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

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

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

International Nuclear Information System (INIS)

Souto, F.J.

1991-06-01

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

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)

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

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

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.

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

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

International Nuclear Information System (INIS)

Hall, P.; Hutt, P.

1994-01-01

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

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

International Nuclear Information System (INIS)

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

1996-11-01

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

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

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

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-07-01

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

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

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)

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)

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

International Nuclear Information System (INIS)

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

2004-03-01

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

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.

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

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.

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.

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.

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

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

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

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)

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)

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

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)

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)

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).

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-03-01

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

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

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

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

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)

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-07-01

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

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

International Nuclear Information System (INIS)

Ipatti, A.

1993-03-01

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

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 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

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

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

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

International Nuclear Information System (INIS)

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

1986-07-01

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

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

International Nuclear Information System (INIS)

Berusch, A.; Gause, E.

1987-01-01

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

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

International Nuclear Information System (INIS)

Bolton, Gary

2013-01-01

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

Current status and future direction of the MONK software package

International Nuclear Information System (INIS)

Smith, Nigel; Armishaw, Malcolm; Cooper, Andrew

2003-01-01

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

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

Calculation of source terms for NUREG-1150

International Nuclear Information System (INIS)

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

1987-10-01

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

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

Science.gov (United States)

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

2016-08-01

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

9975 Shipping package component long-term degradation rates

Energy Technology Data Exchange (ETDEWEB)

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

2017-06-28

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

9 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.

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

International Nuclear Information System (INIS)

Baron, J.; Bastianelli, B.

1997-01-01

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

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.

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

International Nuclear Information System (INIS)

S. Mastilovic

2000-01-01

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

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

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

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

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

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

International Nuclear Information System (INIS)

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

1995-03-01

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

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)

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.

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

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.

WASTE PACKAGE TRANSPORTER DESIGN

International Nuclear Information System (INIS)

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

1998-01-01

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

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)

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-03-01

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

Development of in-vessel source term analysis code, tracer

International Nuclear Information System (INIS)

Miyagi, K.; Miyahara, S.

1996-01-01

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

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.

WASTE PACKAGE TRANSPORTER DESIGN

Energy Technology Data Exchange (ETDEWEB)

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

1998-09-23

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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

Package Formats for Preserved Digital Material

DEFF Research Database (Denmark)

Zierau, Eld

2012-01-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Integrated computer codes for nuclear power plant severe accident analysis

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

Integrated computer codes for nuclear power plant severe accident analysis

International Nuclear Information System (INIS)

Jordanov, I.; Khristov, Y.

1995-01-01

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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...

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

International Nuclear Information System (INIS)

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

1993-01-01

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

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

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

Trends in Food Packaging.

Science.gov (United States)

Ott, Dana B.

1988-01-01

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

AutoDipole - Automated generation of dipole subtraction terms

International Nuclear Information System (INIS)

Hasegawa, K.; Uwer, P.

2009-11-01

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

AutoDipole - Automated generation of dipole subtraction terms

Energy Technology Data Exchange (ETDEWEB)

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

2009-11-15

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

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.

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

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

International Nuclear Information System (INIS)

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

2006-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-12-15

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

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

Energy Technology Data Exchange (ETDEWEB)

Becker, Dirk-Alexander

2016-05-15

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

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)

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

Science.gov (United States)

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

2012-01-01

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

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)

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.

Code-switched English pronunciation modeling for Swahili spoken term detection

CSIR Research Space (South Africa)

Kleynhans, N

2016-05-01

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

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

Applications guide to the MORSE Monte Carlo code

International Nuclear Information System (INIS)

Cramer, S.N.

1985-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

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

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

Computer Package for Graphite Total Cross-Section Calculations

International Nuclear Information System (INIS)

Adib, M.; Fathalla, M.

2008-01-01

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

Blackout sequence modeling for Atucha-I with MARCH3 code

International Nuclear Information System (INIS)

Baron, J.; Bastianelli, B.

1997-01-01

The modeling of a blackout sequence in Atucha I nuclear power plant is presented in this paper, as a preliminary phase for a level II probabilistic safety assessment. Such sequence is analyzed with the code MARCH3 from STCP (Source Term Code Package), based on a specific model developed for Atucha, that takes into accounts it peculiarities. The analysis includes all the severe accident phases, from the initial transient (loss of heat sink), loss of coolant through the safety valves, core uncovered, heatup, metal-water reaction, melting and relocation, heatup and failure of the pressure vessel, core-concrete interaction in the reactor cavity, heatup and failure of the containment building (multi-compartmented) due to quasi-static overpressurization. The results obtained permit to visualize the time sequence of these events, as well as provide the basis for source term studies. (author) [es

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

MODLIB, library of Fortran modules for nuclear reaction codes

International Nuclear Information System (INIS)

Talou, Patrick

2006-01-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Features of the latest version of the PHITS code

International Nuclear Information System (INIS)

Sato, Tatsuhiko; Matsuda, Norihiro; Hashimoto, Shintaro; Iwamoto, Yosuke; Noda, Shusaku; Ogawa, Tatsuhiko; Nakajima, Hiroshi; Fukahori, Tokio; Okumura, Keisuke; Kai, Tetsuya; Chiba, Satoshi; Niita, Koji; Iwase, Hiroshi; Furuta, Takuya

2013-01-01

A Multi-purpose Monte Carlo Particle and Heavy Ion Transport code System, PHITS, developed through collaborations between JAEA and several institutes in Japan and Europe, and upgraded recently and released as PHITS2.52 is presented as an overview. In the new version, higher accuracy of the simulation was achieved by implementing the latest nuclear reaction models such as Liege intra-nuclear cascade version 4.6 (INCL4.6) and a statistical multi-fragmentation model including JAM and JQMD for high-energy regions. The reliability of the simulation code was improved by modifying both the algorithms for the electron-, positron-, and photon-transport simulations in terms of not only the code itself but also the contents of its package, such as the attached data libraries. More than 800 researchers have been registered as PHITS users, and they apply the code to various research and development fields such as nuclear technology, accelerator design, medical physics, and cosmic-ray research. (S. Ohno)

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

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

Science.gov (United States)

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

2011-01-01

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

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)

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.

