exchange (ETDEWEB)

    D.A. Brownson

    2001-09-28

    The objective of this report is to describe the analytical methods and processes used by the Waste Package Design Section to establish the integrity of the various waste package designs, the emplacement pallet, and the drip shield. The scope of this report shall be the methodology used in criticality, risk-informed, shielding, source term, structural, and thermal analyses. The basic features and appropriateness of the methods are illustrated, and the processes are defined whereby input values and assumptions flow through the application of those methods to obtain designs that ensure defense-in-depth as well as satisfy requirements on system performance. Such requirements include those imposed by federal regulation, from both the U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC), and those imposed by the Yucca Mountain Project to meet repository performance goals. The report is to be used, in part, to describe the waste package design methods and techniques to be used for producing input to the License Application Report.

  12. Waste Package Design Methodology Report

    International Nuclear Information System (INIS)

    D.A. Brownson

    2001-01-01

    The objective of this report is to describe the analytical methods and processes used by the Waste Package Design Section to establish the integrity of the various waste package designs, the emplacement pallet, and the drip shield. The scope of this report shall be the methodology used in criticality, risk-informed, shielding, source term, structural, and thermal analyses. The basic features and appropriateness of the methods are illustrated, and the processes are defined whereby input values and assumptions flow through the application of those methods to obtain designs that ensure defense-in-depth as well as satisfy requirements on system performance. Such requirements include those imposed by federal regulation, from both the U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC), and those imposed by the Yucca Mountain Project to meet repository performance goals. The report is to be used, in part, to describe the waste package design methods and techniques to be used for producing input to the License Application Report

  13. Type B Drum packages

    International Nuclear Information System (INIS)

    Edwards, W.S.

    1995-11-01

    The Type B Drum package is a container in which a single drum containing Type B quantities of radioactive material will be packaged for shipment. The Type B Drum containers are being developed to fill a void in the packaging and transportation capabilities of the US Department of Energy (DOE), as no double containment packaging for single drums of Type B radioactive material is currently available. Several multiple-drum containers and shielded casks presently exist. However, the size and weight of these containers present multiple operational challenges for single-drum shipments. The Type B Drum containers will offer one unshielded version and, if needed, two shielded versions, and will provide for the option of either single or double containment. The primary users of the Type B Drum container will be any organization with a need to ship single drums of Type B radioactive material. Those users include laboratories, waste retrieval facilities, emergency response teams, and small facilities

  14. Materials for advanced packaging

    CERN Document Server

    Wong, CP

    2017-01-01

    This second edition continues to be the most comprehensive review on the developments in advanced electronic packaging technologies, with a focus on materials and processing. Recognized experts in the field contribute to 22 updated and new chapters that provide comprehensive coverage on various 3D package architectures, novel bonding and joining techniques, wire bonding, wafer thinning techniques, organic substrates, and novel approaches to make electrical interconnects between integrated circuit and substrates. Various chapters also address advances in several key packaging materials, including: Lead-free solders Flip chip underfills Epoxy molding compounds Conductive adhesives Die attach adhesives/films Thermal interface materials (TIMS) Materials for fabricating embedded passives including capacitors, inductors, and resistors Materials and processing aspects on wafer-level chip scale package (CSP) and MicroElectroMechanical system (MEMS) Contributors also review new and emerging technologies such as Light ...

  15. Food irradiation and packaging

    International Nuclear Information System (INIS)

    Kilcast, David

    1988-01-01

    This outline review was written for 'Food Manufacture'. It deals with the known effects of irradiation on current packaging materials (glass, cellulosics, organic polymers and metals), and their implications for the effective application of the process. (U.K.)

  16. FLEXIBLE FOOD PACKAGING LABORATORY

    Data.gov (United States)

    Federal Laboratory Consortium — This laboratory contains equipment to fabricate and test prototype packages of many types and sizes (e.g., bags, pouches, trays, cartons, etc.). This equipment can...

  17. BCRA R Package

    Science.gov (United States)

    BCRA is an R package that projects absolute risk of invasive breast cancer according to NCI’s Breast Cancer Risk Assessment Tool (BCRAT) algorithm for specified race/ethnic groups and age intervals.

  18. iCosmo: an interactive cosmology package

    Science.gov (United States)

    Refregier, A.; Amara, A.; Kitching, T. D.; Rassat, A.

    2011-04-01

    Aims: The interactive software package iCosmo, designed to perform cosmological calculations is described. Methods: iCosmo is a software package to perfom interactive cosmological calculations for the low-redshift universe. Computing distance measures, the matter power spectrum, and the growth factor is supported for any values of the cosmological parameters. It also computes derived observed quantities for several cosmological probes such as cosmic shear, baryon acoustic oscillations, and type Ia supernovae. The associated errors for these observable quantities can be derived for customised surveys, or for pre-set values corresponding to current or planned instruments. The code also allows for calculation of cosmological forecasts with Fisher matrices, which can be manipulated to combine different surveys and cosmological probes. The code is written in the IDL language and thus benefits from the convenient interactive features and scientific libraries available in this language. iCosmo can also be used as an engine to perform cosmological calculations in batch mode, and forms a convenient adaptive platform for the development of further cosmological modules. With its extensive documentation, it may also serve as a useful resource for teaching and for newcomers to the field of cosmology. Results: The iCosmo package is described with a number of examples and command sequences. The code is freely available with documentation at http://www.icosmo.org, along with an interactive web interface and is part of the Initiative for Cosmology, a common archive for cosmological resources.

  19. Transportation and packaging resource guide

    International Nuclear Information System (INIS)

    Arendt, J.W.; Gove, R.M.; Welch, M.J.

    1994-12-01

    The purpose of this resource guide is to provide a convenient reference document of information that may be useful to the U.S. Department of Energy (DOE) and DOE contractor personnel involved in packaging and transportation activities. An attempt has been made to present the terminology of DOE community usage as it currently exists. DOE's mission is changing with emphasis on environmental cleanup. The terminology or nomenclature that has resulted from this expanded mission is included for the packaging and transportation user for reference purposes. Older terms still in use during the transition have been maintained. The Packaging and Transportation Resource Guide consists of four sections: Sect. 1, Introduction; Sect. 2, Abbreviations and Acronyms; Sect. 3, Definitions; and Sect. 4, References for packaging and transportation of hazardous materials and related activities, and Appendices A and B. Information has been collected from DOE Orders and DOE documents; U.S Department of Transportation (DOT), U.S. Environmental Protection Agency (EPA), and U.S. Nuclear Regulatory Commission (NRC) regulations; and International Atomic Energy Agency (IAEA) standards and other international documents. The definitions included in this guide may not always be a regulatory definition but are the more common DOE usage. In addition, the definitions vary among regulatory agencies. It is, therefore, suggested that if a definition is to be used in a regulatory or a legal compliance issue, the definition should be verified with the appropriate regulation. To assist in locating definitions in the regulations, a listing of all definition sections in the regulations are included in Appendix B. In many instances, the appropriate regulatory reference is indicated in the right-hand margin

  20. Transportation and packaging resource guide

    Energy Technology Data Exchange (ETDEWEB)

    Arendt, J.W.; Gove, R.M.; Welch, M.J.

    1994-12-01

    The purpose of this resource guide is to provide a convenient reference document of information that may be useful to the U.S. Department of Energy (DOE) and DOE contractor personnel involved in packaging and transportation activities. An attempt has been made to present the terminology of DOE community usage as it currently exists. DOE`s mission is changing with emphasis on environmental cleanup. The terminology or nomenclature that has resulted from this expanded mission is included for the packaging and transportation user for reference purposes. Older terms still in use during the transition have been maintained. The Packaging and Transportation Resource Guide consists of four sections: Sect. 1, Introduction; Sect. 2, Abbreviations and Acronyms; Sect. 3, Definitions; and Sect. 4, References for packaging and transportation of hazardous materials and related activities, and Appendices A and B. Information has been collected from DOE Orders and DOE documents; U.S Department of Transportation (DOT), U.S. Environmental Protection Agency (EPA), and U.S. Nuclear Regulatory Commission (NRC) regulations; and International Atomic Energy Agency (IAEA) standards and other international documents. The definitions included in this guide may not always be a regulatory definition but are the more common DOE usage. In addition, the definitions vary among regulatory agencies. It is, therefore, suggested that if a definition is to be used in a regulatory or a legal compliance issue, the definition should be verified with the appropriate regulation. To assist in locating definitions in the regulations, a listing of all definition sections in the regulations are included in Appendix B. In many instances, the appropriate regulatory reference is indicated in the right-hand margin.

  1. Energy and packaging

    Energy Technology Data Exchange (ETDEWEB)

    Boustead, I; Hancock, G F

    1981-01-01

    Information is given on the energy and raw materials required in the production and use of containers used to package beer, cider, and carbonated soft drinks in the United Kingdom. Topics covered include: methodology of energy analysis, primary and secondary fuels, transport, packaging materials, including glass, aluminum, iron, steel, and tinplate, container production, including plastic bottles, distribution of empty containers, filling and packing, distribution, and disposal. (LEW)

  2. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    Washington TRU Solutions LLC

    2003-01-01

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT shipping package, and directly related components. This document complies with the minimum requirements as specified in the TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event of a conflict between this document and the SARP or C of C, the C of C shall govern. The C of Cs state: ''each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.'' They further state: ''each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SARP charges the WIPP management and operating (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 CFR 71.11. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. This document provides the instructions to be followed to operate, maintain, and test the TRUPACT-II and HalfPACT packaging. The intent of these instructions is to standardize operations. All users will follow these instructions or equivalent instructions that assure operations are safe and meet the requirements of the SARPs

  3. CH Packaging Program Guidance

    International Nuclear Information System (INIS)

    Washington TRU Solutions LLC

    2002-01-01

    The purpose of this document is to provide the technical requirements for preparation for use, operation, inspection, and maintenance of a Transuranic Package Transporter Model II (TRUPACT-II), a HalfPACT Shipping Package, and directly related components. This document complies with the minimum requirements as specified in TRUPACT-II Safety Analysis Report for Packaging (SARP), HalfPACT SARP, and Nuclear Regulatory Commission (NRC) Certificates of Compliance (C of C) 9218 and 9279, respectively. In the event there is a conflict between this document and the SARP or C of C, the SARP and/or C of C shall govern. C of Cs state: ''each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, Operating Procedures, of the application.'' They further state: ''each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SAR P charges the WIPP Management and Operation (M and O) contractor with assuring packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC-approved, users need to be familiar with 10 CFR 71.11. Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. This document details the instructions to be followed to operate, maintain, and test the TRUPACT-II and HalfPACT packaging. The intent of these instructions is to standardize these operations. All users will follow these instructions or equivalent instructions that assure operations are safe and meet the requirements of the SARPs

  4. Phonological coding during reading.

    Science.gov (United States)

    Leinenger, Mallorie

    2014-11-01

    The exact role that phonological coding (the recoding of written, orthographic information into a sound based code) plays during silent reading has been extensively studied for more than a century. Despite the large body of research surrounding the topic, varying theories as to the time course and function of this recoding still exist. The present review synthesizes this body of research, addressing the topics of time course and function in tandem. The varying theories surrounding the function of phonological coding (e.g., that phonological codes aid lexical access, that phonological codes aid comprehension and bolster short-term memory, or that phonological codes are largely epiphenomenal in skilled readers) are first outlined, and the time courses that each maps onto (e.g., that phonological codes come online early [prelexical] or that phonological codes come online late [postlexical]) are discussed. Next the research relevant to each of these proposed functions is reviewed, discussing the varying methodologies that have been used to investigate phonological coding (e.g., response time methods, reading while eye-tracking or recording EEG and MEG, concurrent articulation) and highlighting the advantages and limitations of each with respect to the study of phonological coding. In response to the view that phonological coding is largely epiphenomenal in skilled readers, research on the use of phonological codes in prelingually, profoundly deaf readers is reviewed. Finally, implications for current models of word identification (activation-verification model, Van Orden, 1987; dual-route model, e.g., M. Coltheart, Rastle, Perry, Langdon, & Ziegler, 2001; parallel distributed processing model, Seidenberg & McClelland, 1989) are discussed. (PsycINFO Database Record (c) 2014 APA, all rights reserved).

