Group cross sections calculations
International Nuclear Information System (INIS)
Just a few methods have been developped to compute multigroup cross-sections from ENDF data. We have developped an original method in order to get accuracy and to reduce the number of discretization points in the same time; this is why we have tried to use polynomial integration. In this paper, we describe this method: in the first part, we recall some physical hypothesis generally used to solve the linear Boltzmann equation: that is the frame in which the numerical method has been developped. Polynomial methods are really powerfull only if discretization points are suitably chosen. This choice is explained in the next part of this paper. In conclusion, some numerical results are given to illustrate our method
Daskalov, George M; Baker, R S; Rogers, D W O; Williamson, J F
2002-02-01
Our purpose in this work is to demonstrate that the efficiency of dose-rate computations in 125I brachytherapy, using multigroup discrete ordinates radiation transport simulations, can be significantly enhanced using broad energy group cross sections without a loss of accuracy. To this end, the DANTSYS multigroup discrete ordinates neutral particle transport code was used to estimate the absorbed dose-rate distributions around an 125I-model 6702 seed in two-dimensional (2-D) cylindrical R-Z geometry for four different problems spanning the geometries found in clinical practice. First, simulations with a high resolution 210 energy groups library were used to analyze the photon flux spectral distribution throughout this set of problems. These distributions were used to design an energy group structure consisting of three broad groups along with suitable weighting functions from which the three-group cross sections were derived. The accuracy of 2-D DANTSYS dose-rate calculations was benchmarked against parallel Monte Carlo simulations. Ray effects were remedied by using the DANTSYS internal first collision source algorithm. It is demonstrated that the 125I primary photon spectrum leads to inappropriate weighting functions. An accuracy of +/-5% is achieved in the four problem geometries considered using geometry-independent three-group libraries derived from either material-specific weighting functions or a single material-independent weighting function. Agreement between Monte Carlo and the three-group DANTSYS calculations, within three standard Monte Carlo deviations, is observed everywhere except for a limited region along the Z axis of rotational symmetry, where ray effects are difficult to mitigate. The three-group DANTSYS calculations are 10-13 times faster than ones with a 210-group cross section library for 125I dosimetry problems. Compared to 2-D EGS4 Monte Carlo calculations, the 3-group DANTSYS simulations are a 100-fold more efficient. Provided that these
International Nuclear Information System (INIS)
Our purpose in this work is to demonstrate that the efficiency of dose-rate computations in 125I brachytherapy, using multigroup discrete ordinates radiation transport simulations, can be significantly enhanced using broad energy group cross sections without a loss of accuracy. To this end, the DANTSYS multigroup discrete ordinates neutral particle transport code was used to estimate the absorbed dose-rate distributions around an 125I-model 6702 seed in two-dimensional (2-D) cylindrical R-Z geometry for four different problems spanning the geometries found in clinical practice. First, simulations with a high resolution 210 energy groups library were used to analyze the photon flux spectral distribution throughout this set of problems. These distributions were used to design an energy group structure consisting of three broad groups along with suitable weighting functions from which the three-group cross sections were derived. The accuracy of 2-D DANTSYS dose-rate calculations was benchmarked against parallel Monte Carlo simulations. Ray effects were remedied by using the DANTSYS internal first collision source algorithm. It is demonstrated that the 125I primary photon spectrum leads to inappropriate weighting functions. An accuracy of ±5% is achieved in the four problem geometries considered using geometry-independent three-group libraries derived from either material-specific weighting functions or a single material-independent weighting function. Agreement between Monte Carlo and the three-group DANTSYS calculations, within three standard Monte Carlo deviations, is observed everywhere except for a limited region along the Z axis of rotational symmetry, where ray effects are difficult to mitigate. The three-group DANTSYS calculations are 10-13 times faster than ones with a 210-group cross section library for 125I dosimetry problems. Compared to 2-D EGS4 Monte Carlo calculations, the 3-group DANTSYS simulations are a 100-fold more efficient. Provided that these
International Nuclear Information System (INIS)
For both type of reactors, WWER-440 and WWER-1000, two different libraries have been created: BGL440 and BGL1000 respectively. The libraries have been produced by collapsing the American fine-group library VITAMIN-B6 (199 neutron and 42 gamma groups) to 67 group structure (47 neutron and 20 gamma groups). The libraries consider the features (detailed 1D geometry and material compositions) of the appropriate reactor and contain upscattering data for the five thermal energy groups. The order of scattering of the Legendre expansion is P5. Each library consists of 2 parts. The first part consists of neutron/gamma cross section data for all reactor materials: BGL441 consists of neutron/gamma cross section data for 150 isotopes (17 chemical elements which appear with different densities and temperatures in the different reactor materials that comprise the WWER-440 reactor); BGL1001 consists of cross sections for 140 nuclides (22 chemical elements which comprise the materials in the WWER-1000). For collapsing cross-sections (previously energy self-shielded) from the 241 group structure (VITAMIN-B6) to the 67 group structure the appropriate average neutron flux in each reactor zone has been used. These datasets can be used for detailed computations of neutron transport. The second parts of each library, BGL442 and BGL1002, consist of cross sections for all 120 nuclides in the VITAMIN-B6 based on the infinitely dilute values only without energy self-shielding. The neutron spectrum just beyond the Reactor Pressure Vessel (RPV) was used for this collapsing. These second datasets can be used for describing non-reactor materials such as dosimeters, capsules, specimens, etc., which may be inserted in the region behind the RPV. (author). 3 refs, 2 figs, 9 tabs
Energy Technology Data Exchange (ETDEWEB)
White, J.E.
2001-04-19
A revised multigroup cross-section library based on Release 3 of ENDF/B-VI data has been produced and tested for light-water-reactor shielding and reactor pressure vessel dosimetry applications. This new broad-group library, which is designated BUGLE-96, represents an improvement over the BUGLE-93 data library released in February 1994 and replaces the data package for BUGLE-93 in the Radiation Safety Information Computational Center (formerly RSIC). The processing methodology is the same as that used for producing BUGLE-93 and is consistent with ANSI/ANS 6.1.2. The ENDF data were first processed into a fine-group, pseudo-problem-independent format and then collapsed into the final broad-group format. The fine-group library, which is designated VITAMIN-B6, contains 120 nuclides. The BUGLE-96 47-neutron-group/20-gamma-ray-group library contains the same 120 nuclides processed as infinitely dilute and collapsed using a weighting spectrum typical of a concrete shield. Additionally, nuclides processed with resonance self-shielding and weighted using spectra specific to BWR and PWR material compositions and reactor models are available. As an added feature of BUGLE-96, cross-section sets having upscatter data for four thermal neutron groups are included. The upscattering data should improve the application of BUGLE-96 to the calculation of more accurate thermal fluences, although more computer time will be required. Several new dosimetry response functions and kerma factors for all 120 nuclides are also included in the library. The incorporation of feedback from users has resulted in a data library that addresses a wider spectrum of user needs.
Benchmark calculations of 150-group cross section library for LMR's
International Nuclear Information System (INIS)
For the purpose of diversification of selection of cross section library for neutron calculation of LMR, the 150 multi-group cross section library was generated from ENDF-VI release. The set was then examined by analyzing measured reactivity quantities such as control rod worth, Doppler effect and sodium void effect for BFS critical assemblies that we obtained through the critical experiment plan for developing the KALIMER core design. The calculated results based on 9 group structure using the new set were also compared with those of JEF set based on the same group structure and compared with those of the same set based on 25 group structure to find the proper group structure. ENDF-VI-based set shows a small deviation in predicting measured integral quantities in comparison with the previous set and a small group effect
A novel hybrid weighting scheme for multi-group cross section collapsing
International Nuclear Information System (INIS)
Multi-group cross section library generation plays an important role in deterministic transport simulations. In this paper, a new fine-group to broad-group cross section collapsing method is introduced. Rather than a traditional flux weighting, the new method uses a hybrid weighing scheme to collapse the scattering cross section matrix. Based upon a matrix analysis approach, we generalize different weighting schemes and derive the new hybrid weighting scheme, which mathematically shows that it is rational for the scattering cross section to be weighted by the (1) forward fluxes of the incoming/in-bound neutron groups and (2) the adjoint functions of the outgoing/out-bound neutron energy groups. This approach also makes physical sense, since it conserves the “importance flow” of particles through scattering while collapsing cross sections. To conserve the reaction rates at the same time, we re-normalize the hybrid weighted scattering cross section to the original library total scattering reaction rate. We demonstrate that the hybrid weighting scheme is more accurate, especially for the detector response simulation problem in a Dual-Range Coincidence Counter (DRCC) 3-D SN transport model. (author)
2013-01-01
This Report summarizes the results of the activities in 2012 and the first half of 2013 of the LHC Higgs Cross Section Working Group. The main goal of the working group was to present the state of the art of Higgs Physics at the LHC, integrating all new results that have appeared in the last few years. This report follows the first working group report Handbook of LHC Higgs Cross Sections: 1. Inclusive Observables (CERN-2011-002) and the second working group report Handbook of LHC Higgs Cross...