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...

Unit-of-Use Versus Traditional Bulk Packaging

Directory of Open Access Journals (Sweden)

Tiffany So

2012-01-01

Full Text Available Background: The choice between unit-of-use versus traditional bulk packaging in the US has long been a continuous debate for drug manufacturers and pharmacies in order to have the most efficient and safest practices. Understanding the benefits of using unit-of-use packaging over bulk packaging by US drug manufacturers in terms of workflow efficiency, economical costs and medication safety in the pharmacy is sometimes challenging.Methods: A time-saving study comparing the time saved using unit-of-use packaging versus bulk packaging, was examined. Prices between unit-of-use versus bulk packages were compared by using the Red Book: Pharmacy’s Fundamental Reference. Other articles were reviewed on the topics of counterfeiting, safe labeling, and implementation of unit-of-use packaging. Lastly, a cost-saving study was reviewed showing how medication adherence, due to improved packaging, could be cost-effective for patients.Results: When examining time, costs, medication adherence, and counterfeiting arguments, unit-of-use packaging proved to be beneficial for patients in all these terms.

Unit-of-Use Versus Traditional Bulk Packaging

Directory of Open Access Journals (Sweden)

Tiffany So

2012-01-01

Full Text Available Background: The choice between unit-of-use versus traditional bulk packaging in the US has long been a continuous debate for drug manufacturers and pharmacies in order to have the most efficient and safest practices. Understanding the benefits of using unit-of-use packaging over bulk packaging by US drug manufacturers in terms of workflow efficiency, economical costs and medication safety in the pharmacy is sometimes challenging. Methods: A time-saving study comparing the time saved using unit-of-use packaging versus bulk packaging, was examined. Prices between unit-of-use versus bulk packages were compared by using the Red Book: Pharmacy's Fundamental Reference. Other articles were reviewed on the topics of counterfeiting, safe labeling, and implementation of unit-of-use packaging. Lastly, a cost-saving study was reviewed showing how medication adherence, due to improved packaging, could be cost-effective for patients. Results: When examining time, costs, medication adherence, and counterfeiting arguments, unit-of-use packaging proved to be beneficial for patients in all these terms.   Type: Student Project

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

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

Science.gov (United States)

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

2017-09-29

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

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)

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

Dissipation terms in one-dimensional hydrodynamic code for ICF target

International Nuclear Information System (INIS)

Tamba, Moritake; Niu, Keishiro.

1985-01-01

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

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)

Integration of CAM and CNC operation through code editing and manipulation

International Nuclear Information System (INIS)

Rosli Darmawan; Shalina Sheik Muhammad

2004-01-01

The IT technology for engineering design and manufacturing has gone through significant advancement for the last 30 years. It is widely acknowledged that IT would provide competitive advantage for engineering company in term of production cycle, productivity and efficiency. The recent development in this area is on the total system integration. While standard off-shelf CAD/CAM/CNC software and hardware packages would provide solution for system integration, more often than not users will stumble upon compatibility problems. Moreover, most of the integration deals with CAD and CAM systems. CNC integration has not been fully developed. Users always found problems in the integration of CAM and CNC machine due to the different level of technological development. CNC codes have not fundamentally progressed in the last 50 years, while CAD/CAM software packages have undergone massive evolution and improvement. This paper discusses a practical solution of CAM and CNC integration through code editing and manipulation within the CAM system in order to comply with the CNC machine requirements. (Author)

On-line application of the PANTHER advanced nodal code

International Nuclear Information System (INIS)

Hutt, P.K.; Knight, M.P.

1992-01-01

Over the last few years, Nuclear Electric has developed an integrated core performance code package for both light water reactors (LWRs) and advanced gas-cooled reactors (AGRs) that can perform a comprehensive range of calculations for fuel cycle design, safety analysis, and on-line operational support for such plants. The package consists of the following codes: WIMS for lattice physics, PANTHER whole reactor nodal flux and AGR thermal hydraulics, VIPRE for LWR thermal hydraulics, and ENIGMA for fuel performance. These codes are integrated within a UNIX-based interactive system called the Reactor Physics Workbench (RPW), which provides an interactive graphic user interface and quality assurance records/data management. The RPW can also control calculational sequences and data flows. The package has been designed to run both off-line and on-line accessing plant data through the RPW

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

Evaluation of system codes for analyzing naturally circulating gas loop

International Nuclear Information System (INIS)

Lee, Jeong Ik; No, Hee Cheon; Hejzlar, Pavel

2009-01-01

Steady-state natural circulation data obtained in a 7 m-tall experimental loop with carbon dioxide and nitrogen are presented in this paper. The loop was originally designed to encompass operating range of a prototype gas-cooled fast reactor passive decay heat removal system, but the results and conclusions are applicable to any natural circulation loop operating in regimes having buoyancy and acceleration parameters within the ranges validated in this loop. Natural circulation steady-state data are compared to numerical predictions by two system analysis codes: GAMMA and RELAP5-3D. GAMMA is a computational tool for predicting various transients which can potentially occur in a gas-cooled reactor. The code has a capability of analyzing multi-dimensional multi-component mixtures and includes models for friction, heat transfer, chemical reaction, and multi-component molecular diffusion. Natural circulation data with two gases show that the loop operates in the deteriorated turbulent heat transfer (DTHT) regime which exhibits substantially reduced heat transfer coefficients compared to the forced turbulent flow. The GAMMA code with an original heat transfer package predicted conservative results in terms of peak wall temperature. However, the estimated peak location did not successfully match the data. Even though GAMMA's original heat transfer package included mixed-convection regime, which is a part of the DTHT regime, the results showed that the original heat transfer package could not reproduce the data with sufficient accuracy. After implementing a recently developed correlation and corresponding heat transfer regime map into GAMMA to cover the whole range of the DTHT regime, we obtained better agreement with the data. RELAP5-3D results are discussed in parallel.

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-09-28

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

PLOTLIB: a computerized nuclear waste source-term library storage and retrieval system

International Nuclear Information System (INIS)

Marshall, J.R.; Nowicki, J.A.