  5. The octopus burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L.; Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de

    1996-09-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  6. The OCTOPUS burnup and criticality code system

    Energy Technology Data Exchange (ETDEWEB)

    Kloosterman, J.L. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kuijper, J.C. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.).

  7. The octopus burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de.

    1996-01-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional geometries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (author)

  8. The OCTOPUS burnup and criticality code system

    International Nuclear Information System (INIS)

    Kloosterman, J.L.; Kuijper, J.C.; Leege, P.F.A. de

    1996-06-01

    The OCTOPUS burnup and criticality code system is described. This system links the spectrum codes from the SCALE4.1, WIMS7 and MCNP4A packages to the ORIGEN-S and FISPACT4.2 fuel depletion and activation codes, which enables us to perform very accurate burnup calculations in complicated three-dimensional goemetries. The data used by all codes are consistently based on the JEF2.2 evaluated nuclear data file. Some special features of OCTOPUS not available in other codes are described, as well as the validation of the system. (orig.)

  9. Comparative Packaging Study

    Science.gov (United States)

    Perchonok, Michele H.; Oziomek, Thomas V.

    2009-01-01

    Future long duration manned space flights beyond low earth orbit will require the food system to remain safe, acceptable and nutritious. Development of high barrier food packaging will enable this requirement by preventing the ingress and egress of gases and moisture. New high barrier food packaging materials have been identified through a trade study. Practical application of this packaging material within a shelf life test will allow for better determination of whether this material will allow the food system to meet given requirements after the package has undergone processing. The reason to conduct shelf life testing, using a variety of packaging materials, stems from the need to preserve food used for mission durations of several years. Chemical reactions that take place during longer durations may decrease food quality to a point where crew physical or psychological well-being is compromised. This can result in a reduction or loss of mission success. The rate of chemical reactions, including oxidative rancidity and staling, can be controlled by limiting the reactants, reducing the amount of energy available to drive the reaction, and minimizing the amount of water available. Water not only acts as a media for microbial growth, but also as a reactant and means by which two reactants may come into contact with each other. The objective of this study is to evaluate three packaging materials for potential use in long duration space exploration missions.

  10. Validation of thermalhydraulic codes

    International Nuclear Information System (INIS)

    Wilkie, D.

    1992-01-01

    Thermalhydraulic codes require to be validated against experimental data collected over a wide range of situations if they are to be relied upon. A good example is provided by the nuclear industry where codes are used for safety studies and for determining operating conditions. Errors in the codes could lead to financial penalties, to the incorrect estimation of the consequences of accidents and even to the accidents themselves. Comparison between prediction and experiment is often described qualitatively or in approximate terms, e.g. ''agreement is within 10%''. A quantitative method is preferable, especially when several competing codes are available. The codes can then be ranked in order of merit. Such a method is described. (Author)

  11. ARM Data-Oriented Metrics and Diagnostics Package for Climate Model Evaluation Value-Added Product

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Chengzhu [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Xie, Shaocheng [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-10-15

    A Python-based metrics and diagnostics package is currently being developed by the U.S. Department of Energy (DOE) Atmospheric Radiation Measurement (ARM) Infrastructure Team at Lawrence Livermore National Laboratory (LLNL) to facilitate the use of long-term, high-frequency measurements from the ARM Facility in evaluating the regional climate simulation of clouds, radiation, and precipitation. This metrics and diagnostics package computes climatological means of targeted climate model simulation and generates tables and plots for comparing the model simulation with ARM observational data. The Coupled Model Intercomparison Project (CMIP) model data sets are also included in the package to enable model intercomparison as demonstrated in Zhang et al. (2017). The mean of the CMIP model can serve as a reference for individual models. Basic performance metrics are computed to measure the accuracy of mean state and variability of climate models. The evaluated physical quantities include cloud fraction, temperature, relative humidity, cloud liquid water path, total column water vapor, precipitation, sensible and latent heat fluxes, and radiative fluxes, with plan to extend to more fields, such as aerosol and microphysics properties. Process-oriented diagnostics focusing on individual cloud- and precipitation-related phenomena are also being developed for the evaluation and development of specific model physical parameterizations. The version 1.0 package is designed based on data collected at ARM’s Southern Great Plains (SGP) Research Facility, with the plan to extend to other ARM sites. The metrics and diagnostics package is currently built upon standard Python libraries and additional Python packages developed by DOE (such as CDMS and CDAT). The ARM metrics and diagnostic package is available publicly with the hope that it can serve as an easy entry point for climate modelers to compare their models with ARM data. In this report, we first present the input data, which

  12. EQ6 Calculations for Chemical Degradation of Navy Waste Packages

    International Nuclear Information System (INIS)

    S. LeStrange

    1999-01-01

    The Monitored Geologic Repository Waste Package Operations of the Civilian Radioactive Waste Management System Management and Operating Contractor (CRWMS M and O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Navy (Refs. 1 and 2). The Navy SNF has been considered for disposal at the potential Yucca Mountain site. For some waste packages, the containment may breach (Ref. 3), allowing the influx of water. Water in the waste package may moderate neutrons, increasing the likelihood of a criticality event within the waste package. The water may gradually leach the fissile components and neutron absorbers out of the waste package. In addition, the accumulation of silica (SiO 2 ) in the waste package over time may further affect the neutronics of the system. This study presents calculations of the long-term geochemical behavior of waste packages containing the Enhanced Design Alternative (EDA) II inner shell, Navy canister, and basket components. The calculations do not include the Navy SNF in the waste package. The specific study objectives were to determine the chemical composition of the water and the quantity of silicon (Si) and other solid corrosion products in the waste package during the first million years after the waste package is breached. The results of this calculation will be used to ensure that the type and amount of criticality control material used in the waste package design will prevent criticality

  13. EQ6 Calculations for Chemical Degradation of Navy Waste Packages

    Energy Technology Data Exchange (ETDEWEB)

    S. LeStrange

    1999-11-15

    The Monitored Geologic Repository Waste Package Operations of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Navy (Refs. 1 and 2). The Navy SNF has been considered for disposal at the potential Yucca Mountain site. For some waste packages, the containment may breach (Ref. 3), allowing the influx of water. Water in the waste package may moderate neutrons, increasing the likelihood of a criticality event within the waste package. The water may gradually leach the fissile components and neutron absorbers out of the waste package. In addition, the accumulation of silica (SiO{sub 2}) in the waste package over time may further affect the neutronics of the system. This study presents calculations of the long-term geochemical behavior of waste packages containing the Enhanced Design Alternative (EDA) II inner shell, Navy canister, and basket components. The calculations do not include the Navy SNF in the waste package. The specific study objectives were to determine the chemical composition of the water and the quantity of silicon (Si) and other solid corrosion products in the waste package during the first million years after the waste package is breached. The results of this calculation will be used to ensure that the type and amount of criticality control material used in the waste package design will prevent criticality.

  14. Application of GA package in functional packaging

    Science.gov (United States)

    Belousova, D. A.; Noskova, E. E.; Kapulin, D. V.

    2018-05-01

    The approach to application program for the task of configuration of the elements of the commutation circuit for design of the radio-electronic equipment on the basis of the genetic algorithm is offered. The efficiency of the used approach for commutation circuits with different characteristics for computer-aided design on radio-electronic manufacturing is shown. The prototype of the computer-aided design subsystem on the basis of a package GA for R with a set of the general functions for optimization of multivariate models is programmed.

  15. Speaking Code

    DEFF Research Database (Denmark)

    Cox, Geoff

    Speaking Code begins by invoking the “Hello World” convention used by programmers when learning a new language, helping to establish the interplay of text and code that runs through the book. Interweaving the voice of critical writing from the humanities with the tradition of computing and software...

  16. Food Packaging Materials

    Science.gov (United States)

    1978-01-01

    The photos show a few of the food products packaged in Alure, a metallized plastic material developed and manufactured by St. Regis Paper Company's Flexible Packaging Division, Dallas, Texas. The material incorporates a metallized film originally developed for space applications. Among the suppliers of the film to St. Regis is King-Seeley Thermos Company, Winchester, Ma'ssachusetts. Initially used by NASA as a signal-bouncing reflective coating for the Echo 1 communications satellite, the film was developed by a company later absorbed by King-Seeley. The metallized film was also used as insulating material for components of a number of other spacecraft. St. Regis developed Alure to meet a multiple packaging material need: good eye appeal, product protection for long periods and the ability to be used successfully on a wide variety of food packaging equipment. When the cost of aluminum foil skyrocketed, packagers sought substitute metallized materials but experiments with a number of them uncovered problems; some were too expensive, some did not adequately protect the product, some were difficult for the machinery to handle. Alure offers a solution. St. Regis created Alure by sandwiching the metallized film between layers of plastics. The resulting laminated metallized material has the superior eye appeal of foil but is less expensive and more easily machined. Alure effectively blocks out light, moisture and oxygen and therefore gives the packaged food long shelf life. A major packaging firm conducted its own tests of the material and confirmed the advantages of machinability and shelf life, adding that it runs faster on machines than materials used in the past and it decreases product waste; the net effect is increased productivity.

  17. A Method for Determining Optimal Residential Energy Efficiency Packages

    Energy Technology Data Exchange (ETDEWEB)

    Polly, B. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Gestwick, M. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Bianchi, M. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Anderson, R. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Horowitz, S. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Christensen, C. [National Renewable Energy Lab. (NREL), Golden, CO (United States); Judkoff, R. [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2011-04-01

    This report describes an analysis method for determining optimal residential energy efficiency retrofit packages and, as an illustrative example, applies the analysis method to a 1960s-era home in eight U.S. cities covering a range of International Energy Conservation Code (IECC) climate regions. The method uses an optimization scheme that considers average energy use (determined from building energy simulations) and equivalent annual cost to recommend optimal retrofit packages specific to the building, occupants, and location.