Optimization of multi-group cross sections for fast reactor analysis
International Nuclear Information System (INIS)
The selection of the number of broad energy groups, collapsed broad energy group boundaries, and their associated evaluation into collapsed macroscopic cross sections from a general 238-group ENDF/B-VII library dramatically impacted the k eigenvalue for fast reactor analysis. An analysis was undertaken to assess the minimum number of energy groups that would preserve problem physics; this involved studies using the 3D deterministic transport parallel code PENTRAN, the 2D deterministic transport code SCALE6.1, the Monte Carlo based MCNP5 code, and the YGROUP cross section collapsing tool on a spatially discretized MOX fuel pin comprised of 21% PUO2-UO2 with sodium coolant. The various cases resulted in a few hundred pcm difference between cross section libraries that included the 238 multi-group reference, and cross sections rendered using various reaction and adjoint weighted cross sections rendered by the YGROUP tool, and a reference continuous energy MCNP case. Particular emphasis was placed on the higher energies characteristic of fission neutrons in a fast spectrum; adjoint computations were performed to determine the average per-group adjoint fission importance for the MOX fuel pin. This study concluded that at least 10 energy groups for neutron transport calculations are required to accurately predict the eigenvalue for a fast reactor system to within 250 pcm of the 238 group case. In addition, the cross section collapsing/weighting schemes within YGROUP that provided a collapsed library rendering eigenvalues closest to the reference were the contribution collapsed, reaction rate weighted scheme. A brief analysis on homogenization of the MOX fuel pin is also provided, although more work is in progress in this area. (authors)
Cross section probability tables in multi-group transport calculations
International Nuclear Information System (INIS)
The use of cross section probability tables in multigroup transport calculations is presented. Emphasis is placed on how probability table parameters are generated in a multigroup cross section processor and how existing transport codes must be modifed to use them. In order to illustrate the accuracy obtained by using probability tables, results are presented for a variety of neutron and photon transport problems
Ultra-Broad Band Radar Cross Section Reduction of Waveguide Slot Antenna with Metamaterials
Directory of Open Access Journals (Sweden)
Qiang Fu
2016-06-01
Full Text Available To reduce the radar cross section of a waveguide slot antenna, a three-layer metamaterial is presented based on orthogonal double split-ring resonators. The absorption characteristics of three-layer metamaterial are demonstrated by simulation. Moreover, the metamaterials have been loaded on common waveguide slot antenna according to the surface current distribution. The ultra-broad band radar cross section reduction of the antenna with metamaterials had been theoretically and experimentally investigated by radiating and scattering performances. Experimental and simulated results showed that the proposed antenna with metamaterials performed broadband radar cross section reduction from 3.9 GHz to 18 GHz and the gain had been improved due to the coupling effect between slot and the period structure. The maximal radar cross section reduction achieved 17.81 dB at 8.68 GHz for x-polarized incidence and 21.79 dB at 6.25 GHz for y-polarized waves.
Mariotti, C; Passarino, G; Tanaka, R; Andersen, J R; Artoisenet, P; Bagnaschi, E A; Banfi, A; Becher, T; Bernlochner, F U; Bolognesi, S; Bolzoni, P; Boughezal, R; Buarque, D; Campbell, J; Caola, F; Carena, M; Cascioli, F; Chanon, N; Cheng, T; Choi, S Y; David, A; de Aquino, P; Degrassi, G; Del Re, D; Denner, A; van Deurzen, H; Diglio, S; Di Micco, B; Di Nardo, R; Dittmaier, S; Dührssen, M; Ellis, R K; Ferrera, G; Fidanza, N; Flechl, M; de Florian, D; Forte, S; Frederix, R; Frixione, S; Gangal, S; Gao, Y; Garzelli, M V; Gillberg, D; Govoni, P; Grazzini, M; Greiner, N; Griffiths, J; Gritsan, A V; Grojean, C; Hall, D C; Hays, C; Harlander, R; Hernandez-Pinto, R; Höche, S; Huston, J; Jubb, T; Kadastik, M; Kallweit, S; Kardos, A; Kashif, L; Kauer, N; Kim, H; Klees, R; Krämer, M; Krauss, F; Laureys, A; Laurila, S; Lehti, S; Li, Q; Liebler, S; Liu, X; Logan, E; Luisoni, G; Malberti, M; Maltoni, F; Mawatari, K; Maierhoefer, F; Mantler, H; Martin, S; Mastrolia, P; Mattelaer, O; Mazzitelli, J; Mellado, B; Melnikov, K; Meridiani, P; Miller, D J; Mirabella, E; Moch, S O; Monni, P; Moretti, N; Mück, A; Mühlleitner, M; Musella, P; Nason, P; Neu, C; Neubert, M; Oleari, C; Olsen, J; Ossola, G; Peraro, T; Peters, K; Petriello, F; Piacquadio, G; Potter, C T; Pozzorini, S; Prokofiev, K; Puljak, I; Rauch, M; Rebuzzi, D; Reina, L; Rietkerk, R; Rizzi, A; Rotstein-Habarnau, Y; Salam, G P; Sborlini, G; Schissler, F; Schönherr, M; Schulze, M; Schumacher, M; Siegert, F; Slavich, P; Smillie, J M; Stål, O; von Soden-Fraunhofen, J F; Spira, M; Stewart, I W; Tackmann, F J; Taylor, P T E; Tommasini, D; Thompson, J; Thorne, R S; Torrielli, P; Tramontano, F; Tran, N V; Trócsányi, Z; Ubiali, M; Vazquez Acosta, M; Vickey, T; Vicini, A; Waalewijn, W J; Wackeroth, D; Wagner, C; Walsh, J R; Wang, J; Weiglein, G; Whitbeck, A; Williams, C; Yu, J; Zanderighi, G; Zanetti, M; Zaro, M; Zerwas, P M; Zhang, C; Zirke, T J E; Zuberi, S
2013-01-01
This Report summarizes the results of the activities in 2012 and the first half of 2013 of the LHC Higgs Cross Section Working Group. The main goal of the working group was to present the state of the art of Higgs Physics at the LHC, integrating all new results that have appeared in the last few years. This report follows the first working group report Handbook of LHC Higgs Cross Sections: 1. Inclusive Observables (CERN-2011-002) and the second working group report Handbook of LHC Higgs Cross Sections: 2. Differential Distributions (CERN-2012-002). After the discovery of a Higgs boson at the LHC in mid-2012 this report focuses on refined prediction of Standard Model (SM) Higgs phenomenology around the experimentally observed value of 125-126 GeV, refined predictions for heavy SM-like Higgs bosons as well as predictions in the Minimal Supersymmetric Standard Model and first steps to go beyond these models. The other main focus is on the extraction of the characteristics and properties of the newly discovered p...
Calculation and preliminary analysis of group cross sections for gadolinium and its isotopes
International Nuclear Information System (INIS)
Evaluated nuclear data files ENDL 78 and ENDF/B-IV and the cross section generation code FEDGROUP-R were used to produce averaged group cross sections for group systems BNAB and THSIG for gadolinium and its isotopes. Different sets of group cross sections for gadolinium (due to different evaluated data files or different processing codes) were compared and significant differences analyzed. The group capture cross sections for the isotopes of gadolinium were compared with published data. The group cross sections data sets prepared are analyzed as to what extent they can meet the requirements of cell calculations for gadolinium-loaded fuel. Some group cross section tables and cross section plots are presented. (author)
Imaging pore space in tight gas sandstone reservoir: insights from broad ion beam cross-sectioning
Directory of Open Access Journals (Sweden)
Konstanty J.
2010-06-01
Full Text Available Monetization of tight gas reservoirs, which contain significant gas reserves world-wide, represents a challenge for the entire oil and gas industry. The development of new technologies to enhance tight gas reservoir productivity is strongly dependent on an improved understanding of the rock properties and especially the pore framework. Numerous methods are now available to characterize sandstone cores. However, the pore space characterization at pore scale remains difficult due to the fine pore size and delicate sample preparation, and has thus been mostly indirectly inferred until now. Here we propose a new method of ultra high-resolution petrography combining high resolution SEM and argon ion beam cross sectioning (BIB, Broad Ion Beam which prepares smooth and damage free surfaces. We demonstrate this method using the example of Permian (Rotliegend age tight gas sandstone core samples. The combination of Ar-beam cross-sectioning facility and high-resolution SEM imaging has the potential to result in a step change in the understanding of pore geometries, in terms of its morphology, spatial distribution and evolution based on the generation of unprecedented image quality and resolution enhancing the predictive reliability of image analysis.