1978-01-01

The PLOTLIB code was written to provide computer access to the Nuclear Waste Source-Term Library for those users with little previous computer programming experience. The principles of user orientation, quick accessibility, and versatility were extensively employed in the development of the PLOTLIB code to accomplish this goal. The Nuclear Waste Source-Term Library consists of 16 ORIGEN computer runs incorporating a wide variety of differing light water reactor (LWR) fuel cycles and waste streams. The typical isotopic source-term data consist of information on watts, curies, grams, etc., all of which are compiled as a function of time after reactor discharge and unitized on a per metric ton heavy metal basis. The information retrieval code, PLOTLIB, is used to process source-term information requests into computer plots and/or user-specified output tables. This report will serve both as documentation of the current data library and as an operations manual for the PLOTLIB computer code. The accompanying input description, program listing, and sample problems make this code package an easily understood tool for the various nuclear waste studies under way at the Office of Waste Isolation

Thermal Analysis of the SAFKEG Package for Long Term Storage

International Nuclear Information System (INIS)

NARENDRA, GUPTA

2005-01-01

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

“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)

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

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

Design Aspects of the Rayleigh Convection Code

Science.gov (United States)

Featherstone, N. A.

2017-12-01

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

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

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

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.)

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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

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

International Nuclear Information System (INIS)

2004-06-01

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

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.

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.)

The environmental effects of taxes on packages

International Nuclear Information System (INIS)

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

2010-08-01

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

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.

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

Application of polymer nanocomposite materials in food packaging

Directory of Open Access Journals (Sweden)

Amra Odobašić

2015-01-01

Full Text Available The term “nano” refers to nano particle size from 1 to 100 nanometers. The term "nanotechnology" was first introduced by Norio Taniguchi in 1974. Nanotechnology may be used to improve the taste and texture of food and for the production of packaging that maintain fresh product. The primary function of packaging is to maintain the quality and safety of products during transport and storage period, as well as to extend its viability by preventing unwanted effect agents such as microorganisms, chemical contaminants, oxygen, moisture and light. The aim of this paper is to point out the achievements of nanotechnology in terms of food packaging with an overview of polymers that are commonly used in food packaging, as well as strategies to improve the physical properties of polymers, including mechanical strength, thermal stability and barrier to gases. By studing of recently published literature, it was clear that nanomaterials such as nano polymers are trying to replace conventional materials in food packaging. Nanosensors can be used to prove the presence of contaminants, microtoxins and microorganisms in food.

The Barrier code for predicting long-term concrete performance

International Nuclear Information System (INIS)

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

1989-01-01

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

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

The Analytical Repository Source-Term (AREST) model: Description and documentation

International Nuclear Information System (INIS)

Liebetrau, A.M.; Apted, M.J.; Engel, D.W.; Altenhofen, M.K.; Strachan, D.M.; Reid, C.R.; Windisch, C.F.; Erikson, R.L.; Johnson, K.I.

1987-10-01

The geologic repository system consists of several components, one of which is the engineered barrier system. The engineered barrier system interfaces with natural barriers that constitute the setting of the repository. A model that simulates the releases from the engineered barrier system into the natural barriers of the geosphere, called a source-term model, is an important component of any model for assessing the overall performance of the geologic repository system. The Analytical Repository Source-Term (AREST) model being developed is one such model. This report describes the current state of development of the AREST model and the code in which the model is implemented. The AREST model consists of three component models and five process models that describe the post-emplacement environment of a waste package. All of these components are combined within a probabilistic framework. The component models are a waste package containment (WPC) model that simulates the corrosion and degradation processes which eventually result in waste package containment failure; a waste package release (WPR) model that calculates the rates of radionuclide release from the failed waste package; and an engineered system release (ESR) model that controls the flow of information among all AREST components and process models and combines release output from the WPR model with failure times from the WPC model to produce estimates of total release. 167 refs., 40 figs., 12 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.

CACTI: free, open-source software for the sequential coding of behavioral interactions.

Science.gov (United States)

Glynn, Lisa H; Hallgren, Kevin A; Houck, Jon M; Moyers, Theresa B

2012-01-01

The sequential analysis of client and clinician speech in psychotherapy sessions can help to identify and characterize potential mechanisms of treatment and behavior change. Previous studies required coding systems that were time-consuming, expensive, and error-prone. Existing software can be expensive and inflexible, and furthermore, no single package allows for pre-parsing, sequential coding, and assignment of global ratings. We developed a free, open-source, and adaptable program to meet these needs: The CASAA Application for Coding Treatment Interactions (CACTI). Without transcripts, CACTI facilitates the real-time sequential coding of behavioral interactions using WAV-format audio files. Most elements of the interface are user-modifiable through a simple XML file, and can be further adapted using Java through the terms of the GNU Public License. Coding with this software yields interrater reliabilities comparable to previous methods, but at greatly reduced time and expense. CACTI is a flexible research tool that can simplify psychotherapy process research, and has the potential to contribute to the improvement of treatment content and delivery.

Documentation of the seawater intrusion (SWI2) package for MODFLOW

Science.gov (United States)

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

2013-01-01

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

Final report of the inter institutional project ININ-CNSNS 'Source Terms specific for the CNLV'

International Nuclear Information System (INIS)

Anaya M, R.A.

1991-02-01

The purpose of the project inter institutional ININ-CNSNS 'Source Terms Specifies for the CNLV' it is the one of implanting in the computer CYBER (CDC 180-830) of the ININ, the 'Source Term Code Package' (STCP) and to make the operation tests and corresponding operation using the data of the sample problem, for finally to liberate the package, all time that by means of the analysis of the results it is consider appropriate. In this report the results of the are presented simulation of the sequence 'Energy Losses external' (Station blackout) and 'Lost total of CA with failure of the RCIC and success of the HPCS' both with data of the Laguna Verde Central. (Author)

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)

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

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

International Nuclear Information System (INIS)

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

1997-03-01

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

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

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

Warpage of QFN Package in Post Mold Cure Process of integrated circuit packaging

Science.gov (United States)

Sriwithoon, Nattha; Ugsornrat, Kessararat; Srisuwitthanon, Warayoot; Thonglor, Panakamon

