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.

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

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)

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)

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)

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

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

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

Science.gov (United States)

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

2009-01-01

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

Simulation of the OECD Main-Steam-Line-Break Benchmark Exercise 3 Using the Coupled RELAP5/PANTHER Codes

International Nuclear Information System (INIS)

Schneidesch, Christophe R.; Zhang Jinzhao

2004-01-01

The RELAP5 best-estimate thermal-hydraulic system code has been coupled with the PANTHER three-dimensional neutron kinetics code via the TALINK dynamic data exchange control and processing tool. The coupled RELAP5/PANTHER code package has been qualified and will be used at Tractebel Engineering (TE) for analyzing asymmetric pressurized water reactor (PWR) accidents with strong core-system interactions. The Organization for Economic Cooperation and Development/U.S. Nuclear Regulatory Commission PWR main-steam-line-break benchmark problem was analyzed as part of the qualification efforts to demonstrate the capability of the coupled code package of simulating such transients. This paper reports the main results of TE's contribution to the benchmark Exercise 3

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)

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)

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)

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

International Nuclear Information System (INIS)

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

1988-12-01

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

Plasmids and packaging cell lines for use in phage display

Science.gov (United States)

Bradbury, Andrew M.

2012-07-24

The invention relates to a novel phagemid display system for packaging phagemid DNA into phagemid particles which completely avoids the use of helper phage. The system of the invention incorporates the use of bacterial packaging cell lines which have been transformed with helper plasmids containing all required phage proteins but not the packaging signals. The absence of packaging signals in these helper plasmids prevents their DNA from being packaged in the bacterial cell, which provides a number of significant advantages over the use of both standard and modified helper phage. Packaged phagemids expressing a protein or peptide of interest, in fusion with a phage coat protein such as g3p, are generated simply by transfecting phagemid into the packaging cell line.

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

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

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)

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

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

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

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)

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+.

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

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

fgui: A Method for Automatically Creating Graphical User Interfaces for Command-Line R Packages

Science.gov (United States)

Hoffmann, Thomas J.; Laird, Nan M.

2009-01-01

The fgui R package is designed for developers of R packages, to help rapidly, and sometimes fully automatically, create a graphical user interface for a command line R package. The interface is built upon the Tcl/Tk graphical interface included in R. The package further facilitates the developer by loading in the help files from the command line functions to provide context sensitive help to the user with no additional effort from the developer. Passing a function as the argument to the routines in the fgui package creates a graphical interface for the function, and further options are available to tweak this interface for those who want more flexibility. PMID:21625291

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.

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.

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

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

Design of Control System for Flexible Packaging Bags Palletizing Production Line Based on PLC

Science.gov (United States)

Zheng, Huiping; Chen, Lin; Zhao, Xiaoming; Liu, Zhanyang

Flexible packaging bags palletizing production line is to put the bags in the required area according to particular order and size, in order to finish handling, storage, loading and unloading, transportation and other logistics work of goods. Flexible packaging bags palletizing line is composed of turning bags mechanism, shaping mechanism, indexing mechanism, marshalling mechanism, pushing bags mechanism, pressing bags mechanism, laminating mechanism, elevator, tray warehouse, tray conveyor and loaded tray conveyor. Whether the whole production line can smoothly run depends on each of the above equipment and precision control among them. In this paper the technological process and the control logic of flexible packaging bags palletizing production line is introduced. Palletizing process of the production line realized automation by means of a control system based on programmable logic controller (PLC). It has the advantages of simple structure, reliable and easy maintenance etc.

Evaluation of ATLAS 100% DVI Line Break Using TRACE Code

International Nuclear Information System (INIS)

Huh, Byung Gil; Bang, Young Seok; Cheong, Ae Ju; Woo, Sweng Woong

2011-01-01

ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation) is an integral effect test facility in KAERI. It had installed completely to simulate the accident for the OPR1000 and the APR1400 in 2005. After then, several tests for LBLOCA, DVI line break have been performed successfully to resolve the safety issues of the APR1400. Especially, a DVI line break is considered as another spectrum among the SBLOCAs in APR1400 because the DVI line is directly connected to the reactor vessel and the thermal hydraulic behaviors are expected to be different from those for the cold leg injection. However, there are not enough experimental data for the DVI line break. Therefore, integral effect data for the DVI line break of ATLAS is very useful and available for an improvement and validation of safety codes. For the DVI line break in ATLAS, several analyses using MARS and RELAP codes were performed in the ATLAS DSP (Domestic Standard Problem) meetings. However, TRACE code has still not used to simulate a DVI line break. TRACE code has developed as the unified code for the reactor thermal hydraulic analyses in USNRC. In this study, the 100% DVI line break in ATLAS was evaluated by TRACE code. The objectives of this study are to identify the prediction capability of TRACE code for the major thermal hydraulic phenomena of a DVI line break in ATLAS

Efficient topology optimization in MATLAB using 88 lines of code

DEFF Research Database (Denmark)

Andreassen, Erik; Clausen, Anders; Schevenels, Mattias

2011-01-01

The paper presents an efficient 88 line MATLAB code for topology optimization. It has been developed using the 99 line code presented by Sigmund (Struct Multidisc Optim 21(2):120–127, 2001) as a starting point. The original code has been extended by a density filter, and a considerable improvemen...... of the basic code to include recent PDE-based and black-and-white projection filtering methods. The complete 88 line code is included as an appendix and can be downloaded from the web site www.topopt.dtu.dk....

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

Highly parallel line-based image coding for many cores.

Science.gov (United States)

Peng, Xiulian; Xu, Jizheng; Zhou, You; Wu, Feng

2012-01-01

Computers are developing along with a new trend from the dual-core and quad-core processors to ones with tens or even hundreds of cores. Multimedia, as one of the most important applications in computers, has an urgent need to design parallel coding algorithms for compression. Taking intraframe/image coding as a start point, this paper proposes a pure line-by-line coding scheme (LBLC) to meet the need. In LBLC, an input image is processed line by line sequentially, and each line is divided into small fixed-length segments. The compression of all segments from prediction to entropy coding is completely independent and concurrent at many cores. Results on a general-purpose computer show that our scheme can get a 13.9 times speedup with 15 cores at the encoder and a 10.3 times speedup at the decoder. Ideally, such near-linear speeding relation with the number of cores can be kept for more than 100 cores. In addition to the high parallelism, the proposed scheme can perform comparatively or even better than the H.264 high profile above middle bit rates. At near-lossless coding, it outperforms H.264 more than 10 dB. At lossless coding, up to 14% bit-rate reduction is observed compared with H.264 lossless coding at the high 4:4:4 profile.

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

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

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)

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

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)

On-line monitoring and inservice inspection in codes

International Nuclear Information System (INIS)

Bartonicek, J.; Zaiss, W.; Bath, H.R.

1999-01-01

The relevant regulatory codes determine the ISI tasks and the time intervals for recurrent components testing for evaluation of operation-induced damaging or ageing in order to ensure component integrity on the basis of the last available quality data. In-service quality monitoring is carried out through on-line monitoring and recurrent testing. The requirements defined by the engineering codes elaborated by various institutions are comparable, with the KTA nuclear engineering and safety codes being the most complete provisions for quality evaluation and assurance after different, defined service periods. German conventional codes for assuring component integrity provide exclusively for recurrent inspection regimes (mainly pressure tests and optical testing). The requirements defined in the KTA codes however always demanded more specific inspections relying on recurrent testing as well as on-line monitoring. Foreign codes for ensuring component integrity concentrate on NDE tasks at regular time intervals, with time intervals scope of testing activities being defined on the basis of the ASME code, section XI. (orig./CB) [de

Towards Product Lining Model-Driven Development Code Generators

OpenAIRE

Roth, Alexander; Rumpe, Bernhard

2015-01-01

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

A stable murine-based RD114 retroviral packaging line efficiently transduces human hematopoietic cells.

Science.gov (United States)

Ward, Maureen; Sattler, Rose; Grossman, I Robert; Bell, Anthony J; Skerrett, Donna; Baxi, Laxmi; Bank, Arthur

2003-11-01

Several barriers exist to high-efficiency transfer of therapeutic genes into human hematopoietic stem cells (HSCs) using complex oncoretroviral vectors. Human clinical trials to date have used Moloney leukemia virus-based amphotropic and gibbon ape leukemia virus-based envelopes in stable retroviral packaging lines. However, retroviruses pseudotyped with these envelopes have low titers due to the inability to concentrate viral supernatants efficiently by centrifugation without damaging the virus and low transduction efficiencies because of low-level expression of viral target receptors on human HSC. The RD114 envelope from the feline endogenous virus has been shown to transduce human CD34+ cells using transient packaging systems and to be concentrated to high titers by centrifugation. Stable packaging systems have potential advantages over transient systems because greater and more reproducible viral productions can be attained. We have, therefore, constructed and tested a stable RD114-expressing packaging line capable of high-level transduction of human CD34+ cells. Viral particles from this cell line were concentrated up to 100-fold (up to 10(7) viral particles/ml) by ultracentrifugation. Human hematopoietic progenitors from cord blood and sickle cell CD34+ cells were efficiently transduced with a Neo(R)-containing vector after a single exposure to concentrated RD114-pseudotyped virus produced from this cell line. Up to 78% of progenitors from transduced cord blood CD34+ cells and 51% of progenitors from sickle cell CD34+ cells expressed the NeoR gene. We also show transfer of a human beta-globin gene into progenitor cells from CD34+ cells from sickle cell patients with this new RD114 stable packaging system. The results indicate that this packaging line may eventually be useful in human clinical trials of globin gene therapy.

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

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.

LDPC Code Design for Nonuniform Power-Line Channels

Directory of Open Access Journals (Sweden)

Sanaei Ali

2007-01-01

Full Text Available We investigate low-density parity-check code design for discrete multitone channels over power lines. Discrete multitone channels are well modeled as nonuniform channels, that is, different bits experience various channel parameters. We propose a coding system for discrete multitone channels that allows for using a single code over a nonuniform channel. The number of code parameters for the proposed system is much greater than the number of code parameters in conventional channel. Therefore, search-based optimization methods are impractical. We first formulate the problem of optimizing the rate of an irregular low-density parity-check code, with guaranteed convergence over a general nonuniform channel, as an iterative linear programming which is significantly more efficient than search-based methods. Then we use this technique for a typical power-line channel. The methodology of this paper is directly applicable to all decoding algorithms for which a density evolution analysis is possible.

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

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

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

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.

Intercomparison of three microwave/infrared high resolution line-by-line radiative transfer codes

Science.gov (United States)

Schreier, Franz; Milz, Mathias; Buehler, Stefan A.; von Clarmann, Thomas

2018-05-01

An intercomparison of three line-by-line (lbl) codes developed independently for atmospheric radiative transfer and remote sensing - ARTS, GARLIC, and KOPRA - has been performed for a thermal infrared nadir sounding application assuming a HIRS-like (High resolution Infrared Radiation Sounder) setup. Radiances for the 19 HIRS infrared channels and a set of 42 atmospheric profiles from the "Garand dataset" have been computed. The mutual differences of the equivalent brightness temperatures are presented and possible causes of disagreement are discussed. In particular, the impact of path integration schemes and atmospheric layer discretization is assessed. When the continuum absorption contribution is ignored because of the different implementations, residuals are generally in the sub-Kelvin range and smaller than 0.1 K for some window channels (and all atmospheric models and lbl codes). None of the three codes turned out to be perfect for all channels and atmospheres. Remaining discrepancies are attributed to different lbl optimization techniques. Lbl codes seem to have reached a maturity in the implementation of radiative transfer that the choice of the underlying physical models (line shape models, continua etc) becomes increasingly relevant.

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)

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.)

FLP: a field line plotting code for bundle divertor design

International Nuclear Information System (INIS)

Ruchti, C.

1981-01-01

A computer code was developed to aid in the design of bundle divertors. The code can handle discrete toroidal field coils and various divertor coil configurations. All coils must be composed of straight line segments. The code runs on the PDP-10 and displays plots of the configuration, field lines, and field ripple. It automatically chooses the coil currents to connect the separatrix produced by the divertor to the outer edge of the plasma and calculates the required coil cross sections. Several divertor designs are illustrated to show how the code works

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.

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)

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

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.

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

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

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

'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

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.

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

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)

Establishing a store baseline during interim storage of waste packages and a review of potential technologies for base-lining

Energy Technology Data Exchange (ETDEWEB)

McTeer, Jennifer; Morris, Jenny; Wickham, Stephen [Galson Sciences Ltd. Oakham, Rutland (United Kingdom); Bolton, Gary [National Nuclear Laboratory Risley, Warrington (United Kingdom); McKinney, James; Morris, Darrell [Nuclear Decommissioning Authority Moor Row, Cumbria (United Kingdom); Angus, Mike [National Nuclear Laboratory Risley, Warrington (United Kingdom); Cann, Gavin; Binks, Tracey [National Nuclear Laboratory Sellafield (United Kingdom)

2013-07-01

Interim storage is an essential component of the waste management lifecycle, providing a safe, secure environment for waste packages awaiting final disposal. In order to be able to monitor and detect change or degradation of the waste packages, storage building or equipment, it is necessary to know the original condition of these components (the 'waste storage system'). This paper presents an approach to establishing the baseline for a waste-storage system, and provides guidance on the selection and implementation of potential base-lining technologies. The approach is made up of two sections; assessment of base-lining needs and definition of base-lining approach. During the assessment of base-lining needs a review of available monitoring data and store/package records should be undertaken (if the store is operational). Evolutionary processes (affecting safety functions), and their corresponding indicators, that can be measured to provide a baseline for the waste-storage system should then be identified in order for the most suitable indicators to be selected for base-lining. In defining the approach, identification of opportunities to collect data and constraints is undertaken before selecting the techniques for base-lining and developing a base-lining plan. Base-lining data may be used to establish that the state of the packages is consistent with the waste acceptance criteria for the storage facility and to support the interpretation of monitoring and inspection data collected during store operations. Opportunities and constraints are identified for different store and package types. Technologies that could potentially be used to measure baseline indicators are also reviewed. (authors)

GARLIC - A general purpose atmospheric radiative transfer line-by-line infrared-microwave code: Implementation and evaluation

Science.gov (United States)

Schreier, Franz; Gimeno García, Sebastián; Hedelt, Pascal; Hess, Michael; Mendrok, Jana; Vasquez, Mayte; Xu, Jian

2014-04-01

A suite of programs for high resolution infrared-microwave atmospheric radiative transfer modeling has been developed with emphasis on efficient and reliable numerical algorithms and a modular approach appropriate for simulation and/or retrieval in a variety of applications. The Generic Atmospheric Radiation Line-by-line Infrared Code - GARLIC - is suitable for arbitrary observation geometry, instrumental field-of-view, and line shape. The core of GARLIC's subroutines constitutes the basis of forward models used to implement inversion codes to retrieve atmospheric state parameters from limb and nadir sounding instruments. This paper briefly introduces the physical and mathematical basics of GARLIC and its descendants and continues with an in-depth presentation of various implementation aspects: An optimized Voigt function algorithm combined with a two-grid approach is used to accelerate the line-by-line modeling of molecular cross sections; various quadrature methods are implemented to evaluate the Schwarzschild and Beer integrals; and Jacobians, i.e. derivatives with respect to the unknowns of the atmospheric inverse problem, are implemented by means of automatic differentiation. For an assessment of GARLIC's performance, a comparison of the quadrature methods for solution of the path integral is provided. Verification and validation are demonstrated using intercomparisons with other line-by-line codes and comparisons of synthetic spectra with spectra observed on Earth and from Venus.

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)

Colloid transport code-nuclear user's manual

International Nuclear Information System (INIS)

Jain, R.

1992-01-01

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

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)

Comparison of Bit Error Rate of Line Codes in NG-PON2

Directory of Open Access Journals (Sweden)

Tomas Horvath

2016-05-01

Full Text Available This article focuses on simulation and comparison of line codes NRZ (Non Return to Zero, RZ (Return to Zero and Miller’s code for NG-PON2 (Next-Generation Passive Optical Network Stage 2 using. Our article provides solutions with Q-factor, BER (Bit Error Rate, and bandwidth comparison. Line codes are the most important part of communication over the optical fibre. The main role of these codes is digital signal representation. NG-PON2 networks use optical fibres for communication that is the reason why OptSim v5.2 is used for simulation.

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.

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)

A Novel Line Coding Pair for Fully Passive Long Reach {WDM-PON}s

DEFF Research Database (Denmark)

Presi, Marco; Proietti, Roberto; Prince, Kamau

2008-01-01

A novel line coding pair allows to use unsaturated flective-SOAs as upstream remodulator in long-reach WDM-PONs. Full-duplex and symmetric 80 km reach is demonstrated without in-line amplification at 1.25 Gb/s......A novel line coding pair allows to use unsaturated flective-SOAs as upstream remodulator in long-reach WDM-PONs. Full-duplex and symmetric 80 km reach is demonstrated without in-line amplification at 1.25 Gb/s...

The CEBAF test package: A symbolic and dynamic test, histogram and parameter package for on- and off-line particle physics data analysis

International Nuclear Information System (INIS)

Wood, S.A.; Abbott, S.

1995-01-01

Nuclear physics data analysis programs often use packages, such as Q, that allow parameter values, test definitions and histogram booking parameters to be controlled at run time through external files or shared memory. Data within a physics analyzer are usually referenced by indices, leading to a high use of equivalence statements and to extra bookkeeping. In the CEBAF Test Package (CTP), parameters, tests and histogram definitions defined in external files all refer to data elements by the same variable names as used in the C or Fortran source code for the analyzer. This is accomplished by requirieng the analyzer developer to open-quotes registerclose quotes each variable and array that is to be accessible by the package. Any registered variable as well as the test and histogram definitions may be dynamically read and modified by tasks that communicate via standard networking calls. As this package is implemented in C and requires only HBOOK and SUN RPC networking, it is highly portable. CTP works with the CEBAF Online Data Acquisition system (CODA), but may be used with other data acquisition systems or stand alone

Colloid transport code-nuclear user`s manual

Energy Technology Data Exchange (ETDEWEB)

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

1992-04-03

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

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

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

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

GARLIC — A general purpose atmospheric radiative transfer line-by-line infrared-microwave code: Implementation and evaluation

International Nuclear Information System (INIS)

Schreier, Franz; Gimeno García, Sebastián; Hedelt, Pascal; Hess, Michael; Mendrok, Jana; Vasquez, Mayte; Xu, Jian

2014-01-01

A suite of programs for high resolution infrared-microwave atmospheric radiative transfer modeling has been developed with emphasis on efficient and reliable numerical algorithms and a modular approach appropriate for simulation and/or retrieval in a variety of applications. The Generic Atmospheric Radiation Line-by-line Infrared Code — GARLIC — is suitable for arbitrary observation geometry, instrumental field-of-view, and line shape. The core of GARLIC's subroutines constitutes the basis of forward models used to implement inversion codes to retrieve atmospheric state parameters from limb and nadir sounding instruments. This paper briefly introduces the physical and mathematical basics of GARLIC and its descendants and continues with an in-depth presentation of various implementation aspects: An optimized Voigt function algorithm combined with a two-grid approach is used to accelerate the line-by-line modeling of molecular cross sections; various quadrature methods are implemented to evaluate the Schwarzschild and Beer integrals; and Jacobians, i.e. derivatives with respect to the unknowns of the atmospheric inverse problem, are implemented by means of automatic differentiation. For an assessment of GARLIC's performance, a comparison of the quadrature methods for solution of the path integral is provided. Verification and validation are demonstrated using intercomparisons with other line-by-line codes and comparisons of synthetic spectra with spectra observed on Earth and from Venus. - Highlights: • High resolution infrared-microwave radiative transfer model. • Discussion of algorithmic and computational aspects. • Jacobians by automatic/algorithmic differentiation. • Performance evaluation by intercomparisons, verification, validation

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

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

Solving Differential Equations in R: Package deSolve

Directory of Open Access Journals (Sweden)

Karline Soetaert

2010-02-01

Full Text Available In this paper we present the R package deSolve to solve initial value problems (IVP written as ordinary differential equations (ODE, differential algebraic equations (DAE of index 0 or 1 and partial differential equations (PDE, the latter solved using the method of lines approach. The differential equations can be represented in R code or as compiled code. In the latter case, R is used as a tool to trigger the integration and post-process the results, which facilitates model development and application, whilst the compiled code significantly increases simulation speed. The methods implemented are efficient, robust, and well documented public-domain Fortran routines. They include four integrators from the ODEPACK package (LSODE, LSODES, LSODA, LSODAR, DVODE and DASPK2.0. In addition, a suite of Runge-Kutta integrators and special-purpose solvers to efficiently integrate 1-, 2- and 3-dimensional partial differential equations are available. The routines solve both stiff and non-stiff systems, and include many options, e.g., to deal in an efficient way with the sparsity of the Jacobian matrix, or finding the root of equations. In this article, our objectives are threefold: (1 to demonstrate the potential of using R for dynamic modeling, (2 to highlight typical uses of the different methods implemented and (3 to compare the performance of models specified in R code and in compiled code for a number of test cases. These comparisons demonstrate that, if the use of loops is avoided, R code can efficiently integrate problems comprising several thousands of state variables. Nevertheless, the same problem may be solved from 2 to more than 50 times faster by using compiled code compared to an implementation using only R code. Still, amongst the benefits of R are a more flexible and interactive implementation, better readability of the code, and access to R’s high-level procedures. deSolve is the successor of package odesolve which will be deprecated in

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.

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)

Micro packaging of hermetic seal mini dual in line laser diode module for aerospace applications

Science.gov (United States)

Kazemi, Alex A.; Chan, Eric; Koshinz, Dennis

2014-09-01

Normally, reliable, reproducible, high-yield packaging technologies are essential for meeting the cost, performance, and service objectives for the harsh environment of space applications. This paper describes a new improved micro packaging method of hermetic seal mini-DIL (dual in line) laser diode module. The problem of using a softer solder resulted in failure mechanisms observed in the mini-DIL laser diode module based laser firing unit (LFU) for ordinance ignition of a missile system. These failures included: (1) failure in light output pulse power, (2) fiber pigtail damage inside the package snout which caused low LFU production yield. Our distinctive challenge for this project is the micro packaging of mini-DIL. For this package a new technique for the hermetic sealing using a micro-soldering process was developed. The process is able to confine the solder seal to a small region inside the snout near the fiber feed-through hole on the wall of the mini-DIL package. After completing the development, which included temperature and thermal cycling, X-rays analysis showed the new method had no fiber damage after the microsoldering seal. The new process resulted in 100% success in the packaging design and was granted a patent for the innovative development.

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.

The Second Workshop on Lineshape Code Comparison: Isolated Lines

Directory of Open Access Journals (Sweden)

Spiros Alexiou

2014-05-01

Full Text Available In this work, we briefly summarize the theoretical aspects of isolated line broadening. We present and discuss test run comparisons from different participating lineshape codes for the 2s-2p transition for LiI, B III and NV.

The Fourth Workshop on Lineshape Code Comparison: Line Merging

Directory of Open Access Journals (Sweden)

Spiros Alexiou

2018-03-01

Full Text Available For a given set of plasma parameters, along a single series (Lyman, Balmer, etc. the lines with higher principal quantum number (n lines get progressively wider, closer to each other, and start merging for a certain critical n. In the present work, four different codes (with further options are used to calculate the entire Balmer series for moderate and high electron densities. Particular attention is paid to the relevant physics, such as the cutoff criteria, strong and penetrating electron collisions.

Coronal Magnetism and Forward Solarsoft Idl Package

Science.gov (United States)

Gibson, S. E.

2014-12-01

The FORWARD suite of Solar Soft IDL codes is a community resource for model-data comparison, with a particular emphasis on analyzing coronal magnetic fields. FORWARD may be used both to synthesize a broad range of coronal observables, and to access and compare to existing data. FORWARD works with numerical model datacubes, interfaces with the web-served Predictive Science Inc MAS simulation datacubes and the Solar Soft IDL Potential Field Source Surface (PFSS) package, and also includes several analytic models (more can be added). It connects to the Virtual Solar Observatory and other web-served observations to download data in a format directly comparable to model predictions. It utilizes the CHIANTI database in modeling UV/EUV lines, and links to the CLE polarimetry synthesis code for forbidden coronal lines. FORWARD enables "forward-fitting" of specific observations, and helps to build intuition into how the physical properties of coronal magnetic structures translate to observable properties.

Safety evaluation for packaging (onsite) concrete-lined waste packaging

Energy Technology Data Exchange (ETDEWEB)

Romano, T.

1997-09-25

The Pacific Northwest National Laboratory developed a package to ship Type A, non-transuranic, fissile excepted quantities of liquid or solid radioactive material and radioactive mixed waste to the Central Waste Complex for storage on the Hanford Site.

Multiple-valued logic-protected coding for an optical non-quantum communication line

NARCIS (Netherlands)

Antipov, A. L.; Bykovsky, A. Yu.; Vasiliev, N. A.; Egorov, A. A.

2006-01-01

A simple and cheap method of secret coding in an optical line is proposed based on multiple-valued logic. This method is shown to have very high cryptography resources and is designated for bidirectional information exchange in a team of mobile robots, where quantum teleportation coding cannot yet

MAPPIX: A software package for off-line micro-pixe single particle aerosol analysis

International Nuclear Information System (INIS)

Ceccato, D.

2009-01-01

In the framework of a multiannual experiment performed at Baia Terra Nova, Antarctica, size-segregated aerosol samples were collected by using a 12-stage SDI impactor (Hillamo design). Approximately 2800 particles, belonging to the first four supermicrometric SDI stages - 8.39, 4.08, 2.68, 1.66 μm dynamic aerosol diameter cuts - were analyzed at the INFN-LNL micro-PIXE facility, a three lens Oxford Microprobe (OM) product, installed in the early nineties. Four regions on each of the 12 sub-samples were measured; 60 aerosol particles were detected on average in each of the analyzed regions. The off-line single aerosol particle (SAP) analysis of such big amount of data required software that is able to rapidly handle the acquired data, with a simple and fast area selection procedure; the subsequent automated PIXE spectra analysis with a specialized code was also needed. The MAPPIX 2.0 software was designed to make easier and faster the user jobs during the SAP analysis. The package is composed of two separate routines: the first one is devoted to data format conversion (OM-LMF file format to MAPPIX format), while the second one is devoted to micro-PIXE maps graphical presentation and aerosol particle selection procedure. The MAPPIX data format and software features will be discussed; a short report of the speed performances will be presented.