International Nuclear Information System (INIS)
1 - Description of program or function: specified on ORNL-RSIC-25, shielding benchmark problems. - BP-3 (Neutron cross sections): Format: ANISN, DOT and MORSE; Number of groups: 22 neutron / 18 gamma-ray; Nuclides: air; Origin: ENDF/B; Weighting spectrum: 1/E; - BP-6 (neutron and gamma-ray cross sections): Format: ANISN, DOT and MORSE; Number of groups: 22 neutron / 18 gamma-ray; Nuclides: Borated Polyethylene (C-12, H, and B-10); Origin: ENDF/B-II. The cross section data can be used to repeat the Shielding Benchmark Problems 3.0 and 6.0 for testing against the results published in ORNL-RSIC-25. 2 - Method of solution: ZZ-BP-3 neutron cross sections from the CCC-17/05R library were processed into 104 neutron groups using the PSR-9/CSP code. The fine-group neutron cross sections were collapsed to 22 broad groups using CCC-254/ANISN with an equilibrium fission spectrum source. The resulting multigroup cross sections are P5 coefficients punched on cards in format suitable for input to ANISN, DOT, and MORSE. ZZ-BP-6 neutron and gamma-ray cross sections for 12C, H, and 10B were from ENDF/B-II data. The neutron multigroup cross sections were generated into 104 neutron groups using the PSR-13/SUPERTOG code. The fine-group neutron cross sections were collapsed to 22 broad groups using CCC-254/ANISN with an equilibrium fission spectrum source. The gamma-ray multigroup cross sections were generated using PSR-7/MUG. The neutron-gamma-ray coupling utilized yield data from the DLC-12/POPOP4 library (data sets 010101, 060101, 060301, and 05100201). The neutron-gamma-ray coupled multigroup cross-section set was generated using the SAMPLE COUPLING CODE (ASCC). The multigroup cross sections are in a 22-18 group structure with P3 coefficients punched on cards in format suitable for input to ANISN, DOT, and MORSE
Sensitivity coefficients for the 238U neutron-capture shielded-group cross sections
International Nuclear Information System (INIS)
In the unresolved resonance region cross sections are represented with statistical resonance parameters. The average values of these parameters are chosen in order to fit evaluated infinitely dilute group cross sections. The sensitivity of the shielded group cross sections to the choice of mean resonance data has recently been investigated for the case of 235U and 239Pu by Ganesan and by Antsipov et al; similar sensitivity studies for 238U are reported
Group neutron fission and radiative-capture cross-sections for transactinides
International Nuclear Information System (INIS)
A comparison is made between evaluations of radiative-capture and fission cross-sections for the isotopes 236U, 237Np, 238Pu, 241Am, 243Am, 242Cm and 244Cm, and group cross-sections for use in fast-reactor calculations are recommended. Group cross-sections obtained from the HEDL graphical data (evaluation for ENDF/B-V) are shown for 234U, 236Pu, 237Pu, 242Pu, 244Pu, sup(242m)Am, 241Cm, 243Cm and 248Cm. Group cross-sections for 32 isotopes from the ENDL-76 library files are also given. In choosing recommended cross-sections, account was taken of the extent of agreement with experimental data where these are available, the extent to which the cross-sections are documented and the extent to which they have been calculated from a theoretical model. The reliability of evaluations is discussed. An attempt is made to evaluate the error in single-group cross-sections averaged over a typical fast-reactor spectrum. Conclusions are drawn from a study of the literature on the current status of experimental and theoretical research on transactinide cross-sections, and from the spread of the different evaluation data. Finally, the situation with respect to the integral experiments which can be used for correcting transactinide cross-sections is discussed. (author)
JSD1000: multi-group cross section sets for shielding materials
International Nuclear Information System (INIS)
A multi-group cross section library for shielding safety analysis has been produced by using ENDF/B-IV. The library consists of ultra-fine group cross sections, fine-group cross sections, secondary gamma-ray production cross sections and effective macroscopic cross sections for typical shielding materials. Temperature dependent data at 300, 560 and 900 K have been also provided. Angular distributions of the group to group transfer cross section are defined by a new method of ''Direct Angular Representation'' (DAR) instead of the method of finite Legendre expansion. The library designated JSD1000 are stored in a direct access data base named DATA-POOL and data manipulations are available by using the DATA-POOL access package. The 3824 neutron group data of the ultra-fine group cross sections and the 100 neutron, 20 photon group cross sections are applicable to shielding safety analyses of nuclear facilities. This report provides detailed specifications and the access method for the JSD1000 library. (author)
Expanded and applied sixteen-neutron-energy-group cross-section library
International Nuclear Information System (INIS)
The purpose of the work reported in this paper was five-fold: (1) Develop an expanded neutron cross-section library containing ∼1,200 cross-section sets with the Hansen-Roach (H-R) 16-neutron-energy-group structure. (2) Provide an enhanced computational tool on a personal computer for criticality calculations. (3) Provide consistent values of the effective scattering cross sections (σs) for each set of the expanded H-R library for use in the selection of the resonance self-shielded cross sections (σp). (4) Develop a consistent technique for calculating σp in order to select and apply specific self-shielded cross-section sets. (5) Apply the cross sections and the selection technique to a wide variety of criticality calculational benchmarks
SHAMSI, 48 group cross-section library for fusion nucleonics analysis
International Nuclear Information System (INIS)
A P3 48 group coupled neutron gamma-ray (34 N - 14 G) cross-section library is produced and validated for neutronic studies in fusion reactor blanket/shield. This report describes the library content, the procedure adopted and the results of the calculations performed for testing the cross sections
Development of a Multi-Group Neutron Cross Section Library Generation System for PWR
Energy Technology Data Exchange (ETDEWEB)
Kim, Kang Seog; Hong, Ser Gi; Song, Jae Seung; Lee, Kyung Hoon; Cho, Jin Young; Kim, Ha Yong; Koo, Bon Seung; Shim, Hyung Jin; Park, Sang Yoon
2008-10-15
This report describes a generation system of multi-group cross section library which is used in the KARMA lattice calculation code. In particular, the theoretical methodologies, program structures, and input preparations for the constituent programs of the system are described in detail. The library generation system consists of the following five programs : ANJOY, GREDIT, MERIT, SUBDATA, and LIBGEN. ANJOY generates automatically the NJOY input files and two batch files for automatic NJOY run for all the nuclides considered. The automatic NJOY run gives TAPE 23 (PENDF output file of BROADR module of NJOY) and TAPE24 (GENDF output file of GROUPR module of NJOY) files for each nuclide. GREDIT prepares a formatted multi-group cross section file in which the cross sections are tabulated versus temperature and background cross section after reading the TAPE24 file. MERIT generates the hydrogen equivalence factors and the resonance integral tables by solving the slowing down equation with ultra-fine group cross sections which are prepared with the TAPE 23 file. SUBDATA generates the subgroup data including subgroup levels and weights after reading the MERIT output file. Finally, LIBGEN generates the final multi-group library file by assembling the data prepared in the previous steps and by reading the other data such as fission product yield data and decay data.The multi-group cross section library includes general multi-group cross sections, resonance data, subgroup data, fission product yield data, kappa-values (energy release per fission), and all the data which are required in the depletion calculation. The addition or elimination of the cross sections for some nuclides can be easily done by changing the LIBGEN input file if the general multi-group cross section and the subgroup data files are prepared.
Development of a Multi-Group Neutron Cross Section Library Generation System for PWR
International Nuclear Information System (INIS)
This report describes a generation system of multi-group cross section library which is used in the KARMA lattice calculation code. In particular, the theoretical methodologies, program structures, and input preparations for the constituent programs of the system are described in detail. The library generation system consists of the following five programs : ANJOY, GREDIT, MERIT, SUBDATA, and LIBGEN. ANJOY generates automatically the NJOY input files and two batch files for automatic NJOY run for all the nuclides considered. The automatic NJOY run gives TAPE 23 (PENDF output file of BROADR module of NJOY) and TAPE24 (GENDF output file of GROUPR module of NJOY) files for each nuclide. GREDIT prepares a formatted multi-group cross section file in which the cross sections are tabulated versus temperature and background cross section after reading the TAPE24 file. MERIT generates the hydrogen equivalence factors and the resonance integral tables by solving the slowing down equation with ultra-fine group cross sections which are prepared with the TAPE 23 file. SUBDATA generates the subgroup data including subgroup levels and weights after reading the MERIT output file. Finally, LIBGEN generates the final multi-group library file by assembling the data prepared in the previous steps and by reading the other data such as fission product yield data and decay data.The multi-group cross section library includes general multi-group cross sections, resonance data, subgroup data, fission product yield data, kappa-values (energy release per fission), and all the data which are required in the depletion calculation. The addition or elimination of the cross sections for some nuclides can be easily done by changing the LIBGEN input file if the general multi-group cross section and the subgroup data files are prepared
ACT-1000. Group activation cross-section library for WWER-1000 type reactors
International Nuclear Information System (INIS)
The ACT-1000, a problem-oriented library of group-averaged activation cross-sections for WWER-1000 type reactors, is based on evaluated microscopic cross-section data files. The ACT-1000 data library was designed for calculating induced activity for the main dose-generated nuclides contained in WWER-1000 structural materials. In preparing the ACT-1000 library, 47 group-averaged cross-section data for the 10-9-17.33 MeV energy range were used to calculate the spatial-energy neutron flux distribution. (author)
International Nuclear Information System (INIS)
Approximation of few-group neutron cross-sections by functions of burnup and thermal-hydraulics parameters of a fuel cell is considered. The cross-section is written as a sum of two terms: the base cross-section, which depends only on burnup and is computed under the nominal reactor core conditions, and the deviation, which depends on burnup and thermal-hydraulics variables of the cell. A one-dimensional dependence of the base cross-section is interpolated by a cubic spline. Multi-dimensional dependencies of the deviation are approximated by a polynomial. Construction of the polynomial is performed by a best-fitting selection of the polynomial terms using the stepwise regression algorithm. The number of terms to satisfy a user-given accuracy of approximation is minimized. As an example, approximation of a set of two-group macro and micro cross-sections as functions of burnup, coolant and fuel temperature, coolant density and boron concentration is considered for a fuel pin cell of a VVER reactor. The constructed five-dimensional polynomial approximating cross-sections within 0.05% tolerance has about 20 terms for fast group cross-sections and 50 terms for thermal group cross-sections. The error of approximation is verified on the two data sets: the initial data used for approximation and the test data being computed on randomly selected points. Mean square and maximum errors are comparable for all the cross-sections for both sets of data. These results show that the initial data can be applied to control the approximation error
Dorsiflexor muscle-group thickness in children with cerebral palsy: Relation to cross-sectional area
DEFF Research Database (Denmark)
Bandholm, Thomas; Magnusson, Peter; Jensen, Bente R; Sonne-Holm, Stig
2009-01-01
If the thickness and cross-sectional area of the dorsiflexor muscle group are related in children with cerebral palsy, measurements of muscle thickness may be used to monitor changes in muscle size due to training or immobilisation in these patients. We assessed the validity and reliability of...... measurements of dorsiflexor muscle-thickness using the cross-sectional area of the muscle group as the criterion-related muscle-size variable. Muscle thickness was measured using ultrasound, and cross-sectional area using MRI in nine children with spastic cerebral palsy (eight with hemiplegia). Test......-retest reliability of the muscle-thickness measurements was assessed in six healthy subjects. All measurements were made on both legs at 35% lower leg length. In the children with cerebral palsy, dorsiflexor muscle-thickness and cross-sectional area were well correlated (r;{2} = 0.778, P < 0.001), and the...