  18. Food packages for Space Shuttle

    Science.gov (United States)

    Fohey, M. F.; Sauer, R. L.; Westover, J. B.; Rockafeller, E. F.

    1978-01-01

    The paper reviews food packaging techniques used in space flight missions and describes the system developed for the Space Shuttle. Attention is directed to bite-size food cubes used in Gemini, Gemini rehydratable food packages, Apollo spoon-bowl rehydratable packages, thermostabilized flex pouch for Apollo, tear-top commercial food cans used in Skylab, polyethylene beverage containers, Skylab rehydratable food package, Space Shuttle food package configuration, duck-bill septum rehydration device, and a drinking/dispensing nozzle for Space Shuttle liquids. Constraints and testing of packaging is considered, a comparison of food package materials is presented, and typical Shuttle foods and beverages are listed.

  19. Digital Radiography of a Drop Tested 9975 Radioactive Materials Packaging

    International Nuclear Information System (INIS)

    Blanton, P.S.

    2001-01-01

    This paper discusses the use of radiography as a tool for evaluating damage to radioactive material packaging subjected to regulatory accident conditions. The Code of Federal Regulations, 10 CFR 71, presents the performance based requirements that must be used in the development (design, fabrication and testing) of a radioactive material packaging. The use of various non-destructive examination techniques in the fabrication of packages is common. One such technique is the use of conventional radiography in the examination of welds. Radiography is conventional in the sense that images are caught one at a time on film stock. Most recently, digital radiography has been used to characterize internal damage to a package subjected to the 30-foot hypothetical accident conditions (HAC) drop. Digital radiography allows for real time evaluation of the item being inspected. This paper presents a summary discussion of the digital radiographic technique and an example of radiographic results of a 9975 package following the HAC 30-foot drop

  20. 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

  1. Welding robot package; Arc yosetsu robot package

    Energy Technology Data Exchange (ETDEWEB)

    Nishikawa, S. [Yaskawa Electric Corp., Kitakyushu (Japan)

    1998-09-01

    For the conventional high-speed welding robot, the welding current was controlled mainly for reducing the spatters during short circuits and for stabilizing the beads by the periodic short circuits. However, an increase of deposition amount in response to the speed is required for the high-speed welding. Large-current low-spatter welding current region control was added. Units were integrated into a package by which the arc length is kept in short without dispersion of arc length for welding without defects such as undercut and unequal beads. In automobile industry, use of aluminum parts is extended for the light weight. The welding is very difficult, and automation is not so progressing in spite of the poor environment. Buckling of welding wire is easy to occur, and supply of wire is obstructed by the deposition of chipped powders on the torch cable, which stay within the contact chip resulting in the deposition. Dislocation of locus is easy to occur at the corner of rectangular pipe during the welding. By improving these troubles, an aluminum MIG welding robot package has been developed. 13 figs.

  2. Development of codes for physical calculations of WWER

    International Nuclear Information System (INIS)

    Novikov, A.N.

    2000-01-01

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

  3. 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.

  4. Documentation and analysis for packaging limited quantity ice chests

    International Nuclear Information System (INIS)

    Nguyen, P.M.

    1995-01-01

    The purpose of this Documentation and Analysis for Packaging (DAP) is to document that ice chests meet the intent of the International Air Transport Association (IATA) and the U.S. Department of Transportation (DOT) Code of Federal Regulations as strong, tight containers for the packaging of limited quantities for transport. This DAP also outlines the packaging method used to protect the sample bottles from breakage. Because the ice chests meet the DOT requirements, they can be used to ship LTD QTY on the Hanford Site

  5. CTCN: Colloid transport code -- nuclear

    International Nuclear Information System (INIS)

    Jain, R.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Maiorino, J.R.

    1991-01-01

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

  7. TU-AB-BRC-10: Modeling of Radiotherapy Linac Source Terms Using ARCHER Monte Carlo Code: Performance Comparison of GPU and MIC Computing Accelerators

    International Nuclear Information System (INIS)

    Liu, T; Lin, H; Xu, X; Su, L; Shi, C; Tang, X; Bednarz, B

    2016-01-01

    Purpose: (1) To perform phase space (PS) based source modeling for Tomotherapy and Varian TrueBeam 6 MV Linacs, (2) to examine the accuracy and performance of the ARCHER Monte Carlo code on a heterogeneous computing platform with Many Integrated Core coprocessors (MIC, aka Xeon Phi) and GPUs, and (3) to explore the software micro-optimization methods. Methods: The patient-specific source of Tomotherapy and Varian TrueBeam Linacs was modeled using the PS approach. For the helical Tomotherapy case, the PS data were calculated in our previous study (Su et al. 2014 41(7) Medical Physics). For the single-view Varian TrueBeam case, we analytically derived them from the raw patient-independent PS data in IAEA’s database, partial geometry information of the jaw and MLC as well as the fluence map. The phantom was generated from DICOM images. The Monte Carlo simulation was performed by ARCHER-MIC and GPU codes, which were benchmarked against a modified parallel DPM code. Software micro-optimization was systematically conducted, and was focused on SIMD vectorization of tight for-loops and data prefetch, with the ultimate goal of increasing 512-bit register utilization and reducing memory access latency. Results: Dose calculation was performed for two clinical cases, a Tomotherapy-based prostate cancer treatment and a TrueBeam-based left breast treatment. ARCHER was verified against the DPM code. The statistical uncertainty of the dose to the PTV was less than 1%. Using double-precision, the total wall time of the multithreaded CPU code on a X5650 CPU was 339 seconds for the Tomotherapy case and 131 seconds for the TrueBeam, while on 3 5110P MICs it was reduced to 79 and 59 seconds, respectively. The single-precision GPU code on a K40 GPU took 45 seconds for the Tomotherapy dose calculation. Conclusion: We have extended ARCHER, the MIC and GPU-based Monte Carlo dose engine to Tomotherapy and Truebeam dose calculations.

  8. TU-AB-BRC-10: Modeling of Radiotherapy Linac Source Terms Using ARCHER Monte Carlo Code: Performance Comparison of GPU and MIC Computing Accelerators

    Energy Technology Data Exchange (ETDEWEB)

    Liu, T; Lin, H; Xu, X [Rensselaer Polytechnic Institute, Troy, NY (United States); Su, L [John Hopkins University, Baltimore, MD (United States); Shi, C [Saint Vincent Medical Center, Bridgeport, CT (United States); Tang, X [Memorial Sloan Kettering Cancer Center, West Harrison, NY (United States); Bednarz, B [University of Wisconsin, Madison, WI (United States)

    2016-06-15

    Purpose: (1) To perform phase space (PS) based source modeling for Tomotherapy and Varian TrueBeam 6 MV Linacs, (2) to examine the accuracy and performance of the ARCHER Monte Carlo code on a heterogeneous computing platform with Many Integrated Core coprocessors (MIC, aka Xeon Phi) and GPUs, and (3) to explore the software micro-optimization methods. Methods: The patient-specific source of Tomotherapy and Varian TrueBeam Linacs was modeled using the PS approach. For the helical Tomotherapy case, the PS data were calculated in our previous study (Su et al. 2014 41(7) Medical Physics). For the single-view Varian TrueBeam case, we analytically derived them from the raw patient-independent PS data in IAEA’s database, partial geometry information of the jaw and MLC as well as the fluence map. The phantom was generated from DICOM images. The Monte Carlo simulation was performed by ARCHER-MIC and GPU codes, which were benchmarked against a modified parallel DPM code. Software micro-optimization was systematically conducted, and was focused on SIMD vectorization of tight for-loops and data prefetch, with the ultimate goal of increasing 512-bit register utilization and reducing memory access latency. Results: Dose calculation was performed for two clinical cases, a Tomotherapy-based prostate cancer treatment and a TrueBeam-based left breast treatment. ARCHER was verified against the DPM code. The statistical uncertainty of the dose to the PTV was less than 1%. Using double-precision, the total wall time of the multithreaded CPU code on a X5650 CPU was 339 seconds for the Tomotherapy case and 131 seconds for the TrueBeam, while on 3 5110P MICs it was reduced to 79 and 59 seconds, respectively. The single-precision GPU code on a K40 GPU took 45 seconds for the Tomotherapy dose calculation. Conclusion: We have extended ARCHER, the MIC and GPU-based Monte Carlo dose engine to Tomotherapy and Truebeam dose calculations.

  9. Waste package reliability analysis

    International Nuclear Information System (INIS)

    Pescatore, C.; Sastre, C.

    1983-01-01

    Proof of future performance of a complex system such as a high-level nuclear waste package over a period of hundreds to thousands of years cannot be had in the ordinary sense of the word. The general method of probabilistic reliability analysis could provide an acceptable framework to identify, organize, and convey the information necessary to satisfy the criterion of reasonable assurance of waste package performance according to the regulatory requirements set forth in 10 CFR 60. General principles which may be used to evaluate the qualitative and quantitative reliability of a waste package design are indicated and illustrated with a sample calculation of a repository concept in basalt. 8 references, 1 table

  10. RepoTREND. The program package for the integrated long term safety analysis of final repository systems. Version 4.5 (State March 2016); RepoTREND. Das Programmpaket zur integrierten Langzeitsicherheitsanalyse von Endlagersystemen. Version 4.5 (Stand Maerz 2016)

    Energy Technology Data Exchange (ETDEWEB)

    Reiche, Tatiana

    2016-04-15

    The long-term safety analysis is the analysis of final repository behavior after closure includes the spreading of pollutants into the biosphere (mobilization and release of pollutants into the near field, radionuclide migration through the geosphere, radiation exposure in the biosphere) and the radiological consequences. The report describes the program package RepoTREND, the respective modules (near field, GeoTREND, BioTREND and probabilistic analyses), sequencing and postprocessing and the quality management.

  11. Shielding Calculations on Waste Packages – The Limits and Possibilities of different Calculation Methods by the example of homogeneous and inhomogeneous Waste Packages

    OpenAIRE

    Adams Mike; Smalian Silva

    2017-01-01

    For nuclear waste packages the expected dose rates and nuclide inventory are beforehand calculated. Depending on the package of the nuclear waste deterministic programs like MicroShield® provide a range of results for each type of packaging. Stochastic programs like “Monte-Carlo N-Particle Transport Code System” (MCNP®) on the other hand provide reliable results for complex geometries. However this type of program requires a fully trained operator and calculations are time consuming. The prob...