2017-09-01

This research studied about warpage of QFN package in post mold cure process of integrated circuit (IC) packages using pre-plated (PPF) leadframe. For IC package, epoxy molding compound (EMC) are molded by cross linking of compound stiffness but incomplete crosslinked network and leading the fully cured thermoset by post mold cure (PMC) process. The cure temperature of PMC can change microstructure of EMC in term of stress inside the package and effect to warpage of the package due to coefficient of thermal expansion (CTE) between EMC and leadframe. In experiment, cure temperatures were varied to check the effect of internal stress due to different cure temperature after completed post mold cure for TDFN 2×3 8L. The cure temperature were varied with 180 °C, 170 °C, 160 °C, and 150°C with cure time 4 and 6 hours, respectively. For analysis, the TDFN 2×3 8L packages were analyzed the warpage by thickness gauge and scanning acoustic microscope (SAM) after take the test samples out from the oven cure. The results confirmed that effect of different CTE between EMC and leadframe due to different cure temperature resulting to warpage of the TDFN 2×3 8L packages.

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

Directory of Open Access Journals (Sweden)

Song Jeongook

2010-01-01

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

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

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

Aspects of safety assessments for package with additional equipment components

International Nuclear Information System (INIS)

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

2004-01-01

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

actuar: An R Package for Actuarial Science

Directory of Open Access Journals (Sweden)

Christophe Dutang

2008-02-01

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

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

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

Science.gov (United States)

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

2013-03-01

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

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

Low-level radioactive waste performance assessments: Source term modeling

International Nuclear Information System (INIS)

Icenhour, A.S.; Godbee, H.W.; Miller, L.F.

1995-01-01

Low-level radioactive wastes (LLW) generated by government and commercial operations need to be isolated from the environment for at least 300 to 500 yr. Most existing sites for the storage or disposal of LLW employ the shallow-land burial approach. However, the U.S. Department of Energy currently emphasizes the use of engineered systems (e.g., packaging, concrete and metal barriers, and water collection systems). Future commercial LLW disposal sites may include such systems to mitigate radionuclide transport through the biosphere. Performance assessments must be conducted for LUW disposal facilities. These studies include comprehensive evaluations of radionuclide migration from the waste package, through the vadose zone, and within the water table. Atmospheric transport mechanisms are also studied. Figure I illustrates the performance assessment process. Estimates of the release of radionuclides from the waste packages (i.e., source terms) are used for subsequent hydrogeologic calculations required by a performance assessment. Computer models are typically used to describe the complex interactions of water with LLW and to determine the transport of radionuclides. Several commonly used computer programs for evaluating source terms include GWSCREEN, BLT (Breach-Leach-Transport), DUST (Disposal Unit Source Term), BARRIER (Ref. 5), as well as SOURCE1 and SOURCE2 (which are used in this study). The SOURCE1 and SOURCE2 codes were prepared by Rogers and Associates Engineering Corporation for the Oak Ridge National Laboratory (ORNL). SOURCE1 is designed for tumulus-type facilities, and SOURCE2 is tailored for silo, well-in-silo, and trench-type disposal facilities. This paper focuses on the source term for ORNL disposal facilities, and it describes improved computational methods for determining radionuclide transport from waste packages

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)

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.

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.

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

OpenAIRE

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

2012-01-01

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

Naval Waste Package Design Sensitivity

International Nuclear Information System (INIS)

T. Schmitt

2006-01-01

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

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

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.

Physical and chemical characteristics of cashew nut flour stored and packaged with different packages

Directory of Open Access Journals (Sweden)

Bruna Carolina GADANI

Full Text Available Abstract The aim of this study was to evaluate the physical and chemical changes in cashew nut flour under different packaging and storage conditions. The flour samples were characterized according to their chemical composition, packaged in polystyrene trays associated with covering chloride polyvinyl (PVC, plastic pot of polyethylene terephthalate (PET, plastic packaging polyethylene (POL and without coating polystyrene trays (CON. All packages were stored for 225 days, evaluating every 45 days the moisture content of the flour as well as extracting their oil by the cold method for characterization on the following indexes: acidity, saponification, iodine, peroxide and refraction. There was slight and gradual increase in the moisture content from the 45th to the 225th day, especially for products packaged in PVC and without coating polystyrene trays (CON. The oil saponification indexes showed gradual decrease during storage, especially in flour without packaging (CON. There was a gradual increase in the iodine index until the 135th day, followed by decay. The acidity and peroxide indexes increased along with the storage time. However packages with lower oxygen and moisture permeability, such as PET and POL, minimized such changes, being therefore the most suitable for long-term storage.

Particle and heavy ion transport code system, PHITS, version 2.52

International Nuclear Information System (INIS)

Sato, Tatsuhiko; Matsuda, Norihiro; Hashimoto, Shintaro; Iwamoto, Yosuke; Noda, Shusaku; Ogawa, Tatsuhiko; Nakashima, Hiroshi; Fukahori, Tokio; Okumura, Keisuke; Kai, Tetsuya; Niita, Koji; Iwase, Hiroshi; Chiba, Satoshi; Furuta, Takuya; Sihver, Lembit

2013-01-01

An upgraded version of the Particle and Heavy Ion Transport code System, PHITS2.52, was developed and released to the public. The new version has been greatly improved from the previously released version, PHITS2.24, in terms of not only the code itself but also the contents of its package, such as the attached data libraries. In the new version, a higher accuracy of simulation was achieved by implementing several latest nuclear reaction models. The reliability of the simulation was improved by modifying both the algorithms for the electron-, positron-, and photon-transport simulations and the procedure for calculating the statistical uncertainties of the tally results. Estimation of the time evolution of radioactivity became feasible by incorporating the activation calculation program DCHAIN-SP into the new package. The efficiency of the simulation was also improved as a result of the implementation of shared-memory parallelization and the optimization of several time-consuming algorithms. Furthermore, a number of new user-support tools and functions that help users to intuitively and effectively perform PHITS simulations were developed and incorporated. Due to these improvements, PHITS is now a more powerful tool for particle transport simulation applicable to various research and development fields, such as nuclear technology, accelerator design, medical physics, and cosmic-ray research. (author)

COMPASS: A source term code for investigating capillary barrier performance

International Nuclear Information System (INIS)

Zhou, Wei; Apted, J.J.