Few group cross section representation based on sparse grid methods / Danniëll Botes
Botes, Danniëll
2012-01-01
This thesis addresses the problem of representing few group, homogenised neutron cross sections as a function of state parameters (e.g. burn-up, fuel and moderator temperature, etc.) that describe the conditions in the reactor. The problem is multi-dimensional and the cross section samples, required for building the representation, are the result of expensive transport calculations. At the same time, practical applications require high accuracy. The representation method must therefore be eff...
ZZ DLC-13B, Resonance Cross-Section Group Constant Library for Tungsten and Depleted Pu
International Nuclear Information System (INIS)
Nature of physical problem solved: Format: GAM-II; Number of groups: 32-energy-group split (0.4 to 1234 eV). Nuclides: tungsten (W,) and depleted uranium (U,) slabs. Multigroup capture and scatter cross sections in the resolved resonance region were calculated for tungsten and depleted uranium slabs for use in shielding calculations of neutron transport and capture distributions. Slabs of thickness of 1 to 8 centimeters surrounded by hydrogen or lithium hydride were considered. GAROL was used to generate the cross sections, a method previously observed to preserve the total capture rate in a detailed multigroup neutron transport calculation for a thick resonance absorber. Average cross sections were calculated for a 32-energy-group split (0.4 to 1234 eV) compatible with that used by GAM-2. Group fluxes are also presented permitting further group collapsing either by hand calculations or with an included computer program
ZZ VITAMIN-E, 174-Group Neutron, 38-Group Gamma Cross-Section in AMPX Format
International Nuclear Information System (INIS)
1 - Description of program or function: - Format: AMPX; - Number of groups: 174 neutron and 38 gamma-ray groups Nuclides: H, He, Li, Be, B, C, N, O, F, Na, Mg, Al, Si, P, S, Cl, Ar, K, Ca, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Ga, Y, Zr, Nb, Mo, Cd, Sn, I, Cs, Ba, Gd, Hf, Ta, W, Re, Pt, Au, Pb, Bi, Th, Pa, U, Np, Pu, Am, Cm. - Origin: ENDF/B-V; - Weighting spectrum: From 10-5 eV to 0.414 eV → Maxwellian Thermal Spectrum; From 0.414 eV to 2.12 MeV → '1/E' Slowing-Down Spectrum; From 2.12 MeV to 10.0 MeV → Fission Spectrum; From 10.0 MeV to 12.52 MeV → '1/E' Spectrum; From 12.52 MeV to 15.68 MeV → Velocity Exponential Fusion Peak; From 15.68 MeV to 19.64 MeV → '1/E' Spectrum; Photon interaction cross sections → constant in energy. The early phases of this new cross-section library were focused on materials for fast reactor applications and were applied to benchmark testing of ENDF/B-V. More recently, requests have been made for additional materials to be added to the basic library for fusion and weapons radiation transport applications. The library is expected to perform well for radiation transport problems where thermal up-scatter is not important. The energy structure of VITAMIN-E contains 174 neutron and 38 gamma-ray groups and includes the 171 neutron and 36 photon groups of VITAMIN-C as a subset. The group structure has fine detail in the energy region where cross section minima occur for important shielding materials. 2 - Method of solution: The 174 neutron group data were processed with MINXI5; the 174 neutron, 38 photon group data were processed with LAPHNGAS (AMPX III); and the 38 gamma-ray group data with SMUG (AMPX III) from DLC-99/HUGO. The ENDF/B-V special purpose dosimetry, activation, and gas production files have also been processed into the VITAMIN-E group structure using XLACS2, NITAWL, and WORM
KOEBLIB1.0: Two group polynomial cross section library for Koeberg version 1
International Nuclear Information System (INIS)
The mathematical models and engineering data used for the generation of version 1 of the 2-group polynomial cross section library for the two PWR units at the Koeberg Nuclear Power Station, are described. This library was prepared using the few-group coarse mesh cross section generation package of the Reactor Theory Programme at the Atomic Energy Corporation of South Africa Ltd. An overall description of the calculational scheme as well as descriptions of the various modules used for the generation of the cross section library is given. The fuel assembly model is described in detail and the values of the operational parameters used, are given. The methods used to generate the ex-core reflector data are described. Details of the generation of the polynomial library are given and the assembly and reflector engineering data are listed. 2 figs., 6 tabs., 19 refs
Updated multi-group cross sections of minor actinides with improved resonance treatment
International Nuclear Information System (INIS)
The study of minor actinide in transmutation reactors and other future applications makes resonance self-shielding treatment a significant issue for criticality and isotope depletion. Resonance treatment for minor actinides has been carried out by subgroup method with improved interference effect through interference correction. Subgroup data was generated using RMET21 and GENP codes along with multi-group cross section data by NJOY nuclear data processing system. Updated multi-group cross section data library for a neutron transport code nTRACER was compared with solutions from MCNPX. The resonance interaction of uranium with minor actinides has been included by modified interference treatment of interference correction in subgroup methodology. The comparison of cross sections and multiplication factor in pin and assembly problems showed significant improvement from systematic resonance treatment especially for 237Np and 243Am. (author)
ZZ MONTAGE-400, Neutron Activation 100-Group Cross-Section Library of Fusion Reactor Materials
International Nuclear Information System (INIS)
1 - Description of problem or function: Format: GAM-II group structure and ANISN; Number of groups: 100-group cross sections. Nuclides: H, He, Li, Be, B, N, O, F, Na, Al, Si, P, S, Cl, K, Ca, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Y, Zr, Nb, Mo, Tc, Ru, Ag, Sn, Cs, Hf, Ta, W, Re, Au, Pb. Origin: derived from ENDF/B, or calculated at Brookhaven National Laboratory. Weighting spectrum: 1/E except near 14 MeV where a thermally broadened fusion peak, assuming a temperature of 20 MeV, is employed. This data library contains 100- group cross sections, with GAM-II group structure, for 421 neutron activation reactions with fusion reactor structural and coolant materials. The weighting function is 1/E except near 14 MeV where a thermally broadened fusion peak, assuming a temperature of 20 MeV, is employed. The library also contains half life information for the activated nuclei. 2 - Method of solution: The thermal group cross sections were calculated from the 2200 m/s value, when available, otherwise from the group 99 value. The majority of the non-thermal cross sections were derived from pointwise data derived from ENDF/B, or calculated at Brookhaven National Laboratory using the nuclear systematics code THRESH. These were converted to multigroup from using the codes ETOG and NJOY
Amino acids analysis using grouping and parceling of neutrons cross sections techniques
International Nuclear Information System (INIS)
Amino acids used in parenteral administration in hospital patients with special importance in nutritional applications were analyzed to compare with the manufactory data. Individual amino acid samples of phenylalanine, cysteine, methionine, tyrosine and threonine were measured with the neutron crystal spectrometer installed at the J-9 irradiation channel of the 1 kW Argonaut Reactor of the Instituto de Engenharia Nuclear (IEN). Gold and D2O high purity samples were used for the experimental system calibration. Neutron cross section values were calculated from chemical composition, conformation and molecular structure analysis of the materials. Literature data were manipulated by parceling and grouping neutron cross sections. (author)
Optimization of multidimensional cross-section tables for few-group core calculations
International Nuclear Information System (INIS)