Three dimensional field computation software package DE3D and its applications

International Nuclear Information System (INIS)

Fan Mingwu; Zhang Tianjue; Yan Weili

1992-07-01

A software package, DE3D that can be run on PC for three dimensional electrostatic and magnetostatic field analysis has been developed in CIAE (China Institute of Atomic Energy). Two scalar potential method and special numerical techniques have made the code with high precision. It can be used for electrostatic and magnetostatic fields computations with complex boundary conditions. In the most cases, the result accuracy is better than 1% comparing with the measured. In some situations, the results are more acceptable than the other codes because some tricks are used for the current integral. Typical examples, design of a cyclotron magnet and magnetic elements on its beam transport line, given in the paper show how the program helps the designer to improve the design of the product. The software package could bring advantages to the producers and designers

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

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

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

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)

WannierTools: An open-source software package for novel topological materials

Science.gov (United States)

Wu, QuanSheng; Zhang, ShengNan; Song, Hai-Feng; Troyer, Matthias; Soluyanov, Alexey A.

2018-03-01

We present an open-source software package WannierTools, a tool for investigation of novel topological materials. This code works in the tight-binding framework, which can be generated by another software package Wannier90 (Mostofi et al., 2008). It can help to classify the topological phase of a given material by calculating the Wilson loop, and can get the surface state spectrum, which is detected by angle resolved photoemission (ARPES) and in scanning tunneling microscopy (STM) experiments. It also identifies positions of Weyl/Dirac points and nodal line structures, calculates the Berry phase around a closed momentum loop and Berry curvature in a part of the Brillouin zone (BZ).

How to review 4 million lines of ATLAS code

CERN Document Server

AUTHOR|(INSPIRE)INSPIRE-00226135; The ATLAS collaboration; Lampl, Walter

2017-01-01

As the ATLAS Experiment prepares to move to a multi-threaded framework (AthenaMT) for Run3, we are faced with the problem of how to migrate 4 million lines of C++ source code. This code has been written over the past 15 years and has often been adapted, re-written or extended to the changing requirements and circumstances of LHC data taking. The code was developed by different authors, many of whom are no longer active, and under the deep assumption that processing ATLAS data would be done in a serial fashion. In order to understand the scale of the problem faced by the ATLAS software community, and to plan appropriately the significant efforts posed by the new AthenaMT framework, ATLAS embarked on a wide ranging review of our offline code, covering all areas of activity: event generation, simulation, trigger, reconstruction. We discuss the difficulties in even logistically organising such reviews in an already busy community, how to examine areas in sufficient depth to learn key areas in need of upgrade, yet...

How To Review 4 Million Lines of ATLAS Code

CERN Document Server

Stewart, Graeme; The ATLAS collaboration

2016-01-01

As the ATLAS Experiment prepares to move to a multi-threaded framework (AthenaMT) for Run3, we are faced with the problem of how to migrate 4 million lines of C++ source code. This code has been written over the past 15 years and has often been adapted, re-written or extended to the changing requirements and circumstances of LHC data taking. The code was developed by different authors, many of whom are no longer active, and under the deep assumption that processing ATLAS data would be done in a serial fashion. In order to understand the scale of the problem faced by the ATLAS software community, and to plan appropriately the significant efforts posed by the new AthenaMT framework, ATLAS embarked on a wide ranging review of our offline code, covering all areas of activity: event generation, simulation, trigger, reconstruction. We discuss the difficulties in even logistically organising such reviews in an already busy community, how to examine areas in sufficient depth to learn key areas in need of upgrade, yet...

LID: Computer code for identifying atomic and ionic lines below 3500 Angstroms

International Nuclear Information System (INIS)

Peek, J.M.; Dukart, R.J.

1987-08-01

An interactive computer code has been written to search a data base containing information useful for identifying lines in experimentally-observed spectra or for designing experiments. The data base was the basis for the Kelly and Palumbo critical review of well-resolved lines below 2000 Angstroms, includes lines below 3500 Angstroms for atoms and ions of hydrogen through krypton, and was obtained from R.L. Kelly. This code allows the user to search the data base for a user-specified wavelength region, with this search either limited to atoms or ions of the user's choice for all atoms and ions contained in the data base. The line information found in the search is stored in a local file for later reference. A plotting capability is provided to graphically display the lines resulting from the search. Several options are available to control the nature of these graphs. It is also possible to bring in data from another source, such as an experimental spectra, for display along with the lines from the data-base search. Options for manipulating the experimental spectra's background intensity and wavelength scale are also available to the user. The intensities for the lines from each ion found in the data-base search can be scaled by a multiplicative constant to better simulate the observed spectrum

Storage and Retrieval of Encrypted Data Blocks with In-Line Message Authentication Codes

NARCIS (Netherlands)

Bosch, H.G.P.; McLellan Jr, Hubert Rae; Mullender, Sape J.

2007-01-01

Techniques are disclosed for in-line storage of message authentication codes with respective encrypted data blocks. In one aspect, a given data block is encrypted and a message authentication code is generated for the encrypted data block. A target address is determined for storage of the encrypted

LIGHT-WEIGHT SENSOR PACKAGE FOR PRECISION 3D MEASUREMENT WITH MICRO UAVS E.G. POWER-LINE MONITORING

Directory of Open Access Journals (Sweden)

K.-D. Kuhnert

2013-08-01

Full Text Available The paper describes a new sensor package for micro or mini UAVs and one application that has been successfully implemented with this sensor package. It is intended for 3D measurement of landscape or large outdoor structures for mapping or monitoring purposes. The package can be composed modularly into several configurations. It may contain a laser-scanner, camera, IMU, GPS and other sensors as required by the application. Also different products of the same sensor type have been integrated. Always it contains its own computing infrastructure and may be used for intelligent navigation, too. It can be operated in cooperation with different drones but also completely independent of the type of drone it is attached to. To show the usability of the system, an application in monitoring high-voltage power lines that has been successfully realised with the package is described in detail.

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.

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

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

Safety evaluation report for packaging (onsite) concrete-lined waste packaging

International Nuclear Information System (INIS)

Romano, T.

1997-01-01

The Pacific Northwest National Laboratory developed a package to ship Type A, non-transuranic, fissile excepted quantities of liquid or solid radioactive material and radioactive mixed waste to the Central Waste Complex for storage on the Hanford Site

Improvements in the CHART D radiation-hydrodynamic code III: revised analytic equations of state

International Nuclear Information System (INIS)

Thompson, S.L.; Lauson, H.S.

1974-03-01

A revised set of in-line equation-of-state subroutines for the CHART D hydrodynamic code is described. The information generated is thermodynamically complete and self-consistent. The temperature and density range of validity is large. Solids, liquids, vapors, plasmas, and all types of phase mixtures are treated. Energy transport properties are calculated. The set of subroutines form a package which can easily be included in other hydrodynamic codes. (20 figures) (U.S.)

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...

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.

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.

Coupled RELAP5/PANTHER/COBRA steam line break accident analysis in support of licensing DOEL 2 power uprate and steam generator replacement

International Nuclear Information System (INIS)

Zhang, J.; Bosso, S.; Henno, X.; Ouliddren, K.; Schneidesch, C.R.; Hove, W. van

2004-01-01

The nuclear reactor accident analyses using best estimate codes provide a better understanding and more accurate modeling of the key physical phenomena, which allows a more realistic evaluation of the conservatism and margins in the final safety analysis report (FSAR) accident analysis. The use of the best estimate codes and methods is necessary to meet the increasing technical, licensing and regulatory requirements for major plant changes (e.g. steam generator replacement), power uprate, core design optimization (cycle extension), as well as Periodic Safety Review. Since 1992, Tractebel Engineering (TE) has developed and applied a deterministic bounding approach to FASR accident analysis using the best estimate system thermal hydraulic code RELAP5/MOD2.5 and the subchannel thermal hydraulic code COBRA-3C. This approach has been accepted by the Belgian Safety Authorities, and turned out to be cost effective for most of the non-LOCA transient analyses. Since this approach adapts a decoupled modeling of the core responses, the analysis results often involved too large un-quantified conservatisms, due to either simplistic approximations for asymmetric accidents with strong 3D core neutronics - plant thermal hydraulics interactions, or additional penalties introduced from 'incoherent' initial/boundary conditions for separate plant and core analyses. Therefore, an external dynamic coupling between the RELAP5/MOD2.5 code and the 3-D neutronic code PANTHER was implemented since 1997 via the transient analysis code linkage program TALINK. Furthermore, a static linkage between the PANTHER code and the COBRA-3C code was developed for on-line calculation of (Departure from Nucleate Boiling Ratio (DNBR). TE intends to use the coupled code package for licensing non-symmetric FSAR accident analysis. The TE coupled code package has been applied to develop a main steam line break (MSLB) accident analysis methodology [using the TE deterministic bounding approach. The methodology

Generation of a Vero-Based Packaging Cell Line to Produce SV40 Gene Delivery Vectors for Use in Clinical Gene Therapy Studies

Directory of Open Access Journals (Sweden)

Miguel G. Toscano

2017-09-01

Full Text Available Replication-defective (RD recombinant simian virus 40 (SV40-based gene delivery vectors hold a great potential for clinical applications because of their presumed non-immunogenicity and capacity to induce immune tolerance to the transgene products in humans. However, the clinical use of SV40 vectors has been hampered by the lack of a packaging cell line that produces replication-competent (RC free SV40 particles in the vector production process. To solve this problem, we have adapted the current SV40 vector genome used for the production of vector particles and generated a novel Vero-based packaging cell line named SuperVero that exclusively expresses the SV40 large T antigen. SuperVero cells produce similar numbers of SV40 vector particles compared to the currently used packaging cell lines, albeit in the absence of contaminating RC SV40 particles. Our unique SV40 vector platform named SVac paves the way to clinically test a whole new generation of SV40-based therapeutics for a broad range of important diseases.

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

A full-fledged micromagnetic code in fewer than 70 lines of NumPy

International Nuclear Information System (INIS)

Abert, Claas; Bruckner, Florian; Vogler, Christoph; Windl, Roman; Thanhoffer, Raphael; Suess, Dieter

2015-01-01

We present a complete micromagnetic finite-difference code in fewer than 70 lines of Python. The code makes a large use of the NumPy library and computes the exchange field by finite differences and the demagnetization field with a fast convolution algorithm. Since the magnetization in finite-difference micromagnetics is represented by a multi-dimensional array and the NumPy library features a rich interface for this data structure, the code we present is an ideal starting point for the development of novel algorithms. - Highlights: • A very concise but complete micromagnetic code written in Python is presented. • An introduction to finite-difference micromagnetics is provided. • The code is a perfect starting point for the development of novel algorithms

CODE STEM - Moon, Mars, and Beyond; DLESE-Powered On-Line Classroom, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — "CODE (COrps DEvelopment) STEM (Science, Technology, Engineering, and Math) ? Moon Mars and Beyond; DLESE-Powered On-Line Classroom" shares the excitement of...

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.

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

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

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

Dynamic Model for the Z Accelerator Vacuum Section Based on Transmission Line Code%Dynamic Model for the Z Accelerator Vacuum Section Based on Transmission Line Code

Institute of Scientific and Technical Information of China (English)

呼义翔; 雷天时; 吴撼宇; 郭宁; 韩娟娟; 邱爱慈; 王亮平; 黄涛; 丛培天; 张信军; 李岩; 曾正中; 孙铁平

2011-01-01

The transmission-line-circuit model of the Z accelerator, developed originally by W. A. STYGAR, P. A. CORCORAN, et al., is revised. The revised model uses different calculations for the electron loss and flow impedance in the magnetically insulated transmission line system of the Z accelerator before and after magnetic insulation is established. By including electron pressure and zero electric field at the cathode, a closed set of equations is obtained at each time step, and dynamic shunt resistance （used to represent any electron loss to the anode） and flow impedance are solved, which have been incorporated into the transmission line code for simulations of the vacuum section in the Z accelerator. Finally, the results are discussed in comparison with earlier findings to show the effectiveness and limitations of the model.

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

Improving a Power Line Communications Standard with LDPC Codes

Directory of Open Access Journals (Sweden)

Hsu Christine

2007-01-01

Full Text Available We investigate a power line communications (PLC scheme that could be used to enhance the HomePlug 1.0 standard, specifically its ROBO mode which provides modest throughput for the worst case PLC channel. The scheme is based on using a low-density parity-check (LDPC code, in lieu of the concatenated Reed-Solomon and convolutional codes in ROBO mode. The PLC channel is modeled with multipath fading and Middleton's class A noise. Clipping is introduced to mitigate the effect of impulsive noise. A simple and effective method is devised to estimate the variance of the clipped noise for LDPC decoding. Simulation results show that the proposed scheme outperforms the HomePlug 1.0 ROBO mode and has lower computational complexity. The proposed scheme also dispenses with the repetition of information bits in ROBO mode to gain time diversity, resulting in 4-fold increase in physical layer throughput.

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)

Translation of the flavivirus kunjin NS3 gene in cis but not its RNA sequence or secondary structure is essential for efficient RNA packaging.

Science.gov (United States)

Pijlman, Gorben P; Kondratieva, Natasha; Khromykh, Alexander A

2006-11-01

Our previous studies using trans-complementation analysis of Kunjin virus (KUN) full-length cDNA clones harboring in-frame deletions in the NS3 gene demonstrated the inability of these defective complemented RNAs to be packaged into virus particles (W. J. Liu, P. L. Sedlak, N. Kondratieva, and A. A. Khromykh, J. Virol. 76:10766-10775). In this study we aimed to establish whether this requirement for NS3 in RNA packaging is determined by the secondary RNA structure of the NS3 gene or by the essential role of the translated NS3 gene product. Multiple silent mutations of three computer-predicted stable RNA structures in the NS3 coding region of KUN replicon RNA aimed at disrupting RNA secondary structure without affecting amino acid sequence did not affect RNA replication and packaging into virus-like particles in the packaging cell line, thus demonstrating that the predicted conserved RNA structures in the NS3 gene do not play a role in RNA replication and/or packaging. In contrast, double frameshift mutations in the NS3 coding region of full-length KUN RNA, producing scrambled NS3 protein but retaining secondary RNA structure, resulted in the loss of ability of these defective RNAs to be packaged into virus particles in complementation experiments in KUN replicon-expressing cells. Furthermore, the more robust complementation-packaging system based on established stable cell lines producing large amounts of complemented replicating NS3-deficient replicon RNAs and infection with KUN virus to provide structural proteins also failed to detect any secreted virus-like particles containing packaged NS3-deficient replicon RNAs. These results have now firmly established the requirement of KUN NS3 protein translated in cis for genome packaging into virus particles.

Python Radiative Transfer Emission code (PyRaTE): non-LTE spectral lines simulations

Science.gov (United States)

Tritsis, A.; Yorke, H.; Tassis, K.

2018-05-01

We describe PyRaTE, a new, non-local thermodynamic equilibrium (non-LTE) line radiative transfer code developed specifically for post-processing astrochemical simulations. Population densities are estimated using the escape probability method. When computing the escape probability, the optical depth is calculated towards all directions with density, molecular abundance, temperature and velocity variations all taken into account. A very easy-to-use interface, capable of importing data from simulations outputs performed with all major astrophysical codes, is also developed. The code is written in PYTHON using an "embarrassingly parallel" strategy and can handle all geometries and projection angles. We benchmark the code by comparing our results with those from RADEX (van der Tak et al. 2007) and against analytical solutions and present case studies using hydrochemical simulations. The code will be released for public use.

Alternative Line Coding Scheme with Fixed Dimming for Visible Light Communication

Science.gov (United States)

Niaz, M. T.; Imdad, F.; Kim, H. S.

2017-01-01

An alternative line coding scheme called fixed-dimming on/off keying (FD-OOK) is proposed for visible-light communication (VLC). FD-OOK reduces the flickering caused by a VLC transmitter and can maintain a 50% dimming level. Simple encoder and decoder are proposed which generates codes where the number of bits representing one is same as the number of bits representing zero. By keeping the number of ones and zeros equal the change in the brightness of lighting may be minimized and kept constant at 50%, thereby reducing the flickering in VLC. The performance of FD-OOK is analysed with two parameters: the spectral efficiency and power requirement.

Analysis of the OECD main steam line break benchmark using ANC-K/MIDAC code

International Nuclear Information System (INIS)

Aoki, Shigeaki; Tahara, Yoshihisa; Suemura, Takayuki; Ogawa, Junto

2004-01-01

A three-dimensional (3D) neutronics and thermal-and-hydraulics (T/H) coupling code ANC-K/MIDAC has been developed. It is the combination of the 3D nodal kinetic code ANC-K and the 3D drift flux thermal hydraulic code MIDAC. In order to verify the adequacy of this code, we have performed several international benchmark problems. In this paper, we show the calculation results of ''OECD Main Steam Line Break Benchmark (MSLB benchmark)'', which gives the typical local power peaking problem. And we calculated the return-to-power scenario of the Phase II problem. The comparison of the results shows the very good agreement of important core parameters between the ANC-K/MIDAC and other participant codes. (author)

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.

Constrained Convolutional Sparse Coding for Parametric Based Reconstruction of Line Drawings

KAUST Repository

Shaheen, Sara

2017-12-25

Convolutional sparse coding (CSC) plays an essential role in many computer vision applications ranging from image compression to deep learning. In this work, we spot the light on a new application where CSC can effectively serve, namely line drawing analysis. The process of drawing a line drawing can be approximated as the sparse spatial localization of a number of typical basic strokes, which in turn can be cast as a non-standard CSC model that considers the line drawing formation process from parametric curves. These curves are learned to optimize the fit between the model and a specific set of line drawings. Parametric representation of sketches is vital in enabling automatic sketch analysis, synthesis and manipulation. A couple of sketch manipulation examples are demonstrated in this work. Consequently, our novel method is expected to provide a reliable and automatic method for parametric sketch description. Through experiments, we empirically validate the convergence of our method to a feasible solution.

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

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

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

Code Assessment of SPACE 2.19 using LSTF 10% Main Steam-Line-Break Test

Energy Technology Data Exchange (ETDEWEB)

Kim, Minhee; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed in recent years by the Korea Hydro and Nuclear Power Co. through collaborative works with other Korean nuclear industries and research institutes. As a result of the development, the 2.19 version of the code was released through the successive various verification and validation works. In this study, results produced by the SPACE 2.19 code were compared with the experimental data from JAERI's LSTF Test Run SBSL- 01 for a 10% main steam line break transient in a pressurized water reactor. The LSTF 10% main steam line break test were simulated using the SPACE 2.19 for code V and V work. The overall comparisons between the SPACE 2.19 code prediction and the LSTF Test Run SB-SL-01 experimental data are reasonably satisfactory. The comparisons were conducted in terms of the variations of mass flow rate, void fraction, pressure, collapsed liquid level, temperature, and system flow rate for the transient. In addition, the input model was modified for simulation accuracy of PZR pressure based on the calculated results. The correction of PORV setpoint affects to simulate the PORV open and close phenomena similarly with experiments. From the modification, the computed results show a reasonable agreement with experimental data in overall transient time.

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

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.

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 ...

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.

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

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)

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

Nuclear critical safety analysis for UX-30 transport of freight package

International Nuclear Information System (INIS)

Quan Yanhui; Zhou Qi; Yin Shenggui

2014-01-01

The nuclear critical safety analysis and evaluation for UX-30 transport freight package in the natural condition and accident condition were carried out with MONK-9A code and MCNP code. Firstly, the critical benchmark experiment data of public in international were selected, and the deflection and subcritical limiting value with MONK-9A code and MCNP code in calculating same material form were validated and confirmed. Secondly, the neutron efficiency multiplication factors in the natural condition and accident condition were calculated and analyzed, and the safety in transport process was evaluated by taking conservative suppose of nuclear critical safety. The calculation results show that the max value of k eff for UX-30 transport freight package is less than the subcritical limiting value, and the UX-30 transport freight package is in the state of subcritical safety. Moreover, the critical safety index (CSI) for UX-30 package can define zero based on the definition of critical safety index. (authors)

On-line core monitoring with CORE MASTER / PRESTO

International Nuclear Information System (INIS)

Lindahl, S.O.; Borresen, S.; Ovrum, S.

1986-01-01

Advanced calculational tools are instrumental in improving reactor plant capacity factors and fuel utilization. The computer code package CORE MASTER is an integrated system designed to achieve this objective. The system covers all main activities in the area of in-core fuel management for boiling water reactors; design, operation support, and on-line core monitoring. CORE MASTER operates on a common data base, which defines the reactor and documents the operating history of the core and of all fuel bundles ever used

Analysis Of Feedwater Line Break Of APR1400 By MARS Code

International Nuclear Information System (INIS)

Nguyen Thi Thanh Thuy; Le Dai Dien, Hoang Minh Giang

2011-01-01

This paper will deal with analysis of Feed water Line Break problem (FWLB) of the APR 1400 NPP with initial conditions: operation at 100% of power, double-ended break area of 0.058 m 2 and the break location of the feedwater line between the check valve and the steam generator. The analysis was simulated by MARS code through two step: calculation for steady state and calculation for transient state with initial condition mentioned. Some output result were presented with explanation: sequence of events corresponding to the time of the accident, the system behavior as temperature, pressure, steam generator water levels as well as DNBR, etc. before and after the accident. (author)

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

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

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.

Line-Shape Code Comparison through Modeling and Fitting of Experimental Spectra of the C ii 723-nm Line Emitted by the Ablation Cloud of a Carbon Pellet

Directory of Open Access Journals (Sweden)

Mohammed Koubiti

2014-07-01

Full Text Available Various codes of line-shape modeling are compared to each other through the profile of the C ii 723-nm line for typical plasma conditions encountered in the ablation clouds of carbon pellets, injected in magnetic fusion devices. Calculations were performed for a single electron density of 1017 cm−3 and two plasma temperatures (T = 2 and 4 eV. Ion and electron temperatures were assumed to be equal (Te = Ti = T. The magnetic field, B, was set equal to either to zero or 4 T. Comparisons between the line-shape modeling codes and two experimental spectra of the C ii 723-nm line, measured perpendicularly to the B-field in the Large Helical Device (LHD using linear polarizers, are also discussed.

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...

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

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

Development of 'SKYSHINE-CG' code. A line-beam method code equipped with combinatorial geometry routine

Energy Technology Data Exchange (ETDEWEB)

Nakagawa, Takahiro; Ochiai, Katsuharu [Plant and System Planning Department, Toshiba Corporation, Yokohama, Kanagawa (Japan); Uematsu, Mikio; Hayashida, Yoshihisa [Department of Nuclear Engineering, Toshiba Engineering Corporation, Yokohama, Kanagawa (Japan)

2000-03-01

A boiling water reactor (BWR) plant has a single loop coolant system, in which main steam generated in the reactor core proceeds directly into turbines. Consequently, radioactive {sup 16}N (6.2 MeV photon emitter) contained in the steam contributes to gamma-ray skyshine dose in the vicinity of the BWR plant. The skyshine dose analysis is generally performed with the line-beam method code SKYSHINE, in which calculational geometry consists of a rectangular turbine building and a set of isotropic point sources corresponding to an actual distribution of {sup 16}N sources. For the purpose of upgrading calculational accuracy, the SKYSHINE-CG code has been developed by incorporating the combinatorial geometry (CG) routine into the SKYSHINE code, so that shielding effect of in-building equipment can be properly considered using a three-dimensional model composed of boxes, cylinders, spheres, etc. Skyshine dose rate around a 500 MWe BWR plant was calculated with both SKYSHINE and SKYSHINE-CG codes, and the calculated results were compared with measured data obtained with a NaI(Tl) scintillation detector. The C/E values for SKYSHINE-CG calculation were scattered around 4.0, whereas the ones for SKYSHINE calculation were as large as 6.0. Calculational error was found to be reduced by adopting three-dimensional model based on the combinatorial geometry method. (author)

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)

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).

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.

On-line monitoring and inservice inspection in codes; Betriebsueberwachung und wiederkehrende Pruefungen in den Regelwerken

Energy Technology Data Exchange (ETDEWEB)

Bartonicek, J.; Zaiss, W. [Gemeinschaftskernkraftwerk Neckar GmbH, Neckarwestheim (Germany); Bath, H.R. [Bundesamt fuer Strahlenschutz, Salzgitter (Germany). Geschaeftsstelle des Kerntechnischen Ausschusses (KTA)

1999-08-01

The relevant regulatory codes determine the ISI tasks and the time intervals for recurrent components testing for evaluation of operation-induced damaging or ageing in order to ensure component integrity on the basis of the last available quality data. In-service quality monitoring is carried out through on-line monitoring and recurrent testing. The requirements defined by the engineering codes elaborated by various institutions are comparable, with the KTA nuclear engineering and safety codes being the most complete provisions for quality evaluation and assurance after different, defined service periods. German conventional codes for assuring component integrity provide exclusively for recurrent inspection regimes (mainly pressure tests and optical testing). The requirements defined in the KTA codes however always demanded more specific inspections relying on recurrent testing as well as on-line monitoring. Foreign codes for ensuring component integrity concentrate on NDE tasks at regular time intervals, with time intervals scope of testing activities being defined on the basis of the ASME code, section XI. (orig./CB) [Deutsch] Fuer die Komponentenintegritaet sind die Schaedigungsmechanismen mit dem nach den Regelwerken einzuhaltenden Abstand abzusichern. Dabei ist die jeweils vorhandene (Ist-) Qualitaet als Ausgangspunkt entscheidend. Die Absicherung der vorhandenen Qualitaet im weiteren Betrieb erfolgt durch geeignete Betriebsueberwachung und wiederkehrende Pruefungen. Die Anforderungen der Regelwerke sind vergleichbar, wobei die Bestimmung der vorhandenen Qualitaet nach einer bestimmten Betriebszeit sowie deren Absicherung im weiteren Betrieb am vollstaendigsten auf Basis des KTA-Regelwerkes moeglich ist. Die Absicherung der Komponentenintegritaet im Betrieb beruht in deutschen konventionellen Regelwerken nur auf den wiederkehrenden Pruefungen (hauptsaechlich Druckpruefungen und Sichtpruefungen). Das KTA-Regelwerk forderte hier schon immer qualifizierte

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)

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)

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)

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).