1996-01-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

XSOR codes users manual

International Nuclear Information System (INIS)

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

1993-11-01

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

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

actuar: An R Package for Actuarial Science

OpenAIRE

Dutang, Christophe; Goulet, Vincent; Pigeon, Mathieu

2008-01-01

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-10-15

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

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

International Nuclear Information System (INIS)

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

1993-07-01

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

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)

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.

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

Science.gov (United States)

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

1975-03-01

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

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)

ARM Data-Oriented Metrics and Diagnostics Package for Climate Model Evaluation Value-Added Product

Energy Technology Data Exchange (ETDEWEB)

Zhang, Chengzhu [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Xie, Shaocheng [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2017-10-15

A Python-based metrics and diagnostics package is currently being developed by the U.S. Department of Energy (DOE) Atmospheric Radiation Measurement (ARM) Infrastructure Team at Lawrence Livermore National Laboratory (LLNL) to facilitate the use of long-term, high-frequency measurements from the ARM Facility in evaluating the regional climate simulation of clouds, radiation, and precipitation. This metrics and diagnostics package computes climatological means of targeted climate model simulation and generates tables and plots for comparing the model simulation with ARM observational data. The Coupled Model Intercomparison Project (CMIP) model data sets are also included in the package to enable model intercomparison as demonstrated in Zhang et al. (2017). The mean of the CMIP model can serve as a reference for individual models. Basic performance metrics are computed to measure the accuracy of mean state and variability of climate models. The evaluated physical quantities include cloud fraction, temperature, relative humidity, cloud liquid water path, total column water vapor, precipitation, sensible and latent heat fluxes, and radiative fluxes, with plan to extend to more fields, such as aerosol and microphysics properties. Process-oriented diagnostics focusing on individual cloud- and precipitation-related phenomena are also being developed for the evaluation and development of specific model physical parameterizations. The version 1.0 package is designed based on data collected at ARM’s Southern Great Plains (SGP) Research Facility, with the plan to extend to other ARM sites. The metrics and diagnostics package is currently built upon standard Python libraries and additional Python packages developed by DOE (such as CDMS and CDAT). The ARM metrics and diagnostic package is available publicly with the hope that it can serve as an easy entry point for climate modelers to compare their models with ARM data. In this report, we first present the input data, which

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

Waste package characterisation

Energy Technology Data Exchange (ETDEWEB)

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

1998-09-01

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

Waste package characterisation

International Nuclear Information System (INIS)

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

1998-09-01

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

Destructive examination of shipping package 9975-02101

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-01

Destructive and non-destructive examinations have been performed on the components of shipping package 9975-02101 as part of the comprehensive Model 9975 package surveillance program. This package is one of ten high-wattage packages that were selected for field surveillance in FY15, and was identified to contain several non-conforming conditions. Most of these conditions (mold, stains, drum corrosion, calculated fiberboard dimensions and fiberboard damage) relate to the accumulation of water in the outer and lower portions of the cane fiberboard assembly. In the short term, this causes local but reversible changes in the fiberboard properties. Long-term effects can include the permanent loss of fiberboard properties (thus far observed only in the bottom fiberboard layers) and reduced drum integrity due to corrosion. The observed conditions must be fully evaluated by KAC to ensure the safety function of the package is being maintained. Three of the other nine FY15 high-wattage packages examined in the K-Area Complex showed similar behavior. Corrosion of the overpack drum has been seen primarily in those packages with relatively severe fiberboard degradation. Visual examination of the drums in storage for external corrosion should be considered as a screening tool to identify additional packages with potential fiberboard degradation. Where overpack drum corrosion has been observed, it is typically heaviest adjacent to the stitch welds along the bottom edge. It is possible that changes to the stitch weld design would reduce the degree of corrosion in this area, but would not eliminate it. Several factors can contribute to the concentration of moisture in the fiberboard, including higher than average initial moisture content, higher internal temperature (due to internal heat load and placement with the array of packages), and the creation of additional moisture as the fiberboard begins to degrade.

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.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-06-01

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

Speech coding

Energy Technology Data Exchange (ETDEWEB)

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

1998-05-08

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

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)

Magnus: A New Resistive MHD Code with Heat Flow Terms

Science.gov (United States)

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

2017-07-01

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

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)

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.)

VizieR Online Data Catalog: FARGO_THORIN 1.0 hydrodynamic code (Chrenko+, 2017)

Science.gov (United States)

Chrenko, O.; Broz, M.; Lambrechts, M.

2017-07-01

This archive contains the source files, documentation and example simulation setups of the FARGO_THORIN 1.0 hydrodynamic code. The program was introduced, described and used for simulations in the paper. It is built on top of the FARGO code (Masset, 2000A&AS..141..165M, Baruteau & Masset, 2008ApJ...672.1054B) and it is also interfaced with the REBOUND integrator package (Rein & Liu, 2012A&A...537A.128R). THORIN stands for Two-fluid HydrOdynamics, the Rebound integrator Interface and Non-isothermal gas physics. The program is designed for self-consistent investigations of protoplanetary systems consisting of a gas disk, a disk of small solid particles (pebbles) and embedded protoplanets. Code features: I) Non-isothermal gas disk with implicit numerical solution of the energy equation. The implemented energy source terms are: Compressional heating, viscous heating, stellar irradiation, vertical escape of radiation, radiative diffusion in the midplane and radiative feedback to accretion heating of protoplanets. II) Planets evolved in 3D, with close encounters allowed. The orbits are integrated using the IAS15 integrator (Rein & Spiegel, 2015MNRAS.446.1424R). The code detects the collisions among planets and resolve them as mergers. III) Refined treatment of the planet-disk gravitational interaction. The code uses a vertical averaging of the gravitational potential, as outlined in Muller & Kley (2012A&A...539A..18M). IV) Pebble disk represented by an Eulerian, presureless and inviscid fluid. The pebble dynamics is affected by the Epstein gas drag and optionally by the diffusive effects. We also implemented the drag back-reaction term into the Navier-Stokes equation for the gas. Archive summary: ------------------------------------------------------------------------- directory/file Explanation ------------------------------------------------------------------------- /in_relax Contains setup of the first example simulation /in_wplanet Contains setup of the second

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

International Nuclear Information System (INIS)

Kress, T.S.