  12. Evaluation of system codes for analyzing naturally circulating gas loop

    International Nuclear Information System (INIS)

    Lee, Jeong Ik; No, Hee Cheon; Hejzlar, Pavel

    2009-01-01

    Steady-state natural circulation data obtained in a 7 m-tall experimental loop with carbon dioxide and nitrogen are presented in this paper. The loop was originally designed to encompass operating range of a prototype gas-cooled fast reactor passive decay heat removal system, but the results and conclusions are applicable to any natural circulation loop operating in regimes having buoyancy and acceleration parameters within the ranges validated in this loop. Natural circulation steady-state data are compared to numerical predictions by two system analysis codes: GAMMA and RELAP5-3D. GAMMA is a computational tool for predicting various transients which can potentially occur in a gas-cooled reactor. The code has a capability of analyzing multi-dimensional multi-component mixtures and includes models for friction, heat transfer, chemical reaction, and multi-component molecular diffusion. Natural circulation data with two gases show that the loop operates in the deteriorated turbulent heat transfer (DTHT) regime which exhibits substantially reduced heat transfer coefficients compared to the forced turbulent flow. The GAMMA code with an original heat transfer package predicted conservative results in terms of peak wall temperature. However, the estimated peak location did not successfully match the data. Even though GAMMA's original heat transfer package included mixed-convection regime, which is a part of the DTHT regime, the results showed that the original heat transfer package could not reproduce the data with sufficient accuracy. After implementing a recently developed correlation and corresponding heat transfer regime map into GAMMA to cover the whole range of the DTHT regime, we obtained better agreement with the data. RELAP5-3D results are discussed in parallel.

  13. SPHINX experimenters information package

    International Nuclear Information System (INIS)

    Zarick, T.A.

    1996-08-01

    This information package was prepared for both new and experienced users of the SPHINX (Short Pulse High Intensity Nanosecond X-radiator) flash X-Ray facility. It was compiled to help facilitate experiment design and preparation for both the experimenter(s) and the SPHINX operational staff. The major areas covered include: Recording Systems Capabilities,Recording System Cable Plant, Physical Dimensions of SPHINX and the SPHINX Test cell, SPHINX Operating Parameters and Modes, Dose Rate Map, Experiment Safety Approval Form, and a Feedback Questionnaire. This package will be updated as the SPHINX facilities and capabilities are enhanced

  14. Nirex - The complete package

    International Nuclear Information System (INIS)

    Carr, N.A.; Rarok, M.

    2002-01-01

    Through continuous interaction with its customers, Nirex had identified the need for a comprehensive range of waste containers, reflecting the variety of wastes and operational undertakings. The current range consists of five standard containers. Standardisation is introduced across all waste packages to enable the safe and efficient operation of future waste management facilities. The practical lessons learned during the development of standard containers are in turn reflected in the container design work that Nirex has undertaken. They are also fed into the advice given to customers during evaluation of waste packaging proposals. (author)

  15. Ada Namelist Package

    Science.gov (United States)

    Klumpp, Allan R.

    1991-01-01

    Ada Namelist Package, developed for Ada programming language, enables calling program to read and write FORTRAN-style namelist files. Features are: handling of any combination of types defined by user; ability to read vectors, matrices, and slices of vectors and matrices; handling of mismatches between variables in namelist file and those in programmed list of namelist variables; and ability to avoid searching entire input file for each variable. Principle benefits derived by user: ability to read and write namelist-readable files, ability to detect most file errors in initialization phase, and organization keeping number of instantiated units to few packages rather than to many subprograms.

  16. Evaluation on the structural soundness of the transport package for low-level radioactive waste for subsurface disposal against aircraft impact by finite element method

    International Nuclear Information System (INIS)

    Itoh, Chihiro

    2009-01-01

    The structural analysis of aircraft crush on the transport package for low-level radioactive waste was performed using the impact force which was already used for the evaluation of the high-level waste transport package by LSDYNA code. The transport package was deformed, and stresses due to the crush exceeded elastic range. However, plastic strains yieled in the package were far than the elongation of the materials and the body of the package did not contact the disposal packages due to the deformation of the package. Therefore, it was confirmed that the package keeps its integrity against aircraft crush. (author)

  17. Development of a numerical code for the prediction of the long-term behavior of the underground facilities for the high-level radioactive waste disposal

    International Nuclear Information System (INIS)

    Sawada, Masataka; Okada, Tetsuji; Hasegawa, Takuma

    2006-01-01

    Complicated phenomena originated by thermo-hydro-mechanical coupling behavior will occur in the near-field of geological disposal of nuclear waste. Development of a numerical evaluation method for such phenomena is important in order to make a reasonable repository design and a safety assessment. In order to achieve the objective above, a numerical model using the equations which can evaluate the swelling characteristics of buffer materials based on the diffusive double layer theory is proposed, and a numerical scheme for the thermo-hydro-mechanical coupled analysis including the swelling model is constructed. The proposed swelling model can reproduce the behavior observed during both swelling pressure tests and swelling deformation tests. When the developed numerical code is applied to the laboratory heater test using a bentonite specimen, it can reproduce the thermal gradient, the distribution of saturation rate and the variation of porosity. The developed numerical code will be applied to well-controlled laboratory tests and full-scale in-situ tests in the future work. In order to apply to the various geochemical conditions around the engineered barrier, chemical component will be coupled to the present numerical code. (author)

  18. Physical and chemical characteristics of cashew nut flour stored and packaged with different packages

    Directory of Open Access Journals (Sweden)

    Bruna Carolina GADANI

    Full Text Available Abstract The aim of this study was to evaluate the physical and chemical changes in cashew nut flour under different packaging and storage conditions. The flour samples were characterized according to their chemical composition, packaged in polystyrene trays associated with covering chloride polyvinyl (PVC, plastic pot of polyethylene terephthalate (PET, plastic packaging polyethylene (POL and without coating polystyrene trays (CON. All packages were stored for 225 days, evaluating every 45 days the moisture content of the flour as well as extracting their oil by the cold method for characterization on the following indexes: acidity, saponification, iodine, peroxide and refraction. There was slight and gradual increase in the moisture content from the 45th to the 225th day, especially for products packaged in PVC and without coating polystyrene trays (CON. The oil saponification indexes showed gradual decrease during storage, especially in flour without packaging (CON. There was a gradual increase in the iodine index until the 135th day, followed by decay. The acidity and peroxide indexes increased along with the storage time. However packages with lower oxygen and moisture permeability, such as PET and POL, minimized such changes, being therefore the most suitable for long-term storage.

  19. Warpage of QFN Package in Post Mold Cure Process of integrated circuit packaging

    Science.gov (United States)

    Sriwithoon, Nattha; Ugsornrat, Kessararat; Srisuwitthanon, Warayoot; Thonglor, Panakamon

    2017-09-01

    This research studied about warpage of QFN package in post mold cure process of integrated circuit (IC) packages using pre-plated (PPF) leadframe. For IC package, epoxy molding compound (EMC) are molded by cross linking of compound stiffness but incomplete crosslinked network and leading the fully cured thermoset by post mold cure (PMC) process. The cure temperature of PMC can change microstructure of EMC in term of stress inside the package and effect to warpage of the package due to coefficient of thermal expansion (CTE) between EMC and leadframe. In experiment, cure temperatures were varied to check the effect of internal stress due to different cure temperature after completed post mold cure for TDFN 2×3 8L. The cure temperature were varied with 180 °C, 170 °C, 160 °C, and 150°C with cure time 4 and 6 hours, respectively. For analysis, the TDFN 2×3 8L packages were analyzed the warpage by thickness gauge and scanning acoustic microscope (SAM) after take the test samples out from the oven cure. The results confirmed that effect of different CTE between EMC and leadframe due to different cure temperature resulting to warpage of the TDFN 2×3 8L packages.

  20. Packaging Solutions : Delivering customer value through Logistical Packaging: A Case Study at Stora Enso Packaging

    OpenAIRE

    Shan, Kun; Julius, Joezer

    2015-01-01

    AbstractBackground;Despite of the significant role of packaging within logistics and supply chain management, packaging is infrequently studied as focal point in supply chain. Most of the previous logistics research studies tend to explain the integration between packaging and logistics through logistical packaging. In very rare cases, the studies mentioned about customer value. Therefore the major disadvantage of these studies is that, they didn’t consider logistical packaging and customer v...

  1. Coding Labour

    Directory of Open Access Journals (Sweden)

    Anthony McCosker

    2014-03-01

    Full Text Available As well as introducing the Coding Labour section, the authors explore the diffusion of code across the material contexts of everyday life, through the objects and tools of mediation, the systems and practices of cultural production and organisational management, and in the material conditions of labour. Taking code beyond computation and software, their specific focus is on the increasingly familiar connections between code and labour with a focus on the codification and modulation of affect through technologies and practices of management within the contemporary work organisation. In the grey literature of spreadsheets, minutes, workload models, email and the like they identify a violence of forms through which workplace affect, in its constant flux of crisis and ‘prodromal’ modes, is regulated and governed.

  2. Nuclear-waste-package materials degradation modes and accelerated testing

    International Nuclear Information System (INIS)

    1981-09-01

    This report reviews the materials degradation modes that may affect the long-term behavior of waste packages for the containment of nuclear waste. It recommends an approach to accelerated testing that can lead to the qualification of waste package materials in specific repository environments in times that are short relative to the time period over which the waste package is expected to provide containment. This report is not a testing plan but rather discusses the direction for research that might be considered in developing plans for accelerated testing of waste package materials and waste forms

  3. AN ADA NAMELIST PACKAGE

    Science.gov (United States)

    Klumpp, A. R.

    1994-01-01

    The Ada Namelist Package, developed for the Ada programming language, enables a calling program to read and write FORTRAN-style namelist files. A namelist file consists of any number of assignment statements in any order. Features of the Ada Namelist Package are: the handling of any combination of user-defined types; the ability to read vectors, matrices, and slices of vectors and matrices; the handling of mismatches between variables in the namelist file and those in the programmed list of namelist variables; and the ability to avoid searching the entire input file for each variable. The principle user benefits of this software are the following: the ability to write namelist-readable files, the ability to detect most file errors in the initialization phase, a package organization that reduces the number of instantiated units to a few packages rather than to many subprograms, a reduced number of restrictions, and an increased execution speed. The Ada Namelist reads data from an input file into variables declared within a user program. It then writes data from the user program to an output file, printer, or display. The input file contains a sequence of assignment statements in arbitrary order. The output is in namelist-readable form. There is a one-to-one correspondence between namelist I/O statements executed in the user program and variables read or written. Nevertheless, in the input file, mismatches are allowed between assignment statements in the file and the namelist read procedure statements in the user program. The Ada Namelist Package itself is non-generic. However, it has a group of nested generic packages following the nongeneric opening portion. The opening portion declares a variety of useraccessible constants, variables and subprograms. The subprograms are procedures for initializing namelists for reading, reading and writing strings. The subprograms are also functions for analyzing the content of the current dataset and diagnosing errors. Two nested

  4. MEMS packaging: state of the art and future trends

    Science.gov (United States)

    Bossche, Andre; Cotofana, Carmen V. B.; Mollinger, Jeff R.