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)

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

An Integration of the Restructured Melcor for the Midas Computer Code

International Nuclear Information System (INIS)

Sunhee Park; Dong Ha Kim; Ko-Ryu Kim; Song-Won Cho

2006-01-01

The developmental need for a localized severe accident analysis code is on the rise. KAERI is developing a severe accident code called MIDAS, which is based on MELCOR. In order to develop the localized code (MIDAS) which simulates a severe accident in a nuclear power plant, the existing data structure is reconstructed for all the packages in MELCOR, which uses pointer variables for data transfer between the packages. During this process, new features in FORTRAN90 such as a dynamic allocation are used for an improved data saving and transferring method. Hence the readability, maintainability and portability of the MIDAS code have been enhanced. After the package-wise restructuring, the newly converted packages are integrated together. Depending on the data usage in the package, two types of packages can be defined: some use their own data within the package (let's call them independent packages) and the others share their data with other packages (dependent packages). For the independent packages, the integration process is simple to link the already converted packages together. That is, the package-wise structuring does not require further conversion of variables for the integration process. For the dependent packages, extra conversion is necessary to link them together. As the package-wise restructuring converts only the corresponding package's variables, other variables defined from other packages are not touched and remain as it is. These variables are to be converted into the new types of variables simultaneously as well as the main variables in the corresponding package. Then these dependent packages are ready for integration. In order to check whether the integration process is working well, the results from the integrated version are verified against the package-wise restructured results. Steady state runs and station blackout sequences are tested and the major variables are found to be the same each other. In order to verify the results, the integrated

The Conversion of Wiswesser Line Notations to Ring Codes. I. The Conversion of Ring Systems

Science.gov (United States)

Granito, Charles E.; And Others

1972-01-01

The computerized conversion of Wiswesser Line Notations to Ring Codes, using a two-part approach, and the set of computer programs generated for the conversion of ring systems are described. (9 references) (Author)

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.

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

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)

An evaluation of TRAC-PF1/MOD1 computer code performance during posttest simulations of Semiscale MOD-2C feedwater line break transients

International Nuclear Information System (INIS)

Hall, D.G.; Watkins, J.C.

1987-01-01

This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In addition to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use

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

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.

SPACE code simulation of ATLAS DVI line break accident test (SB DVI 08 Test)

Energy Technology Data Exchange (ETDEWEB)

Lim, Sang Gyu [KHNP, Daejeon (Korea, Republic of)

2012-10-15

APR1400 has adopted new safety design features which are 4 mechanically independent DVI (Direct Vessel Injection) systems and fluidic device in the safety injection tanks (SITs). Hence, DVI line break accident has to be evaluated as one of the small break LOCA (SBLOCA) to ensure the safety of APR1400. KAERI has been performed for DVI line break test (SB DVI 08) using ATLAS (Advanced Thermal Hydraulic Test Loop for Accident Simulation) facility which is an integral effect test facility for APR1400. The test result shows that the core collapsed water level decreased before a loop seal clearance, so that a core uncover occurred. At this time, the peak cladding temperature (PCT) is rapidly increased even though the emergency core cooling (ECC) water is injected from safety injection pump (SIP). This test result is useful for supporting safety analysis using thermal hydraulic safety analysis code and increases the understanding of SBLOCA phenomena in APR1400. The SBLOCA evaluation methodology for APR1400 is now being developed using SPACE code. The object of the development of this methodology is to set up a conservative evaluation methodology in accordance with appendix K of 10 CFR 50. ATLAS SB DVI 08 test is selected for the evaluation of SBLOCA methodology using SPACE code. Before applying the conservative models and correlations, benchmark calculation of the test is performed with the best estimate models and correlations to verify SPACE code capability. This paper deals with benchmark calculations results of ATLAS SB DVI 08 test. Calculation results of the major hydraulics variables are compared with measured data. Finally, this paper carries out the SPACE code performances for simulating the integral effect test of SBLOCA.

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

Recognition of facial expressions by cortical multi-scale line and edge coding

OpenAIRE

Sousa, R.; Rodrigues, J. M. F.; du Buf, J. M. H.

2010-01-01

Face-to-face communications between humans involve emotions, which often are unconsciously conveyed by facial expressions and body gestures. Intelligent human-machine interfaces, for example in cognitive robotics, need to recognize emotions. This paper addresses facial expressions and their neural correlates on the basis of a model of the visual cortex: the multi-scale line and edge coding. The recognition model links the cortical representation with Paul Ekman's Action Units which are relate...

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

A line code with quick-resynchronization capability and low latency for the optical data links of LHC experiments

International Nuclear Information System (INIS)

Deng, B; He, M; Chen, J; Guo, D; Hou, S; Teng, P-K; Li, X; Liu, C; Xiang, A C; Ye, J; Gong, D; Liu, T; You, Y

2014-01-01

We propose a line code that has fast resynchronization capability and low latency. Both the encoder and decoder have been implemented in FPGAs. The encoder has also been implemented in an ASIC. The latency of the whole optical link (not including the optical fiber) is estimated to be less than 73.9 ns. In the case of radiation-induced link synchronization loss, the decoder can recover the synchronization in 25 ns. The line code will be used in the ATLAS liquid argon calorimeter Phase-I trigger upgrade and can also be potentially used in other LHC experiments

Feasibility analysis of the modified ATHLET code for supercritical water cooled systems

Energy Technology Data Exchange (ETDEWEB)

Zhou Chong, E-mail: ch.zhou@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Priessnitz-Str. 3, 76131 Karlsruhe (Germany); Yang Yanhua [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Cheng Xu [Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Priessnitz-Str. 3, 76131 Karlsruhe (Germany)

2012-09-15

Highlights: Black-Right-Pointing-Pointer Modification of system code ATHLET for supercritical water application. Black-Right-Pointing-Pointer Development and assessment of a heat transfer package for supercritical water. Black-Right-Pointing-Pointer Validation of the modified code at supercritical pressures with the theoretical point-hydraulics model and the SASC code. Black-Right-Pointing-Pointer Application of the modified code to LOCA analysis of a supercritical water cooled in-pile fuel qualification test loop. - Abstract: Since the existing thermal-hydraulic computer codes for light water reactors are not applicable to supercritical water cooled reactors (SCWRs) owing to the limitation of physical models and numerical treatments, the development of a reliable thermal-hydraulic computer code is very important to design analysis and safety assessment of SCWRs. Based on earlier modification of ATHLET for SCWR, a general interface is implemented to the code, which serves as the platform for information exchange between ATHLET and the external independent physical modules. A heat transfer package containing five correlations for supercritical water is connected to the ATHLET code through the interface. The correlations are assessed with experimental data. To verify the modified ATHLET code, the Edwards-O'Brian blow-down test is simulated. As first validation at supercritical pressures, a simplified supercritical water cooled loop is modeled and its stability behavior is analyzed. Results are compared with that of the theoretical model and SASC code in the reference and show good agreement. To evaluate its feasibility, the modified ATHLET code is applied to a supercritical water cooled in-pile fuel qualification test loop. Loss of coolant accidents (LOCAs) due to break of coolant supply lines are calculated for the loop. Sensitivity analysis of some safety system parameters is performed to get further knowledge about their influence on the function of the

CONDOP: an R package for CONdition-Dependent Operon Predictions.

Science.gov (United States)

Fortino, Vittorio; Tagliaferri, Roberto; Greco, Dario

2016-10-15

The use of high-throughput RNA sequencing to predict dynamic operon structures in prokaryotic genomes has recently gained popularity in bioinformatics. We provide the R implementation of a novel method that uses transcriptomic features extracted from RNA-seq transcriptome profiles to develop ensemble classifiers for condition-dependent operon predictions. The CONDOP package provides a deeper insight into RNA-seq data analysis and allows scientists to highlight the operon organization in the context of transcriptional regulation with a few lines of code. CONDOP is implemented in R and is freely available at CRAN. vittorio.fortino@helsinki.fiSupplementary information: Supplementary data are available at Bioinformatics online. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please e-mail: journals.permissions@oup.com.

Regulatory compliance in the design of packages used to transport radioactive materials

International Nuclear Information System (INIS)

Raske, D.T.

1993-01-01

Shipments of radioactive materials within the regulatory jurisdiction of the US Department of Energy (DOE) must meet the package design requirements contained in Title 10 of the Code of Federal Regulations, Part 71, and DOE Order 5480.3. These regulations do not provide design criteria requirements, but only detail the approval standards, structural performance criteria, and package integrity requirements that must be met during transport. The DOE recommended design criterion for high-level Category I radioactive packagings is Section III, Division 1, of the ASME Boiler and Pressure Vessel Code. However, alternative design criteria may be used if all the design requirements are satisfied. The purpose of this paper is to review alternatives to the Code criteria and discuss their applicability to the design of containment vessels in packages for high-level radioactive materials. Issues such as design qualification by physical testing, the use of scale models, and problems encountered using a non-ASME design approach are addressed

A 2.9 ps equivalent resolution interpolating time counter based on multiple independent coding lines

International Nuclear Information System (INIS)

Szplet, R; Jachna, Z; Kwiatkowski, P; Rozyc, K

2013-01-01

We present the design, operation and test results of a time counter that has an equivalent resolution of 2.9 ps, a measurement uncertainty at the level of 6 ps, and a measurement range of 10 s. The time counter has been implemented in a general-purpose reprogrammable device Spartan-6 (Xilinx). To obtain both high precision and wide measurement range the counting of periods of a reference clock is combined with a two-stage interpolation within a single period of the clock signal. The interpolation involves a four-phase clock in the first interpolation stage (FIS) and an equivalent coding line (ECL) in the second interpolation stage (SIS). The ECL is created as a compound of independent discrete time coding lines (TCL). The number of TCLs used to create the virtual ECL has an effect on its resolution. We tested ECLs made from up to 16 TCLs, but the idea may be extended to a larger number of lines. In the presented time counter the coarse resolution of the counting method equal to 2 ns (period of the 500 MHz reference clock) is firstly improved fourfold in the FIS and next even more than 400 times in the SIS. The proposed solution allows us to overcome the technological limitation in achievable resolution and improve the precision of conversion of integrated interpolators based on tapped delay lines. (paper)

Atmospheric stellar parameters for large surveys using FASMA, a new spectral synthesis package

Science.gov (United States)

Tsantaki, M.; Andreasen, D. T.; Teixeira, G. D. C.; Sousa, S. G.; Santos, N. C.; Delgado-Mena, E.; Bruzual, G.

2018-02-01

In the era of vast spectroscopic surveys focusing on Galactic stellar populations, astronomers want to exploit the large quantity and good quality of data to derive their atmospheric parameters without losing precision from automatic procedures. In this work, we developed a new spectral package, FASMA, to estimate the stellar atmospheric parameters (namely effective temperature, surface gravity and metallicity) in a fast and robust way. This method is suitable for spectra of FGK-type stars in medium and high resolution. The spectroscopic analysis is based on the spectral synthesis technique using the radiative transfer code, MOOG. The line list is comprised of mainly iron lines in the optical spectrum. The atomic data are calibrated after the Sun and Arcturus. We use two comparison samples to test our method, (i) a sample of 451 FGK-type dwarfs from the high-resolution HARPS spectrograph; and (ii) the Gaia-ESO benchmark stars using both high and medium resolution spectra. We explore biases in our method from the analysis of synthetic spectra covering the parameter space of our interest. We show that our spectral package is able to provide reliable results for a wide range of stellar parameters, different rotational velocities, different instrumental resolutions and for different spectral regions of the VLT-GIRAFFE spectrographs, used amongst others for the Gaia-ESO survey. FASMA estimates stellar parameters in less than 15 m for high-resolution and 3 m for medium-resolution spectra. The complete package is publicly available to the community.

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.)

A predictive transport modeling code for ICRF-heated tokamaks

International Nuclear Information System (INIS)

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

1992-02-01

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

FLOC: Field Line and Orbit Code for the study of ripple beam injection into tokamaks

Energy Technology Data Exchange (ETDEWEB)

Fowler, R. H.; Lee, D. K.; Gaffney, P. W.; Rome, J. A.

1978-06-01

The computer code described is used to study ripple beam injection into a tokamak plasma. The collisionless guiding center equations of motion are integrated to find the orbits of single particles in realistic magnetic fields for ripple injection. In order to determine if the ripple is detrimental to the plasma, the magnetic flux surfaces are constructed by integration of the field line equations. The numerical techniques are described, and use of the code is outlined. A program listing is provided, and the results of sample cases are presented.

FLOC: Field Line and Orbit Code for the study of ripple beam injection into tokamaks

International Nuclear Information System (INIS)

Fowler, R.H.; Lee, D.K.; Gaffney, P.W.; Rome, J.A.

1978-06-01

The computer code described is used to study ripple beam injection into a tokamak plasma. The collisionless guiding center equations of motion are integrated to find the orbits of single particles in realistic magnetic fields for ripple injection. In order to determine if the ripple is detrimental to the plasma, the magnetic flux surfaces are constructed by integration of the field line equations. The numerical techniques are described, and use of the code is outlined. A program listing is provided, and the results of sample cases are presented

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

A 1ST Step Integration of the Restructured MELCOR for the MIDAS Computer Code

International Nuclear Information System (INIS)

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

2006-01-01

KAERI is developing a localized severe accident code, MIDAS, based on MELCOR. MELCOR uses pointer variables for a fixed-size storage management to save the data. It passes data through two depths, its meaning is not understandable by variable itself. So it is needed to understand the methods for data passing. This method 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 for each package was developed and tested. And then integration of each restructured package was being processed one by one. In this paper, the integrating scope includes the BUR, CF, CVH, DCH, EDF, ESF, MP, SPR, TF and TP packages. As most of them use data within each package and a few packages share data with other packages. The verification was done through comparing the results before and after the restructuring

GRILLIX. A 3D turbulence code for magnetic fusion devices based on a field line map

International Nuclear Information System (INIS)

Stegmeir, Andreas Korbinian

2015-01-01

The complex geometry in the scrape-off layer of tokamaks poses problems to existing turbulence codes. The usually employed field aligned coordinates become ill defined at the separatrix. Therefore the parallel code GRILLIX was developed, which is based on a field line map. This allows simulations in additional complex geometries, especially across the separatrix. A new discretisation, based on the support operator method, for the highly anisotropic diffusion was developed and applied to a simple turbulence model (Hasegawa-Wakatani).

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

Article I. Multi-platform Automated Software Building and Packaging

International Nuclear Information System (INIS)

Rodriguez, A Abad; Gomes Gouveia, V E; Meneses, D; Capannini, F; Aimar, A; Di Meglio, A

2012-01-01

One of the major goals of the EMI (European Middleware Initiative) project is the integration of several components of the pre-existing middleware (ARC, gLite, UNICORE and dCache) into a single consistent set of packages with uniform distributions and repositories. Those individual middleware projects have been developed in the last decade by tens of development teams and before EMI were all built and tested using different tools and dedicated services. The software, millions of lines of code, is written in several programming languages and supports multiple platforms. Therefore a viable solution ought to be able to build and test applications on multiple programming languages using common dependencies on all selected platforms. It should, in addition, package the resulting software in formats compatible with the popular Linux distributions, such as Fedora and Debian, and store them in repositories from which all EMI software can be accessed and installed in a uniform way. Despite this highly heterogeneous initial situation, a single common solution, with the aim of quickly automating the integration of the middleware products, had to be selected and implemented in a few months after the beginning of the EMI project. Because of the previous knowledge and the short time available in which to provide this common solution, the ETICS service, where the gLite middleware was already built for years, was selected. This contribution describes how the team in charge of providing a common EMI build and packaging infrastructure to the whole project has developed a homogeneous solution for releasing and packaging the EMI components from the initial set of tools used by the earlier middleware projects. An important element of the presentation is the developers experience and feedback on converging on ETICS and on the on-going work in order to finally add more widely used and supported build and packaging solutions of the Linux platforms

Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

International Nuclear Information System (INIS)

Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J.

2001-01-01

A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

Energy Technology Data Exchange (ETDEWEB)

Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J. [Tractebel Energy Engineering, Brussels (Belgium)

2001-07-01

A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

ENDF utility codes version 6.4 for ENDF-5 and ENDF-6

International Nuclear Information System (INIS)

McLaughlin, P.K.

1988-09-01

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, GRABLIB, VERSAT; and the file maintenance and retrieval codes LISTEF, SETMDC, GETMAT, STANEF. This program package can be obtained on magnetic tape or floppy diskette, free of charge, from the IAEA Nuclear Data Section. (author)

eXtended CASA Line Analysis Software Suite (XCLASS)

Science.gov (United States)

Möller, T.; Endres, C.; Schilke, P.

2017-02-01

The eXtended CASA Line Analysis Software Suite (XCLASS) is a toolbox for the Common Astronomy Software Applications package (CASA) containing new functions for modeling interferometric and single dish data. Among the tools is the myXCLASS program which calculates synthetic spectra by solving the radiative transfer equation for an isothermal object in one dimension, whereas the finite source size and dust attenuation are considered as well. Molecular data required by the myXCLASS program are taken from an embedded SQLite3 database containing entries from the Cologne Database for Molecular Spectroscopy (CDMS) and JPL using the Virtual Atomic and Molecular Data Center (VAMDC) portal. Additionally, the toolbox provides an interface for the model optimizer package Modeling and Analysis Generic Interface for eXternal numerical codes (MAGIX), which helps to find the best description of observational data using myXCLASS (or another external model program), that is, finding the parameter set that most closely reproduces the data. http://www.astro.uni-koeln.de/projects/schilke/myXCLASSInterface A copy of the code is available at the CDS via anonymous ftp to http://cdsarc.u-strasbg.fr (http://130.79.128.5) or via http://cdsarc.u-strasbg.fr/viz-bin/qcat?J/A+A/598/A7

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1992-06-01

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

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

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.

Stabilizing And Packaging Pu Materials Per 3013 At SRS

International Nuclear Information System (INIS)

STEVE, HENSEL

2005-01-01

The Savannah River Site (SRS) began packaging Pu metals into 3013 containers in April, 2003 and oxides in October, 2003. A total of 919 outer 3013 containers were made in the FB-Line at SRS when stabilization and packaging was completed in January, 2005. Experiences, lessons learned, and an overview of packaging activities are presented

ENDF utility codes release 6.10. Description and operating instructions

International Nuclear Information System (INIS)

Dunford, C.L.

1995-01-01

Description and operating instructions are given for a package of utility codes operating on evaluated nuclear data files in the formats ENDF-6 (and ENDF-5). Included are the data checking codes CHECKER, FIZCON, PSYCHE; the code INTER for retrieving thermal cross-sections and some other data; graphical plotting subroutines PLOTEF, GRALIB, 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)

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...

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...

Design and tests of a package for the transport of radioactive sources

International Nuclear Information System (INIS)

Santos, Paulo de Oliveira

2011-01-01

The Type A package was designed for transportation of seven cobalt-60 sources with total activity of 1 GBq. The shield thickness to accomplish the dose rate and the transport index established by the radioactive transport regulation was calculated by the code MCNP (Monte Carlo N-Particle Transport Code Version 5). The sealed cobalt-60 sources were tested for leakages. according to the regulation ISO 9978:1992 (E). The package was tested according to regulation Radioactive Material Transport CNEN. The leakage tests results pf the sources, and the package tests demonstrate that the transport can be safe performed from the CDTN to the steelmaking industries

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

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

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.

Applications of the Discrete ordinates of Oak ridge System (DOORS) package to Nuclear Engineering problems

Energy Technology Data Exchange (ETDEWEB)

Azmy, Y.Y. [The Pennsylvania State University, 229 Reber Building, University Park, PA 16802 (United States)]. e-mail: yya3@psu.edu

2004-07-01

Particle transport problems are notorious for their difficulty. This fact requires that production level computer codes designed to address realistic engineering problems possess three important features: (i) high computational efficiency as measured by solution accuracy for a fixed computational cost; (ii) a wide variety of options to enhance robustness of the transport solver; and (iii) a broad collection of support codes that extend the reach of the transport solver to a wide variety of applications. The Discrete Ordinates of Oak Ridge System (DOORS) code package was designed with these features in mind. In this paper, capabilities of member codes in the DOORS package are overviewed with particular emphasis on two newly developed peripheral codes: BOT3P the mesh-generation and visualization code package, and GipGui the graphical user interface for the cross section manipulation code, GIP. Two large applications are used to illustrate the tight coupling between the peripheral codes and the DORT and TORT transport solvers in two and three dimensional geometries, respectively. These are: (i) criticality calculations for the C5G7MOX core benchmark; and (ii) dose distribution calculations for the Target Service Cell (TSC) of the Spallation Neutron Source (SNS). (Author)

Applications of the Discrete ordinates of Oak ridge System (DOORS) package to Nuclear Engineering problems

International Nuclear Information System (INIS)

Azmy, Y.Y.

2004-01-01

Particle transport problems are notorious for their difficulty. This fact requires that production level computer codes designed to address realistic engineering problems possess three important features: (i) high computational efficiency as measured by solution accuracy for a fixed computational cost; (ii) a wide variety of options to enhance robustness of the transport solver; and (iii) a broad collection of support codes that extend the reach of the transport solver to a wide variety of applications. The Discrete Ordinates of Oak Ridge System (DOORS) code package was designed with these features in mind. In this paper, capabilities of member codes in the DOORS package are overviewed with particular emphasis on two newly developed peripheral codes: BOT3P the mesh-generation and visualization code package, and GipGui the graphical user interface for the cross section manipulation code, GIP. Two large applications are used to illustrate the tight coupling between the peripheral codes and the DORT and TORT transport solvers in two and three dimensional geometries, respectively. These are: (i) criticality calculations for the C5G7MOX core benchmark; and (ii) dose distribution calculations for the Target Service Cell (TSC) of the Spallation Neutron Source (SNS). (Author)

Labelling and marking of packages, for the transport of radioactive materials

International Nuclear Information System (INIS)

1977-09-01

It is the responsibility of the consignor, even when he is also the carrier, to ensure that every package of dangerous materials is correctly labelled and marked before dispatch. The purpose of this Code of Practice is to amplify the provisions, embodied in various regulations and codes for the safe transport of radioactive materials, relating to the labelling of packages of such materials, and to provide detailed instructions that will ensure fulfilment of the relevant requirements. The model regulations published by the International Atomic Energy Agency are referred to in this Code as 'the IAEA regulations'. It has been assumed that those using the Code will be familiar with the international and national transport regulations, which are based on the IAEA regulations and that they will have experience of transport procedures. (author)

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

International Nuclear Information System (INIS)

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

1985-07-01

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

A restructuring proposal based on MELCOR for severe accident analysis code development

Energy Technology Data Exchange (ETDEWEB)

Park, Sun Hee; Song, Y. M.; Kim, D. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-03-01

In order to develop a template based on existing MELCOR code, current data saving and transferring methods used in MELCOR are addressed first. Then a naming convention for the constructed module is suggested and an automatic program to convert old variables into new derived type variables has been developed. Finally, a restructured module for the SPR package has been developed to be applied to MELCOR. The current MELCOR code ensures a fixed-size storage for four different data types, and manages the variable-sized data within the storage limit by storing the data on the stacked packages. It uses pointer to identify the variables between the packages. This technique causes a difficult grasping of the meaning of the variables as well as memory waste. New features of FORTRAN90, however, make it possible to allocate the storage dynamically, and to use the user-defined data type which lead to a restructured module development for the SPR package. An efficient memory treatment and as easy understanding of the code are allowed in this developed module. The validation of the template 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. The template for the SPR package suggested in this report hints the extension of the template to the entire code. It is expected that the template will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models. 3 refs., 15 figs., 16 tabs. (Author)

Review of criticality safety and shielding analysis issues for transportation packages

International Nuclear Information System (INIS)

Parks, C.V.; Broadhead, B.L.

1995-01-01

The staff of the Nuclear Engineering Applications Section (NEAS) at Oak Ridge National Laboratory (ORNL) have been involved for over 25 years with the development and application of computational tools for use in analyzing the criticality safety and shielding features of transportation packages carrying radioactive material (RAM). The majority of the computational tools developed by ORNL/NEAS have been included within the SCALE modular code system (SCALE 1995). This code system has been used throughout the world for the evaluation of nuclear facility and package designs. With this development and application experience as a basis, this paper highlights a number of criticality safety and shielding analysis issues that confront the designer and reviewer of a new RAM package. Changes in the types and quantities of material that need to be shipped will keep these issues before the technical community and provide challenges to future package design and certification

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)

Radcalc: A computer program to calculate the radiolytic production of hydrogen gas from radioactive wastes in packages

International Nuclear Information System (INIS)

Green, J.R.; Schwarz, R.A.; Hillesland, K.E.; Roetman, V.E.; Field, J.G.

1995-11-01

Radcalc for Windows' is a menu-driven Microsoft2 Windows-compatible computer code that calculates the radiolytic production of hydrogen gas in high- and low-level radioactive waste. In addition, the code also determines US Department of Transportation (DOT) transportation classifications, calculates the activities of parent and daughter isotopes for a specified period of time, calculates decay heat, and calculates pressure buildup from the production of hydrogen gas in a given package geometry. Radcalc for Windows was developed by Packaging Engineering, Transportation and Packaging, Westinghouse Hanford Company, Richland, Washington, for the US Department of Energy (DOE). It is available from Packaging Engineering and is issued with a user's manual and a technical manual. The code has been verified and validated

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

X-ray emission lines from photoionized plasmas

International Nuclear Information System (INIS)

Liedahl, D.A.