Highlights: • Optimization of tabulated cross-sections libraries for multi-group diffusion codes. • Sensitivity coefficients using perturbation theory are determined. • A non-uniform grid satisfying a given target accuracy in k-effective is built. • Satisfactory results are obtained using libraries with different accuracy level. - Abstract: Multigroup diffusion codes for three dimensional LWR core analysis use as input data pre-generated homogenized few group cross sections and discontinuity factors for certain combinations of state variables, such as temperatures or densities. The simplest way of compiling those data are tabulated libraries, where a grid covering the domain of state variables is defined and the homogenized cross sections are computed at the grid points. Then, during the core calculation, an interpolation algorithm is used to compute the cross sections from the table values. Since interpolation errors depend on the distance between the grid points, a determined refinement of the mesh is required to reach a target accuracy, which could lead to large data storage volume and a large number of lattice transport calculations. In this paper, a simple and effective procedure to optimize the distribution of grid points for tabulated libraries is presented. Optimality is considered in the sense of building a non-uniform point distribution with the minimum number of grid points for each state variable satisfying a given target accuracy in k-effective. The procedure consists of determining the sensitivity coefficients of k-effective to cross sections using perturbation theory; and estimating the interpolation errors committed with different mesh steps for each state variable. These results allow evaluating the influence of interpolation errors of each cross section on k-effective for any combination of state variables, and estimating the optimal distance between grid points
ZZ COV-15GROUP-2006, 15-group cross section covariance matrix library
International Nuclear Information System (INIS)
Description: ZZ-COV-15GROUP is a 15-group cross section covariance matrix library presenting a general overview of the presently available data. Number of groups: 15 neutron. Nuclides: H-1, Li-6, Li-7, Be-9, B-10, C-12, N-14, O-16, F-19, Na-23, Al-27, Si-28, Si-nat, Cr-52, Mn-55, Fe-56, Fe-57, Ni-58, Zr-90, Pb-nat, Pb-206, Pb-207, Pb-208, Th-232, U-235, U-238, Np-237, Pu-239, Pu-240, Pu-241, Am-241. Origin: ENDF/B-V, /B-VI.8, JENDL-3.3, JEFF-3.0, IRDF-2002 and IAEA Version 02 differs from version 01 in the following features: The input files (original BOXER format covariance libraries and ANGELO inputs instructions) have been included thus allowing to convert the covariance matrices to a user-defined energy group structure. Examples of output for the 15 group structure are provided. The code LAMBDA for verification of mathematical properties of the matrices (e. g. eigenvalues) is also included. This verification is highly recommended before using any covariance matrices. Version 03 differs from version 02 in the following features: The library file covfils2.lib was corrected (energy group structure was provided only for one isotope), as well as the corresponding test case outputs
On the use of the Serpent Monte Carlo code for few-group cross section generation
International Nuclear Information System (INIS)
Research highlights: → B1 methodology was used for generation of leakage-corrected few-group cross sections in the Serpent Monte-Carlo code. → Few-group constants generated by Serpent were compared with those calculated by Helios deterministic lattice transport code. → 3D analysis of a PWR core was performed by a nodal diffusion code DYN3D employing two-group cross section sets generated by Serpent and Helios. → An excellent agreement in the results of 3D core calculations obtained with Helios and Serpent generated cross-section libraries was observed. - Abstract: Serpent is a recently developed 3D continuous-energy Monte Carlo (MC) reactor physics burnup calculation code. Serpent is specifically designed for lattice physics applications including generation of homogenized few-group constants for full-core core simulators. Currently in Serpent, the few-group constants are obtained from the infinite-lattice calculations with zero neutron current at the outer boundary. In this study, in order to account for the non-physical infinite-lattice approximation, B1 methodology, routinely used by deterministic lattice transport codes, was considered for generation of leakage-corrected few-group cross sections in the Serpent code. A preliminary assessment of the applicability of the B1 methodology for generation of few-group constants in the Serpent code was carried out according to the following steps. Initially, the two-group constants generated by Serpent were compared with those calculated by Helios deterministic lattice transport code. Then, a 3D analysis of a Pressurized Water Reactor (PWR) core was performed by the nodal diffusion code DYN3D employing two-group cross section sets generated by Serpent and Helios. At this stage thermal-hydraulic (T-H) feedback was neglected. The DYN3D results were compared with those obtained from the 3D full core Serpent MC calculations. Finally, the full core DYN3D calculations were repeated taking into account T-H feedback and
International Nuclear Information System (INIS)
1 - Description of problem or function: Format: 'data base' for subsequent collapsing into both fine and broad group data in various formats (working and/or weighted ANISN, CCCC, etc.). Number of groups: AMPX-2/123 → 123 group structure; AMPX-2/219 → 219 group structure. Nuclides: H, He, Li, Be, B, C, N, O, F, Na, Mg, Al, Si, Cl, K, Ca, Ti, V, Cr, Mn, Fe, Ni, Cu, Kr, Zirc, Mo, Tc, Rh, Ag, Cd, Xe, Sm, Eu, Gd, Dy, Cu, Ta, W, Re, Pb, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDF/B-IV. Weighting spectrum: Most data were generated using a standard flux over three energy ranges (fission - 1/E - Maxwellian) as point-to-fine-group cross sections weighing function. The AMPX-2 P3 123- and 219- Group Neutron Cross-Section Master Interface Libraries may be considered as 'data bases' for subsequent collapsing into both fine and broad group data in various formats (working and/or weighted ANISN, CCCC, etc.). The built-in 123 and 219 group structures have been used to process all available data of ENDF/B-IV. 2 - Method of solution: The program AMPX-2 has been used to generate the data. By various executions of the module XLACS-2 (XLACS for bound H-1 in some materials) a number of independent libraries were generated which then were combined using the AMPX-2 module AJAX. Most data were generated using a standard flux over three energy ranges (fission - 1/E - Maxwellian) as point-to-fine-group cross sections weighing function. For some structural materials (e.g. Fe, Cr,...) different master data sets were produced using a weighting function fission - 1/E sigma T(SS-304) - Maxwellian, and the three parts of the spectrum were joined at properly selected energies. For some nuclides (e.g. 238U and 240Pu) various master data sets have been produced which contain problem-dependent unresolved cross sections characterized by the associated potential scattering cross sections. Some data sets contain P3 thermal scattering matrices, for which ENDF/B File 7 S(alpha, beta) data were used, e
Analysis of human hair cross sections from two different population groups by Nuclear Microscopy
International Nuclear Information System (INIS)
Trace element analysis of hair is used as a screening technique to assess body-nutrient levels and/or toxicity due to environmental pollutants. With the aim to compare element content and spatial distribution within scalp hair-shaft cross sections of two distinct human population groups, and to assess possible similarities and/or differences, hair samples from Sudan and South Africa were collected. Proton backscattering and Micro-PIXE were used to determine the matrix composition and content of light and middle transition elements, with beam energies of 1.5 and 3.0 MeV. Mapping analysis showed a relatively similar content distribution for S, Cl, K and Ca within each group. However significant differences, particularly for heavier metals, such as Fe and Zn were also found. Correspondence Analysis of the data showed a clear separation between the two groups when the total content over the hair cross section was considered.
Energy Technology Data Exchange (ETDEWEB)
Wilson, W.B.; England, T.R.; LaBauve, R.J.
1978-02-01
The ENDF/B-IV fission-product data file includes data describing 824 nuclides. Cross sections, given for 181 of these nuclides, were processed into 154 neutron energy groups. The production of the data file is described. The TOAFEW code, useful in collapsing the multigroup values to few-group cross sections, is presented with instructions and examples of its use. The file of multigroup cross sections is available on request. 3 figures, 11 tables.