    1998-07-01

    Now that the technology for Integrated sensor and MEMS devices has become sufficiently mature to allow mass production, it is expected that the prices of bare chips will drop dramatically. This means that the package prices will become a limiting factor in market penetration, unless low cost packaging solutions become available. This paper will discuss the developments in packaging technology. Both single-chip and multi-chip packaging solutions will be addressed. It first starts with a discussion on the different requirements that have to be met; both from a device point of view (open access paths to the environment, vacuum cavities, etc.) and from the application point of view (e.g. environmental hostility). Subsequently current technologies are judged on their applicability for MEMS and sensor packaging and a forecast is given for future trends. It is expected that the large majority of sensing devices will be applied in relative friendly environments for which plastic packages would suffice. Therefore, on the short term an important role is foreseen for recently developed plastic packaging techniques such as precision molding and precision dispensing. Just like in standard electronic packaging, complete wafer level packaging methods for sensing devices still have a long way to go before they can compete with the highly optimized and automated plastic packaging processes.

  5. Tritium immobilization and packaging using metal hydrides

    International Nuclear Information System (INIS)

    Holtslander, W.J.; Yaraskavitch, J.M.

    1981-04-01

    Tritium recovered from CANDU heavy water reactors will have to be packaged and stored in a safe manner. Tritium will be recovered in the elemental form, T 2 . Metal tritides are effective compounds in which to immobilize the tritium as a stable non-reactive solid with a high tritium capacity. The technology necessary to prepare hydrides of suitable metals, such as titanium and zirconium, have been developed and the properties of the prepared materials evaluated. Conceptual designs of packages for containing metal tritides suitable for transportation and long-term storage have been made and initial testing started. (author)

  6. Unit-of-Use Versus Traditional Bulk Packaging

    Directory of Open Access Journals (Sweden)

    Tiffany So

    2012-01-01

    Full Text Available Background: The choice between unit-of-use versus traditional bulk packaging in the US has long been a continuous debate for drug manufacturers and pharmacies in order to have the most efficient and safest practices. Understanding the benefits of using unit-of-use packaging over bulk packaging by US drug manufacturers in terms of workflow efficiency, economical costs and medication safety in the pharmacy is sometimes challenging.Methods: A time-saving study comparing the time saved using unit-of-use packaging versus bulk packaging, was examined. Prices between unit-of-use versus bulk packages were compared by using the Red Book: Pharmacy’s Fundamental Reference. Other articles were reviewed on the topics of counterfeiting, safe labeling, and implementation of unit-of-use packaging. Lastly, a cost-saving study was reviewed showing how medication adherence, due to improved packaging, could be cost-effective for patients.Results: When examining time, costs, medication adherence, and counterfeiting arguments, unit-of-use packaging proved to be beneficial for patients in all these terms.

  7. Unit-of-Use Versus Traditional Bulk Packaging

    Directory of Open Access Journals (Sweden)

    Tiffany So

    2012-01-01

    Full Text Available Background: The choice between unit-of-use versus traditional bulk packaging in the US has long been a continuous debate for drug manufacturers and pharmacies in order to have the most efficient and safest practices. Understanding the benefits of using unit-of-use packaging over bulk packaging by US drug manufacturers in terms of workflow efficiency, economical costs and medication safety in the pharmacy is sometimes challenging. Methods: A time-saving study comparing the time saved using unit-of-use packaging versus bulk packaging, was examined. Prices between unit-of-use versus bulk packages were compared by using the Red Book: Pharmacy's Fundamental Reference. Other articles were reviewed on the topics of counterfeiting, safe labeling, and implementation of unit-of-use packaging. Lastly, a cost-saving study was reviewed showing how medication adherence, due to improved packaging, could be cost-effective for patients. Results: When examining time, costs, medication adherence, and counterfeiting arguments, unit-of-use packaging proved to be beneficial for patients in all these terms.   Type: Student Project

  8. Transportation package design using numerical optimization

    International Nuclear Information System (INIS)

    Harding, D.C.; Witkowski, W.R.

    1993-01-01

    Since the design of transportation packages involves a complex coupling of structural, thermal and radiation shielding analyses and must follow very strict design constraints, numerical optimization provides the potential for more efficient container designs. In numerical optimization, the requirements of the design problem are mathematically formulated through the use of an objective function and constraints. The objective function(s), e.g., package weight, cost, volume, or combination thereof, is the function to be minimized or maximized by altering a set of design variables that define the package's shape and dimensions. Constraints are limitations on the performance of the system, such as resisting structural and thermal accident environments. Two constraints defined for an example wire mesh composite Type B package are: 1) deformation in the containment vessel seal region remains small enough throughout the 10 CFR-71 accident conditions to meet containment criteria, and 2) the elastomeric seal region remains below its operational temperature limit to guarantee seal integrity in the fire environment. The first constraint of a minimum energy absorbing layer thickness is evaluated with finite element analyses of the proposed dynamic crush accident criteria. The second constraint is evaluated with a 1-D transient thermal finite difference code parametrized for variable composite layer thicknesses, and is integrated with the optimization process. (J.P.N.)

  9. Module 13: Bulk Packaging Shipments by Highway

    International Nuclear Information System (INIS)

    Przybylski, J.L.

    1994-07-01

    The Hazardous Materials Modular Training Program provides participating United States Department of Energy (DOE) sites with a basic, yet comprehensive, hazardous materials transportation training program for use onsite. This program may be used to assist individual program entities to satisfy the general awareness, safety training, and function specific training requirements addressed in Code of Federal Regulation (CFR), Title 49, Part 172, Subpart H -- ''Training.'' Module 13 -- Bulk Packaging Shipments by Highway is a supplement to the Basic Hazardous Materials Workshop. Module 13 -- Bulk Packaging Shipments by Highway focuses on bulk shipments of hazardous materials by highway mode, which have additional or unique requirements beyond those addressed in the ten module core program. Attendance in this course of instruction should be limited to those individuals with work experience in transporting hazardous materials utilizing bulk packagings and who have completed the Basic Hazardous Materials Workshop or an equivalent. Participants will become familiar with the rules and regulations governing the transportation by highway of hazardous materials in bulk packagings and will demonstrate the application of these requirements through work projects and examination

  10. Radioactive waste disposal package

    Science.gov (United States)

    Lampe, Robert F.

    1986-11-04

    A radioactive waste disposal package comprising a canister for containing vitrified radioactive waste material and a sealed outer shell encapsulating the canister. A solid block of filler material is supported in said shell and convertible into a liquid state for flow into the space between the canister and outer shell and subsequently hardened to form a solid, impervious layer occupying such space.

  11. ROBATEL RS 24 packaging

    International Nuclear Information System (INIS)

    Robatel, M.; Bochard, C.

    1986-01-01

    The ROBATEL RS 24 packaging for light-water reactor fuel assemblies is discussed. The cask uses a strength frame of carbon steel sheets, lead gamma ray shielding, neutron shielding of aluminous concrete, a thermal protection system for the gamma shielding, and a natural convection thermal transfer system

  12. The Swarm Magnetometry Package

    DEFF Research Database (Denmark)

    Merayo, José M.G.; Jørgensen, John Leif; Friis-Christensen, Eigil

    2008-01-01

    The Swarm mission under the ESA's Living Planet Programme is planned for launch in 2010 and consists of a constellation of three satellites at LEO. The prime objective of Swarm is to measure the geomagnetic field with unprecedented accuracy in space and time. The magnetometry package consists...

  13. Learning Activity Package, Algebra.

    Science.gov (United States)

    Evans, Diane

    A set of ten teacher-prepared Learning Activity Packages (LAPs) in beginning algebra and nine in intermediate algebra, these units cover sets, properties of operations, number systems, open expressions, solution sets of equations and inequalities in one and two variables, exponents, factoring and polynomials, relations and functions, radicals,…

  14. Geothermal Greenhouse Information Package

    Energy Technology Data Exchange (ETDEWEB)

    Rafferty, K. [P.E.; Boyd, T. [ed.

    1997-01-01

    This package of information is intended to provide a foundation of background information for developers of geothermal greenhouses. The material is divided into seven sections covering such issues as crop culture and prices, operating costs for greenhouses, heating system design, vendors and a list of other sources of information.

  15. Openability of tamperproof packaging

    NARCIS (Netherlands)

    Del Castillo C., A.; Wever, R.; Buijs, P.J.; Stevels, A.

    2007-01-01

    Communication, product protection and presentation are three key aspects in the world of packaging nowadays. Due to a retail landscape consisting of large stores, displaying packed products on the shelves in self-service environments, these aspects become increasingly important, not only for Fast

  16. Packaging LLW and ILW

    International Nuclear Information System (INIS)

    Flowers, R.H.; Owen, R.G.

    1991-01-01

    Low level waste (LLW) accounts for 70-80% by volume of all radioactive wastes produced by the nuclear industry. It has low specific activity, negligible actinide content and requires little, if any, shielding to protect workers. Volume reduction for LLW of high volume but low density may be achieved by incineration and compaction as appropriate, before packaging for disposal by near surface burial. Intermediate level waste (ILW) is treated and packed to convert it into a stable form to minimize any release of activity and make handling easier. The matrix chosen for immobilization, usually cement, polymers or bitumen, depends on the nature of the waste and the acceptance criteria of the disposal facility. The special case of LLW and ILW which will arise from reactor decommissioning is discussed. Packaging methods adopted by individual countries are reviewed. The range of costs involved for packaging ILW is indicated. There is no international consensus on the performance required from packaged waste to ensure its suitability both for interim storage and final disposal. (UK)

  17. Aquaculture Information Package

    Energy Technology Data Exchange (ETDEWEB)

    Boyd, T.; Rafferty, K. [editors

    1998-01-01

    This package of information is intended to provide background to developers of geothermal aquaculture projects. The material is divided into eight sections and includes information on market and price information for typical species, aquaculture water quality issues, typical species culture information, pond heat loss calculations, an aquaculture glossary, regional and university aquaculture offices and state aquaculture permit requirements.

  18. CH Packaging Maintenance Manual

    International Nuclear Information System (INIS)

    Washington TRU Solutions

    2002-01-01

    This procedure provides instructions for performing inner containment vessel (ICV) and outer containment vessel (OCV) maintenance and periodic leakage rate testing on the following packaging seals and corresponding seal surfaces using a nondestructive helium (He) leak test. In addition, this procedure provides instructions for performing ICV and OCV structural pressure tests

  19. Printer Graphics Package

    Science.gov (United States)

    Blanchard, D. C.

    1986-01-01

    Printer Graphics Package (PGP) is tool for making two-dimensional symbolic plots on line printer. PGP created to support development of Heads-Up Display (HUD) simulation. Standard symbols defined with HUD in mind. Available symbols include circle, triangle, quadrangle, window, line, numbers, and text. Additional symbols easily added or built up from available symbols.

  20. Type B drum packages

    International Nuclear Information System (INIS)

    McCoy, J.C.