1992-11-01

Plasma emission codes have become a standard tool for the analysis of spectroscopic data from cosmic X-ray sources. However, the assumption of collisional equilibrium, typically invoked in these codes, renders them inapplicable to many important astrophysical situations, particularly those involving X-ray photoionized nebulae, which are likely to exist in the circumsource environments of compact X-ray sources. X-ray line production in a photoionized plasma is primarily the result of radiative cascades following recombination. Through the development of atomic models of several highly-charged ions, this work extends the range of applicability of discrete spectral models to plasmas dominated by recombination. Assuming that ambient plasma conditions lie in the temperature range 10 5 --10 6 K and the density range 10 11 --10 16 cm -3 , X-ray line spectra are calculated over the wavelength range 5--45 angstrom using the HULLAC atomic physics package. Most of the work focuses on the Fe L-shell ions. Line ratios of the form (3s-2p)/(3d-2p) are shown to characterize the principal mode of line excitation, thereby providing a simple signature of photoionization. At electron densities exceeding 10 12 cm -3 , metastable state populations in the ground configurations approach their LTE value, resulting in the enrichment of the Fe L-shell recombination spectrum and a set of density-sensitive X-ray line ratios. Radiative recombination continua and emission lines produced selectively by Δn = 0 dielectronic recombination are shown to provide two classes of temperature diagnostics. Because of the extreme overionization, the recombination continua are expected to be narrow (ΔE/E much-lt 1), with ΔE = kT. Dielectronic recombination selectively drives radiative transitions that originate on states with vacancies in the 2s subshell, states that are inaccessible under pure RR population kinetics

Improvement and validation of the wall heat transfer package of RELAP5/MOD3.3

International Nuclear Information System (INIS)

Wu, Pan; Xiong, Xiaofei; Shan, Jianqiang; Gou, Junli; Zhang, Bin; Zhang, Bo

2016-01-01

Highlights: • A new heat transfer package has been developed. • It has been incorporated into RELAP5/MOD3.3 to verify its advantages. • The results of modified code were compared with available experimental data. • The results showed that higher prediction accuracy was achieved. - Abstract: The process of energy transfer from heat structure to control volume is determined by the wall-to-fluid heat transfer package, which is crucial for nuclear reactor safety analysis codes. The current logic for selection of heat transfer modes of RELAP5/MOD3.3 code is too complex and may result in incorrect heat transfer mode judgment. Also, the narrow application scope of film boiling heat transfer correlations may result in large errors in film boiling region which is of paramount importance for the predicted peak clad temperatures during hypothetical LB-LOCAs in PWRs. In this study, a new heat transfer package has been developed and incorporated into the RELAP5/MOD3.3 code. Differing from the original package, the modified one consists of twelve heat transfer modes and proposes a new logic for selection of heat transfer modes. For each mode, the models in the existing safety analysis codes and the leading models in literature have been reviewed in order to determine the best model which can easily be applicable to the RELAP5/MOD3.3 code. Particularly (1) a new package of heat transfer correlations are produced; (2) a new logic for selection of film boiling and transition boiling heat transfer modes is proposed which use minimum film boiling temperature and critical heat flux temperature as distinguished points. The modified code has been validated by comparing the analysis results with available experimental data from tube post dryout experiments and loss-of-fluid test (LOFT) facility. The calculation results showed that the improved package could better predict the experimental phenomena with higher prediction accuracy.

Improvement and validation of the wall heat transfer package of RELAP5/MOD3.3

Energy Technology Data Exchange (ETDEWEB)

Wu, Pan; Xiong, Xiaofei; Shan, Jianqiang, E-mail: jqshan@mail.xjtu.edu.cn; Gou, Junli; Zhang, Bin; Zhang, Bo

2016-12-15

Highlights: • A new heat transfer package has been developed. • It has been incorporated into RELAP5/MOD3.3 to verify its advantages. • The results of modified code were compared with available experimental data. • The results showed that higher prediction accuracy was achieved. - Abstract: The process of energy transfer from heat structure to control volume is determined by the wall-to-fluid heat transfer package, which is crucial for nuclear reactor safety analysis codes. The current logic for selection of heat transfer modes of RELAP5/MOD3.3 code is too complex and may result in incorrect heat transfer mode judgment. Also, the narrow application scope of film boiling heat transfer correlations may result in large errors in film boiling region which is of paramount importance for the predicted peak clad temperatures during hypothetical LB-LOCAs in PWRs. In this study, a new heat transfer package has been developed and incorporated into the RELAP5/MOD3.3 code. Differing from the original package, the modified one consists of twelve heat transfer modes and proposes a new logic for selection of heat transfer modes. For each mode, the models in the existing safety analysis codes and the leading models in literature have been reviewed in order to determine the best model which can easily be applicable to the RELAP5/MOD3.3 code. Particularly (1) a new package of heat transfer correlations are produced; (2) a new logic for selection of film boiling and transition boiling heat transfer modes is proposed which use minimum film boiling temperature and critical heat flux temperature as distinguished points. The modified code has been validated by comparing the analysis results with available experimental data from tube post dryout experiments and loss-of-fluid test (LOFT) facility. The calculation results showed that the improved package could better predict the experimental phenomena with higher prediction accuracy.

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

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

“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.

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

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

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)

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.

Review of waste package verification tests. Semiannual report, April 1985-September 1985

International Nuclear Information System (INIS)

Soo, P.

1986-01-01

Several studies were completed this period to evaluate experimental and analytical methodologies being used in the DOE waste package program. The first involves a determination of the relevance of the test conditions being used by DOE to characterize waste package component behavior in a salt repository system. Another study focuses on the testing conditions and procedures used to measure radionuclide solubility and colloid formation in repository groundwaters. An attempt was also made to evaluate the adequacy of selected waste package performance codes. However, the latter work was limited by an inability to obtain several codes from DOE. Nevertheless, it was possible to comment briefly on the structures and intents of the codes based on publications in the open literature. The final study involved an experimental program to determine the likelihood of stress-corrosion cracking of austenitic stainless steels and Incoloy 825 in simulated tuff repository environments. Tests for six-month exposure periods in water and air-steam conditions are described. 52 figs., 48 tabs

Evaluation of radiation packages type A from the center of isotopes in Cuba

International Nuclear Information System (INIS)

Balbona, Zayda Amador; Pijuan, Saul Perez; Gual, Maritza Rodriguez

2013-01-01

The Isotope Center (CENTIS) of the Republic of Cuba makes the transportation of its products mainly in packaged type A. To undertake the design of packages, packaging components from 6 producing firms (including those found Amersham, CISBIO and IZOTOP) are studied. From the applicable regulations, security features and requirements are established as well as the technical characteristics of the packaging components. This study evaluated according each radioisotope, product and specific activity, high activity that can be included in a Type A package with the limitation that the dose rate on their surfaces is less than or equal to 2 mSv/h. In addition, each package is characterized taking into account the value of the maximum dose rate at maximum contact and the transport index for the day of transport. For this, the Microshield code using version 5.0.3. The dose rate in contact with the package of 90 Y is calculated using the Monte Carlo code MCNPX version 2.6.0. The maximum possible activity values are obtained for each shielding transport radionuclides CENTIS produced, namely 131 I, 125 I, 32 P, 99 Mo/ 99m Tc, 99m Tc, 188 Re and 90 Y and 69 radioactive packages type A are evaluated

Development and verification of a coupled code system RETRAN-MASTER-TORC

International Nuclear Information System (INIS)

Cho, J.Y.; Song, J.S.; Joo, H.G.; Zee, S.Q.

2004-01-01

Recently, coupled thermal-hydraulics (T-H) and three-dimensional kinetics codes have been widely used for the best-estimate simulations such as the main steam line break (MSLB) and locked rotor problems. This work is to develop and verify one of such codes by coupling the system T-H code RETRAN, the 3-D kinetics code MASTER and sub-channel analysis code TORC. The MASTER code has already been applied to such simulations after coupling with the MARS or RETRAN-3D multi-dimensional system T-H codes. The MASTER code contains a sub-channel analysis code COBRA-III C/P, and the coupled systems MARSMASTER-COBRA and RETRAN-MASTER-COBRA had been already developed and verified. With these previous studies, a new coupled system of RETRAN-MASTER-TORC is to be developed and verified for the standard best-estimate simulation code package in Korea. The TORC code has already been applied to the thermal hydraulics design of the several ABB/CE type plants and Korean Standard Nuclear Power Plants (KSNP). This justifies the choice of TORC rather than COBRA. Because the coupling between RETRAN and MASTER codes are already established and verified, this work is simplified to couple the TORC sub-channel T-H code with the MASTER neutronics code. The TORC code is a standalone code that solves the T-H equations for a given core problem from reading the input file and finally printing the converged solutions. However, in the coupled system, because TORC receives the pin power distributions from the neutronics code MASTER and transfers the T-H results to MASTER iteratively, TORC needs to be controlled by the MASTER code and does not need to solve the given problem completely at each iteration step. By this reason, the coupling of the TORC code with the MASTER code requires several modifications in the I/O treatment, flow iteration and calculation logics. The next section of this paper describes the modifications in the TORC code. The TORC control logic of the MASTER code is then followed. The

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

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

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

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.)

49 CFR 178.522 - Standards for composite packagings with inner plastic receptacles.

Science.gov (United States)

2010-10-01

... plastic receptacles. 178.522 Section 178.522 Transportation Other Regulations Relating to Transportation... Standards for composite packagings with inner plastic receptacles. (a) The following are the identification codes for composite packagings with inner plastic receptacles: (1) 6HA1 for a plastic receptacle within...

Review of DOE waste package program. Subtask 1.1. National waste package program, April-September 1983. Volume 5

Energy Technology Data Exchange (ETDEWEB)

Soo, P. (ed.)

1984-08-01

The current effort is part of an ongoing task to review the national high-level waste package effort. It includes evaluations of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt, basalt, and tuff repositories. In the current Biannual Report a section on carbon steel container corrosion has been included to complement prior work on TiCode-12 and Type 304 stainless steel. The use of crushed tuff as a packing material is discussed and waste package component interaction test data are included. Licensing data requirements to estimate the degree of compliance with NRC performance objectives are specified. 41 figures, 24 tables.

Review of DOE waste package program. Subtask 1.1. National waste package program, April-September 1983. Volume 5

International Nuclear Information System (INIS)

Soo, P.

1984-08-01

The current effort is part of an ongoing task to review the national high-level waste package effort. It includes evaluations of reference waste form, container, and packing material components with respect to determining how they may contribute to the containment and controlled release of radionuclides after waste packages have been emplaced in salt, basalt, and tuff repositories. In the current Biannual Report a section on carbon steel container corrosion has been included to complement prior work on TiCode-12 and Type 304 stainless steel. The use of crushed tuff as a packing material is discussed and waste package component interaction test data are included. Licensing data requirements to estimate the degree of compliance with NRC performance objectives are specified. 41 figures, 24 tables

Fuel management and core design code systems for pressurized water reactor neutronic calculations

International Nuclear Information System (INIS)

Ahnert, C.; Arayones, J.M.

1985-01-01

A package of connected code systems for the neutronic calculations relevant in fuel management and core design has been developed and applied for validation to the startup tests and first operating cycle of a 900MW (electric) PWR. The package includes the MARIA code system for the modeling of the different types of PWR fuel assemblies, the CARMEN code system for detailed few group diffusion calculations for PWR cores at operating and burnup conditions, and the LOLA code system for core simulation using onegroup nodal theory parameters explicitly calculated from the detailed solutions

Sensitivity analysis of the RESRAD, a dose assessment code

International Nuclear Information System (INIS)

Yu, C.; Cheng, J.J.; Zielen, A.J.

1991-01-01

The RESRAD code is a pathway analysis code that is designed to calculate radiation doses and derive soil cleanup criteria for the US Department of Energy's environmental restoration and waste management program. the RESRAD code uses various pathway and consumption-rate parameters such as soil properties and food ingestion rates in performing such calculations and derivations. As with any predictive model, the accuracy of the predictions depends on the accuracy of the input parameters. This paper summarizes the results of a sensitivity analysis of RESRAD input parameters. Three methods were used to perform the sensitivity analysis: (1) Gradient Enhanced Software System (GRESS) sensitivity analysis software package developed at oak Ridge National Laboratory; (2) direct perturbation of input parameters; and (3) built-in graphic package that shows parameter sensitivities while the RESRAD code is operational

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.

Supplement to the approved requirements for the packaging, labelling and carriage of radioactive material by rail. Packaging, Labelling and Carriage of Radioactive Material by Rail Regulations 1996

International Nuclear Information System (INIS)

1999-01-01

The ADR and RID Framework Directives require EC member states' arrangements for the carriage of dangerous goods on domestic road and rail journeys to align with the existing ADR and RID agreements which cover international journeys by road and rail. Because ADR and RID are updated every two years in line with technical and scientific developments, the ADR/RID Framework Directives are also revised on a two-year cycle, to require member states to amend their implementing legislation accordingly. In Great Britain, these two Directives were initially implemented on 1 September 1996 via regulations (usually referred to as the 'carriage regulations'), containing the general legal duties, supported by approved documents, and an Approved Code of Practice containing the detailed technical requirements. The following approved documents have been updated: (a) Approved Vehicle Requirements (AVR) - L89; (b) Approved Requirements and test methods for the classification and packaging of dangerous goods for carriage (ARTM) - L88; (c) Approved Requirements for the packaging, labelling and carriage of radioactive material by rail (ARCRR) - L94; (d) Approved Requirements for the construction of vehicles intended for the carriage of explosives by road (AEVR) - L92; and (e) Approved Carriage List (ACL) - L90

Applications guide to the RSIC-distributed version of the MCNP code (coupled Monte Carlo neutron-photon Code)

International Nuclear Information System (INIS)

Cramer, S.N.

1985-09-01

An overview of the RSIC-distributed version of the MCNP code (a soupled Monte Carlo neutron-photon code) is presented. All general features of the code, from machine hardware requirements to theoretical details, are discussed. The current nuclide cross-section and other libraries available in the standard code package are specified, and a realistic example of the flexible geometry input is given. Standard and nonstandard source, estimator, and variance-reduction procedures are outlined. Examples of correct usage and possible misuse of certain code features are presented graphically and in standard output listings. Finally, itemized summaries of sample problems, various MCNP code documentation, and future work are given

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

International Nuclear Information System (INIS)

Marr, D.R.

1975-01-01

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

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

BUMS--Bonner sphere Unfolding Made Simple an HTML based multisphere neutron spectrometer unfolding package

CERN Document Server

Sweezy, J; Veinot, K

2002-01-01

A new multisphere neutron spectrometer unfolding package, Bonner sphere Unfolding Made Simple (BUMS) has been developed that uses an HTML interface to simplify data input and code execution for the novice and the advanced user. This new unfolding package combines the unfolding algorithms contained in other popular unfolding codes under one easy to use interface. The interface makes use of web browsing software to provide a graphical user interface to the unfolding algorithms. BUMS integrates the SPUNIT, BON, MAXIET, and SAND-II unfolding algorithms into a single package. This package also includes a library of 14 response matrices, 58 starting spectra, and 24 dose and detector responses. BUMS has several improvements beyond the addition of unfolding algorithms. It has the ability to search for the most appropriate starting spectra. Also, plots of the unfolded neutron spectra are automatically generated. The BUMS package runs via a web server and may be accessed by any computer with access to the Internet at h...

BUMS--Bonner sphere Unfolding Made Simple: an HTML based multisphere neutron spectrometer unfolding package

International Nuclear Information System (INIS)

Sweezy, Jeremy; Hertel, Nolan; Veinot, Ken

2002-01-01

A new multisphere neutron spectrometer unfolding package, Bonner sphere Unfolding Made Simple (BUMS) has been developed that uses an HTML interface to simplify data input and code execution for the novice and the advanced user. This new unfolding package combines the unfolding algorithms contained in other popular unfolding codes under one easy to use interface. The interface makes use of web browsing software to provide a graphical user interface to the unfolding algorithms. BUMS integrates the SPUNIT, BON, MAXIET, and SAND-II unfolding algorithms into a single package. This package also includes a library of 14 response matrices, 58 starting spectra, and 24 dose and detector responses. BUMS has several improvements beyond the addition of unfolding algorithms. It has the ability to search for the most appropriate starting spectra. Also, plots of the unfolded neutron spectra are automatically generated. The BUMS package runs via a web server and may be accessed by any computer with access to the Internet at http://nukeisit.gatech.edu/bums

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)

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.).

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)

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.)

Structural Evaluation on HIC Transport Packaging under Accident Conditions

International Nuclear Information System (INIS)

Chung, Sung Hwan; Kim, Duck Hoi; Jung, Jin Se; Yang, Ke Hyung; Lee, Heung Young

2005-01-01

HIC transport packaging to transport a high integrity container(HIC) containing dry spent resin generated from nuclear power plants is to comply with the regulatory requirements of Korea and IAEA for Type B packaging due to the high radioactivity of the content, and to maintain the structural integrity under normal and accident conditions. It must withstand 9 m free drop impact onto an unyielding surface and 1 m drop impact onto a mild steel bar in a position causing maximum damage. For the conceptual design of a cylindrical HIC transport package, three dimensional dynamic structural analysis to ensure that the integrity of the package is maintained under all credible loads for 9 m free drop and 1 m puncture conditions were carried out using ABAQUS code.

Eddylicious: A Python package for turbulent inflow generation

Science.gov (United States)

Mukha, Timofey; Liefvendahl, Mattias

2018-01-01

A Python package for generating inflow for scale-resolving computer simulations of turbulent flow is presented. The purpose of the package is to unite existing inflow generation methods in a single code-base and make them accessible to users of various Computational Fluid Dynamics (CFD) solvers. The currently existing functionality consists of an accurate inflow generation method suitable for flows with a turbulent boundary layer inflow and input/output routines for coupling with the open-source CFD solver OpenFOAM.

Measurement value analysis overall equipment effectiveness (OEE) packaging process in line 2 (Case Study of PT. MBI Tbk)

Science.gov (United States)

Rimawan, Erry; Kholil, Muhammad; Hendri

2018-03-01

PT. MBI Tbk is engaged in the manufacture of beverage industry, where the company’s production is based on the magnitude of customer demand that is marketing offices that had been scattered in various regions of Indonesia. In the packaging process steps in PT.MBI through the line 3 lines including racking, canning line, bottling line. In the canning process to existing packing on Line 2 (canning line), there are some machines that are used continuously, among other Depalletizer machine, filler machine, can seamer machine, pasteurizer machine, machine FLD, Wrap Around engine, engine Shrink Wrap. Due to the large demand from customers that is relentless, therefore the calculation of overall equipment effectiveness (OEE) as a whole on line 2 (canning line) is needed in order to make improvements continuously (Continuous Improvement) at line 2 (canning line). This study aims to determine the value of overall equipment effectiveness (OEE) and Losses of the most influential of the big six OEE Losses focused on equipment or machinery as a whole into a single unit that is on the line 2, which will then be known root cause of the losses that occur from the research over the field. From the calculation of overall equipment effectiveness (OEE), there are two ratios are still poor and under world-class standards, while the ratio of the availability of 88.85% of the world-class standards by 90% and the performance ratio of 78.51% of the standard world class by 95%, whereas for quality ratio has entered the world-class standard that is equal to 99.90%. Thus the value of OEE on Line 2 line is below world class standards. In this study there were only five losses, which can be identified, and while the losses were very influential, namely the Speed Reduced Losses, losses, these losses accounted for the largest percentage of the value of the rate of 19.12%, of the results of this study losses occurred due to poor surveillance systems (less good) that causes the employee or

Code Development in Coupled PARCS/RELAP5 for Supercritical Water Reactor

Directory of Open Access Journals (Sweden)

Po Hu

2014-01-01

Full Text Available The new capability is added to the existing coupled code package PARCS/RELAP5, in order to analyze SCWR design under supercritical pressure with the separated water coolant and moderator channels. This expansion is carried out on both codes. In PARCS, modification is focused on extending the water property tables to supercritical pressure, modifying the variable mapping input file and related code module for processing thermal-hydraulic information from separated coolant/moderator channels, and modifying neutronics feedback module to deal with the separated coolant/moderator channels. In RELAP5, modification is focused on incorporating more accurate water properties near SCWR operation/transient pressure and temperature in the code. Confirming tests of the modifications is presented and the major analyzing results from the extended codes package are summarized.

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.

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

Safety Analysis Report for Packaging, Y-12 National Security Complex, Model ES-3100 Package with Bulk HEU Contents

Energy Technology Data Exchange (ETDEWEB)

Anderson, James [Y-12 National Security Complex, Oak Ridge, TN (United States); Goins, Monty [Y-12 National Security Complex, Oak Ridge, TN (United States); Paul, Pran [Y-12 National Security Complex, Oak Ridge, TN (United States); Wilkinson, Alan [Y-12 National Security Complex, Oak Ridge, TN (United States); Wilson, David [Y-12 National Security Complex, Oak Ridge, TN (United States)

2015-09-03

This safety analysis report for packaging (SARP) presents the results of the safety analysis prepared in support of the Consolidated Nuclear Security, LLC (CNS) request for licensing of the Model ES-3100 package with bulk highly enriched uranium (HEU) contents and issuance of a Type B(U) Fissile Material Certificate of Compliance. This SARP, published in the format specified in the Nuclear Regulatory Commission (NRC) Regulatory Guide 7.9 and using information provided in UCID-21218 and NRC Regulatory Guide 7.10, demonstrates that the Y-12 National Security Complex (Y-12) ES-3100 package with bulk HEU contents meets the established NRC regulations for packaging, preparation for shipment, and transportation of radioactive materials given in Title 10, Part 71, of the Code of Federal Regulations (CFR) [10 CFR 71] as well as U.S. Department of Transportation (DOT) regulations for packaging and shipment of hazardous materials given in Title 49 CFR. To protect the health and safety of the public, shipments of adioactive materials are made in packaging that is designed, fabricated, assembled, tested, procured, used, maintained, and repaired in accordance with the provisions cited above. Safety requirements addressed by the regulations that must be met when transporting radioactive materials are containment of radioactive materials, radiation shielding, and assurance of nuclear subcriticality.

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...

Evaluation on the structural soundness of the package for subsurface disposal by finite element method

International Nuclear Information System (INIS)

Itoh, Chihiro

2009-01-01

The structural analysis of the disposal package for low-level radioactive wastes with relatively high activities (called L1 waste in Japan) were performed against normal and hypothetical conditions. As a normal condition the external load due to lifting, stacking of the package and filling the space of disposal pit with mortar or something were considered. On the other hand, drop incident during handling and pressure due to some external force were taken up as hypothetical conditions. Using finite element code ABAQUS and three dimensional finite element model, structural analyses were carried out for the normal conditions. The results show that the maximum stresses occurred at the package due to the loads above mentioned were far less than the yield strength for all conditions. Therefore, it is confirmed that the disposal package keeps its integrity under the normal conditions. Analyses for load cases of 9 m drop onto the reinforced concrete slab and 5.9 m drop onto the embedded disposal package were performed by using finite element code LS-DYNA. Both results show that the strains at the impact zone of the package exceeded the fracture strain of the material but the damaged area was limited in the vicinity of impact zone. As a maximum external pressure, 4MPa was applied to the surface of the packages which were piled up in four layered in the disposal tunnel. According to the results of analyses by ABAQUS code the maximum strain occurred at the contact surfaces close to the welding zone between lid and body of the top package. However, the package stays in sound because the value of the maximum strain was less than the fracture strain of the materials. (author)

treeman: an R package for efficient and intuitive manipulation of phylogenetic trees.

Science.gov (United States)

Bennett, Dominic J; Sutton, Mark D; Turvey, Samuel T

2017-01-07

Phylogenetic trees are hierarchical structures used for representing the inter-relationships between biological entities. They are the most common tool for representing evolution and are essential to a range of fields across the life sciences. The manipulation of phylogenetic trees-in terms of adding or removing tips-is often performed by researchers not just for reasons of management but also for performing simulations in order to understand the processes of evolution. Despite this, the most common programming language among biologists, R, has few class structures well suited to these tasks. We present an R package that contains a new class, called TreeMan, for representing the phylogenetic tree. This class has a list structure allowing phylogenetic trees to be manipulated more efficiently. Computational running times are reduced because of the ready ability to vectorise and parallelise methods. Development is also improved due to fewer lines of code being required for performing manipulation processes. We present three use cases-pinning missing taxa to a supertree, simulating evolution with a tree-growth model and detecting significant phylogenetic turnover-that demonstrate the new package's speed and simplicity.

DANTSYS: A diffusion accelerated neutral particle transport code system

Energy Technology Data Exchange (ETDEWEB)

Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Marr, D.R.; O`Dell, R.D.; Walters, W.F.

1995-06-01

The DANTSYS code package includes the following transport codes: ONEDANT, TWODANT, TWODANT/GQ, TWOHEX, and THREEDANT. The DANTSYS code package is a modular computer program package designed to solve the time-independent, multigroup discrete ordinates form of the boltzmann transport equation in several different geometries. The modular construction of the package separates the input processing, the transport equation solving, and the post processing (or edit) functions into distinct code modules: the Input Module, one or more Solver Modules, and the Edit Module, respectively. The Input and Edit Modules are very general in nature and are common to all the Solver Modules. The ONEDANT Solver Module contains a one-dimensional (slab, cylinder, and sphere), time-independent transport equation solver using the standard diamond-differencing method for space/angle discretization. Also included in the package are solver Modules named TWODANT, TWODANT/GQ, THREEDANT, and TWOHEX. The TWODANT Solver Module solves the time-independent two-dimensional transport equation using the diamond-differencing method for space/angle discretization. The authors have also introduced an adaptive weighted diamond differencing (AWDD) method for the spatial and angular discretization into TWODANT as an option. The TWOHEX Solver Module solves the time-independent two-dimensional transport equation on an equilateral triangle spatial mesh. The THREEDANT Solver Module solves the time independent, three-dimensional transport equation for XYZ and RZ{Theta} symmetries using both diamond differencing with set-to-zero fixup and the AWDD method. The TWODANT/GQ Solver Module solves the 2-D transport equation in XY and RZ symmetries using a spatial mesh of arbitrary quadrilaterals. The spatial differencing method is based upon the diamond differencing method with set-to-zero fixup with changes to accommodate the generalized spatial meshing.

DANTSYS: A diffusion accelerated neutral particle transport code system

International Nuclear Information System (INIS)

Alcouffe, R.E.; Baker, R.S.; Brinkley, F.W.; Marr, D.R.; O'Dell, R.D.; Walters, W.F.

1995-06-01

The DANTSYS code package includes the following transport codes: ONEDANT, TWODANT, TWODANT/GQ, TWOHEX, and THREEDANT. The DANTSYS code package is a modular computer program package designed to solve the time-independent, multigroup discrete ordinates form of the boltzmann transport equation in several different geometries. The modular construction of the package separates the input processing, the transport equation solving, and the post processing (or edit) functions into distinct code modules: the Input Module, one or more Solver Modules, and the Edit Module, respectively. The Input and Edit Modules are very general in nature and are common to all the Solver Modules. The ONEDANT Solver Module contains a one-dimensional (slab, cylinder, and sphere), time-independent transport equation solver using the standard diamond-differencing method for space/angle discretization. Also included in the package are solver Modules named TWODANT, TWODANT/GQ, THREEDANT, and TWOHEX. The TWODANT Solver Module solves the time-independent two-dimensional transport equation using the diamond-differencing method for space/angle discretization. The authors have also introduced an adaptive weighted diamond differencing (AWDD) method for the spatial and angular discretization into TWODANT as an option. The TWOHEX Solver Module solves the time-independent two-dimensional transport equation on an equilateral triangle spatial mesh. The THREEDANT Solver Module solves the time independent, three-dimensional transport equation for XYZ and RZΘ symmetries using both diamond differencing with set-to-zero fixup and the AWDD method. The TWODANT/GQ Solver Module solves the 2-D transport equation in XY and RZ symmetries using a spatial mesh of arbitrary quadrilaterals. The spatial differencing method is based upon the diamond differencing method with set-to-zero fixup with changes to accommodate the generalized spatial meshing

Benchmark of the CASMO-3G/MICROBURN-B codes for Commonwealth Edison boiling water reactors

International Nuclear Information System (INIS)

Wheeler, J.K.; Pallotta, A.S.