1985-04-01

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

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.

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.

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)

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

Development of polymer packaging for power cable

Directory of Open Access Journals (Sweden)

S. Sremac

2014-10-01

Full Text Available This paper discusses the issues of product design and the procedure of developing polymer packaging as one of the most important engineering tasks. For the purpose of packing power cables a polymer packaging has been designed in the form of drum. Packaging and many other consumer products are largely produced using polymeric materials due to many positive features. High Density Polyethylene is the type of polyethylene proposed for packaging purposes due to its low degree of branching and strong intermolecular forces. Transport and storage processes were automated based on the radio-frequency identification technology. The proposed system is flexible in terms of its possibility of accepting and processing different types of cables and other products.

Peer Review of the Waste Package Material Performance Interim Report

International Nuclear Information System (INIS)

J. A. Beavers; T. M. Devine, Jr.; G. S. Frankel; R. H. Jones; R. G. Kelly; R. M. Latanision; J. H. Payer

2001-01-01

At the request of the U.S. Department of Energy, Bechtel SAIC Company, LLC, formed the Waste Package Materials Performance Peer Review Panel (the Panel) to review the technical basis for evaluating the long-term performance of waste package materials in a proposed repository at Yucca Mountain, Nevada. This is the interim report of the Panel; a final report will be issued in February 2002. In its work to date, the Panel has identified important issues regarding waste package materials performance. In the remainder of its work, the Panel will address approaches and plans to resolve these issues. In its review to date, the Panel has not found a technical basis to conclude that the waste package materials are unsuitable for long-term containment at the proposed Yucca Mountain Repository. Nevertheless, significant technical issues remain unsettled and, primarily because of the extremely long life required for the waste packages, there will always be some uncertainty in the assessment. A significant base of scientific and engineering knowledge for assessing materials performance does exist and, therefore, the likelihood is great that uncertainty about the long-term performance can be substantially reduced through further experiments and analysis

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.

Development of a general model to predict the rate of radionuclide release (source term) from a low-level waste shallow land burial facility

International Nuclear Information System (INIS)

Sullivan, T.M.; Kempf, C.R.; Suen, C.J.; Mughabghab, S.M.

1988-01-01

Federal Code of Regulations 10 CFR 61 requires that any near surface disposal site be capable of being characterized, analyzed, and modeled. The objective of this program is to assist NRC in developing the ability to model a disposal site that conforms to these regulations. In particular, a general computer model capable of predicting the quantity and rate of radionuclide release from a shallow land burial trench, i.e., the source term, is being developed. The framework for this general model has been developed and consists of four basic compartments that represent the major processes that influence release. These compartments are: water flow, container degradation, release from the waste packages, and radionuclide transport. Models for water flow and radionuclide transport rely on the use of the computer codes FEMWATER and FEMWASTE. These codes are generally regarded as being state-of-the-art and required little modification for their application to this project. Models for container degradation and release from waste packages have been specifically developed for this project. This paper provides a brief description of the models being used in the source term project and examples of their use over a range of potential conditions. 13 refs

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

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)

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

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

Software and the Scientist: Coding and Citation Practices in Geodynamics

Science.gov (United States)

Hwang, Lorraine; Fish, Allison; Soito, Laura; Smith, MacKenzie; Kellogg, Louise H.

2017-11-01

In geodynamics as in other scientific areas, computation has become a core component of research, complementing field observation, laboratory analysis, experiment, and theory. Computational tools for data analysis, mapping, visualization, modeling, and simulation are essential for all aspects of the scientific workflow. Specialized scientific software is often developed by geodynamicists for their own use, and this effort represents a distinctive intellectual contribution. Drawing on a geodynamics community that focuses on developing and disseminating scientific software, we assess the current practices of software development and attribution, as well as attitudes about the need and best practices for software citation. We analyzed publications by participants in the Computational Infrastructure for Geodynamics and conducted mixed method surveys of the solid earth geophysics community. From this we learned that coding skills are typically learned informally. Participants considered good code as trusted, reusable, readable, and not overly complex and considered a good coder as one that participates in the community in an open and reasonable manor contributing to both long- and short-term community projects. Participants strongly supported citing software reflected by the high rate a software package was named in the literature and the high rate of citations in the references. However, lacking are clear instructions from developers on how to cite and education of users on what to cite. In addition, citations did not always lead to discoverability of the resource. A unique identifier to the software package itself, community education, and citation tools would contribute to better attribution practices.

Nuclear-waste-package materials degradation modes and accelerated testing

International Nuclear Information System (INIS)

1981-09-01

This report reviews the materials degradation modes that may affect the long-term behavior of waste packages for the containment of nuclear waste. It recommends an approach to accelerated testing that can lead to the qualification of waste package materials in specific repository environments in times that are short relative to the time period over which the waste package is expected to provide containment. This report is not a testing plan but rather discusses the direction for research that might be considered in developing plans for accelerated testing of waste package materials and waste forms

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.

Young people's perceptions of cigarette packaging and plain packaging: an online survey.

Science.gov (United States)

Moodie, Crawford; Ford, Allison; Mackintosh, Anne Marie; Hastings, Gerard

2012-01-01

In the United Kingdom, with most marketing channels prohibited, packaging is one of the few remaining ways that tobacco companies can promote their products. An online survey with young people aged 10-17 years (N = 658) was used to explore why youth choose cigarettes, perceptions of pack color, and perceptions of plain (nonbranded) cigarette packaging. Young people were also shown an image of 3 plain packs, which differed by shape and method of opening, and asked which they liked most and thought others their age would smoke. Price and what significant others smoke were key factors for choosing cigarettes, with packaging also an important influence. More than a third of the sample associated lighter pack color with weak tasting and less harmful cigarettes. Plain packs were rated negatively as were perceptions of plain pack users. One in 3 showed a preference for either a narrow "perfume type" plain pack or a plain "slide" pack that opened from the side, and 1 in 3 also thought that young people would smoke these packs. Packaging appears to both attract young people and mislead them about product strength and relative harm. Innovative pack construction (novel pack shape and method of opening) and the use of color are instrumental in these effects. The findings therefore suggest that any move to plain packaging should not only consider the benefits of removing branding (including color) but also of standardizing pack construction in terms of shape and method of opening.