    1994-08-01

    The Type B drum packages (TBD) are conceptualized as a family of containers in which a single 208 L or 114 L (55 gal or 30 gal) drum containing Type B quantities of radioactive material (RAM) can be packaged for shipment. The TBD containers are being developed to fill a void in the packaging and transportation capabilities of the U.S. Department of Energy as no container packaging single drums of Type B RAM exists offering double containment. Several multiple-drum containers currently exist, as well as a number of shielded casks, but the size and weight of these containers present many operational challenges for single-drum shipments. As an alternative, the TBD containers will offer up to three shielded versions (light, medium, and heavy) and one unshielded version, each offering single or optional double containment for a single drum. To reduce operational complexity, all versions will share similar design and operational features where possible. The primary users of the TBD containers are envisioned to be any organization desiring to ship single drums of Type B RAM, such as laboratories, waste retrieval activities, emergency response teams, etc. Currently, the TBD conceptual design is being developed with the final design and analysis to be completed in 1995 to 1996. Testing and certification of the unshielded version are planned to be completed in 1996 to 1997 with production to begin in 1997 to 1998

  1. Waste disposal package

    Science.gov (United States)

    Smith, M.J.

    1985-06-19

    This is a claim for a waste disposal package including an inner or primary canister for containing hazardous and/or radioactive wastes. The primary canister is encapsulated by an outer or secondary barrier formed of a porous ceramic material to control ingress of water to the canister and the release rate of wastes upon breach on the canister. 4 figs.

  2. Innovations and Changes of Procedure Civil Code, and Maintenance of the Subjectivism Term "Resources Insufficient" for Granting of Justice of Gratuity

    Directory of Open Access Journals (Sweden)

    Juliane Dziubate Krefta

    2016-10-01

    Full Text Available This work is focused on the analysis of Gratuity of Justice as an important mechanism of access to justice. It aims to address the news about the Gratuity of Justice in the current Civil Procedure Code, recently amended by Law nº. 13,105/2015, bringing out important changes that fixed the obsolete Law nº. 1,060/50. Aims to reflect about the requirement for granting the benefit of free justice, namely the lack of resources, and its high degree of subjectivity, emphasizing some positive and negative issues related to it.

  3. Benchmark problems for radiological assessment codes. Final report

    International Nuclear Information System (INIS)

    Mills, M.; Vogt, D.; Mann, B.

    1983-09-01

    This report describes benchmark problems to test computer codes used in the radiological assessment of high-level waste repositories. The problems presented in this report will test two types of codes. The first type of code calculates the time-dependent heat generation and radionuclide inventory associated with a high-level waste package. Five problems have been specified for this code type. The second code type addressed in this report involves the calculation of radionuclide transport and dose-to-man. For these codes, a comprehensive problem and two subproblems have been designed to test the relevant capabilities of these codes for assessing a high-level waste repository setting

  4. Hazardous materials package performance regulations

    International Nuclear Information System (INIS)

    Russell, N.A.; Glass, R.E.; McClure, J.D.; Finley, N.C.

    1992-01-01

    The hazardous materials (hazmat) packaging development and certification process is currently defined by two different regulatory philosophies, one based on specification packagings and the other based on performance standards. With specification packagings, a packaging is constructed according to an agreed set of design specifications. In contrast, performance standards do not specify the packaging design; they specify performance standards that a packaging design must be able to pass before it can be certified for transport. The packaging can be designed according to individual needs as long as it meets these performance standards. Performance standards have been used nationally and internationally for about 40 years to certify radioactive materials (RAM) packagings. It is reasonable to state that for RAM transport, performance specifications have maintained transport safety. A committee of United Nation's experts recommended the performance standard philosophy as the preferred regulation method for hazmat packaging. Performance standards for hazmat packagings smaller than 118 gallons have been adopted in 49CFR178. Packagings for materials that are classified as toxic-by-inhalation must comply with the performance standards by October 1, 1993, and packagings for all other classes of hazardous materials covered must comply by October 1, 1996. For packages containing bulk (in excess of 188 gallons) quantities of materials that are extremely toxic by inhalation, there currently are no performance requirements. This paper discusses a Hazmat Packaging Performance Evaluation (HPPE) project to look at the subset of bulk packagings that are larger than 2000 gallons. The objectives of this project are the evaluate current hazmat specification packagings and develop supporting documentation for determining performance requirements for packagings in excess of 2000 gallons that transport hazardous materials that have been classified as extremely toxic by inhalation (METBI)

  5. Symmetric Rock Fall on Waste Package

    International Nuclear Information System (INIS)

    Sreten Mastilovic

    2001-01-01

    The objective of this calculation is to determine the structural response of the Naval SNF (spent nuclear fuel) Waste Package (WP) and the emplacement pallet (EP) subjected to the rock fall DBE (design basis event) dynamic loads. The scope of this calculation is limited to reporting the calculation results in terms of stress intensities and residual stresses in the WP, and stress intensities and maximum permanent downward displacements of the EP-lifting surface. The information provided by the sketches (Attachment I) is that of the potential design of the type of WP and EP considered in this calculation, and all obtained results are valid for those designs only. This calculation is associated with the waste package design and is performed by the Waste Package Design Section in accordance with Reference 24. AP-3.124, ''Calculations'', is used to perform the calculation and develop the document

  6. In-Package Chemistry Abstraction

    Energy Technology Data Exchange (ETDEWEB)

    P.S. Domski

    2003-07-21

    The work associated with the development of this model report was performed in accordance with the requirements established in ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA'' (BSC 2002a). The in-package chemistry model and in-package chemistry model abstraction are developed to predict the bulk chemistry inside of a failed waste package and to provide simplified expressions of that chemistry. The purpose of this work is to provide the abstraction model to the Performance Assessment Project and the Waste Form Department for development of geochemical models of the waste package interior. The scope of this model report is to describe the development and validation of the in-package chemistry model and in-package chemistry model abstraction. The in-package chemistry model will consider chemical interactions of water with the waste package materials and the waste form for commercial spent nuclear fuel (CSNF) and codisposed high-level waste glass (HLWG) and N Reactor spent fuel (CDNR). The in-package chemistry model includes two sub-models, the first a water vapor condensation (WVC) model, where water enters a waste package as vapor and forms a film on the waste package components with subsequent film reactions with the waste package materials and waste form--this is a no-flow model, the reacted fluids do not exit the waste package via advection. The second sub-model of the in-package chemistry model is the seepage dripping model (SDM), where water, water that may have seeped into the repository from the surrounding rock, enters a failed waste package and reacts with the waste package components and waste form, and then exits the waste package with no accumulation of reacted water in the waste package. Both of the submodels of the in-package chemistry model are film models in contrast to past in-package chemistry models where all of the waste package pore space was filled with water. The

  7. In-Package Chemistry Abstraction

    International Nuclear Information System (INIS)

    P.S. Domski

    2003-01-01

    The work associated with the development of this model report was performed in accordance with the requirements established in ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA'' (BSC 2002a). The in-package chemistry model and in-package chemistry model abstraction are developed to predict the bulk chemistry inside of a failed waste package and to provide simplified expressions of that chemistry. The purpose of this work is to provide the abstraction model to the Performance Assessment Project and the Waste Form Department for development of geochemical models of the waste package interior. The scope of this model report is to describe the development and validation of the in-package chemistry model and in-package chemistry model abstraction. The in-package chemistry model will consider chemical interactions of water with the waste package materials and the waste form for commercial spent nuclear fuel (CSNF) and codisposed high-level waste glass (HLWG) and N Reactor spent fuel (CDNR). The in-package chemistry model includes two sub-models, the first a water vapor condensation (WVC) model, where water enters a waste package as vapor and forms a film on the waste package components with subsequent film reactions with the waste package materials and waste form--this is a no-flow model, the reacted fluids do not exit the waste package via advection. The second sub-model of the in-package chemistry model is the seepage dripping model (SDM), where water, water that may have seeped into the repository from the surrounding rock, enters a failed waste package and reacts with the waste package components and waste form, and then exits the waste package with no accumulation of reacted water in the waste package. Both of the submodels of the in-package chemistry model are film models in contrast to past in-package chemistry models where all of the waste package pore space was filled with water. The current in-package

  8. Optimal codes as Tanner codes with cyclic component codes

    DEFF Research Database (Denmark)

    Høholdt, Tom; Pinero, Fernando; Zeng, Peng

    2014-01-01

    In this article we study a class of graph codes with cyclic code component codes as affine variety codes. Within this class of Tanner codes we find some optimal binary codes. We use a particular subgraph of the point-line incidence plane of A(2,q) as the Tanner graph, and we are able to describe ...

  9. Modular assembly of chimeric phi29 packaging RNAs that support DNA packaging.

    Science.gov (United States)

    Fang, Yun; Shu, Dan; Xiao, Feng; Guo, Peixuan; Qin, Peter Z

    2008-08-08

    The bacteriophage phi29 DNA packaging motor is a protein/RNA complex that can produce strong force to condense the linear-double-stranded DNA genome into a pre-formed protein capsid. The RNA component, called the packaging RNA (pRNA), utilizes magnesium-dependent inter-molecular base-pairing interactions to form ring-shaped complexes. The pRNA is a class of non-coding RNA, interacting with phi29 motor proteins to enable DNA packaging. Here, we report a two-piece chimeric pRNA construct that is fully competent in interacting with partner pRNA to form ring-shaped complexes, in packaging DNA via the motor, and in assembling infectious phi29 virions in vitro. This is the first example of a fully functional pRNA assembled using two non-covalently interacting fragments. The results support the notion of modular pRNA architecture in the phi29 packaging motor.

  10. Packaging design criteria for the Hanford Ecorok Packaging

    International Nuclear Information System (INIS)

    Mercado, M.S.

    1996-01-01

    The Hanford Ecorok Packaging (HEP) will be used to ship contaminated water purification filters from K Basins to the Central Waste Complex. This packaging design criteria documents the design of the HEP, its intended use, and the transportation safety criteria it is required to meet. This information will serve as a basis for the safety analysis report for packaging

  11. ARC Code TI: Optimal Alarm System Design and Implementation

    Data.gov (United States)

    National Aeronautics and Space Administration — An optimal alarm system can robustly predict a level-crossing event that is specified over a fixed prediction horizon. The code contained in this packages provides...

  12. NORTICA—a new code for cyclotron analysis

    Science.gov (United States)

    Gorelov, D.; Johnson, D.; Marti, F.

    2001-12-01

    The new package NORTICA (Numerical ORbit Tracking In Cyclotrons with Analysis) of computer codes for beam dynamics simulations is under development at NSCL. The package was started as a replacement for the code MONSTER [1] developed in the laboratory in the past. The new codes are capable of beam dynamics simulations in both CCF (Coupled Cyclotron Facility) accelerators, the K500 and K1200 superconducting cyclotrons. The general purpose of this package is assisting in setting and tuning the cyclotrons taking into account the main field and extraction channel imperfections. The computer platform for the package is Alpha Station with UNIX operating system and X-Windows graphic interface. A multiple programming language approach was used in order to combine the reliability of the numerical algorithms developed over the long period of time in the laboratory and the friendliness of modern style user interface. This paper describes the capability and features of the codes in the present state.