1992-01-01

The Commonwealth Edison Company has performed an extensive benchmark against measured data from three boiling water reactors using the Studsvik lattice physics code CASMO-3G and the Siemens Nuclear Power three-dimensional simulator code MICROBURN-B. The measured data of interest for this benchmark are the hot and cold reactivity, and the core power distributions as measured by the traversing incore probe system and gamma scan data for fuel pins and assemblies. A total of nineteen unit-cycles were evaluated. The database included fuel product lines manufactured by General Electric and Siemens Nuclear Power, wit assemblies containing 7 x 7 to 9 x 9 pin configurations, several water rod designs, various enrichments and gadolina loadings, and axially varying lattice designs throughout the enriched portion of the bundle. The results of the benchmark present evidence that the CASMO-3G/MICROBURN-B code package can adequately model the range of fuel and core types in the benchmark, and the codes are acceptable for performing neutronic analyses of Commonwealth Edison's boiling water reactors

ZEST: A Fast Code for Simulating Zeeman-Stark Line-Shape Functions

Directory of Open Access Journals (Sweden)

Franck Gilleron

2018-03-01

Full Text Available We present the ZEST code, dedicated to the calculation of line shapes broadened by Zeeman and Stark effects. As concerns the Stark effect, the model is based on the Standard Lineshape Theory in which ions are treated in the quasi-static approximation, whereas the effects of electrons are represented by weak collisions in the framework of a binary collision relaxation theory. A static magnetic field may be taken into account in the radiator Hamiltonian in the dipole approximation, which leads to additional Zeeman splitting patterns. Ion dynamics effects are implemented using the fast Frequency-Fluctuation Model. For fast calculations, the static ion microfield distribution in the plasma is evaluated using analytic fits of Monte-Carlo simulations, which depend only on the ion-ion coupling parameter and the electron-ion screening factor.

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

Mechanisms of human immunodeficiency virus type 2 RNA packaging

DEFF Research Database (Denmark)

Ni, Na; Nikolaitchik, Olga A; Dilley, Kari A

2011-01-01

do not support the cis-packaging hypothesis but instead indicate that trans packaging is the major mechanism of HIV-2 RNA packaging. To further characterize the mechanisms of HIV-2 RNA packaging, we visualized HIV-2 RNA in individual particles by using fluorescent protein-tagged RNA-binding proteins......Human immunodeficiency virus type 2 (HIV-2) has been reported to have a distinct RNA packaging mechanism, referred to as cis packaging, in which Gag proteins package the RNA from which they were translated. We examined the progeny generated from dually infected cell lines that contain two HIV-2...... proviruses, one with a wild-type gag/gag-pol and the other with a mutant gag that cannot express functional Gag/Gag-Pol. Viral titers and RNA analyses revealed that mutant viral RNAs can be packaged at efficiencies comparable to that of viral RNA from which wild-type Gag/Gag-Pol is translated. These results...

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

International Nuclear Information System (INIS)

Deng Li

1990-01-01

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

Potential benefits of a CAD package for designing multivariable control systems

International Nuclear Information System (INIS)

Mensah, S.

1983-01-01

An open-ended CAD package, MVPACK, has been developed and implemented on a PDP-11/45 minicomputer at the Chalk River Nuclear Laboratories. The package is fully interactive, and includes a comprehensive state-of-the-art methematical library to support development of complex multi-variable control algorithms. Coded in RATFOR, MVPACK operates with a flexible data structure which makes efficient use of minicomputer resources and provides a standard framework for program generation. The existence of a help mechanism enhances the simplicity of package utilization. The capability of the package as a tool for designing control systems is illustrated with the design of a regulating system for an evaporator

The appearance of beam lines

International Nuclear Information System (INIS)

Carey, D.C.

1993-05-01

The combination of an existing graphics package with a large program like TRANSPORT has often resulted in considerable modification to the large program. Use of other graphics package has resulted in essentially having to repeat the work. This difficulty has been avoided in a modification of TRANSPORT which produce layouts of beam lines. Drawings of the reference trajectory and three-dimensional images of all magnets are made by the graphics package TOP DRAWER. Nothing specific to TOP DRAWER or any other graphics has been incorporated into TRANSPORT. If a user is with a different graphics package he or she can then begin usage of this alternate package essentially immediately

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.

The GeoSteiner software package for computing Steiner trees in the plane

DEFF Research Database (Denmark)

Juhl, Daniel; Warme, David M.; Winter, Pawel

The GeoSteiner software package has for more than 10 years been the fastest (publicly available) program for computing exact solutions to Steiner tree problems in the plane. The computational study by Warme, Winter and Zachariasen, published in 2000, documented the performance of the GeoSteiner...... approach --- allowing the exact solution of Steiner tree problems with more than a thousand terminals. Since then, a number of algorithmic enhancements have improved the performance of the software package significantly. In this computational study we run the current code on the largest problem instances...... from the 2000-study, and on a number of larger problem instances. The computational study is performed using both the publicly available GeoSteiner 3.1 code base, and the commercial GeoSteiner 4.0 code base....

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

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

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

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

A new coupling of the 3D thermal-hydraulic code THYC and the thermo-mechanical code CYRANO3 for PWR calculations

Energy Technology Data Exchange (ETDEWEB)

Marguet, S.D. [Electricite de France (EDF), 92 - Clamart (France)

1997-12-31

Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF`s neutronic code COCCINELLE uses the Rowland`s formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission`s products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on `low` configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.) 7 refs.

A new coupling of the 3D thermal-hydraulic code THYC and the thermo-mechanical code CYRANO3 for PWR calculations

International Nuclear Information System (INIS)

Marguet, S.D.

1997-01-01

Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF's neutronic code COCCINELLE uses the Rowland's formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission's products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on 'low' configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.)

TRIGLAV-W a Windows computer program package with graphical users interface for TRIGA reactor core management calculations

International Nuclear Information System (INIS)

Zagar, T.; Zefran, B.; Slavic, S.; Snoj, L.; Ravnik, M.

2006-01-01

TRIGLAV-W is a program package for reactor calculations of TRIGA Mark II research reactor cores. This program package runs under Microsoft Windows operating system and has new friendly graphical user interface (GUI). The main part of the package is the TRIGLAV code based on two dimensional diffusion approximation for flux distribution calculation. The new GUI helps the user to prepare the input files, runs the main code and displays the output files. TRIGLAV-W has a user friendly GUI also for the visualisation of the calculation results. Calculation results can be visualised using 2D and 3D coloured graphs for easy presentations and analysis. In the paper the many options of the new GUI are presented along with the results of extensive testing of the program. The results of the TRIGLAV-W program package were compared with the results of WIMS-D and MCNP code for calculations of TRIGA benchmark. TRIGLAV-W program was also tested using several libraries developed under IAEA WIMS-D Library Update Project. Additional literature and application form for TRIGLAV-W program package beta testing can be found at http://www.rcp.ijs.si/triglav/. (author)

Implementation of the KASKAD computer code system for WWER-440 at Kozloduy NPP

International Nuclear Information System (INIS)

Antonov, A.; Georgieva, N.; Spasova, V.

2003-01-01

Since 2002 at Kozloduy NPP - EP1 the code package KASKAD is used for WWER-440 reactor core calculations. The main codes entering this package are: BIPR-7A: 3-D diffusion and core analysis code; PERMAK-A: 2-D fine mesh diffusion code. The burnup calculations were performed for all cycles of the Kozloduy NPP Unit 1, Unit 2, Unit 3 and Unit 4. For the last 4-5 cycles of the Units were calculated control rods worth, critical boron concentration at zero power, reactivity coefficients and linear power. These results were analysed and were compared with experimental data. Some results were given in this paper

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.

Soft-edged magnet models for higher-order beam-optics map codes

International Nuclear Information System (INIS)

Walstrom, P.L.

2004-01-01

Continuously varying surface and volume source-density distributions are used to model magnetic fields inside of cylindrical volumes. From these distributions, a package of subroutines computes on-axis generalized gradients and their derivatives at arbitrary points on the magnet axis for input to the numerical map-generating subroutines of the Lie-algebraic map code Marylie. In the present version of the package, the magnet menu includes: (1) cylindrical current-sheet or radially thick current distributions with either open boundaries or with a surrounding cylindrical boundary with normal field lines (which models high-permeability iron), (2) Halbach-type permanent multipole magnets, either as sheet magnets or as radially thick magnets, (3) modeling of arbitrary fields inside a cylinder by use of a fictitious current sheet. The subroutines provide on-axis gradients and their z derivatives to essentially arbitrary order, although in the present third- and fifth-order Marylie only the zeroth through sixth derivatives are needed. The formalism is especially useful in beam-optics applications, such as magnetic lenses, where realistic treatment of fringe-field effects is needed

Data visualization for ONEDANT and TWODANT discrete ordinates codes

International Nuclear Information System (INIS)

Lee, C.L.

1993-01-01

Effective graphical display of code calculations allow for efficient analysis of results. This is especially true in the case of discrete ordinates transport codes, which can generate thousands of flux or reaction rate data points per calculation. For this reason, a package of portable interface programs called OTTUI (ONEDANT-TWODANT-Tecplot trademark Unix-Based Interface) has been developed at Los Alamos National Laboratory to permit rapid visualization of ONEDANT and TWODANT discrete ordinates results using the graphics package Tecplot. This paper describes the various uses of OTTUI for display of ONEDANT and TWODANT problem geometries and calculational results

BatTool: an R package with GUI for assessing the effect of White-nose syndrome and other take events on Myotis spp. of bats

Science.gov (United States)

Erickson, Richard A.; Thogmartin, Wayne E.; Szymanski, Jennifer A.

2014-01-01

Background: Myotis species of bats such as the Indiana Bat and Little Brown Bat are facing population declines because of White-nose syndrome (WNS). These species also face threats from anthropogenic activities such as wind energy development. Population models may be used to provide insights into threats facing these species. We developed a population model, BatTool, as an R package to help decision makers and natural resource managers examine factors influencing the dynamics of these species. The R package includes two components: 1) a deterministic and stochastic model that are accessible from the command line and 2) a graphical user interface (GUI). Results: BatTool is an R package allowing natural resource managers and decision makers to understand Myotis spp. population dynamics. Through the use of a GUI, the model allows users to understand how WNS and other take events may affect the population. The results are saved both graphically and as data files. Additionally, R-savvy users may access the population functions through the command line and reuse the code as part of future research. This R package could also be used as part of a population dynamics or wildlife management course. Conclusions: BatTool provides access to a Myotis spp. population model. This tool can help natural resource managers and decision makers with the Endangered Species Act deliberations for these species and with issuing take permits as part of regulatory decision making. The tool is available online as part of this publication.

BatTool: an R package with GUI for assessing the effect of White-nose syndrome and other take events on Myotis spp. of bats.

Science.gov (United States)

Erickson, Richard A; Thogmartin, Wayne E; Szymanski, Jennifer A

2014-01-01

Myotis species of bats such as the Indiana Bat and Little Brown Bat are facing population declines because of White-nose syndrome (WNS). These species also face threats from anthropogenic activities such as wind energy development. Population models may be used to provide insights into threats facing these species. We developed a population model, BatTool, as an R package to help decision makers and natural resource managers examine factors influencing the dynamics of these species. The R package includes two components: 1) a deterministic and stochastic model that are accessible from the command line and 2) a graphical user interface (GUI). BatTool is an R package allowing natural resource managers and decision makers to understand Myotis spp. population dynamics. Through the use of a GUI, the model allows users to understand how WNS and other take events may affect the population. The results are saved both graphically and as data files. Additionally, R-savvy users may access the population functions through the command line and reuse the code as part of future research. This R package could also be used as part of a population dynamics or wildlife management course. BatTool provides access to a Myotis spp. population model. This tool can help natural resource managers and decision makers with the Endangered Species Act deliberations for these species and with issuing take permits as part of regulatory decision making. The tool is available online as part of this publication.

Package and Assisted Travel Arrangement

Directory of Open Access Journals (Sweden)

Ivan Tot

2015-01-01

Full Text Available In the ordinary legislative procedure before the European Parliament and the Council, there is a proposal of the European Commission for the adoption of a new directive that would bring the regulation of the contract on organized tours into line with current market development of organized trips. The proposal is intended to regulate the various combinations of travel services that are today offered to passengers, particularly online, which are identical or comparable to the travel services provided in a classic pre-arranged package. The subject of the paper are the provisions of the proposal of the directive which govern the field of application of the proposed directive, in particular the proposed changes regarding the concept of "package" contained in the European Commission proposal and amendments of the European Parliament, as well as the analysis of the proposed new concept of "assisted travel arrangements." The paper also critically refers to the method of targeted maximum harmonization as a proposed new intensity of the harmonization. The conclusion is that, despite the welcome updating of an outdated text of the directive on package travel which is line with the current market needs, the proposed text of the new directive is burdened with technical and complex definitions that could lead to significant difficulties in their transposition into the provisions of national law of the Member States.

Unique features in the ARIES glovebox line

International Nuclear Information System (INIS)

Martinez, H.E.; Brown, W.G.; Flamm, B.; James, C.A.; Laskie, R.; Nelson, T.O.; Wedman, D.E.

1998-01-01

A series of unique features have been incorporated into the Advanced Recovery and Integrated Extraction System (ARIES) at the Los Alamos National Laboratory, TA-55 Plutonium Facility. The features enhance the material handling in the process of the dismantlement of nuclear weapon primaries in the glovebox line. Incorporated into these features are the various plutonium process module's different ventilation zone requirements that the material handling systems must meet. These features include a conveyor system that consists of a remotely controlled cart that transverses the length of the conveyor glovebox, can be operated from a remote location and can deliver process components to the entrance of any selected module glovebox. Within the modules there exists linear motion material handling systems with lifting hoist, which are controlled via an Allen Bradley control panel or local control panels. To remove the packaged products from the hot process line, the package is processed through an air lock/electrolytic decontamination process that removes the radioactive contamination from the outside of the package container and allows the package to be removed from the process line

Swedish analysis of NEA/CSNI benchmark problems for criticality codes

International Nuclear Information System (INIS)

Mennerdahl, D.

1984-08-01

The Monte Carlo methods used by the members of the working group are adequate for calculations on large arrays. The differences in the results from different codes are probably caused by the differences in cross sections. The previous difficulties in obtaining good results for bare arrays of UNH-solution are, at least to some extent, explained by incomplete information. The neutron reflection from walls, ceiling and floor has earlier been neglected. The inclusion of these in the input to the Monte Carlo codes appears to lead to adequate results. The basis for the IAEA rules of calculating allowable numbers of fissile packages mixed with other packages (fissile or not) during transport, does not seem justified. This has been demonstrated for theoretical package designs. It has not been confirmed by the other members of the working group and no conclusion was drawn by the group. It is very likely that a mix of real packages can be found that supports the mentioned theoretical demonstration. (author)

Simulations of the broad line region of NGC 5548 with CLOUDY code: Temperature determination

Directory of Open Access Journals (Sweden)

Ilić D.

2007-01-01

Full Text Available In this paper an analysis of the physical properties of the Broad Line Region (BLR of the active galaxy NGC 5548 is presented. Using the photoionization code CLOUDY and the measurements of Peterson et al. (2002, the physical conditions of the BLR are simulated and the BLR temperature is obtained. This temperature was compared to the temperature estimated with the Boltzmann-Plot (BP method (Popović et al. 2007. It was shown that the measured variability in the BLR temperature could be due to the change in the hydrogen density.

The EQ3/6 software package for geochemical modeling: Current status

International Nuclear Information System (INIS)

Worlery, T.J.; Jackson, K.J.; Bourcier, W.L.; Bruton, C.J.; Viani, B.E.; Knauss, K.G.; Delany, J.M.

1988-07-01

EQ3/6 is a software package for modeling chemical and mineralogic interactions in aqueous geochemical systems. The major components of the package are EQ3NR (a speciation-solubility code), EQ6 (a reaction path code), EQLIB (a supporting library), and a supporting thermodynamic data base. EQ3NR calculates aqueous speciation and saturation indices from analytical data. It can also be used to calculate compositions of buffer solutions for use in laboratory experiments. EQ6 computes reaction path models of both equilibrium step processes and kinetic reaction processes. These models can be computed for closed systems and relatively simple open systems. EQ3/6 is useful in making purely theoretical calculations, in designing, interpreting, and extrapolating laboratory experiments, and in testing and developing submodels and supporting data used in these codes. The thermodynamic data base supports calculations over the range 0-300 degree C. 60 refs., 2 figs

The EQ3/6 software package for geochemical modeling: Current status

Energy Technology Data Exchange (ETDEWEB)

Wolery, T.J.; Jackson, K.J.; Bourcier, W.L.; Bruton, C.J.; Viani, B.E.; Knauss, K.G.; Delany, J.M.

1988-07-01

EQ3/6 is a software package for modeling chemical and mineralogic interactions in aqueous geochemical systems. The major components of the package are EQ3NR (a speciation-solubility code), EQ6 (a reaction path code), EQLIB (a supporting library), and a supporting thermodynamic data base. EQ3NR calculates aqueous speciation and saturation indices from analytical data. It can also be used to calculate compositions of buffer solutions for use in laboratory experiments. EQ6 computes reaction path models of both equilibrium step processes and kinetic reaction processes. These models can be computed for closed systems and relatively simple open systems. EQ3/6 is useful in making purely theoretical calculations, in designing, interpreting, and extrapolating laboratory experiments, and in testing and developing submodels and supporting data used in these codes. The thermodynamic data base supports calculations over the range 0-300{degree}C. 60 refs., 2 figs.

Cine: Line excitation by infrared fluorescence in cometary atmospheres

Science.gov (United States)

de Val-Borro, Miguel; Cordiner, Martin A.; Milam, Stefanie N.; Charnley, Steven B.

2017-03-01

CINE is a Python module for calculating infrared pumping efficiencies that can be applied to the most common molecules found in cometary comae such as water, hydrogen cyanide or methanol. Excitation by solar radiation of vibrational bands followed by radiative decay to the ground vibrational state is one of the main mechanisms for molecular excitation in comets. This code calculates the effective pumping rates for rotational levels in the ground vibrational state scaled by the heliocentric distance of the comet. Line transitions are queried from the latest version of the HITRAN spectroscopic repository using the astroquery affiliated package of astropy. Molecular data are obtained from the LAMDA database. These coefficients are useful for modeling rotational emission lines observed in cometary spectra at sub-millimeter wavelengths. Combined with computational methods to solve the radiative transfer equations based, e.g., on the Monte Carlo algorithm, this model can retrieve production rates and rotational temperatures from the observed emission spectrum.

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.

Coding for dummies

CERN Document Server

Abraham, Nikhil

2015-01-01

Hands-on exercises help you learn to code like a pro No coding experience is required for Coding For Dummies,your one-stop guide to building a foundation of knowledge inwriting computer code for web, application, and softwaredevelopment. It doesn't matter if you've dabbled in coding or neverwritten a line of code, this book guides you through the basics.Using foundational web development languages like HTML, CSS, andJavaScript, it explains in plain English how coding works and whyit's needed. Online exercises developed by Codecademy, a leading online codetraining site, help hone coding skill

AlgoRun: a Docker-based packaging system for platform-agnostic implemented algorithms.

Science.gov (United States)

Hosny, Abdelrahman; Vera-Licona, Paola; Laubenbacher, Reinhard; Favre, Thibauld

2016-08-01

There is a growing need in bioinformatics for easy-to-use software implementations of algorithms that are usable across platforms. At the same time, reproducibility of computational results is critical and often a challenge due to source code changes over time and dependencies. The approach introduced in this paper addresses both of these needs with AlgoRun, a dedicated packaging system for implemented algorithms, using Docker technology. Implemented algorithms, packaged with AlgoRun, can be executed through a user-friendly interface directly from a web browser or via a standardized RESTful web API to allow easy integration into more complex workflows. The packaged algorithm includes the entire software execution environment, thereby eliminating the common problem of software dependencies and the irreproducibility of computations over time. AlgoRun-packaged algorithms can be published on http://algorun.org, a centralized searchable directory to find existing AlgoRun-packaged algorithms. AlgoRun is available at http://algorun.org and the source code under GPL license is available at https://github.com/algorun laubenbacher@uchc.edu Supplementary data are available at Bioinformatics online. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please e-mail: journals.permissions@oup.com.

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.

Performance analysis of conceptual waste package designs in salt repositories

International Nuclear Information System (INIS)

Jansen, G. Jr.; Raines, G.E.; Kircher, J.F.

1984-01-01

A performance analysis of commercial high-level waste and spent fuel conceptual package designs in reference repositories in three salt formations was conducted with the WAPPA waste package code. Expected conditions for temperature, stress, brine composition, radiation level, and brine flow rate were used as boundary conditions to compute expected corrosion of a thick-walled overpack of 1025 wrought steel. In all salt formations corrosion by low Mg salt-dissolution brines typical of intrusion scenarios was too slow to cause the package to fail for thousands of years after burial. In high Mg brines judged typical of thermally migrating brines in bedded salt formations, corrosion rates which would otherwise have caused the packages to fail within a few hundred years were limited by brine availability. All of the brine reaching the package was consumed by reaction with the iron in the overpack, thus preventing further corrosion. Uniform brine distribution over the package surface was an important factor in predicting long package lifetimes for the high Mg brines. 14 references, 15 figures

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

Alpharetroviral self-inactivating vectors produced by a superinfection-resistant stable packaging cell line allow genetic modification of primary human T lymphocytes.

Science.gov (United States)

Labenski, Verena; Suerth, Julia D; Barczak, Elke; Heckl, Dirk; Levy, Camille; Bernadin, Ornellie; Charpentier, Emmanuelle; Williams, David A; Fehse, Boris; Verhoeyen, Els; Schambach, Axel

2016-08-01

Primary human T lymphocytes represent an important cell population for adoptive immunotherapies, including chimeric-antigen and T-cell receptor applications, as they have the capability to eliminate non-self, virus-infected and tumor cells. Given the increasing numbers of clinical immunotherapy applications, the development of an optimal vector platform for genetic T lymphocyte engineering, which allows cost-effective high-quality vector productions, remains a critical goal. Alpharetroviral self-inactivating vectors (ARV) have several advantages compared to other vector platforms, including a more random genomic integration pattern and reduced likelihood for inducing aberrant splicing of integrated proviruses. We developed an ARV platform for the transduction of primary human T lymphocytes. We demonstrated functional transgene transfer using the clinically relevant herpes-simplex-virus thymidine kinase variant TK.007. Proof-of-concept of alpharetroviral-mediated T-lymphocyte engineering was shown in vitro and in a humanized transplantation model in vivo. Furthermore, we established a stable, human alpharetroviral packaging cell line in which we deleted the entry receptor (SLC1A5) for RD114/TR-pseudotyped ARVs to prevent superinfection and enhance genomic integrity of the packaging cell line and viral particles. We showed that superinfection can be entirely prevented, while maintaining high recombinant virus titers. Taken together, this resulted in an improved production platform representing an economic strategy for translating the promising features of ARVs for therapeutic T-lymphocyte engineering. Copyright © 2016 Elsevier Ltd. All rights reserved.

LISA - a powerful program package for LIstmode and Spectral data Analysis

International Nuclear Information System (INIS)

Oberstedt, A.; Hambsch, F.J.

1994-01-01

LISA is a graphical program package which enables both off-line listmode and spectral data evaluation as well as on-line monitoring while multi-parameter experiments are running. It can be executed on every computer with a UNIX operating system and an X-WINDOW environment, running PV-WAVE from Visual Numerics Incorporation. This package is basically written in the language PV-WAVE CL, but integration of procedures written in the C-language and execution of UNIX shell commands lead to an additional increase of performance. (orig.)

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

Structure and operation of the ITS code system

International Nuclear Information System (INIS)

Halbleib, J.

1988-01-01

The TIGER series of time-independent coupled electron-photon Monte Carlo transport codes is a group of multimaterial and multidimensional codes designed to provide a state-of-the-art description of the production and transport of the electron-photon cascade by combining microscopic photon transport with a macroscopic random walk for electron transport. Major contributors to its evolution are listed. The author and his associates are primarily code users rather than code developers, and have borrowed freely from existing work wherever possible. Nevertheless, their efforts have resulted in various software packages for describing the production and transport of the electron-photon cascade that they found sufficiently useful to warrant dissemination through the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory. The ITS system (Integrated TIGER Series) represents the organization and integration of this combined software, along with much additional capability from previously unreleased work, into a single convenient package of exceptional user friendliness and portability. Emphasis is on simplicity and flexibility of application without sacrificing the rigor or sophistication of the physical model

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

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.

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.

Letter report: status on code maintenance (EQ3/6)

International Nuclear Information System (INIS)

Wolery, T.J.

1995-01-01

EQ3/6 is a software package for geochemical modeling of aqueous systems, such as water/rock or waste/water rock. It is being developed for a variety of applications in geochemical studies for the Yucca Mountain Site Characterization Project. Version 7.2a was the first version of this software to be certified for use in quality- affecting work (originally issued for use in non-quality-affecting work only on 12/28/93; certified on S/17/94). In the past year, the Version 7 line software has been maintained while the new Version 8 line has been developed. In this period, sixteen defect reports have been logged and resolved. Corrected software is being released as Version 7.2b. Defect reporting and resolution for the Version 7 line will continue until all released versions in this line are retired, perhaps six months to a year after Version 8.0 is released later this year. The Version 7 software is written in Fortran 77, technically speaking, but incorporates many aspects of older Fortran. The Version 8 software is written in a much more modern Fortran, technically somewhere between Fortran 77 and Fortran 90. Future code maintenance activities will include a more complete move to Fortran 90, as well as continued maintaining of defect reporting and resolution

Evaluation on operation of liquid relief valves for steam line break accidents by RELAP5/CANDU+ code

International Nuclear Information System (INIS)

Yang, C. Y.; Bang, Y. S.; Kim, H. J.