Uncertainty Analysis of Few Group Cross Sections Based on Generalized Perturbation Theory
International Nuclear Information System (INIS)
In this paper, the methodology of the sensitivity and uncertainty analysis code based on GPT was described and the preliminary verification calculations on the PMR200 pin cell problem were carried out. As a result, they are in a good agreement when compared with the results by TSUNAMI. From this study, it is expected that MUSAD code based on GPT can produce the uncertainty of the homogenized few group microscopic cross sections for a core simulator. For sensitivity and uncertainty analyses for general core responses, a two-step method is available and it utilizes the generalized perturbation theory (GPT) for homogenized few group cross sections in the first step and stochastic sampling method for general core responses in the second step. The uncertainty analysis procedure based on GPT in the first step needs the generalized adjoint solution from a cell or lattice code. For this, the generalized adjoint solver has been integrated into DeCART in our previous work. In this paper, MUSAD (Modues of Uncertainty and Sensitivity Analysis for DeCART) code based on the classical perturbation theory was expanded to the function of the sensitivity and uncertainty analysis for few group cross sections based on GPT. First, the uncertainty analysis method based on GPT was described and, in the next section, the preliminary results of the verification calculation on a VHTR pin cell problem were compared with the results by TSUNAMI of SCALE 6.1
Uncertainty Analysis of Few Group Cross Sections Based on Generalized Perturbation Theory
Energy Technology Data Exchange (ETDEWEB)
Han, Tae Young; Lee, Hyun Chul; Noh, Jae Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2014-05-15
In this paper, the methodology of the sensitivity and uncertainty analysis code based on GPT was described and the preliminary verification calculations on the PMR200 pin cell problem were carried out. As a result, they are in a good agreement when compared with the results by TSUNAMI. From this study, it is expected that MUSAD code based on GPT can produce the uncertainty of the homogenized few group microscopic cross sections for a core simulator. For sensitivity and uncertainty analyses for general core responses, a two-step method is available and it utilizes the generalized perturbation theory (GPT) for homogenized few group cross sections in the first step and stochastic sampling method for general core responses in the second step. The uncertainty analysis procedure based on GPT in the first step needs the generalized adjoint solution from a cell or lattice code. For this, the generalized adjoint solver has been integrated into DeCART in our previous work. In this paper, MUSAD (Modues of Uncertainty and Sensitivity Analysis for DeCART) code based on the classical perturbation theory was expanded to the function of the sensitivity and uncertainty analysis for few group cross sections based on GPT. First, the uncertainty analysis method based on GPT was described and, in the next section, the preliminary results of the verification calculation on a VHTR pin cell problem were compared with the results by TSUNAMI of SCALE 6.1.
ECNJEFI. A JEFI based 219-group neutron cross-section library: User's manual
International Nuclear Information System (INIS)
This manual describes the contents of the ECNJEF1 library. The ECNJEF1 library is a JEF1.1 based 219-group AMPX-Master library for reactor calculations with the AMPX/SCALE-system, e.g. the PASC-3 system as implemented at the Netherlands Energy Research Foundation in Petten, Netherlands. The group cross-section data were generated with NJOY and NPTXS/XLACS-2 from the AMPX system. The data on the ECNJEF1 library allows resolved-resonance treatment by NITAWL and/or unresolved resonance self-shielding by BONAMI. These codes are based upon the Nordheim and Bondarenko methods, respectively. (author). 10 refs., 7 tabs
EJ1: a JEF1 based 219-group neutron cross-section library
International Nuclear Information System (INIS)
This manual describes the contents of the EJ1 library. The EJ1 library is a JEF1.1 based 219-group AMPX-Master library for reactor calculations with the AMPX/SCALE-3 system, e.g. the PASC-3 system, as implemented at ECN-Petten. The group cross-section data were generated with NJOY. The data on the EJ1 library allow resolved-resonance treatment by NITAWL and unresolved resonance self-shielding by BONAMI. These codes are based upon the Nordheim and Bondarenko methods, respectively. (author). 24 refs., 8 tabs
Consistent Generation and Verification of 190 Group Cross Section Library Data for Primary Nuclides
International Nuclear Information System (INIS)
The multigroup cross section data used in the lattice transport or the direct whole core transport codes such as HELIOS and DeCART have a significant impact on the accuracy of the criticality prediction. If a large discrepancy is noted in the analysis of critical experiments, it is customary to adjust the resonance integral (RI) data of U-238 given in the cross section library in order to match the measurement. In case of HELIOS, the unadjusted library gives about 300∼550 pcm lower reactivity than the adjusted one. The sole adjustment of the U238 RI, however, is to blame only U238 for all the discrepancies that can originate from various sources. One of the sources of the error would be the inaccuracy of subgroup parameters used in the a group codes which employ the subgroup method for resonance treatment. The inconsistency problem noted in the subgroup parameter generation and usage steps which was reported in our previous work can be smeared out by the RI adjustment. Thus such blind adjustment of the resonance integral is to be avoided. In this work, we examine a new procedure for generating multigroup cross section data from the ENDF/B files, which would not require any forced adjustment. One of the distinct steps in the procedure is to employ a consistent method of generating subgroup parameters formulated by imposing a shielded cross section conservation principle rather than the resonance integral conservation. In order to check the validity of the procedure, multigroup data are generated only for a group of primary nuclides which appear in a fresh fuel UO2 pin cell, namely, U-235, U-238, H-1, O-16, and Zr. The accuracy of the new library is assessed by comparing the reactivity with those obtained from corresponding continuous energy Monte Carlo calculations. Since recently the ENDF/B-VII file was released which reflects improvements in the U238 resonance data, the difference between the multigroup cross section libraries generated from the new ENDF file
Consistent Generation and Verification of 190 Group Cross Section Library Data for Primary Nuclides
Energy Technology Data Exchange (ETDEWEB)
Kim, Gwan Young; Joo, Han Gyu [Seoul National University, Seoul (Korea, Republic of)
2008-05-15
The multigroup cross section data used in the lattice transport or the direct whole core transport codes such as HELIOS and DeCART have a significant impact on the accuracy of the criticality prediction. If a large discrepancy is noted in the analysis of critical experiments, it is customary to adjust the resonance integral (RI) data of U-238 given in the cross section library in order to match the measurement. In case of HELIOS, the unadjusted library gives about 300{approx}550 pcm lower reactivity than the adjusted one. The sole adjustment of the U238 RI, however, is to blame only U238 for all the discrepancies that can originate from various sources. One of the sources of the error would be the inaccuracy of subgroup parameters used in the a group codes which employ the subgroup method for resonance treatment. The inconsistency problem noted in the subgroup parameter generation and usage steps which was reported in our previous work can be smeared out by the RI adjustment. Thus such blind adjustment of the resonance integral is to be avoided. In this work, we examine a new procedure for generating multigroup cross section data from the ENDF/B files, which would not require any forced adjustment. One of the distinct steps in the procedure is to employ a consistent method of generating subgroup parameters formulated by imposing a shielded cross section conservation principle rather than the resonance integral conservation. In order to check the validity of the procedure, multigroup data are generated only for a group of primary nuclides which appear in a fresh fuel UO{sub 2} pin cell, namely, U-235, U-238, H-1, O-16, and Zr. The accuracy of the new library is assessed by comparing the reactivity with those obtained from corresponding continuous energy Monte Carlo calculations. Since recently the ENDF/B-VII file was released which reflects improvements in the U238 resonance data, the difference between the multigroup cross section libraries generated from the new
Method for calculation of global sensitivity indices for few-group cross-section-dependent problems
International Nuclear Information System (INIS)