  13. 49 CFR 178.502 - Identification codes for packagings.

    Science.gov (United States)

    2010-10-01

    ... construction, as follows: (i) “A” means steel (all types and surface treatments). (ii) “B” means aluminum. (iii) “C” means natural wood. (iv) “D” means plywood. (v) “F” means reconstituted wood. (vi) “G” means...

  14. ETR/ITER systems code

    Energy Technology Data Exchange (ETDEWEB)

    Barr, W.L.; Bathke, C.G.; Brooks, J.N.; Bulmer, R.H.; Busigin, A.; DuBois, P.F.; Fenstermacher, M.E.; Fink, J.; Finn, P.A.; Galambos, J.D.; Gohar, Y.; Gorker, G.E.; Haines, J.R.; Hassanein, A.M.; Hicks, D.R.; Ho, S.K.; Kalsi, S.S.; Kalyanam, K.M.; Kerns, J.A.; Lee, J.D.; Miller, J.R.; Miller, R.L.; Myall, J.O.; Peng, Y-K.M.; Perkins, L.J.; Spampinato, P.T.; Strickler, D.J.; Thomson, S.L.; Wagner, C.E.; Willms, R.S.; Reid, R.L. (ed.)

    1988-04-01

    A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs.

  15. ETR/ITER systems code

    International Nuclear Information System (INIS)

    Barr, W.L.; Bathke, C.G.; Brooks, J.N.

    1988-04-01

    A tokamak systems code capable of modeling experimental test reactors has been developed and is described in this document. The code, named TETRA (for Tokamak Engineering Test Reactor Analysis), consists of a series of modules, each describing a tokamak system or component, controlled by an optimizer/driver. This code development was a national effort in that the modules were contributed by members of the fusion community and integrated into a code by the Fusion Engineering Design Center. The code has been checked out on the Cray computers at the National Magnetic Fusion Energy Computing Center and has satisfactorily simulated the Tokamak Ignition/Burn Experimental Reactor II (TIBER) design. A feature of this code is the ability to perform optimization studies through the use of a numerical software package, which iterates prescribed variables to satisfy a set of prescribed equations or constraints. This code will be used to perform sensitivity studies for the proposed International Thermonuclear Experimental Reactor (ITER). 22 figs., 29 tabs

  16. SUSTAINABLE PACKAGING SOLUTIONS FOR ORGANIC FRESH BERRIES

    Directory of Open Access Journals (Sweden)

    Elisabeta Elena TĂNASE

    2017-12-01

    Full Text Available Climate changes and particularly global warming are topics carefully treated by specialists already since decades. The most pregnant factor that influences climate change is pollution, namely the high level carbon dioxide emissions. Besides other substances used by the most of the industries (oil, charcoal, fertilizers, etc., plastics are not to be ignored when talking about pollution. Plastic waste affects animals and humans, as well as their habitat. In this respect, food industry engages in preserving the good functioning of the environment by developing and using biodegradable and bio-based resources for food packaging. The aim of this literature review was to identify the optimal sustainable packaging solution used for berries. The results of the study pointed out that the most used environmentally friendly packaging technique is the one that involves modified atmosphere. In terms of packaging materials, the literature is limited when it comes to biodegradable/bio-based solutions. However, active packaging gains popularity among researchers, considering the endless possibilities to include sustainable compounds in a biopolymer based matrix, in order to prolong the shelf-life of berries or fruits in general.

  17. Aztheca Code

    International Nuclear Information System (INIS)

    Quezada G, S.; Espinosa P, G.; Centeno P, J.; Sanchez M, H.

    2017-09-01

    This paper presents the Aztheca code, which is formed by the mathematical models of neutron kinetics, power generation, heat transfer, core thermo-hydraulics, recirculation systems, dynamic pressure and level models and control system. The Aztheca code is validated with plant data, as well as with predictions from the manufacturer when the reactor operates in a stationary state. On the other hand, to demonstrate that the model is applicable during a transient, an event occurred in a nuclear power plant with a BWR reactor is selected. The plant data are compared with the results obtained with RELAP-5 and the Aztheca model. The results show that both RELAP-5 and the Aztheca code have the ability to adequately predict the behavior of the reactor. (Author)

  18. Contribution to study and design of PWR plant simulation code

    International Nuclear Information System (INIS)

    Delourme, Didier.

    1980-11-01

    This paper presents an improvement of PICOLO, a package for PWR plants simulation. Its describes principally the integration to the code of a primary loop and pressurizer model and the corresponding control loops. Fast transients are tested on the packages and results are compared with real transients obtained on plants [fr

  19. Dual Use Packaging, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA seeks down-weighted packaging compatible with microwave preparation and perhaps high hydrostatic pressure processing. New packaging must satisfy NASA's 3-year...

  20. Food packaging history and innovations.

    Science.gov (United States)

    Risch, Sara J

    2009-09-23

    Food packaging has evolved from simply a container to hold food to something today that can play an active role in food quality. Many packages are still simply containers, but they have properties that have been developed to protect the food. These include barriers to oxygen, moisture, and flavors. Active packaging, or that which plays an active role in food quality, includes some microwave packaging as well as packaging that has absorbers built in to remove oxygen from the atmosphere surrounding the product or to provide antimicrobials to the surface of the food. Packaging has allowed access to many foods year-round that otherwise could not be preserved. It is interesting to note that some packages have actually allowed the creation of new categories in the supermarket. Examples include microwave popcorn and fresh-cut produce, which owe their existence to the unique packaging that has been developed.

  1. How to avoid errors in the design and fabrication of transportation packages

    International Nuclear Information System (INIS)

    Raske, D.T.

    1996-01-01

    The purpose of this paper is to discuss the errors and omissions most often identified when reviewing the design and fabrication of a packaging to transport high-level radioactive materials. The design and fabrication criteria recommended by the U.S. Department of Energy, Office of Facility Safety Analysis, for containment vessels of Type B commercial packagings containing high-level radioactive materials is based on the requirements of Section III, Division 1, Subsection NB of the ASME Boiler and Pressure Vessel Code. However, most packaging designers, engineers, and fabricators are intimidated by the sheer volume of requirements contained in the Code; as a result, the Code is not always followed and many requirements that do apply are often overlooked during preparation of the Safety Analysis Report for Packaging that constitutes the basis for evaluating the packaging for certification

  2. Type B plutonium transport package development that uses metallic filaments and composite materials

    International Nuclear Information System (INIS)

    Pierce, J.D.; Moya, J.L.; McClure, J.D.; Hohnstreiter, G.F.; Golliher, K.G.

    1992-01-01

    A new design concept for a Type B transport packaging for transporting plutonium and uranium has been developed by the Transportation Systems Department at Sandia National Laboratories (SNL). The new design came about following a review of current packagings, projected future transportation needs, and current and future regulatory requirements. United States packaging, regulations specified in Title 49, Code of Federal Regulations Parts 173.416 and 173.417 (for fissile materials) offer parallel paths under the heading of authorized Type B packages for the transport of greater than A 1 or A 2 quantities of radioactive material. These pathways are for certified Type B packagings and specification packagings. Consequently, a review was made of both type B and specification packages. A request for comment has been issued by the US Nuclear Regulatory Commission (NRC) for proposed changes to Title 10, Code of Federal Regulations Part 71. These regulations may therefore change in the near future. The principle proposed regulation change that would affect this type of package is the addition of a dynamic crush requirement for certain packagings. The US Department of Transportation (DOT) may also re-evaluate the specifications in 49 CFR that authorize the fabrication and use of specification packagings. Therefore, packaging, options were considered that will meet expected new regulations and provide shipment capability for the US Department of Energy well into the future

  3. Vocable Code

    DEFF Research Database (Denmark)

    Soon, Winnie; Cox, Geoff

    2018-01-01

    a computational and poetic composition for two screens: on one of these, texts and voices are repeated and disrupted by mathematical chaos, together exploring the performativity of code and language; on the other, is a mix of a computer programming syntax and human language. In this sense queer code can...... be understood as both an object and subject of study that intervenes in the world’s ‘becoming' and how material bodies are produced via human and nonhuman practices. Through mixing the natural and computer language, this article presents a script in six parts from a performative lecture for two persons...

  4. NSURE code

    International Nuclear Information System (INIS)

    Rattan, D.S.

    1993-11-01

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

  5. Use of computer codes for system reliability analysis

    International Nuclear Information System (INIS)

    Sabek, M.; Gaafar, M.; Poucet, A.

    1989-01-01

    This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author)

  6. Use of computer codes for system reliability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Sabek, M.; Gaafar, M. (Nuclear Regulatory and Safety Centre, Atomic Energy Authority, Cairo (Egypt)); Poucet, A. (Commission of the European Communities, Ispra (Italy). Joint Research Centre)

    1989-01-01

    This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author).

  7. Multiple component codes based generalized LDPC codes for high-speed optical transport.

    Science.gov (United States)

    Djordjevic, Ivan B; Wang, Ting

    2014-07-14

    A class of generalized low-density parity-check (GLDPC) codes suitable for optical communications is proposed, which consists of multiple local codes. It is shown that Hamming, BCH, and Reed-Muller codes can be used as local codes, and that the maximum a posteriori probability (MAP) decoding of these local codes by Ashikhmin-Lytsin algorithm is feasible in terms of complexity and performance. We demonstrate that record coding gains can be obtained from properly designed GLDPC codes, derived from multiple component codes. We then show that several recently proposed classes of LDPC codes such as convolutional and spatially-coupled codes can be described using the concept of GLDPC coding, which indicates that the GLDPC coding can be used as a unified platform for advanced FEC enabling ultra-high speed optical transport. The proposed class of GLDPC codes is also suitable for code-rate adaption, to adjust the error correction strength depending on the optical channel conditions.