2001-01-01

A development of RELAP5/CANDU+ code for regulatory audits of accident analysis of CANDU nuclear power plants is on progress. This paper is undertaken in a procedure of a verification and validation for RELAP5/CANDU+ code by analyzing main steam line break accidents of WS 2/3/4. Following the ECC injection in sequence of the steam line breaks, the mismatch in heat transfer between the primary and the secondary systems makes pressure of the primary system instantly peaked to the open setpoint of liquid relief valves. The event sequence follows the result of WS 2/3/4 FSAR, but there is a difference in pressure transient after ECC injection. Sensitivity analysis for main factors dependent on the peak pressure such as control logics of liquid relief valves. ECC flow path and feedwater flow is performed. Because the pressure increase is continued for a long time and its peaking is high, open and close of the liquid relief valves are repeated several times, which is obviously different from those of WS 2/3/4 FSAR. As a result, it is evaluated that conservative modeling for the above variables is required in the analysis

The 9th international symposium on the packaging and transportation of radioactive materials

Energy Technology Data Exchange (ETDEWEB)

None

1989-06-01

This three-volume document contains the papers and poster sessions presented at the symposium. Volume 3 contains 87 papers on topics such as structural codes and benchmarking, shipment of plutonium by air, spent fuel shipping, planning, package design and risk assessment, package testing, OCRWN operations experience and regulations. Individual papers were processed separately for the data base. (TEM)

Plato: A localised orbital based density functional theory code

Science.gov (United States)

Kenny, S. D.; Horsfield, A. P.

2009-12-01

The Plato package allows both orthogonal and non-orthogonal tight-binding as well as density functional theory (DFT) calculations to be performed within a single framework. The package also provides extensive tools for analysing the results of simulations as well as a number of tools for creating input files. The code is based upon the ideas first discussed in Sankey and Niklewski (1989) [1] with extensions to allow high-quality DFT calculations to be performed. DFT calculations can utilise either the local density approximation or the generalised gradient approximation. Basis sets from minimal basis through to ones containing multiple radial functions per angular momenta and polarisation functions can be used. Illustrations of how the package has been employed are given along with instructions for its utilisation. Program summaryProgram title: Plato Catalogue identifier: AEFC_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEFC_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 219 974 No. of bytes in distributed program, including test data, etc.: 1 821 493 Distribution format: tar.gz Programming language: C/MPI and PERL Computer: Apple Macintosh, PC, Unix machines Operating system: Unix, Linux and Mac OS X Has the code been vectorised or parallelised?: Yes, up to 256 processors tested RAM: Up to 2 Gbytes per processor Classification: 7.3 External routines: LAPACK, BLAS and optionally ScaLAPACK, BLACS, PBLAS, FFTW Nature of problem: Density functional theory study of electronic structure and total energies of molecules, crystals and surfaces. Solution method: Localised orbital based density functional theory. Restrictions: Tight-binding and density functional theory only, no exact exchange. Unusual features: Both atom centred and uniform meshes available

Food packaging and radiation sterilization

International Nuclear Information System (INIS)

Kawamura, Yoko

1998-01-01

Radiation sterilization has several merits that it is a positively effective sterilization method, it can be used to sterilize low heat-resistant containers and high gas barrier films, and there is no possibility of residual chemicals being left in the packages. It has been commercially used in 'Bag in a Box' and some food containers. The γ ray and an electron beam are commonly used in radiation sterilization. The γ ray can sterilize large size containers and containers with complex shapes or sealed containers due to its strong transmission capability. However, since the equipment tends to be large and expensive, it is generally used in off production lines. On the other hand, it is possible to install and electron beam system on food production lines since the food can be processed in a short time due to its high beam coefficient and its ease of maintenance, even though an electron beam has limited usage such as sterilizing relatively thin materials and surface sterilization due to the weak transmission. A typical sterilization dose is approximately 10-30 kGy. Direct effects impacting packaging materials, particularly plastics, include scission of polymer links, cross-linkage between polymers, and generating radiolysis products such as hydrogen, methane, aliphatic hydrocarbons, etc. Furthermore, under the existence of oxygen, the oxygen radicals generated by the radiation will oxidize and peroxidize polymer chains and will generate alcohol and carbonyl groups, which shear polymer links, and generate oxygen containing low molecular compounds. As a result, degradation of physical strength such as elongation and seal strength, generating foreign odor, and an increase in global migration values shown in an elution test are sometimes evident. The food packages have different shapes, materials, additives, number of microorganisms and purpose. Therefor the effects of radiation, the optimum dose and so on must be investigated on the individual package. (J.P.N.)

Interfacial and Wall Transport Models for SPACE-CAP Code

International Nuclear Information System (INIS)

Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul; Choi, Hoon; Ha, Sang Jun

2009-01-01

The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code

Interfacial and Wall Transport Models for SPACE-CAP Code

Energy Technology Data Exchange (ETDEWEB)

Hong, Soon Joon; Choo, Yeon Joon; Han, Tae Young; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech., Seoul (Korea, Republic of); Choi, Hoon; Ha, Sang Jun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2009-10-15

The development project for the domestic design code was launched to be used for the safety and performance analysis of pressurized light water reactors. And CAP (Containment Analysis Package) code has been also developed for the containment safety and performance analysis side by side with SPACE. The CAP code treats three fields (gas, continuous liquid, and dispersed drop) for the assessment of containment specific phenomena, and is featured by its multidimensional assessment capabilities. Thermal hydraulics solver was already developed and now under testing of its stability and soundness. As a next step, interfacial and wall transport models was setup. In order to develop the best model and correlation package for the CAP code, various models currently used in major containment analysis codes, which are GOTHIC, CONTAIN2.0, and CONTEMPT-LT, have been reviewed. The origins of the selected models used in these codes have also been examined to find out if the models have not conflict with a proprietary right. In addition, a literature survey of the recent studies has been performed in order to incorporate the better models for the CAP code. The models and correlations of SPACE were also reviewed. CAP models and correlations are composed of interfacial heat/mass, and momentum transport models, and wall heat/mass, and momentum transport models. This paper discusses on those transport models in the CAP code.

The equation of state package FEOS for high energy density matter

Science.gov (United States)

Faik, Steffen; Tauschwitz, Anna; Iosilevskiy, Igor

2018-06-01

Adequate equation of state (EOS) data is of high interest in the growing field of high energy density physics and especially essential for hydrodynamic simulation codes. The semi-analytical method used in the newly developed Frankfurt equation of state (FEOS) package provides an easy and fast access to the EOS of - in principle - arbitrary materials. The code is based on the well known QEOS model (More et al., 1988; Young and Corey, 1995) and is a further development of the MPQeos code (Kemp and Meyer-ter Vehn, 1988; Kemp and Meyer-ter Vehn, 1998) from Max-Planck-Institut für Quantenoptik (MPQ) in Garching Germany. The list of features contains the calculation of homogeneous mixtures of chemical elements and the description of the liquid-vapor two-phase region with or without a Maxwell construction. Full flexibility of the package is assured by its structure: A program library provides the EOS with an interface designed for Fortran or C/C++ codes. Two additional software tools allow for the generation of EOS tables in different file output formats and for the calculation and visualization of isolines and Hugoniot shock adiabats. As an example the EOS of fused silica (SiO2) is calculated and compared to experimental data and other EOS codes.

Piping data retrieval system (PDRS): An integrated package to aid piping layout

International Nuclear Information System (INIS)

Vyas, K.N.; Sharma, A.; Susandhi, R.; Basu, S.

1986-01-01

An integrated package to aid piping layout has been developed and implemented on PDP-11/34 system at Hall 7. The package allows various equipments to be modelled, consisting of primitive equipment components. The equipment layout for the plant can then be reproduced in the form of drawings such as plan, elevation, isometric or perspective. The package has the built in function to perform hidden line removal among equipments. Once the equipment layout is finalised, the package aids in superimposing the piping as per the specified pipe routine. The report discusses the general capabilities and the major input requirements for the package. (author)

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

Development of the processing software package for RPV neutron fluence determination methodology

International Nuclear Information System (INIS)

Belousov, S.; Kirilova, K.; Ilieva, K.

2001-01-01

According to the INRNE methodology the neutron transport calculation is carried out by two steps. At the first step reactor core eigenvalue calculation is performed. This calculation is used for determination of the fixed source for the next step calculation of neutron transport from the reactor core to the RPV. Both calculation steps are performed by state of the art and tested codes. The interface software package DOSRC developed at INRNE is used as a link between these two calculations. The package transforms reactor core calculation results to neutron source input data in format appropriate for the neutron transport codes (DORT, TORT and ASYNT) based on the discrete ordinates method. These codes are applied for calculation of the RPV neutron flux and its responses - induced activity, radiation damage, neutron fluence etc. Fore more precise estimation of the neutron fluence, the INRNE methodology has been supplemented by the next improvements: - implementation of more advanced codes (PYTHIA/DERAB) for neutron-physics parameter calculations; - more detailed neutron source presentation; - verification of neutron fluence by statistically treated experimental data. (author)

Impact testing and analysis for structural code benchmarking

International Nuclear Information System (INIS)

Glass, R.E.

1989-01-01

Sandia National Laboratories, in cooperation with industry and other national laboratories, has been benchmarking computer codes (''Structural Code Benchmarking for the Analysis of Impact Response of Nuclear Material Shipping Cask,'' R.E. Glass, Sandia National Laboratories, 1985; ''Sample Problem Manual for Benchmarking of Cask Analysis Codes,'' R.E. Glass, Sandia National Laboratories, 1988; ''Standard Thermal Problem Set for the Evaluation of Heat Transfer Codes Used in the Assessment of Transportation Packages, R.E. Glass, et al., Sandia National Laboratories, 1988) used to predict the structural, thermal, criticality, and shielding behavior of radioactive materials packages. The first step in the benchmarking of the codes was to develop standard problem sets and to compare the results from several codes and users. This step for structural analysis codes has been completed as described in ''Structural Code Benchmarking for the Analysis of Impact Response of Nuclear Material Shipping Casks,'' R.E. Glass, Sandia National Laboratories, 1985. The problem set is shown in Fig. 1. This problem set exercised the ability of the codes to predict the response to end (axisymmetric) and side (plane strain) impacts with both elastic and elastic/plastic materials. The results from these problems showed that there is good agreement in predicting elastic response. Significant differences occurred in predicting strains for the elastic/plastic models. An example of the variation in predicting plastic behavior is given, which shows the hoop strain as a function of time at the impacting end of Model B. These differences in predicting plastic strains demonstrated a need for benchmark data for a cask-like problem. 6 refs., 5 figs

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

Software Package for Optics Measurement and Correction in the LHC

CERN Document Server

Aiba, M; Tomas, R; Vanbavinckhove, G

2010-01-01

A software package has been developed for the LHC on-line optics measurement and correction. This package includes several different algorithms to measure phase advance, beta functions, dispersion, coupling parameters and even some non-linear terms. A Graphical User Interface provides visualization tools to compare measurements to model predictions, fit analytical formula, localize error sources and compute and send corrections to the hardware.

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.)

OECD/NEZ Main Steam Line Break Benchmark Problem Exercise I Simulation Using the SPACE Code with the Point Kinetics Model

International Nuclear Information System (INIS)

Kim, Yohan; Kim, Seyun; Ha, Sangjun

2014-01-01

The Safety and Performance Analysis Code for Nuclear Power Plants (SPACE) has been developed in recent years by the Korea Nuclear Hydro and Nuclear Power Co. (KHNP) through collaborative works with other Korean nuclear industries. The SPACE is a best-estimated two-phase three-field thermal-hydraulic analysis code to analyze the safety and performance of pressurized water reactors (PWRs). The SPACE code has sufficient features to replace outdated vendor supplied codes and to be used for the safety analysis of operating PWRs and the design of advanced reactors. As a result of the second phase of the development, the 2.14 version of the code was released through the successive various V and V works. The topical reports on the code and related safety analysis methodologies have been prepared for license works. In this study, the OECD/NEA Main Steam Line Break (MSLB) Benchmark Problem Exercise I was simulated as a V and V work. The results were compared with those of the participants in the benchmark project. The OECD/NEA MSLB Benchmark Problem Exercise I was simulated using the SPACE code. The results were compared with those of the participants in the benchmark project. Through the simulation, it was concluded that the SPACE code can effectively simulate PWR MSLB accidents

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

International Nuclear Information System (INIS)

Liu Zhanquan; Chen Yixue; Yang Chao; Dang Halei

2013-01-01

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

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

MED101: a laser-plasma simulation code. User guide

International Nuclear Information System (INIS)

Rodgers, P.A.; Rose, S.J.; Rogoyski, A.M.

1989-12-01

Complete details for running the 1-D laser-plasma simulation code MED101 are given including: an explanation of the input parameters, instructions for running on the Rutherford Appleton Laboratory IBM, Atlas Centre Cray X-MP and DEC VAX, and information on three new graphics packages. The code, based on the existing MEDUSA code, is capable of simulating a wide range of laser-produced plasma experiments including the calculation of X-ray laser gain. (author)

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.

COOLOD-N2: a computer code, for the analyses of steady-state thermal-hydraulics in research reactors

International Nuclear Information System (INIS)

Kaminaga, Masanori

1994-03-01

The COOLOD-N2 code provides a capability for the analyses of the steady-state thermal-hydraulics of research reactors. This code is revised version of the COOLOD-N code, and is applicable not only for research reactors in which plate-type fuel is adopted, but also for research reactors in which rod-type fuel is adopted. In the code, subroutines to calculate temperature distribution in rod-type fuel have been newly added to the COOLOD-N code. The COOLOD-N2 code can calculate fuel temperatures under both forced convection cooling mode and natural convection cooling mode as well as COOLOD-N code. In the COOLOD-N2 code, a 'Heat Transfer package' is used for calculating heat transfer coefficient, DNB heat flux etc. The 'Heat Transfer package' is subroutine program and is especially developed for research reactors in which plate-type fuel is adopted. In case of rod-type fuel, DNB heat flux is calculated by both the 'Heat Transfer package' and Lund DNB heat flux correlation which is popular for TRIGA reactor. The COOLOD-N2 code also has a capability of calculating ONB temperature, the heat flux at onset of flow instability as well as DNB heat flux. (author)

Simulations of X-ray synchrotron beams using the EGS4 code system in medical applications

International Nuclear Information System (INIS)

Orion, I.; Henn, A.; Sagi, I.; Dilmanian, F.A.; Pena, L.; Rosenfeld, A.B.

2001-01-01

X-ray synchrotron beams are commonly used in biological and medical research. The availability of intense, polarized low-energy photons from the synchrotron beams provides a high dose transfer to biological materials. The EGS4 code system, which includes the photoelectron angular distribution, electron motion inside a magnetic field, and the LSCAT package, found to be the appropriate Monte Carlo code for synchrotron-produced X-ray simulations. The LSCAT package was developed in 1995 for the EGS4 code to contain the routines to simulate the linear polarization, the bound Compton, and the incoherent scattering functions. Three medical applications were demonstrated using the EGS4 Monte Carlo code as a proficient simulation code system for the synchrotron low-energy X-ray source. (orig.)

Effects of Secondary Circuit Modeling on Results of Pressurized Water Reactor Main Steam Line Break Benchmark Calculations with New Coupled Code TRAB-3D/SMABRE

International Nuclear Information System (INIS)

Daavittila, Antti; Haemaelaeinen, Anitta; Kyrki-Rajamaeki, Riitta

2003-01-01

All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finland. For the first exercise, the plant simulation with point-kinetic neutronics, the thermal-hydraulics code SMABRE was used. The second exercise was calculated with the three-dimensional reactor dynamics code TRAB-3D, and the third exercise with the combination TRAB-3D/SMABRE. VTT has over ten years' experience of coupling neutronic and thermal-hydraulic codes, but this benchmark was the first time these two codes, both developed at VTT, were coupled together. The coupled code system is fast and efficient; the total computation time of the 100-s transient in the third exercise was 16 min on a modern UNIX workstation. The results of all the exercises are similar to those of the other participants. In order to demonstrate the effect of secondary circuit modeling on the results, three different cases were calculated. In case 1 there is no phase separation in the steam lines and no flow reversal in the aspirator. In case 2 the flow reversal in the aspirator is allowed, but there is no phase separation in the steam lines. Finally, in case 3 the drift-flux model is used for the phase separation in the steam lines, but the aspirator flow reversal is not allowed. With these two modeling variations, it is possible to cover a remarkably broad range of results. The maximum power level reached after the reactor trip varies from 534 to 904 MW, the range of the time of the power maximum being close to 30 s. Compared to the total calculated transient time of 100 s, the effect of the secondary side modeling is extremely important

Fielding of the on-axis diagnostic package at Z

International Nuclear Information System (INIS)

Hurst, M.J.; Nash, T.J.; Derzon, M.; Kellogg, J.W.; Torres, J.; McGurn, J.; Seaman, J.; Jobe, D.; Lazier, S.E.

1998-06-01

The authors have developed a comprehensive diagnostic package for observing z-pinch radiation along the pinch axis on the Z accelerator. The instrumentation, which was fielded on the axial package, are x-ray diagnostics requiring direct lines of sight to the target. The diagnostics require vacuum access to the center of the accelerator. The environment is a hostile one, where one must deal with an intense, energetic photon flux (>100 keV), EMP, debris (e.g. bullets or shrapnel), and mechanical shock in order for the diagnostics to survive. In addition, practical constraints require the package be refurbished and utilized on a once a day shot schedule. In spite of this harsh environment, the authors have successfully fielded the diagnostic package with a high survivability of the data and the instruments. In this paper, they describe the environment and issues related to the re-entrant diagnostic package's implementation and maintenance

On-Line Generation of 3D-Waves

DEFF Research Database (Denmark)

Frigaard, Peter

1992-01-01

The paper describes the technique of filtering white noise for on-line generation of 3D-waves on a small computer in the laboratory. The wave generation package is implemented and tested in the 3D-wave basin at the University of Aalborg.......The paper describes the technique of filtering white noise for on-line generation of 3D-waves on a small computer in the laboratory. The wave generation package is implemented and tested in the 3D-wave basin at the University of Aalborg....

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.

HB-Line Plutonium Oxide Data Collection Strategy

Energy Technology Data Exchange (ETDEWEB)

Watkins, R. [Savannah River Nuclear Solutions; Varble, J. [Savannah River Nuclear Solutions; Jordan, J. [Savannah River Nuclear Solutions

2015-05-26

HB-Line and H-Canyon will handle and process plutonium material to produce plutonium oxide for feed to the Mixed Oxide Fuel Fabrication Facility (MFFF). However, the plutonium oxide product will not be transferred to the MFFF directly from HB-Line until it is packaged into a qualified DOE-STD-3013-2012 container. In the interim, HB-Line will load plutonium oxide into an inner, filtered can. The inner can will be placed in a filtered bag, which will be loaded into a filtered outer can. The outer can will be loaded into a certified 9975 with getter assembly in compliance with onsite transportation requirement, for subsequent storage and transfer to the K-Area Complex (KAC). After DOE-STD-3013-2012 container packaging capabilities are established, the product will be returned to HB-Line to be packaged into a qualified DOE-STD-3013-2012 container. To support the transfer of plutonium oxide to KAC and then eventually to MFFF, various material and packaging data will have to be collected and retained. In addition, data from initial HB-Line processing operations will be needed to support future DOE-STD-3013-2012 qualification as amended by the HB-Line DOE Standard equivalency. As production increases, the volume of data to collect will increase. The HB-Line data collected will be in the form of paper copies and electronic media. Paper copy data will, at a minimum, consist of facility procedures, nonconformance reports (NCRs), and DCS print outs. Electronic data will be in the form of Adobe portable document formats (PDFs). Collecting all the required data for each plutonium oxide can will be no small effort for HB-Line, and will become more challenging once the maximum annual oxide production throughput is achieved due to the sheer volume of data to be collected. The majority of the data collected will be in the form of facility procedures, DCS print outs, and laboratory results. To facilitate complete collection of this data, a traveler form will be developed which

Browndye: A software package for Brownian dynamics

Science.gov (United States)

Huber, Gary A.; McCammon, J. Andrew

2010-11-01

A new software package, Browndye, is presented for simulating the diffusional encounter of two large biological molecules. It can be used to estimate second-order rate constants and encounter probabilities, and to explore reaction trajectories. Browndye builds upon previous knowledge and algorithms from software packages such as UHBD, SDA, and Macrodox, while implementing algorithms that scale to larger systems. Program summaryProgram title: Browndye Catalogue identifier: AEGT_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEGT_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: MIT license, included in distribution No. of lines in distributed program, including test data, etc.: 143 618 No. of bytes in distributed program, including test data, etc.: 1 067 861 Distribution format: tar.gz Programming language: C++, OCaml ( http://caml.inria.fr/) Computer: PC, Workstation, Cluster Operating system: Linux Has the code been vectorised or parallelized?: Yes. Runs on multiple processors with shared memory using pthreads RAM: Depends linearly on size of physical system Classification: 3 External routines: uses the output of APBS [1] ( http://www.poissonboltzmann.org/apbs/) as input. APBS must be obtained and installed separately. Expat 2.0.1, CLAPACK, ocaml-expat, Mersenne Twister. These are included in the Browndye distribution. Nature of problem: Exploration and determination of rate constants of bimolecular interactions involving large biological molecules. Solution method: Brownian dynamics with electrostatic, excluded volume, van der Waals, and desolvation forces. Running time: Depends linearly on size of physical system and quadratically on precision of results. The included example executes in a few minutes.

The spectral code Apollo2: from lattice to 2D core calculations

International Nuclear Information System (INIS)

Coste-Delclaux, M.; Santandrea, S.; Damian, F.; Blanc-Tranchant, P.; Zmijarevic, I.; Santamarina, A.

2005-01-01

Apollo2 is a powerful code dedicated to neutron transport, it is a highly qualified tool for a wide range of applications from research and development studies to industrial applications. Today Apollo2 is part of several advanced 3-dimensional nuclear code packages dedicated to reactor physics, fuel cycle, criticality and safety analysis. The presentations have been organized into 7 topics: -) an introduction to Apollo2, -) cross-sections, -) flux calculation, -) advanced applications, -) Apollo2 users, specialized packages, -) qualification program, and -) the future of Apollo2. This document gathers only the slides of the presentations

The spectral code Apollo2: from lattice to 2D core calculations

Energy Technology Data Exchange (ETDEWEB)

Coste-Delclaux, M.; Santandrea, S.; Damian, F.; Blanc-Tranchant, P.; Zmijarevic, I. [CEA Saclay (DEN/DANS/SERMA), 91 - Gif-sur-Yvette (France); Santamarina, A. [CEA Cadarache (CEA/DEN/DER/SPRC), 13 - Saint Paul lez Durance (France)

2005-07-01

Apollo2 is a powerful code dedicated to neutron transport, it is a highly qualified tool for a wide range of applications from research and development studies to industrial applications. Today Apollo2 is part of several advanced 3-dimensional nuclear code packages dedicated to reactor physics, fuel cycle, criticality and safety analysis. The presentations have been organized into 7 topics: -) an introduction to Apollo2, -) cross-sections, -) flux calculation, -) advanced applications, -) Apollo2 users, specialized packages, -) qualification program, and -) the future of Apollo2. This document gathers only the slides of the presentations.

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Coded Modulation in C and MATLAB

Science.gov (United States)

Hamkins, Jon; Andrews, Kenneth S.

2011-01-01

This software, written separately in C and MATLAB as stand-alone packages with equivalent functionality, implements encoders and decoders for a set of nine error-correcting codes and modulators and demodulators for five modulation types. The software can be used as a single program to simulate the performance of such coded modulation. The error-correcting codes implemented are the nine accumulate repeat-4 jagged accumulate (AR4JA) low-density parity-check (LDPC) codes, which have been approved for international standardization by the Consultative Committee for Space Data Systems, and which are scheduled to fly on a series of NASA missions in the Constellation Program. The software implements the encoder and decoder functions, and contains compressed versions of generator and parity-check matrices used in these operations.

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

The HELIOS-2 lattice physics code

International Nuclear Information System (INIS)

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

2008-01-01

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

Antimicrobial food packaging: potential and pitfalls

Science.gov (United States)

Malhotra, Bhanu; Keshwani, Anu; Kharkwal, Harsha

2015-01-01

Nowadays food preservation, quality maintenance, and safety are major growing concerns of the food industry. It is evident that over time consumers’ demand for natural and safe food products with stringent regulations to prevent food-borne infectious diseases. Antimicrobial packaging which is thought to be a subset of active packaging and controlled release packaging is one such promising technology which effectively impregnates the antimicrobial into the food packaging film material and subsequently delivers it over the stipulated period of time to kill the pathogenic microorganisms affecting food products thereby increasing the shelf life to severe folds. This paper presents a picture of the recent research on antimicrobial agents that are aimed at enhancing and improving food quality and safety by reduction of pathogen growth and extension of shelf life, in a form of a comprehensive review. Examination of the available antimicrobial packaging technologies is also presented along with their significant impact on food safety. This article entails various antimicrobial agents for commercial applications, as well as the difference between the use of antimicrobials under laboratory scale and real time applications. Development of resistance amongst microorganisms is considered as a future implication of antimicrobials with an aim to come up with actual efficacies in extension of shelf life as well as reduction in bacterial growth through the upcoming and promising use of antimicrobials in food packaging for the forthcoming research down the line. PMID:26136740

CH-TRU Waste Content Codes

Energy Technology Data Exchange (ETDEWEB)

Washington TRU Solutions LLC

2008-01-16

The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

Functional Coding Variation in Recombinant Inbred Mouse Lines Reveals Novel Serotonin Transporter-Associated Phenotypes

Energy Technology Data Exchange (ETDEWEB)

Carneiro, Ana [Vanderbilt University; Airey, David [University of Tennessee Health Science Center, Memphis; Thompson, Brent [Vanderbilt University; Zhu, C [Vanderbilt University; Rinchik, Eugene M [ORNL; Lu, Lu [University of Tennessee Health Science Center, Memphis; Chesler, Elissa J [ORNL; Erikson, Keith [University of North Carolina; Blakely, Randy [Vanderbilt University

2009-01-01

The human serotonin (5-hydroxytryptamine, 5-HT) transporter (hSERT, SLC6A4) figures prominently in the etiology or treatment of many prevalent neurobehavioral disorders including anxiety, alcoholism, depression, autism and obsessive-compulsive disorder (OCD). Here we utilize naturally occurring polymorphisms in recombinant inbred (RI) lines to identify novel phenotypes associated with altered SERT function. The widely used mouse strain C57BL/6J, harbors a SERT haplotype defined by two nonsynonymous coding variants (Gly39 and Lys152 (GK)). At these positions, many other mouse lines, including DBA/2J, encode Glu39 and Arg152 (ER haplotype), assignments found also in hSERT. Synaptosomal 5-HT transport studies revealed reduced uptake associated with the GK variant. Heterologous expression studies confirmed a reduced SERT turnover rate for the GK variant. Experimental and in silico approaches using RI lines (C57Bl/6J X DBA/2J=BXD) identifies multiple anatomical, biochemical and behavioral phenotypes specifically impacted by GK/ER variation. Among our findings are multiple traits associated with anxiety and alcohol consumption, as well as of the control of dopamine (DA) signaling. Further bioinformatic analysis of BXD phenotypes, combined with biochemical evaluation of SERT knockout mice, nominates SERT-dependent 5-HT signaling as a major determinant of midbrain iron homeostasis that, in turn, dictates ironregulated DA phenotypes. Our studies provide a novel example of the power of coordinated in vitro, in vivo and in silico approaches using murine RI lines to elucidate and quantify the system-level impact of gene variation.