The variance based global sensitivity analysis technique is robust, has a wide range of applicability and provides accurate sensitivity information for most models. However, it requires input variables to be mutually independent. A modification to this technique that allows one to deal with input variables that are block-wise correlated and normally distributed is presented. The focus of this study is the application of the modified global sensitivity analysis technique to calculations of reactor parameters that are dependent on multigroup or few-group neutron cross-sections. The result of the sensitivity analysis is obtained in terms of the global sensitivity indices, which can be used for characterising the contribution of uncertainties from the input cross-sections or their groups to the uncertainty of the calculated reactor parameter. The main effort in this work, besides presenting the theoretical background, is in establishing a method for a practical numerical calculation of the global sensitivity indices. The implementation of the method involves the calculation of multi-dimensional integrals, which can be prohibitively expensive to compute. Numerical techniques specifically suited to the evaluation of multidimensional integrals namely Monte Carlo and sparse grids methods are used, and their efficiency is compared. The method is illustrated and tested on a two-group cross-section dependent problem. In all the cases considered the results obtained with sparse grids achieved much better accuracy while using a significantly smaller number of samples. This aspect is addressed in a mini-study and a preliminary explanation of the results obtained is given. (author)
ZZ ABBN, 26 Group Cross-Sections and Self Shielding Factors for Fast Reactors
International Nuclear Information System (INIS)
1 - Description of program or function: Format: special format; Number of groups: 26 group X-section and resonance self-shielding factor library. Nuclides: H, D, Li-6, Li-7, Be, B-10, B-11, C, N, O, Na, Mg, Al, Si, K, Ca, Ti, V, Cr, Fe, Ni, Cu, Zr, Nb, Mo, Ta, W, Re, Pb, Bi, Th-232, U-233, U-234, U-235, U-236, U-238, Pu-239, Pu-240, Pu-241, Pu-242, FP-U-233, FP-U-235, FP-Pu-239. Origin: Multiple experimental sources; Weighting spectrum: yes; 26 group cross section and resonance self-shielding factor library for the following materials: H, D, Li-6, Li-7, Be, B-10, B-11, C, N, O, Na, Mg, Al, Si, K, Ca, Ti, V, Cr, Fe, Ni, Cu, Zr, Nb, Mo, Ta, W, Re, Pb, Bi, Th-232, U-233, U-234, U-235, U-236, U-238, Pu-239, Pu-240, Pu-241, Pu-242, FP-U-233, FP-U-235, FP-Pu-239. 2 - Restrictions on the complexity of the problem: This group cross section library has been developed for fast and intermediate reactors
Damage cross-section library DAMSIG84 (in a 640 group structure of the SAND-II type)
International Nuclear Information System (INIS)
The damage cross-section library DAMSIG84 is an updated and extended version of the damage cross-section library DAMSIG81. The library contains energy dependent group cross-section data for a number of materials to facilitate the calculations of damage production (based on displacements of atoms), the calculations of probable zones and the calculation of gas production due to (n,α) and (n,p) reactions. The group cross-section data are given for a fine group structure of the SAND-II type with 640 groups. This library contains for some materials more than one cross-section set originating from different evaluations. Cross-section data sets for the activation reactions 54Fe(n,p), 58Ni(n,p), 59Co(n,γ), and 63Cu(n,α), which reactions are commonly used to determine thermal and fast neutron fluences, have been included also. Moreover also some artificial cross-sections are incorporated in this library which can be used to calculate values for some physical quantities characterizing neutron spectra, such as mean lethargy , and mean energy . Also cross-sections for B, Al and Cd are included; these are required to reach compatibility with other libraries in the SAND-II format
International Nuclear Information System (INIS)
Allowing Monte Carlo (MC) codes to perform fuel cycle calculations requires coupling to a point depletion solver. In order to perform depletion calculations, one-group (1-g) cross sections must be provided in advance. This paper focuses on generating accurate 1-g cross section values that are necessary for evaluation of nuclide densities as a function of burnup. The proposed method is an alternative to the conventional direct reaction rate tally approach, which requires extensive computational efforts. The method presented here is based on the multi-group (MG) approach, in which pre-generated MG sets are collapsed with MC calculated flux. In our previous studies, we showed that generating accurate 1-g cross sections requires their tabulation against the background cross-section (σ0) to account for the self-shielding effect. However, in previous studies, the model that was used to calculate σ0 was simplified by fixing Bell and Dancoff factors. This work demonstrates that 1-g values calculated under the previous simplified model may not agree with the tallied values. Therefore, the original background cross section model was extended by implicitly accounting for the Dancoff and bell factors. The method developed here reconstructs the correct value of σ0 by utilizing statistical data generated within the MC transport calculation by default. The proposed method was implemented into BGCore code system. The 1-g cross section values generated by BGCore were compared with those tallied directly from the MCNP code. Very good agreement (<0.05%) in the 1-g cross values was observed. The method dose not carry any additional computational burden and it is universally applicable to the analysis of thermal as well as fast reactor systems. (author)
International Nuclear Information System (INIS)
This paper presents a synthesis of the ENEA-Bologna Nuclear Data Group programme dedicated to generate and validate group-wise cross section libraries for shielding and radiation damage deterministic calculations in nuclear fission reactors, following the data processing methodology recommended in the ANSI/ANS-6.1.2-1999 (R2009) American Standard. The VITJEFF311.BOLIB and VITENDF70.BOLIB fine group coupled n-γ (199 n + 42 γ - VITAMIN-B6 structure) multi-purpose cross section libraries, based on the Bondarenko method for neutron resonance self-shielding and respectively on JEFF-3.1.1 and ENDF/B-VII.0 evaluated nuclear data, were produced in AMPX format using the NJOY-99.259 and the ENEA-Bologna 2007 Revision of the SCAMPI nuclear data processing systems. Two derived broad-group coupled n-γ (47 n + 20 γ - BUGLE-96 structure) working cross section libraries in FIDO-ANISN format for LWR shielding and pressure vessel dosimetry calculations, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, were generated by the revised version of SCAMPI, through problem-dependent cross section collapsing and self-shielding from the cited fine-group libraries. The validation results on the criticality safety benchmark experiments for the fine-group libraries and the preliminary validation results for the broad-group working libraries on the PCA-Replica and VENUS-3 engineering neutron shielding benchmark experiments are reported in synthesis. (authors)
Pescarini, M.; Sinitsa, V.; Orsi, R.; Frisoni, M.
2013-03-01
This paper presents a synthesis of the ENEA-Bologna Nuclear Data Group programme dedicated to generate and validate group-wise cross section libraries for shielding and radiation damage deterministic calculations in nuclear fission reactors, following the data processing methodology recommended in the ANSI/ANS-6.1.2-1999 (R2009) American Standard. The VITJEFF311.BOLIB and VITENDF70.BOLIB finegroup coupled n-γ (199 n + 42 γ - VITAMIN-B6 structure) multi-purpose cross section libraries, based on the Bondarenko method for neutron resonance self-shielding and respectively on JEFF-3.1.1 and ENDF/B-VII.0 evaluated nuclear data, were produced in AMPX format using the NJOY-99.259 and the ENEA-Bologna 2007 Revision of the SCAMPI nuclear data processing systems. Two derived broad-group coupled n-γ (47 n + 20 γ - BUGLE-96 structure) working cross section libraries in FIDO-ANISN format for LWR shielding and pressure vessel dosimetry calculations, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, were generated by the revised version of SCAMPI, through problem-dependent cross section collapsing and self-shielding from the cited fine-group libraries. The validation results on the criticality safety benchmark experiments for the fine-group libraries and the preliminary validation results for the broad-group working libraries on the PCA-Replica and VENUS-3 engineering neutron shielding benchmark experiments are reported in synthesis.
Directory of Open Access Journals (Sweden)
Orsi R.
2013-03-01
Full Text Available This paper presents a synthesis of the ENEA-Bologna Nuclear Data Group programme dedicated to generate and validate group-wise cross section libraries for shielding and radiation damage deterministic calculations in nuclear fission reactors, following the data processing methodology recommended in the ANSI/ANS-6.1.2-1999 (R2009 American Standard. The VITJEFF311.BOLIB and VITENDF70.BOLIB finegroup coupled n-γ (199 n + 42 γ – VITAMIN-B6 structure multi-purpose cross section libraries, based on the Bondarenko method for neutron resonance self-shielding and respectively on JEFF-3.1.1 and ENDF/B-VII.0 evaluated nuclear data, were produced in AMPX format using the NJOY-99.259 and the ENEA-Bologna 2007 Revision of the SCAMPI nuclear data processing systems. Two derived broad-group coupled n-γ (47 n + 20 γ – BUGLE-96 structure working cross section libraries in FIDO-ANISN format for LWR shielding and pressure vessel dosimetry calculations, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, were generated by the revised version of SCAMPI, through problem-dependent cross section collapsing and self-shielding from the cited fine-group libraries. The validation results on the criticality safety benchmark experiments for the fine-group libraries and the preliminary validation results for the broad-group working libraries on the PCA-Replica and VENUS-3 engineering neutron shielding benchmark experiments are reported in synthesis.