  8. White LED with High Package Extraction Efficiency

    International Nuclear Information System (INIS)

    Yi Zheng; Stough, Matthew

    2008-01-01

    The goal of this project is to develop a high efficiency phosphor converting (white) Light Emitting Diode (pcLED) 1-Watt package through an increase in package extraction efficiency. A transparent/translucent monolithic phosphor is proposed to replace the powdered phosphor to reduce the scattering caused by phosphor particles. Additionally, a multi-layer thin film selectively reflecting filter is proposed between blue LED die and phosphor layer to recover inward yellow emission. At the end of the project we expect to recycle approximately 50% of the unrecovered backward light in current package construction, and develop a pcLED device with 80 lm/W e using our technology improvements and commercially available chip/package source. The success of the project will benefit luminous efficacy of white LEDs by increasing package extraction efficiency. In most phosphor-converting white LEDs, the white color is obtained by combining a blue LED die (or chip) with a powdered phosphor layer. The phosphor partially absorbs the blue light from the LED die and converts it into a broad green-yellow emission. The mixture of the transmitted blue light and green-yellow light emerging gives white light. There are two major drawbacks for current pcLEDs in terms of package extraction efficiency. The first is light scattering caused by phosphor particles. When the blue photons from the chip strike the phosphor particles, some blue light will be scattered by phosphor particles. Converted yellow emission photons are also scattered. A portion of scattered light is in the backward direction toward the die. The amount of this backward light varies and depends in part on the particle size of phosphors. The other drawback is that yellow emission from phosphor powders is isotropic. Although some backward light can be recovered by the reflector in current LED packages, there is still a portion of backward light that will be absorbed inside the package and further converted to heat. Heat generated

  9. Advanced Sensor Arrays and Packaging

    Energy Technology Data Exchange (ETDEWEB)

    Ryter, John Wesley [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Romero, Christopher J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Ramaiyan, Kannan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Brosha, Eric L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-08-11

    Novel sensor packaging elements were designed, fabricated, and tested in order to facilitate the transition of electrochemical mixed-potential sensors toward commercialization. Of the two designs completed, the first is currently undergoing field trials, taking direct measurements within vehicle exhaust streams, while the second is undergoing preliminary laboratory testing. The sensors’ optimal operating conditions, sensitivity to hydrogen, and long-­term baseline stability were also investigated. The sensing capabilities of lanthanum chromite (La0.8Sr0.2CrO3) and indium-­doped tin oxide (ITO) working electrodes were compared, and the ITO devices were selected for pre-­commercial field trials testing at a hydrogen fuel cell vehicle fueling station in California. Previous data from that fueling station were also analyzed, and the causes of anomalous baseline drift were identified.

  10. ISOLA II - A FORTRAN IV code for the calculation of the long-term α- and β-dose distributions in the vicinity of nuclear installations

    International Nuclear Information System (INIS)

    Huebschmann, W.; Nagel, D.

    1975-12-01

    The computer code ISOLA is used to calculate the annual radiation doses caused by α- and β-active off-gases in the environment of the Karlsruhe Nuclear Research Center. In the revised version ISOLA II the double Gaussian distribution model is strictly observed. As a consequence, the contribution of activity from neighbour sectors is taken into account. Up to 15 emitters may be coped with simultaneously. The emission rates are considered to be constant during the given time interval. Optionally either the isodoses chart of a specified area (for instance a square 20 by 20 km) or a list of doses calculated at up to 2,000 locations (for instance the living areas) in the environment may be set up. Input and output are shown for a specific case. (orig.) [de

  11. Fair Package Assignment

    Science.gov (United States)

    Lahaie, Sébastien; Parkes, David C.

    We consider the problem of fair allocation in the package assignment model, where a set of indivisible items, held by single seller, must be efficiently allocated to agents with quasi-linear utilities. A fair assignment is one that is efficient and envy-free. We consider a model where bidders have superadditive valuations, meaning that items are pure complements. Our central result is that core outcomes are fair and even coalition-fair over this domain, while fair distributions may not even exist for general valuations. Of relevance to auction design, we also establish that the core is equivalent to the set of anonymous-price competitive equilibria, and that superadditive valuations are a maximal domain that guarantees the existence of anonymous-price competitive equilibrium. Our results are analogs of core equivalence results for linear prices in the standard assignment model, and for nonlinear, non-anonymous prices in the package assignment model with general valuations.

  12. Anticounterfeit packaging technologies

    Directory of Open Access Journals (Sweden)

    Ruchir Y Shah

    2010-01-01

    Full Text Available Packaging is the coordinated system that encloses and protects the dosage form. Counterfeit drugs are the major cause of morbidity, mortality, and failure of public interest in the healthcare system. High price and well-known brands make the pharma market most vulnerable, which accounts for top priority cardiovascular, obesity, and antihyperlipidemic drugs and drugs like sildenafil. Packaging includes overt and covert technologies like barcodes, holograms, sealing tapes, and radio frequency identification devices to preserve the integrity of the pharmaceutical product. But till date all the available techniques are synthetic and although provide considerable protection against counterfeiting, have certain limitations which can be overcome by the application of natural approaches and utilization of the principles of nanotechnology.

  13. Package materials, waste form

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    The schedules for waste package development for the various host rocks were presented. The waste form subtask activities were reviewed, with the papers focusing on high-level waste, transuranic waste, and spent fuel. The following ten papers were presented: (1) Waste Package Development Approach; (2) Borosilicate Glass as a Matrix for Savannah River Plant Waste; (3) Development of Alternative High-Level Waste Forms; (4) Overview of the Transuranic Waste Management Program; (5) Assessment of the Impacts of Spent Fuel Disassembly - Alternatives on the Nuclear Waste Isolation System; (6) Reactions of Spent Fuel and Reprocessing Waste Forms with Water in the Presence of Basalt; (7) Spent Fuel Stabilizer Screening Studies; (8) Chemical Interactions of Shale Rock, Prototype Waste Forms, and Prototype Canister Metals in a Simulated Wet Repository Environment; (9) Impact of Fission Gas and Volatiles on Spent Fuel During Geologic Disposal; and (10) Spent Fuel Assembly Decay Heat Measurement and Analysis

  14. RH Packaging Program Guidance

    International Nuclear Information System (INIS)

    Washington TRU Solutions, LLC

    2003-01-01

    The purpose of this program guidance document is to provide technical requirements for use, operation, inspection, and maintenance of the RH-TRU 72-B Waste Shipping Package and directly related components. This document complies with the requirements as specified in the RH-TRU 72-B Safety Analysis Report for Packaging (SARP), and Nuclear Regulatory Commission (NRC) Certificate of Compliance (C of C) 9212. If there is a conflict between this document and the SARP and/or C of C, the SARP and/or C of C shall govern. The C of C states: ''...each package must be prepared for shipment and operated in accordance with the procedures described in Chapter 7.0, ''Operating Procedures,'' of the application.'' It further states: ''...each package must be tested and maintained in accordance with the procedures described in Chapter 8.0, ''Acceptance Tests and Maintenance Program of the Application.'' Chapter 9.0 of the SARP tasks the Waste Isolation Pilot Plant (WIPP) Management and Operating (M and O) contractor with assuring the packaging is used in accordance with the requirements of the C of C. Because the packaging is NRC approved, users need to be familiar with 10 CFR (section) 71.11, ''Deliberate Misconduct.'' Any time a user suspects or has indications that the conditions of approval in the C of C were not met, the Carlsbad Field Office (CBFO) shall be notified immediately. CBFO will evaluate the issue and notify the NRC if required. This document details the instructions to be followed to operate, maintain, and test the RH-TRU 72-B packaging. This Program Guidance standardizes instructions for all users. Users shall follow these instructions. Following these instructions assures that operations are safe and meet the requirements of the SARP. This document is available on the Internet at: ttp://www.ws/library/t2omi/t2omi.htm. Users are responsible for ensuring they are using the current revision and change notices. Sites may prepare their own document using the word

  15. Irradiation of packaged food

    International Nuclear Information System (INIS)

    Kilcast, D.

    1990-01-01

    Food irradiation is used to improve the safety of food by killing insects and microorganisms, to inhibit sprouting in crops such as onions and potatoes and to control ripening in agricultural produce. In order to prevent re-infestation and re-contamination it is essential that the food is suitably packed. Consequently, the packaging material is irradiated whilst in contact with the food, and it is important that the material is resistant to radiation-induced changes. In this paper the nature of the irradiation process is reviewed briefly, together with the known effects of irradiation on packaging materials and their implications for the effective application of food irradiation. Recent research carried out at the Leatherhead Food RA on the possibility of taint transfer into food is described. (author)

  16. Technology transfer packages

    International Nuclear Information System (INIS)

    Mizon, G.A.; Bleasdale, P.A.

    1994-01-01

    Nuclear power is firmly established in many developed countries'energy policies and is being adopted by emerging nations as an attractive way of gaining energy self sufficiency. The early users of nuclear power had to develop the technology that they needed, which now, through increasing world wide experience, has been rationalised to meet demanding economic and environmental pressures. These justifiable pressures, can lead to existing suppliers of nuclear services to consider changing to more appropriate technologies and for new suppliers to consider licensing proven technology rather then incurring the cost of developing new alternatives. The transfer of technology, under license, is made more straight forward if the owner conveniently groups appropriate technology into packages. This paper gives examples of 'Technology Packages' and suggests criteria for the specification, selection and contractual requirements to ensure successful licensing

  17. The Aster code; Code Aster

    Energy Technology Data Exchange (ETDEWEB)

    Delbecq, J.M

    1999-07-01

    The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)

  18. Software and the Scientist: Coding and Citation Practices in Geodynamics

    Science.gov (United States)

    Hwang, Lorraine; Fish, Allison; Soito, Laura; Smith, MacKenzie; Kellogg, Louise H.

    2017-11-01

    In geodynamics as in other scientific areas, computation has become a core component of research, complementing field observation, laboratory analysis, experiment, and theory. Computational tools for data analysis, mapping, visualization, modeling, and simulation are essential for all aspects of the scientific workflow. Specialized scientific software is often developed by geodynamicists for their own use, and this effort represents a distinctive intellectual contribution. Drawing on a geodynamics community that focuses on developing and disseminating scientific software, we assess the current practices of software development and attribution, as well as attitudes about the need and best practices for software citation. We analyzed publications by participants in the Computational Infrastructure for Geodynamics and conducted mixed method surveys of the solid earth geophysics community. From this we learned that coding skills are typically learned informally. Participants considered good code as trusted, reusable, readable, and not overly complex and considered a good coder as one that participates in the community in an open and reasonable manor contributing to both long- and short-term community projects. Participants strongly supported citing software reflected by the high rate a software package was named in the literature and the high rate of citations in the references. However, lacking are clear instructions from developers on how to cite and education of users on what to cite. In addition, citations did not always lead to discoverability of the resource. A unique identifier to the software package itself, community education, and citation tools would contribute to better attribution practices.

  19. Amdahl 470 Chip Package

    CERN Multimedia

    1975-01-01

    In the late 70s the larger IBM computers were water cooled. Amdahl, an IBM competitor, invented an air cooling technology for it's computers. His company worked hard, developing a computer that was faster and less expensive than the IBM System/360 mainframe computer systems. This object contains an actual Amdahl series 470 computer logic chip with an air cooling device mounted on top. The package leads and cooling tower are gold-plated.

  20. Aquaculture information package

    Energy Technology Data Exchange (ETDEWEB)

    Boyd, T.; Rafferty, K.

    1998-08-01

    This package of information is intended to provide background information to developers of geothermal aquaculture projects. The material is divided into eight sections and includes information on market and price information for typical species, aquaculture water quality issues, typical species culture information, pond heat loss calculations, an aquaculture glossary, regional and university aquaculture offices and state aquaculture permit requirements. A bibliography containing 68 references is also included.