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.

A numerical simulation package for analysis of neutronics and thermal fluids of space nuclear power and propulsion systems

International Nuclear Information System (INIS)

Anghaie, S.; Feller, G.J.; Peery, S.D.; Parsley, R.C.

1993-01-01

A system of computer codes for engineering simulation and in-depth analysis of nuclear and thermal fluid design of nuclear thermal rockets is developed. The computational system includes a neutronic solver package, a thermal fluid solver package and a propellant and materials property package. The Rocket Engine Transient Simulation (ROCETS) system code is incorporated with computational modules specific to nuclear powered engines. ROCETS features a component based performance architecture that interfaces component modules into the user designed configuration, interprets user commands, creates an executable FORTRAN computer program, and executes the program to provide output to the user. Basic design features of the Pratt ampersand Whitney XNR2000 nuclear rocket concept and its operational performance are analyzed and simulated

TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES

International Nuclear Information System (INIS)

DOE

1997-01-01

A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k eff , of a spent nuclear fuel package. Fifty-seven UO 2 , UO 2 /Gd 2 O 3 , and UO 2 /PuO 2 critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k eff (which can be a function of the trending parameters) such that the biased k eff , when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection

MAGNETOHYDRODYNAMIC EQUATIONS (MHD GENERATION CODE

Directory of Open Access Journals (Sweden)

Francisco Frutos Alfaro

2017-04-01

Full Text Available A program to generate codes in Fortran and C of the full magnetohydrodynamic equations is shown. The program uses the free computer algebra system software REDUCE. This software has a package called EXCALC, which is an exterior calculus program. The advantage of this program is that it can be modified to include another complex metric or spacetime. The output of this program is modified by means of a LINUX script which creates a new REDUCE program to manipulate the magnetohydrodynamic equations to obtain a code that can be used as a seed for a magnetohydrodynamic code for numerical applications. As an example, we present part of the output of our programs for Cartesian coordinates and how to do the discretization.

Testing efficiency transfer codes for equivalence

International Nuclear Information System (INIS)

Vidmar, T.; Celik, N.; Cornejo Diaz, N.; Dlabac, A.; Ewa, I.O.B.; Carrazana Gonzalez, J.A.; Hult, M.; Jovanovic, S.; Lepy, M.-C.; Mihaljevic, N.; Sima, O.; Tzika, F.; Jurado Vargas, M.; Vasilopoulou, T.; Vidmar, G.

2010-01-01

Four general Monte Carlo codes (GEANT3, PENELOPE, MCNP and EGS4) and five dedicated packages for efficiency determination in gamma-ray spectrometry (ANGLE, DETEFF, GESPECOR, ETNA and EFFTRAN) were checked for equivalence by applying them to the calculation of efficiency transfer (ET) factors for a set of well-defined sample parameters, detector parameters and energies typically encountered in environmental radioactivity measurements. The differences between the results of the different codes never exceeded a few percent and were lower than 2% in the majority of cases.

Contributions to the validation of the ASTEC V1 code

International Nuclear Information System (INIS)

Constantin, Marin; Rizoiu, Andrei; Turcu, Ilie

2004-01-01

In the frame of PHEBEN2 project (Validation of the severe accidents codes for applications to nuclear power plants, based on the PHEBUS FP experiments), a project developed within the EU research Frame Program 5 (FP5), the INR-Pitesti's team has received the task of determining the ASTEC code sensitivity. The PHEBEN2 project has been initiated in 1998 and gathered 13 partners from 6 EU member states. To the project 4 partners from 3 candidate states (Hungary, Bulgaria and Romania) joined later. The works were contracted with the European Commission (under FIKS-CT1999-00009 contract) that supports financially the research effort up to about 50%. According to the contract provisions, INR's team participated in developing the Working Package 1 (WP1) which refers to validation of the integral computation codes that use the PHOEBUS experimental data and the Working Package 3 (WP3) referring to the evaluation of the codes to be applied in nuclear power plants for risk evaluation, nuclear safety margin evaluation and determination/evaluation of the measures to be adopted in case of severe accident. The present work continues the efforts to validate preliminarily the ASTEC code. Focused are the the stand-alone sensitivity analyses applied to two most important modules of the code, namely DIVA and SOPHAEROS

Status report on the 'Merging' of the Electron-Cloud Code POSINST with the 3-D Accelerator PIC CODE WARP

International Nuclear Information System (INIS)

Vay, J.-L.; Furman, M.A.; Azevedo, A.W.; Cohen, R.H.; Friedman, A.; Grote, D.P.; Stoltz, P.H.

2004-01-01

We have integrated the electron-cloud code POSINST [1] with WARP [2]--a 3-D parallel Particle-In-Cell accelerator code developed for Heavy Ion Inertial Fusion--so that the two can interoperate. Both codes are run in the same process, communicate through a Python interpreter (already used in WARP), and share certain key arrays (so far, particle positions and velocities). Currently, POSINST provides primary and secondary sources of electrons, beam bunch kicks, a particle mover, and diagnostics. WARP provides the field solvers and diagnostics. Secondary emission routines are provided by the Tech-X package CMEE

MODEL 9977 B(M)F-96 SAFETY ANALYSIS REPORT FOR PACKAGING

Energy Technology Data Exchange (ETDEWEB)

Abramczyk, G; Paul Blanton, P; Kurt Eberl, K

2006-05-18

This Safety Analysis Report for Packaging (SARP) documents the analysis and testing performed on and for the 9977 Shipping Package, referred to as the General Purpose Fissile Package (GPFP). The performance evaluation presented in this SARP documents the compliance of the 9977 package with the regulatory safety requirements for Type B packages. Per 10 CFR 71.59, for the 9977 packages evaluated in this SARP, the value of ''N'' is 50, and the Transport Index based on nuclear criticality control is 1.0. The 9977 package is designed with a high degree of single containment. The 9977 complies with 10 CFR 71 (2002), Department of Energy (DOE) Order 460.1B, DOE Order 460.2, and 10 CFR 20 (2003) for As Low As Reasonably Achievable (ALARA) principles. The 9977 also satisfies the requirements of the Regulations for the Safe Transport of Radioactive Material--1996 Edition (Revised)--Requirements. IAEA Safety Standards, Safety Series No. TS-R-1 (ST-1, Rev.), International Atomic Energy Agency, Vienna, Austria (2000). The 9977 package is designed, analyzed and fabricated in accordance with Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, 1992 edition.

A new package: MySAS for small angle scattering data analysis

International Nuclear Information System (INIS)

Huang Chaoqiang; Xia Qingzhong; Yan Guanyun; Sun Guang'ai; Chen Bo

2010-01-01

In this paper, A MySAS package, which is verified on Windows XP, can easily convert two-dimensional data in small angle neutron and X-ray scattering analysis, operate individually and execute one particular operation as numerical data reduction or analysis, and graphical visualization. This MySAS package can implement the input and output routines via scanning certain properties, thus recalling completely sets of repetition input and selecting the input files. On starting from the two-dimensional files, the MySAS package can correct the anisotropic or isotropic data for physical interpretation and select the relevant pixels. Over 50 model functions are fitted by the POWELL code using χ 2 as the figure of merit function. (authors)

IMAP: A complete Ion Micro-Analysis Package for the nuclear microprobe

International Nuclear Information System (INIS)

Antolak, A.J.; Hildner, M.L.; Morse, D.H.; Bench, G.S.

1993-01-01

Microprobe techniques using scanned, focused MeV ions are routinely used in Livermore for materials characterization. Comprehensive data analysis with these techniques is accomplished with the computer software package IMAP, for Ion Micro-Analysis Package. IMAP consists of a set of command language procedures for data processing and quantitative spectral analysis. Deconvolution of the data is achieved by spawning sub-processes within IMAP which execute analysis codes for each specific microprobe technique. IMAP is structured to rapidly analyze individual spectra or multi-dimensional data blocks which classify individual events by the two scanning dimensions, the energy of the detected radiation and, when necessary, one sample rotation dimension. Several examples are presented to demonstrate the utility of the package

Computer-assisted Particle-in-Cell code development

International Nuclear Information System (INIS)

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

1997-12-01

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

SURE: a system of computer codes for performing sensitivity/uncertainty analyses with the RELAP code

International Nuclear Information System (INIS)

Bjerke, M.A.

1983-02-01

A package of computer codes has been developed to perform a nonlinear uncertainty analysis on transient thermal-hydraulic systems which are modeled with the RELAP computer code. Using an uncertainty around the analyses of experiments in the PWR-BDHT Separate Effects Program at Oak Ridge National Laboratory. The use of FORTRAN programs running interactively on the PDP-10 computer has made the system very easy to use and provided great flexibility in the choice of processing paths. Several experiments simulating a loss-of-coolant accident in a nuclear reactor have been successfully analyzed. It has been shown that the system can be automated easily to further simplify its use and that the conversion of the entire system to a base code other than RELAP is possible

Simplified analytical solutions for free drops during NCT for radioactive material packagings

International Nuclear Information System (INIS)

Gupta, N.K.

1997-01-01

To ensure structural integrity during normal conditions of transport (NCT), Federal regulations in 10CFR71.71 require that the nuclear material package designs be evaluated for the effects of free drops. The vessel stress acceptance criteria for these drops are given in Regulatory Guide 7.6 and ASME Section III Code. During initial phases of the package design, the effects of the NCT free drops can be evaluated by simplified analytical solutions which will ensure that the safety margins specified in R. G. 7.6 are met. These safety margins can be verified during the final stages of the package design with dynamic analyses using finite element methods. This paper calculates the maximum impact open-quotes gclose quotes loading on the vessels using single degree of freedom models for different drop orientations. Only end, bottom, and corner drops are analyzed for cylindrical packages or packages with cylindrical ends

SCALE: A modular code system for performing standardized computer analyses for licensing evaluation

International Nuclear Information System (INIS)

1997-03-01

This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This manual covers an array of modules written for the SCALE package, consisting of drivers, system libraries, cross section and materials properties libraries, input/output routines, storage modules, and help files

SCALE: A modular code system for performing standardized computer analyses for licensing evaluation

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-03-01

This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This manual covers an array of modules written for the SCALE package, consisting of drivers, system libraries, cross section and materials properties libraries, input/output routines, storage modules, and help files.

The Monte Carlo SRNA-VOX code for 3D proton dose distribution in voxelized geometry using CT data

International Nuclear Information System (INIS)

Ilic, Radovan D; Spasic-Jokic, Vesna; Belicev, Petar; Dragovic, Milos

2005-01-01

This paper describes the application of the SRNA Monte Carlo package for proton transport simulations in complex geometry and different material compositions. The SRNA package was developed for 3D dose distribution calculation in proton therapy and dosimetry and it was based on the theory of multiple scattering. The decay of proton induced compound nuclei was simulated by the Russian MSDM model and our own using ICRU 63 data. The developed package consists of two codes: the SRNA-2KG, which simulates proton transport in combinatorial geometry and the SRNA-VOX, which uses the voxelized geometry using the CT data and conversion of the Hounsfield's data to tissue elemental composition. Transition probabilities for both codes are prepared by the SRNADAT code. The simulation of the proton beam characterization by multi-layer Faraday cup, spatial distribution of positron emitters obtained by the SRNA-2KG code and intercomparison of computational codes in radiation dosimetry, indicate immediate application of the Monte Carlo techniques in clinical practice. In this paper, we briefly present the physical model implemented in the SRNA package, the ISTAR proton dose planning software, as well as the results of the numerical experiments with proton beams to obtain 3D dose distribution in the eye and breast tumour

Survey of computer codes applicable to waste facility performance evaluations

International Nuclear Information System (INIS)

Alsharif, M.; Pung, D.L.; Rivera, A.L.; Dole, L.R.

1988-01-01

This study is an effort to review existing information that is useful to develop an integrated model for predicting the performance of a radioactive waste facility. A summary description of 162 computer codes is given. The identified computer programs address the performance of waste packages, waste transport and equilibrium geochemistry, hydrological processes in unsaturated and saturated zones, and general waste facility performance assessment. Some programs also deal with thermal analysis, structural analysis, and special purposes. A number of these computer programs are being used by the US Department of Energy, the US Nuclear Regulatory Commission, and their contractors to analyze various aspects of waste package performance. Fifty-five of these codes were identified as being potentially useful on the analysis of low-level radioactive waste facilities located above the water table. The code summaries include authors, identification data, model types, and pertinent references. 14 refs., 5 tabs

Vector-matrix-quaternion, array and arithmetic packages: All HAL/S functions implemented in Ada

Science.gov (United States)

Klumpp, Allan R.; Kwong, David D.

1986-01-01

The HAL/S avionics programmers have enjoyed a variety of tools built into a language tailored to their special requirements. Ada is designed for a broader group of applications. Rather than providing built-in tools, Ada provides the elements with which users can build their own. Standard avionic packages remain to be developed. These must enable programmers to code in Ada as they have coded in HAL/S. The packages under development at JPL will provide all of the vector-matrix, array, and arithmetic functions described in the HAL/S manuals. In addition, the linear algebra package will provide all of the quaternion functions used in Shuttle steering and Galileo attitude control. Furthermore, using Ada's extensibility, many quaternion functions are being implemented as infix operations; equivalent capabilities were never implemented in HAL/S because doing so would entail modifying the compiler and expanding the language. With these packages, many HAL/S expressions will compile and execute in Ada, unchanged. Others can be converted simply by replacing the implicit HAL/S multiply operator with the Ada *. Errors will be trapped and identified. Input/output will be convenient and readable.

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...

Computer code FIT

International Nuclear Information System (INIS)

Rohmann, D.; Koehler, T.

1987-02-01

This is a description of the computer code FIT, written in FORTRAN-77 for a PDP 11/34. FIT is an interactive program to decude position, width and intensity of lines of X-ray spectra (max. length of 4K channels). The lines (max. 30 lines per fit) may have Gauss- or Voigt-profile, as well as exponential tails. Spectrum and fit can be displayed on a Tektronix terminal. (orig.) [de

Sustainable evolution of product line infrastructure code

OpenAIRE

Patzke, T.

2011-01-01

A major goal in many software development organizations today is to reduce development effort and cost, while improving their products' quality and diversity by developing reusable software. An organization takes advantage of its products' similarities, exploits what they have in common and manages what varies among them by building a product line infrastructure. A product line infrastructure is a reuse repository that contains exactly those common and variable artifacts, such as requirements...

Parallel processing Monte Carlo radiation transport codes

International Nuclear Information System (INIS)

McKinney, G.W.

1994-01-01

Issues related to distributed-memory multiprocessing as applied to Monte Carlo radiation transport are discussed. Measurements of communication overhead are presented for the radiation transport code MCNP which employs the communication software package PVM, and average efficiency curves are provided for a homogeneous virtual machine

graphkernels: R and Python packages for graph comparison.

Science.gov (United States)

Sugiyama, Mahito; Ghisu, M Elisabetta; Llinares-López, Felipe; Borgwardt, Karsten

2018-02-01

Measuring the similarity of graphs is a fundamental step in the analysis of graph-structured data, which is omnipresent in computational biology. Graph kernels have been proposed as a powerful and efficient approach to this problem of graph comparison. Here we provide graphkernels, the first R and Python graph kernel libraries including baseline kernels such as label histogram based kernels, classic graph kernels such as random walk based kernels, and the state-of-the-art Weisfeiler-Lehman graph kernel. The core of all graph kernels is implemented in C ++ for efficiency. Using the kernel matrices computed by the package, we can easily perform tasks such as classification, regression and clustering on graph-structured samples. The R and Python packages including source code are available at https://CRAN.R-project.org/package=graphkernels and https://pypi.python.org/pypi/graphkernels. mahito@nii.ac.jp or elisabetta.ghisu@bsse.ethz.ch. Supplementary data are available online at Bioinformatics. © The Author(s) 2017. Published by Oxford University Press.

Analysis of ATLAS FLB-EC6 Experiment using SPACE Code

International Nuclear Information System (INIS)

Lee, Donghyuk; Kim, Yohan; Kim, Seyun

2013-01-01

The new code is named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). As a part of code validation effort, simulation of ATLAS FLB(Feedwater Line Break) experiment using SPACE code has been performed. The FLB-EC6 experiment is economizer break of a main feedwater line. The calculated results using the SPACE code are compared with those from the experiment. The ATLAS FLB-EC6 experiment, which is economizer feedwater line break, was simulated using the SPACE code. The calculated results were compared with those from the experiment. The comparisons of break flow rate and steam generator water level show good agreement with the experiment. The SPACE code is capable of predicting physical phenomena occurring during ATLAS FLB-EC6 experiment

Evaluation on the structural soundness of the package with the lid bolted for subsurface disposal by finite element method

International Nuclear Information System (INIS)

Ito, Chihiro

2011-01-01

The structural analysis of the disposal package for low-level radioactive wastes with relatively high activities (called L1 waste in Japan) were performed against normal and hypothetical conditions. As a normal condition the external load due to lifting, stacking of the package and filling the space of disposal pit with mortar or something were considered. Drop incident during handling and pressure due to some external force were taken up as hypothetical conditions. Using finite element code ABAQUS and three dimensional finite element model, structural analyses were carried out for the normal conditions. The results show that the maximum stresses occurred at the package due to the loads above mentioned were far less than the yield strength for all conditions. Therefore, it is confirmed that the disposal package keeps its integrity against the normal conditions. Analyses for load cases of 8 m drop onto the reinforced concrete slab were performed by using finite element code LS-DYNA. The results show that the strains at the impact zone of the package exceeded the fracture strain of the material and the opening of the lid at the vicinity on the impact zone was observed but the damaged area was limited in the vicinity of impact zone. As a maximum external pressure, 4 MPa was applied to the surface of the packages which were piled up in four layered in the disposal tunnel. According to the results of analyses by ABAQUS code the maximum strain occurred at the contact surfaces between lid and body of the top package. However, the package stays in sound because the plastic zone was so small and the value of the maximum strain was less than the fracture strain of the materials. (author)

PORTABLE ACOUSTIC MONITORING PACKAGE (PAMP)

Energy Technology Data Exchange (ETDEWEB)

John l. Loth; Gary J. Morris; George M. Palmer; Richard Guiler; Deepak Mehra

2003-07-01

The 1st generation acoustic monitoring package was designed to detect and analyze weak acoustic signals inside natural gas transmission lines. Besides a microphone it housed a three-inch diameter aerodynamic acoustic signal amplifier to maximize sensitivity to leak induced {Delta}p type signals. The theory and test results of this aerodynamic signal amplifier was described in the master's degree thesis of our Research Assistant Deepak Mehra who is about to graduate. To house such a large three-inch diameter sensor required the use of a steel 300-psi rated 4 inch weld neck flange, which itself weighed already 29 pounds. The completed 1st generation Acoustic Monitoring Package weighed almost 100 pounds. This was too cumbersome to mount in the field, on an access port at a pipeline shut-off valve. Therefore a 2nd generation and truly Portable Acoustic Monitor was built. It incorporated a fully self-contained {Delta}p type signal sensor, rated for line pressures up to 1000 psi with a base weight of only 6 pounds. This is the Rosemont Inc. Model 3051CD-Range 0, software driven sensor, which is believed to have industries best total performance. Its most sensitive unit was purchased with a {Delta}p range from 0 to 3 inch water. This resulted in the herein described 2nd generation: Portable Acoustic Monitoring Package (PAMP) for pipelines up to 1000 psi. Its 32-pound total weight includes an 18-volt battery. Together with a 3 pound laptop with its 4-channel data acquisition card, completes the equipment needed for field acoustic monitoring of natural gas transmission pipelines.

Beyond filtered backprojection: A reconstruction software package for ion beam microtomography data

Science.gov (United States)

Habchi, C.; Gordillo, N.; Bourret, S.; Barberet, Ph.; Jovet, C.; Moretto, Ph.; Seznec, H.

2013-01-01

A new version of the TomoRebuild data reduction software package is presented, for the reconstruction of scanning transmission ion microscopy tomography (STIMT) and particle induced X-ray emission tomography (PIXET) images. First, we present a state of the art of the reconstruction codes available for ion beam microtomography. The algorithm proposed here brings several advantages. It is a portable, multi-platform code, designed in C++ with well-separated classes for easier use and evolution. Data reduction is separated in different steps and the intermediate results may be checked if necessary. Although no additional graphic library or numerical tool is required to run the program as a command line, a user friendly interface was designed in Java, as an ImageJ plugin. All experimental and reconstruction parameters may be entered either through this plugin or directly in text format files. A simple standard format is proposed for the input of experimental data. Optional graphic applications using the ROOT interface may be used separately to display and fit energy spectra. Regarding the reconstruction process, the filtered backprojection (FBP) algorithm, already present in the previous version of the code, was optimized so that it is about 10 times as fast. In addition, Maximum Likelihood Expectation Maximization (MLEM) and its accelerated version Ordered Subsets Expectation Maximization (OSEM) algorithms were implemented. A detailed user guide in English is available. A reconstruction example of experimental data from a biological sample is given. It shows the capability of the code to reduce noise in the sinograms and to deal with incomplete data, which puts a new perspective on tomography using low number of projections or limited angle.

The fastclime Package for Linear Programming and Large-Scale Precision Matrix Estimation in R.

Science.gov (United States)

Pang, Haotian; Liu, Han; Vanderbei, Robert

2014-02-01

We develop an R package fastclime for solving a family of regularized linear programming (LP) problems. Our package efficiently implements the parametric simplex algorithm, which provides a scalable and sophisticated tool for solving large-scale linear programs. As an illustrative example, one use of our LP solver is to implement an important sparse precision matrix estimation method called CLIME (Constrained L 1 Minimization Estimator). Compared with existing packages for this problem such as clime and flare, our package has three advantages: (1) it efficiently calculates the full piecewise-linear regularization path; (2) it provides an accurate dual certificate as stopping criterion; (3) it is completely coded in C and is highly portable. This package is designed to be useful to statisticians and machine learning researchers for solving a wide range of problems.

Solving Differential Equations in R: Package deSolve

Science.gov (United States)

In this paper we present the R package deSolve to solve initial value problems (IVP) written as ordinary differential equations (ODE), differential algebraic equations (DAE) of index 0 or 1 and partial differential equations (PDE), the latter solved using the method of lines appr...

User’s guide for GcClust—An R package for clustering of regional geochemical data

Science.gov (United States)

Ellefsen, Karl J.; Smith, David B.

2016-04-08

GcClust is a software package developed by the U.S. Geological Survey for statistical clustering of regional geochemical data, and similar data such as regional mineralogical data. Functions within the software package are written in the R statistical programming language. These functions, their documentation, and a copy of the user’s guide are bundled together in R’s unit of sharable code, which is called a “package.” The user’s guide includes step-by-step instructions showing how the functions are used to cluster data and to evaluate the clustering results. These functions are demonstrated in this report using test data, which are included in the package.

CAFE: A Computer Tool for Accurate Simulation of the Regulatory Pool Fire Environment for Type B Packages

International Nuclear Information System (INIS)

Gritzo, L.A.; Koski, J.A.; Suo-Anttila, A.J.

1999-01-01

The Container Analysis Fire Environment computer code (CAFE) is intended to provide Type B package designers with an enhanced engulfing fire boundary condition when combined with the PATRAN/P-Thermal commercial code. Historically an engulfing fire boundary condition has been modeled as σT 4 where σ is the Stefan-Boltzman constant, and T is the fire temperature. The CAFE code includes the necessary chemistry, thermal radiation, and fluid mechanics to model an engulfing fire. Effects included are the local cooling of gases that form a protective boundary layer that reduces the incoming radiant heat flux to values lower than expected from a simple σT 4 model. In addition, the effect of object shape on mixing that may increase the local fire temperature is included. Both high and low temperature regions that depend upon the local availability of oxygen are also calculated. Thus the competing effects that can both increase and decrease the local values of radiant heat flux are included in a reamer that is not predictable a-priori. The CAFE package consists of a group of computer subroutines that can be linked to workstation-based thermal analysis codes in order to predict package performance during regulatory and other accident fire scenarios

MVPACK: a package for the computer-aided design of multivariable control systems

International Nuclear Information System (INIS)

Mensah, S.

1984-01-01

The design and analysis of high performing controllers for large complex plants require a collection of interactive, powerful computer software. MVPACK, an open-ended package for the computer aided design of control systems has been developed in the Reactor Control Branch of the Chalk River Nuclear Laboratories. The package is fully interactive, and includes a comprehensive state-of-the-art mathematical library to support development of complex multivariable control algorithms. Coded in RATFOR, MVPACK operates with a flexible data structure which makes efficient use of minicomputer resources and provides a standard framework for program generation. The existence of a help mechanism enhances the simplicity of package utilization. This report provides the technical description of the package. It reviews the specifications used in the design and implementation of the package. The database structure, the supporting libraries and the design and analysis modules of MVPACK are described. The report includes several application examples to illustrate the capability of the package. Experience with MVPACK shows that the package provides a synergistic environment for control and regulation systems design, and that it is a unique tool in training of control system engineers

Interface thermal characteristics of flip chip packages - A numerical study

International Nuclear Information System (INIS)

Kandasamy, Ravi; Mujumdar, A.S.