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.

SCALE: A modular code system for performing standardized computer analyses for licensing evaluation: Functional modules F1-F8

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 volume consists of the section of the manual dealing with eight of the functional modules in the code. Those are: BONAMI - resonance self-shielding by the Bondarenko method; NITAWL-II - SCALE system module for performing resonance shielding and working library production; XSDRNPM - a one-dimensional discrete-ordinates code for transport analysis; XSDOSE - a module for calculating fluxes and dose rates at points outside a shield; KENO IV/S - an improved monte carlo criticality program; COUPLE; ORIGEN-S - SCALE system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms; ICE

SCALE: A modular code system for performing standardized computer analyses for licensing evaluation: Functional modules F1-F8

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 volume consists of the section of the manual dealing with eight of the functional modules in the code. Those are: BONAMI - resonance self-shielding by the Bondarenko method; NITAWL-II - SCALE system module for performing resonance shielding and working library production; XSDRNPM - a one-dimensional discrete-ordinates code for transport analysis; XSDOSE - a module for calculating fluxes and dose rates at points outside a shield; KENO IV/S - an improved monte carlo criticality program; COUPLE; ORIGEN-S - SCALE system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms; ICE.

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...

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.

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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)

Cigarette package design: opportunities for disease prevention.

Science.gov (United States)

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

2002-06-15

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

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.

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.

Short-term memory coding in children with intellectual disabilities.

Science.gov (United States)

Henry, Lucy

2008-05-01

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

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

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

Recent developments in intelligent packaging for enhancing food quality and safety.

Science.gov (United States)

Sohail, Muhammad; Sun, Da-Wen; Zhu, Zhiwei

2018-03-07

The role of packaging cannot be denied in the life cycle of any food product. Intelligent packaging is an emerging technology in the food packaging sector. Although it still needs its full emergence in the market, its importance has been proved for the maintenance of food quality and safety. The present review describes several aspects of intelligent packaging. It first highlights different tools used in intelligent packaging and elucidates the role of these packaging devices for maintaining the quality of different food items in terms of controlling microbial growth and gas concentration, and for providing convenience and easiness to its users in the form of time temperature indication. This review also discusses other intelligent packaging solutions in supply chain management of food products to control theft and counterfeiting conducts and broaden the image of the food companies in terms of branding and marketing. Overall, intelligent packaging can ensure food quality and safety in the food industry, however there are still some concerns over this emerging technology including high cost and legal aspects, and thus future work should be performed to overcome these problems for further promoting its applications in the food industry. Moreover, work should also be carried out to combine several single intelligent packaging devices into a single one, so that most of the benefits from this emerging technology can be achieved.

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.

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

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

Progress toward NuPack, the ASME code for Type B containments

International Nuclear Information System (INIS)

Turula, P.

1995-01-01

This paper presented a brief status report on the development of an ASME Code Division for nuclear packaging and discussed some of the more interesting policy decisions as to what is and is not covered in terms of analytical methods, criteria, scope, and other aspects. The process of the development of this Division has been very slow and inconsistent. There were many participants with many diverse interests. The Division 3 rules are close to being ready to be issued. They are a compromise between many needs and the result is certainly not perfect. Opportunities for fine tuning and expanding this document will present themselves after it is issued as future needs become clear

Crisis, Stimulus Package and Migration in China

NARCIS (Netherlands)

Csanádi, Maria; Nie, Zihan; Li, Shi

2015-01-01

This paper analyzes the short-term and long-term effects that the global economic crisis and the investment priorities of the Chinese Government's stimulus package had on Chinese migrant flows between 2008 and 2014. Combining micro-level household survey data and macro-level statistics, the

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.

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

Science.gov (United States)

Schultz, Wolfram

2004-04-01

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

Cigarette package design: opportunities for disease prevention

Directory of Open Access Journals (Sweden)

Pollay RW

2003-01-01

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

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

Energy conversion of source separated packaging; Energiutvinning ur kaellsorterade foerpackningsfraktioner

Energy Technology Data Exchange (ETDEWEB)

Blidholm, O; Wiklund, S E [AaF-Energikonsult (Sweden); Bauer, A C [Energikonsult A. Bauer (Sweden)

1997-02-01

The basic idea of this project is to study the possibilities to use source separated combustible material for energy conversion in conventional solid fuel boilers (i.e. not municipal waste incineration plants). The project has been carried out in three phases. During phase 1 and 2 a number of fuel analyses of different fractions were carried out. During phase 3 two combustion tests were carried out; (1) a boiler with grate equipped with cyclone, electrostatic precipitator and flue gas condenser, and (2) a bubbling fluidized bed boiler with electrostatic precipitator and flue gas condenser. During the tests source separated paper and plastic packagings were co-fired with biomass fuels. The mixing rate of packagings was approximately 15%. This study reports the results of phase 3 and the conclusions of the whole project. The technical terms of using packaging as fuel are good. The technique is available for shredding both paper and plastic packaging. The material can be co-fired with biomass. The economical terms of using source separated packaging for energy conversion can be very advantageous, but can also form obstacles. The result is to a high degree guided by such facts as how the fuel is collected, transported, reduced in size and handled at the combustion plant. The results of the combustion tests show that the environmental terms of using source separated packaging for energy conversion are good. The emissions of heavy metals into the atmosphere are very low. The emissions are well below the emission standards for waste incineration plants. 35 figs, 13 tabs, 8 appendices

Topical report on actinide-only burnup credit for PWR spent nuclear fuel packages. Revision 1

Energy Technology Data Exchange (ETDEWEB)

None, None

1997-04-01

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{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub 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

airGR: an R-package suitable for large sample hydrology presenting a suite of lumped hydrological models

Science.gov (United States)

Thirel, G.; Delaigue, O.; Coron, L.; Perrin, C.; Andreassian, V.