2009-01-01

Flip chip ball grid array (FC-BGA) packages are commonly used for high inputs/outputs (I/O) ICs; they have been proven to provide good solutions for a variety of applications to maximize thermal and electrical performance. A fundamental limitation to such devices is the thermal resistance at the top of the package, which is characterized θ JC parameter. The die-to-lid interface thermal resistance is identified as a critical issue for the thermal management of electronic packages. This paper focuses on the effect of the interface material property changes on the interface thermal resistance. The effect of package's junction to case (Theta-JC or θ JC ) thermal performance is investigated for bare die, flat lid and cup lid packages using a validated thermal model. Thermal performance of a cup or flat lid attached and bare die packages were investigated for different interface materials. Improved Theta-JC performance was observed for the large die as compared to the smaller die. Several parametric studies were carried out to understand the effects of interface bond line thickness (BLT), different die sizes, the average void size during assembly and thermal conductivity of interface materials on package thermal resistance

A comparison and benchmark of two electron cloud packages

Energy Technology Data Exchange (ETDEWEB)

Lebrun, Paul L.G.; Amundson, James F; Spentzouris, Panagiotis G; Veitzer, Seth A

2012-01-01

We present results from precision simulations of the electron cloud (EC) problem in the Fermilab Main Injector using two distinct codes. These two codes are (i)POSINST, a F90 2D+ code, and (ii)VORPAL, a 2D/3D electrostatic and electromagnetic code used for self-consistent simulations of plasma and particle beam problems. A specific benchmark has been designed to demonstrate the strengths of both codes that are relevant to the EC problem in the Main Injector. As differences between results obtained from these two codes were bigger than the anticipated model uncertainties, a set of changes to the POSINST code were implemented. These changes are documented in this note. This new version of POSINST now gives EC densities that agree with those predicted by VORPAL, within {approx}20%, in the beam region. The root cause of remaining differences are most likely due to differences in the electrostatic Poisson solvers. From a software engineering perspective, these two codes are very different. We comment on the pros and cons of both approaches. The design(s) for a new EC package are briefly discussed.

Design of a type - a transport package for 99Mo-99mTc Coltech generator

International Nuclear Information System (INIS)

Kothalkar, Chetan; Suryanarayana, G.V.; Dey, A.C.; Sachdev, S.S.; Choughule, N.; Murali, S.

2012-01-01

BRIT is launching a new product called 99 Mo- 99m Tc Coltech generator. The Coltech generator is a devise designed for the transport of 99 Mo radioisotope adsorbed on the acidic alumina in a sealed glass column (max dimensions: 13 mm diameter, 70 mm height) as the primary containment. At hospital end, 99m Tc, the daughter product of 99 Mo, can be eluted out from the generator using saline. The active column is fitted with a leak proof network of stainless steel needles. The glass column carrying 99 Mo is housed inside a lead shielding having minimum thickness of 50 mm all around, which serves as secondary containment. The shielding is housed inside the ABS shell which acts as tertiary containment, also provides protection to the needles, filters etc. Total weight of the generator is 16 kg. Based on the AERB code SC/TR-1 (being revised), 99 Mo- 99m Tc Coltech generator will be transported in a Type-A transport container. A transport package has been designed by following the code SC/TR-1. Principle design of the package is based on the package for transportation of the similar generator produced by POLATOM, Poland and the package is approved by the Polish regulatory authority. Components are manufactured locally taking care of lndian conditions. The package comprised of a MS drum (HOBBOCK) with tamper proof lockable MS lid and a handle to assist in lifting. For absorbing the shock during transportation, the generator assembly is packed inside the two pieces EPS top and bottom support. The package has been designed for transportation by all modes of transport. Since radioactive material is solid in form and sealed a glass column, it has been designed to sustain a free drop test of 1.2 m, in addition to other tests specified in SC/TR-1. During trial batches upto ∼ 1 Ci of 99 Mo generators were produced, packed in the same Type-A package and supplied to local nuclear medicine center RMC, Mumbai in BRIT vehicle in consultation with AERB. The radiometry of the packages

The CHEASE code for toroidal MHD equilibria

International Nuclear Information System (INIS)

Luetjens, H.

1996-03-01

CHEASE solves the Grad-Shafranov equation for the MHD equilibrium of a Tokamak-like plasma with pressure and current profiles specified by analytic forms or sets of data points. Equilibria marginally stable to ballooning modes or with a prescribed fraction of bootstrap current can be computed. The code provides a mapping to magnetic flux coordinates, suitable for MHD stability calculations or global wave propagation studies. The code computes equilibrium quantities for the stability codes ERATO, MARS, PEST, NOVA-W and XTOR and for the global wave propagation codes LION and PENN. The two-dimensional MHD equilibrium (Grad-Shafranov) equation is solved in variational form. The discretization uses bicubic Hermite finite elements with continuous first order derivates for the poloidal flux function Ψ. The nonlinearity of the problem is handled by Picard iteration. The mapping to flux coordinates is carried out with a method which conserves the accuracy of the cubic finite elements. The code uses routines from the CRAY libsci.a program library. However, all these routines are included in the CHEASE package itself. If CHEASE computes equilibrium quantities for MARS with fast Fourier transforms, the NAG library is required. CHEASE is written in standard FORTRAN-77, except for the use of the input facility NAMELIST. CHEASE uses variable names with up to 8 characters, and therefore violates the ANSI standard. CHEASE transfers plot quantities through an external disk file to a plot program named PCHEASE using the UNIRAS or the NCAR plot package. (author) figs., tabs., 34 refs

The CHEASE code for toroidal MHD equilibria

Energy Technology Data Exchange (ETDEWEB)

Luetjens, H. [Ecole Polytechnique, 91 - Palaiseau (France). Centre de Physique Theorique; Bondeson, A. [Chalmers Univ. of Technology, Goeteborg (Sweden). Inst. for Electromagnetic Field Theory and Plasma Physics; Sauter, O. [ITER-San Diego, La Jolla, CA (United States)

1996-03-01

CHEASE solves the Grad-Shafranov equation for the MHD equilibrium of a Tokamak-like plasma with pressure and current profiles specified by analytic forms or sets of data points. Equilibria marginally stable to ballooning modes or with a prescribed fraction of bootstrap current can be computed. The code provides a mapping to magnetic flux coordinates, suitable for MHD stability calculations or global wave propagation studies. The code computes equilibrium quantities for the stability codes ERATO, MARS, PEST, NOVA-W and XTOR and for the global wave propagation codes LION and PENN. The two-dimensional MHD equilibrium (Grad-Shafranov) equation is solved in variational form. The discretization uses bicubic Hermite finite elements with continuous first order derivates for the poloidal flux function {Psi}. The nonlinearity of the problem is handled by Picard iteration. The mapping to flux coordinates is carried out with a method which conserves the accuracy of the cubic finite elements. The code uses routines from the CRAY libsci.a program library. However, all these routines are included in the CHEASE package itself. If CHEASE computes equilibrium quantities for MARS with fast Fourier transforms, the NAG library is required. CHEASE is written in standard FORTRAN-77, except for the use of the input facility NAMELIST. CHEASE uses variable names with up to 8 characters, and therefore violates the ANSI standard. CHEASE transfers plot quantities through an external disk file to a plot program named PCHEASE using the UNIRAS or the NCAR plot package. (author) figs., tabs., 34 refs.

CH-TRU Content Codes (CH-TRUCON)

International Nuclear Information System (INIS)

2005-01-01

The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes 'shipping categories' that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the 'General Case,' which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for 'Close-Proximity Shipments' (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for 'Controlled Shipments

CH-TRU Content Codes (CH-TRUCON)

Energy Technology Data Exchange (ETDEWEB)

Washington TRU Solutions LLC

2005-10-15

The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

Functional Interaction of the Adenovirus IVa2 Protein with Adenovirus Type 5 Packaging Sequences

OpenAIRE

Ostapchuk, Philomena; Yang, Jihong; Auffarth, Ece; Hearing, Patrick

2005-01-01

Adenovirus type 5 (Ad5) DNA packaging is initiated in a polar fashion from the left end of the genome. The packaging process is dependent on the cis-acting packaging domain located between nucleotides 230 and 380. Seven AT-rich repeats that direct packaging have been identified within this domain. A1, A2, A5, and A6 are the most important repeats functionally and share a bipartite sequence motif. Several lines of evidence suggest that there is a limiting trans-acting factor(s) that plays a ro...

GAME: GAlaxy Machine learning for Emission lines

Science.gov (United States)

Ucci, G.; Ferrara, A.; Pallottini, A.; Gallerani, S.

2018-06-01

We present an updated, optimized version of GAME (GAlaxy Machine learning for Emission lines), a code designed to infer key interstellar medium physical properties from emission line intensities of ultraviolet /optical/far-infrared galaxy spectra. The improvements concern (a) an enlarged spectral library including Pop III stars, (b) the inclusion of spectral noise in the training procedure, and (c) an accurate evaluation of uncertainties. We extensively validate the optimized code and compare its performance against empirical methods and other available emission line codes (PYQZ and HII-CHI-MISTRY) on a sample of 62 SDSS stacked galaxy spectra and 75 observed HII regions. Very good agreement is found for metallicity. However, ionization parameters derived by GAME tend to be higher. We show that this is due to the use of too limited libraries in the other codes. The main advantages of GAME are the simultaneous use of all the measured spectral lines and the extremely short computational times. We finally discuss the code potential and limitations.

GAME: GAlaxy Machine learning for Emission lines

Science.gov (United States)

Ucci, G.; Ferrara, A.; Pallottini, A.; Gallerani, S.

2018-03-01

We present an updated, optimized version of GAME (GAlaxy Machine learning for Emission lines), a code designed to infer key interstellar medium physical properties from emission line intensities of UV/optical/far infrared galaxy spectra. The improvements concern: (a) an enlarged spectral library including Pop III stars; (b) the inclusion of spectral noise in the training procedure, and (c) an accurate evaluation of uncertainties. We extensively validate the optimized code and compare its performance against empirical methods and other available emission line codes (pyqz and HII-CHI-mistry) on a sample of 62 SDSS stacked galaxy spectra and 75 observed HII regions. Very good agreement is found for metallicity. However, ionization parameters derived by GAME tend to be higher. We show that this is due to the use of too limited libraries in the other codes. The main advantages of GAME are the simultaneous use of all the measured spectral lines, and the extremely short computational times. We finally discuss the code potential and limitations.

SPEXTRA: Optimal extraction code for long-slit spectra in crowded fields

Science.gov (United States)

Sarkisyan, A. N.; Vinokurov, A. S.; Solovieva, Yu. N.; Sholukhova, O. N.; Kostenkov, A. E.; Fabrika, S. N.

2017-10-01

We present a code for the optimal extraction of long-slit 2D spectra in crowded stellar fields. Its main advantage and difference from the existing spectrum extraction codes is the presence of a graphical user interface (GUI) and a convenient visualization system of data and extraction parameters. On the whole, the package is designed to study stars in crowded fields of nearby galaxies and star clusters in galaxies. Apart from the spectrum extraction for several stars which are closely located or superimposed, it allows the spectra of objects to be extracted with subtraction of superimposed nebulae of different shapes and different degrees of ionization. The package can also be used to study single stars in the case of a strong background. In the current version, the optimal extraction of 2D spectra with an aperture and the Gaussian function as PSF (point spread function) is proposed. In the future, the package will be supplemented with the possibility to build a PSF based on a Moffat function. We present the details of GUI, illustrate main features of the package, and show results of extraction of the several interesting spectra of objects from different telescopes.

Method for Determining Optimal Residential Energy Efficiency Retrofit Packages

Energy Technology Data Exchange (ETDEWEB)

Polly, B.; Gestwick, M.; Bianchi, M.; Anderson, R.; Horowitz, S.; Christensen, C.; Judkoff, R.

2011-04-01

Businesses, government agencies, consumers, policy makers, and utilities currently have limited access to occupant-, building-, and location-specific recommendations for optimal energy retrofit packages, as defined by estimated costs and energy savings. 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. Energy savings and incremental costs are calculated relative to a minimum upgrade reference scenario, which accounts for efficiency upgrades that would occur in the absence of a retrofit because of equipment wear-out and replacement with current minimum standards.

Clinical information in drug package inserts in India

Directory of Open Access Journals (Sweden)

Shivkar Y

2009-01-01

Full Text Available Background: It is widely recognized that accurate and reliable product information is essential for the safe and effective use of medications. Pharmaceutical companies are the primary source of most drug information, including package inserts. Package inserts are printed leaflets accompanying marketed drug products and contain information approved by the regulatory agencies. Studies on package inserts in India, in 1996, had shown that crucial information was often missing and they lacked uniformity. Aim: To assess the presentation and completeness of clinically important information provided in the currently available package inserts in India. Materials and Methods: Package inserts accompanying allopathic drug products marketed by pharmaceutical companies in India were collected. These package inserts were analyzed for the content of clinically important information in various sections. Statistical Analysis: The results were expressed as absolute numbers and percentages. Results: Preliminary analyses revealed that most package inserts did contain information under headings, such as, therapeutic indications, contraindications, undesirable effects, etc., listed in the Drugs and Cosmetics Rules 1945. The findings indicated considerable improvement in package inserts since 1996. However, on critical evaluation it was revealed that clinically important information was not well presented and was often incomplete. Information with regard to pediatric and geriatric use was present in only 44% and 13% of the package inserts, respectively. Only five of the inserts had information on the most frequent adverse drug reactions associated with the drug. Also, information on interactions and overdosage was often missing. Conclusion: Although the package inserts appear to have improved over the past decade there is still a definite need to further refine the clinical information contained, to minimize the risks to patients. This could be brought about by self

Response analysis on nonuniform transmission line

Directory of Open Access Journals (Sweden)

Cvetković Zlata

2005-01-01

Full Text Available Transients on a loss less exponential transmission line with a pure resistance load are presented in this paper. The approach is based on the two-port presentation of the transmission line. Using Picard-Carson's method the transmission line equations are solved. The relationship between source voltage and the load voltage in s-domain is derived. All the results are plotted using program package Mathematica 3.0.

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

Scilab software package for the study of dynamical systems

Science.gov (United States)

Bordeianu, C. C.; Beşliu, C.; Jipa, Al.; Felea, D.; Grossu, I. V.

2008-05-01

This work presents a new software package for the study of chaotic flows and maps. The codes were written using Scilab, a software package for numerical computations providing a powerful open computing environment for engineering and scientific applications. It was found that Scilab provides various functions for ordinary differential equation solving, Fast Fourier Transform, autocorrelation, and excellent 2D and 3D graphical capabilities. The chaotic behaviors of the nonlinear dynamics systems were analyzed using phase-space maps, autocorrelation functions, power spectra, Lyapunov exponents and Kolmogorov-Sinai entropy. Various well known examples are implemented, with the capability of the users inserting their own ODE. Program summaryProgram title: Chaos Catalogue identifier: AEAP_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEAP_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 885 No. of bytes in distributed program, including test data, etc.: 5925 Distribution format: tar.gz Programming language: Scilab 3.1.1 Computer: PC-compatible running Scilab on MS Windows or Linux Operating system: Windows XP, Linux RAM: below 100 Megabytes Classification: 6.2 Nature of problem: Any physical model containing linear or nonlinear ordinary differential equations (ODE). Solution method: Numerical solving of ordinary differential equations. The chaotic behavior of the nonlinear dynamical system is analyzed using Poincaré sections, phase-space maps, autocorrelation functions, power spectra, Lyapunov exponents and Kolmogorov-Sinai entropies. Restrictions: The package routines are normally able to handle ODE systems of high orders (up to order twelve and possibly higher), depending on the nature of the problem. Running time: 10 to 20 seconds for problems that do not

Solving Differential Equations in R: Package deSolve

NARCIS (Netherlands)

Soetaert, K.E.R.; Petzoldt, T.; Setzer, R.W.

2010-01-01

In this paper we present the R package deSolve to solve initial value problems (IVP) written as ordinary differential equations (ODE), differential algebraic equations (DAE) of index 0 or 1 and partial differential equations (PDE), the latter solved using the method of lines approach. The

DOUBLE code simulations of emissivities of fast neutrals for different plasma observation view-lines of neutral particle analyzers on the COMPASS tokamak

Science.gov (United States)

Mitosinkova, K.; Tomes, M.; Stockel, J.; Varju, J.; Stano, M.

2018-03-01

Neutral particle analyzers (NPA) measure line-integrated energy spectra of fast neutral atoms escaping the tokamak plasma, which are a product of charge-exchange (CX) collisions of plasma ions with background neutrals. They can observe variations in the ion temperature T i of non-thermal fast ions created by additional plasma heating. However, the plasma column which a fast atom has to pass through must be sufficiently short in comparison with the fast atom’s mean-free-path. Tokamak COMPASS is currently equipped with one NPA installed at a tangential mid-plane port. This orientation is optimal for observing non-thermal fast ions. However, in this configuration the signal at energies useful for T i derivation is lost in noise due to the too long fast atoms’ trajectories. Thus, a second NPA is planned to be connected for the purpose of measuring T i. We analyzed different possible view-lines (perpendicular mid-plane, tangential mid-plane, and top view) for the second NPA using the DOUBLE Monte-Carlo code and compared the results with the performance of the present NPA with tangential orientation. The DOUBLE code provides fast-atoms’ emissivity functions along the NPA view-line. The position of the median of these emissivity functions is related to the location from where the measured signal originates. Further, we compared the difference between the real central T i used as a DOUBLE code input and the T iCX derived from the exponential decay of simulated energy spectra. The advantages and disadvantages of each NPA location are discussed.

Software Design Document for the AMP Nuclear Fuel Performance Code

International Nuclear Information System (INIS)

Philip, Bobby; Clarno, Kevin T.; Cochran, Bill

2010-01-01

The purpose of this document is to describe the design of the AMP nuclear fuel performance code. It provides an overview of the decomposition into separable components, an overview of what those components will do, and the strategic basis for the design. The primary components of a computational physics code include a user interface, physics packages, material properties, mathematics solvers, and computational infrastructure. Some capability from established off-the-shelf (OTS) packages will be leveraged in the development of AMP, but the primary physics components will be entirely new. The material properties required by these physics operators include many highly non-linear properties, which will be replicated from FRAPCON and LIFE where applicable, as well as some computationally-intensive operations, such as gap conductance, which depends upon the plenum pressure. Because there is extensive capability in off-the-shelf leadership class computational solvers, AMP will leverage the Trilinos, PETSc, and SUNDIALS packages. The computational infrastructure includes a build system, mesh database, and other building blocks of a computational physics package. The user interface will be developed through a collaborative effort with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Capability Transfer program element as much as possible and will be discussed in detail in a future document.

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

19 CFR 142.49 - Deletion of C-4 Code.

Science.gov (United States)

2010-04-01

.... Entry filers may delete C-4 Codes from Line Release by notifying the port director in writing on a Deletion Data Loading Sheet. Such notification shall state the C-4 Code which is to be deleted, the port... TREASURY (CONTINUED) ENTRY PROCESS Line Release § 142.49 Deletion of C-4 Code. (a) By Customs. A port...

Status report on the 'Merging' of the Electron-Cloud Code POSINST with the 3-D Accelerator PIC CODE WARP

Energy Technology Data Exchange (ETDEWEB)

Vay, J.-L.; Furman, M.A.; Azevedo, A.W.; Cohen, R.H.; Friedman, A.; Grote, D.P.; Stoltz, P.H.

2004-04-19

We have integrated the electron-cloud code POSINST [1] with WARP [2]--a 3-D parallel Particle-In-Cell accelerator code developed for Heavy Ion Inertial Fusion--so that the two can interoperate. Both codes are run in the same process, communicate through a Python interpreter (already used in WARP), and share certain key arrays (so far, particle positions and velocities). Currently, POSINST provides primary and secondary sources of electrons, beam bunch kicks, a particle mover, and diagnostics. WARP provides the field solvers and diagnostics. Secondary emission routines are provided by the Tech-X package CMEE.

A flexible R package for nonnegative matrix factorization

Directory of Open Access Journals (Sweden)

Seoighe Cathal

2010-07-01

Full Text Available Abstract Background Nonnegative Matrix Factorization (NMF is an unsupervised learning technique that has been applied successfully in several fields, including signal processing, face recognition and text mining. Recent applications of NMF in bioinformatics have demonstrated its ability to extract meaningful information from high-dimensional data such as gene expression microarrays. Developments in NMF theory and applications have resulted in a variety of algorithms and methods. However, most NMF implementations have been on commercial platforms, while those that are freely available typically require programming skills. This limits their use by the wider research community. Results Our objective is to provide the bioinformatics community with an open-source, easy-to-use and unified interface to standard NMF algorithms, as well as with a simple framework to help implement and test new NMF methods. For that purpose, we have developed a package for the R/BioConductor platform. The package ports public code to R, and is structured to enable users to easily modify and/or add algorithms. It includes a number of published NMF algorithms and initialization methods and facilitates the combination of these to produce new NMF strategies. Commonly used benchmark data and visualization methods are provided to help in the comparison and interpretation of the results. Conclusions The NMF package helps realize the potential of Nonnegative Matrix Factorization, especially in bioinformatics, providing easy access to methods that have already yielded new insights in many applications. Documentation, source code and sample data are available from CRAN.

Status of shielding analysis methods for transport packages

International Nuclear Information System (INIS)

Parks, C.V.; Broadhead, B.L.; Brady, M.C.

1991-01-01

Shielding analysis methods for transport packages are becoming more important to the cask designer because optimized cask designs with higher payloads can yield doses near the limits set by regulatory authorities. Uncertainty arising from generation of radiation sources, selection of cross-section data, and the radiation transport methodology must be considered. Recent comparison studies using popular US codes illustrate calculational discrepancies arising from each of these areas

PSRPOPPy: an open-source package for pulsar population simulations

Science.gov (United States)

Bates, S. D.; Lorimer, D. R.; Rane, A.; Swiggum, J.

2014-04-01

We have produced a new software package for the simulation of pulsar populations, PSRPOPPY, based on the PSRPOP package. The codebase has been re-written in Python (save for some external libraries, which remain in their native Fortran), utilizing the object-oriented features of the language, and improving the modularity of the code. Pre-written scripts are provided for running the simulations in `standard' modes of operation, but the code is flexible enough to support the writing of personalised scripts. The modular structure also makes the addition of experimental features (such as new models for period or luminosity distributions) more straightforward than with the previous code. We also discuss potential additions to the modelling capabilities of the software. Finally, we demonstrate some potential applications of the code; first, using results of surveys at different observing frequencies, we find pulsar spectral indices are best fitted by a normal distribution with mean -1.4 and standard deviation 1.0. Secondly, we model pulsar spin evolution to calculate the best fit for a relationship between a pulsar's luminosity and spin parameters. We used the code to replicate the analysis of Faucher-Giguère & Kaspi, and have subsequently optimized their power-law dependence of radio luminosity, L, with period, P, and period derivative, Ṗ. We find that the underlying population is best described by L ∝ P-1.39±0.09 Ṗ0.48±0.04 and is very similar to that found for γ-ray pulsars by Perera et al. Using this relationship, we generate a model population and examine the age-luminosity relation for the entire pulsar population, which may be measurable after future large-scale surveys with the Square Kilometre Array.

Software supervisor: extension to the on-line codes utilization in order to help the process control

International Nuclear Information System (INIS)

Thomas, J.B.; Dumas, M.; Evrard, J.M.

1988-01-01

Calculation is a complex problem, which is usually solved by human experts. The complexity and the potentiality of the software increases. The introduction of the calculations in real time systems needs additional specifications. These aims cab be achieved by means of the control of the knowledge based systems, as well as by the introduction of the software techniques in the existing computer world. The following examples are given: the automatic generation of the calculation methods (in control language) in the modular code systems; the calculations monitoring by the expert systems, in order to help the on-line operations [fr

A package for environmental impact assessment of nuclear installations (NGLAR)

International Nuclear Information System (INIS)

Yang Yin; Chen Xiaoqiu; Ding Jinhou; Zhao Hui; Xi Xiaojun; Li Hongsheng

1996-09-01

The main contents, designing strategies and properties of the microcomputer-based software package NGLAR are described for environmental impact assessment of nuclear installations. The package consists of the following components: NGAS and NACC, the codes for routine and accidental airborne releases respectively; NLIQ, the code for both routine and accidental liquid releases; and NRED, environmental database system of nuclear installations. NGAS and NACC are used for evaluating atmosphere dispersion and doses to public of radioactive materials released from nuclear facilities, giving the concentrations around the facilities of radionuclides in air, on ground surface, and in varieties of animal foods and farm produces, and further estimating collective doses and doses to critical group around the facilities. NLIQ is suitable for liquid effluence released to non-tide rivers, and is modelled to calculate firstly the concentration of radionuclides concerned in the polluted rivers, and then to estimate the resulting doses to public. Under routine releases, the doses obtained from NGAS and NLIQ can be appropriately categorized and summed up together. NRED can be run independently, also used to provide some input data for above programs and save data permanently for them. Having both English and Chinese versions, the package, which was fabricated of multiple functions can be run on IBM 386 or higher and its compatible microcomputers. (3 figs., 1 tabs)

Experience gained in running the EPRI MMS code with an in-house simulation language

International Nuclear Information System (INIS)

Weber, D.S.

1987-01-01

The EPRI Modular Modeling System (MMS) code represents a collection of component models and a steam/water properties package. This code has undergone extensive verification and validation testing. Currently, the code requires a commercially available simulation language to run. The Philadelphia Electric Company (PECO) has been modeling power plant systems for over the past sixteen years. As a result, an extensive number of models have been developed. In addition, an extensive amount of experience has been developed and gained using an in-house simulation language. The objective of this study was to explore the possibility of developing an MMS pre-processor which would allow the use of the MMS package with other simulation languages such as the PECO in-house simulation language

Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

Energy Technology Data Exchange (ETDEWEB)

Page, R.; Jones, J.R.

1997-07-01

Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell `B` Loss of offsite power fault transient.

Development of an integrated thermal-hydraulics capability incorporating RELAP5 and PANTHER neutronics code

International Nuclear Information System (INIS)

Page, R.; Jones, J.R.

1997-01-01

Ensuring that safety analysis needs are met in the future is likely to lead to the development of new codes and the further development of existing codes. It is therefore advantageous to define standards for data interfaces and to develop software interfacing techniques which can readily accommodate changes when they are made. Defining interface standards is beneficial but is necessarily restricted in application if future requirements are not known in detail. Code interfacing methods are of particular relevance with the move towards automatic grid frequency response operation where the integration of plant dynamic, core follow and fault study calculation tools is considered advantageous. This paper describes the background and features of a new code TALINK (Transient Analysis code LINKage program) used to provide a flexible interface to link the RELAP5 thermal hydraulics code with the PANTHER neutron kinetics and the SIBDYM whole plant dynamic modelling codes used by Nuclear Electric. The complete package enables the codes to be executed in parallel and provides an integrated whole plant thermal-hydraulics and neutron kinetics model. In addition the paper discusses the capabilities and pedigree of the component codes used to form the integrated transient analysis package and the details of the calculation of a postulated Sizewell 'B' Loss of offsite power fault transient