2016-12-01

Lumped hydrological models are useful and convenient tools for research, engineering and educational purposes. They propose catchment-scale representations of the precipitation-discharge relationship. Thanks to their limited data requirements, they can be easily implemented and run. With such models, it is possible to simulate a number of hydrological key processes over the catchment with limited structural and parametric complexity, typically evapotranspiration, runoff, underground losses, etc. The Hydrology Group at Irstea (Antony) has been developing a suite of rainfall-runoff models over the past 30 years with the main objectives of designing models as efficient as possible in terms of streamflow simulation, applicable to a wide range of catchments and having low data requirements. This resulted in a suite of models running at different time steps (from hourly to annual) applicable for various issues including water balance estimation, forecasting, simulation of impacts and scenario testing. Recently, Irstea has developed an easy-to-use R-package (R Core Team, 2015; Coron et al., 2016), called airGR, to make these models widely available. It includes: - the water balance annual GR1A (Mouehli et al., 2006), - the monthly GR2M (Mouehli, 2003) models, - three versions of the daily model, namely GR4J (Perrin et al., 2003), GR5J (Le Moine, 2008) and GR6J (Pushpalatha et al., 2011), - the hourly GR4H model (Mathevet, 2005), - a degree-day snow module CemaNeige (Valéry et al., 2014). The airGR package has been designed to facilitate the use by non-expert users and allow the addition of evaluation criteria, models or calibration algorithm selected by the end-user. Each model core is coded in FORTRAN to ensure low computational time. The other package functions (i.e. mainly the calibration algorithm and the efficiency criteria) are coded in R. The package is already used for educational purposes. It allows for convenient implementation of model inter-comparisons and

network: A Package for Managing Relational Data in R

Directory of Open Access Journals (Sweden)

Carter T. Butts

2007-12-01

Full Text Available Effective memory structures for relational data within R must be capable of representing a wide range of data while keeping overhead to a minimum. The network package provides an class which may be used for encoding complex relational structures composed a vertex set together with any combination of undirected/directed, valued/unvalued, dyadic/hyper, and single/multiple edges; storage requirements are on the order of the number of edges involved. Some simple constructor, interface, and visualization functions are provided, as well as a set of operators to facilitate employment by end users. The package also supports a C-language API, which allows developers to work directly with network objects within backend code.

Packaging supplier inspection guide

International Nuclear Information System (INIS)

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

1991-05-01

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

Grandfathering of competent authority approved packages

International Nuclear Information System (INIS)

Osgood, N.L.

2004-01-01

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

EQ3/6, a software package for geochemical modeling of aqueous systems: Package overview and installation guide (Version 7.0)

International Nuclear Information System (INIS)

Wolery, T.J.

1992-01-01

EQ3/6 is a software package for geochemical modeling of aqueous systems. This report describes version 7.0. The major components of the package include: EQ3NR, a speciation-solubility code; EQ6, a reaction path code which models water/rock interaction or fluid mixing in either a pure reaction progress mode or a time mode; EQPT, a data file preprocessor, EQLIB, a supporting software library; and five supporting thermodynamic data files. The software deals with the concepts of thermodynamic equilibrium, thermodynamic disequilibrium, and reaction kinetics. The five supporting data files contain both standard state and activity coefficient-related data. Three support the use of the Davies or B equations for the activity coefficients; the other two support the use of Pitzer's equations. The temperature range of the thermodynamic data on the data files varies from 25 degree C only to 0--300 degree C. EQPT takes a formatted data file (a data0 file) and writes an unformatted near-equivalent called a datal file, which is actually the form read by EQ3NR and EQ6. EQ3NR is useful for analyzing groundwater chemistry data, calculating solubility limits, and determining whether certain reactions are in states of partial equilibrium or disequilibrium. It is also required to initialize an EQ6 calculation. EQ6 models the consequences of reacting an aqueous solution with a set of reactants which react irreversibly. It can also model fluid mixing and the consequences of changes in temperature. This code operates both in a pure reaction progress frame and in a time frame

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

Rapid installation of numerical models in multiple parent codes

Energy Technology Data Exchange (ETDEWEB)

Brannon, R.M.; Wong, M.K.

1996-10-01

A set of``model interface guidelines``, called MIG, is offered as a means to more rapidly install numerical models (such as stress-strain laws) into any parent code (hydrocode, finite element code, etc.) without having to modify the model subroutines. The model developer (who creates the model package in compliance with the guidelines) specifies the model`s input and storage requirements in a standardized way. For portability, database management (such as saving user inputs and field variables) is handled by the parent code. To date, NUG has proved viable in beta installations of several diverse models in vectorized and parallel codes written in different computer languages. A NUG-compliant model can be installed in different codes without modifying the model`s subroutines. By maintaining one model for many codes, MIG facilitates code-to-code comparisons and reduces duplication of effort potentially reducing the cost of installing and sharing models.

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Powerlaw: a Python package for analysis of heavy-tailed distributions.

Science.gov (United States)

Alstott, Jeff; Bullmore, Ed; Plenz, Dietmar

2014-01-01

Power laws are theoretically interesting probability distributions that are also frequently used to describe empirical data. In recent years, effective statistical methods for fitting power laws have been developed, but appropriate use of these techniques requires significant programming and statistical insight. In order to greatly decrease the barriers to using good statistical methods for fitting power law distributions, we developed the powerlaw Python package. This software package provides easy commands for basic fitting and statistical analysis of distributions. Notably, it also seeks to support a variety of user needs by being exhaustive in the options available to the user. The source code is publicly available and easily extensible.

Powerlaw: a Python package for analysis of heavy-tailed distributions.

Directory of Open Access Journals (Sweden)

Jeff Alstott

Full Text Available Power laws are theoretically interesting probability distributions that are also frequently used to describe empirical data. In recent years, effective statistical methods for fitting power laws have been developed, but appropriate use of these techniques requires significant programming and statistical insight. In order to greatly decrease the barriers to using good statistical methods for fitting power law distributions, we developed the powerlaw Python package. This software package provides easy commands for basic fitting and statistical analysis of distributions. Notably, it also seeks to support a variety of user needs by being exhaustive in the options available to the user. The source code is publicly available and easily extensible.