Cross-section library DOSCROS81 (in a 640 group structure of the SAND-II type)
International Nuclear Information System (INIS)
The cross section library DOSCROS81 is an updated and extended version of the dosimetry cross section library DOSCROS77. The library contains energy dependent fine group cross section values for a number of reactions which are applied in neutron metrology and in neutron activation spectrometry. The library contains data from the ENDF/B-V file supplemented with information from the ENDF/B-IV and from the INDL/V. The total number of reaction cross section sets incorporated in this library is 70 (+ 3 cover cross section sets). A documentation of the library is presented. The library is written in the SAND-II format. The numerical data are available on microfiche upon request to ECN. The library will be available in a computer compatible form from the OECD NEA Data Bank and from the RSIC at Oak Ridge
WLUP3.0, 69 and 172 Group Cross Section Libraries for WIMS
International Nuclear Information System (INIS)
Description or function: WLUP contains validated WIMS-D formatted cross section libraries in 69 and 172 energy group structures for nuclear reactor calculations. Materials from recently released evaluated nuclear data libraries are included. The NJOY nuclear data processing system was applied for generating the cross section files following the models and conventions built into the WIMS-D lattice code. The relevant features for the WIMS users are: - Energy group structures: 69 and 172 energy groups. - List of materials: WIMS ID, general information, source of data. - Cross sections: 69 and 172 group plots. - Resonance data: WIMS ID, temperature, background cross sections. - Goldstein-Cohen factors: Goldstein-Cohen lambda values. - Thermal scattering data: thermal scattering laws and P1 matrixes. - Fission spectrum: fission spectrum data. - Burnup data: burnup chains. - Fission product yields: fission yield tables. - Pseudo lumped fission product: Description of pseudo fission product. - Energy release by fission: table of energy released by fission. - Dosimetry data: dosimetry reactions, source of data. - Averaging flux and current spectra: flux and current spectra plots (Numerical data on NJOY inputs). - WIMSD5B updates: WIMSD5B extensions and updates. - Processing methods: Brief description on processing methods. Moderators: 1-H-H2O, 1-H-ZrH, 1-D-D2O, 4-Be, 6-C, 8-O-16. Structural materials: 2-He-3, 2-He-4, 3-Li-6, 3-Li-7, 5-B-10, 5-B, 7-N, 9-F, 11-Na, 12-Mg, 13-Al, 14-Si, 15-P, 16-S, 17-Cl, 20-Ca, 22-Ti, 23-V, 24-Cr, 25-Mn, 26-Fe, 27-Co-59, 28-Ni, 29-Cu, 40-Zr, 41-Nb-93, 42-Mo, 47-Ag, 48-Cd, 49-In, 50-Sn, 51-Sb-121, 51-Sb-123, 63-Eu, 72-Hf, 73-Ta, 74-W, 82-Pb. Burnable materials: 5-B-10, 5-B-11, 72-Hf-176, 72-Hf-177, 72-Hf-178, 72-Hf-179, 72-Hf-180. Fission products: 36-Kr-83, 42-Mo-95, 43-Tc-99, 44-Ru-101, 44-Ru-103, 44-Ru-106, 45-Rh-103, 45-Rh-105, 46-Pd-105, 46-Pd-107, 46-Pd-108, 47-Ag-109, 48-Cd-113, 49-In-115, 51-Sb-125, 52-Te-127, 53-I-127, 53-I-135, 54-Xe
FOURACES, MultiGroup Cross-Sections, Resonance Calculation from ENDF/B, KEDAK, UKNDL
International Nuclear Information System (INIS)
1 - Description of problem or function: FOURACES produces spectrum weighted, group averaged nuclear cross sections and related parameters for nuclear reactor calculations. ENDF/B-IV, ENDF/B-V, KEDAK or UKNDL libraries may be used as basic input data. The weighting function and energy group structure are arbitrary, and are specified by the user. The code can deal with single or multi-level Breit-Wigner Adler- Adler and Reich-Moore resonance formalisms, and includes a Doppler broadening option. 4. Method of solution: If the weighting function is simple enough group averaged quantities are computed from the point data and interpolation rule read from the evaluated data library using analytic formulae. Otherwise the integrations are performed using the trapezium rule. Resonance data are converted into point data using subroutines written primarily for the program CRESO (abstract NEA 0719), then Doppler broadened, and finally group averaged. 5. Restrictions on the complexity of the problem: A maximum of 256 energy groups can be dealt with
Measurement committee of the US cross section evaluation working group. Annual report, 1995
International Nuclear Information System (INIS)
The Cross Section Evaluation Working Group is a long-standing committee charged with the responsibility for organizing and overseeing the U.S. cross-section evaluation effort. It's main product is the official U.S. evaluated nuclear data file, ENDF; the current version of this file is Version VI. All evaluations included in ENDF are reviewed and approved by CSEWG and issued by the U.S. Nuclear Data Center, Brookhaven National Laboratory. CSEWG is comprised of volunteers from the U.S. nuclear data community who possess expertise in evaluation methodologies and who collectively have been responsible for producing most of the evaluations included in ENDF. In 1992 CSEWG added the Measurements Committee to its list of standing committees and subcommittees. This was based on recognition of the importance of experimental data in the evaluation process as well as the realization that measurement activities in the U.S. were declining at an alarming rate. The mission of the Committee is to establish a network of experimentalists in the U.S. which would provide encouragement to the national nuclear data measurement effort through improved communication and facilitation of collaborative activities. The Committee currently has 19 members, and interested scientists are welcome to join the network simply by contacting the Chairman. For reference, the names of the current members and contact information are contained in this report. This annual report is the first such document issued by the Committee. It contains voluntary contributions from 10 laboratories in the U.S. which have been prepared by members of the Committee and submitted to the Chairman for compilation and editing. This report is being distributed in hard copy and is also available on-line via the National Nuclear Data Center, Brookhaven National Laboratory. It is hoped that the information provided here on the work that is going on at the reporting laboratories will prove interesting and stimulating to the readers
A 39 neutron group self-shielded cross section library for the Lotus fusion-fission test facility
International Nuclear Information System (INIS)
A 39 neutron group cross section library for fusion fission blanket calculations and especially for the analysis of the LOTUS experiment has been processed using the NJOY system. The library has been generated mostly using the ENDF/B-IV basic files at 296 K. All cross sections were self-shielded using the Bondarenko method. 5 background cross sections, namely 1010, 104, 102, 10 and 1 barns respectively were considered. The tabulated dilution dependent cross sections have been interpolated with the code TRANSX-CTR which is adequate for fusion applications. The fission spectrum of the fissionable material thorium has been collapsed from the fission matrices using the Bondarenko weighting scheme. The correct geometry of the LOTUS blanket and the cell specifications were correctly considered in the interpolation scheme. Some reaction cross sections for dosimetry applications have been included into the library. These base on the more recent ENDF/B-V evaluation. Transport and response edit cross sections have been coupled in the usual way to form P0 - P3 card image tables. Furthermore they have been converted into a binary file suitable to our RSYST computational system. Moreover the cross section card image tables have been reformatted and fitted into a BXSLIB binary library for the LANL-ONEDANT transport module. (Auth.)
International Nuclear Information System (INIS)
In the group cross section libraries usually applied for reactor calculations, the energy dependent probabilities of interactions between neutrons and the materials existing in the reactor are represented by weighted average values over certain energy ranges with a neutron energy spectrum regarded as representative. The influence of the resonance structure of the cross sections via the neutron spectrum and the resultant effect on the averaged group cross sections is taken into account in an approximate way by so-called resonance self-shielding factors. The approximations indicated are of considerable importance for the elastic down scattering. They can be improved by the so-called REMO correction, which takes into account the neutron energy distribution existing in the reactor model. Because such detailed neutron distributions are very expensive to prepare, especially in multi-dimensional models, automatic program runs were established which, in some cases by simplifications of the model, allow collision densities to be made available at relatively little expenditure which permit many nuclear quantities to be calculated with a sufficient degree of accuracy. This report describes the program runs set up and the experience acquired in testing them by the examples of the MASURCA 3B experiment and the SNEAK 11B2 assembly. This report deals especially with the influence of the collision density used for the REMO correction on the ksub(eff) value and other parameters of the reactor models considered. (orig.)
Validation of Nuclear Criticality Safety Software and 27 energy group ENDF/B-IV cross sections
International Nuclear Information System (INIS)
The validation documented in this report is based on calculations that were executed during June through August 1992, and was completed in June 1993. The statistical analyses in Appendix C and Appendix D were completed in October 1993. This validation gives Portsmouth NCS personnel a basis for performing computerized KENO V.a calculations using the Martin Marietta Nuclear Criticality Safety Software. The first portion of the document outlines basic information in regard to validation of NCSS using ENDF/B-IV 27-group cross sections on the IBM 3090 at ORNL. A basic discussion of the NCSS system is provided, some discussion on the validation database and validation in general. Then follows a detailed description of the statistical analysis which was applied. The results of this validation indicate that the NCSS software may be used with confidence for criticality calculations at the Portsmouth Gaseous Diffusion Plant. When the validation results are treated as a single group, there is 95% confidence that 99.9% of future calculations of similar critical systems will have a calculated Keff > 0.9616. Based on this result the Portsmouth Nuclear Criticality Safety Department has adopted the calculational acceptance criteria that a keff + 2σ ≤ 0.95 is safety subcritical. The validation of NCSS on the IBM 3090 at ORNL was extended to include NCSS on the IBM 3090 at K-25
A cross-sectional study of food group intake and C-reactive protein among children
Directory of Open Access Journals (Sweden)
Moore Lynn L
2009-10-01
Full Text Available Abstract Background C-reactive protein (CRP, a marker of sub-clinical inflammation, is a predictor of future cardiovascular diseases. Dietary habits affect serum CRP level however the relationship between consumption of individual food groups and CRP levels has not been established. Methods This study was designed to explore the relation between food intake and CRP levels in children using data from the cross-sectional 1999-2002 National Health and Nutrition Examination Surveys. CRP level was classified as low, average or high (3.0 mg/L, respectively. Adjusted mean daily intakes of dairy, grains, fruit, vegetables, and meat/other proteins in each CRP category were estimated using multivariate analysis of covariance modeling. The effect modification by age (5-11 years vs. 12-16 years, gender and race/ethnicity was explored. We examined whether total or central body fat (using BMI Z-scores and waist circumference explained any of the observed associations. Results A total of 4,010 children and adolescents had complete information on diet, CRP and all covariates of interest and were included in the analyses. Individuals with high CRP levels had significantly lower intake of grains (p Conclusion Children and adolescents with higher CRP levels had significantly lower intakes of grains and vegetables. The associations between selected childhood dietary patterns and CRP levels seem largely mediated through effects on body composition.
New approach to the adjustment of group cross-sections fitting integral measurements
International Nuclear Information System (INIS)
The adjustment of group cross-sections fitting integral measurements is viewed as a process of uncovering theoretical and/or experimental negligence errors to bring statistical consistency to the integral and differential data so that they can be combined to form an enlarged ensemble, on which an improved estimation of the physical constants can be based. An approach with three steps is suggested, and its formalism of general validity is developed. In step one, the data of negligence error are extracted from the given integral and differential data. The method of extraction is based on the concepts of prior probability and information entropy. It automatically leads to vanishing negligence error as the two sets of data are statistically consistent. The second step is to identify the sources of negligence error and adjust the data by an amount compensating the extracted negligence discrepancy. In the last step the two data sets, already adjusted to mutual consistency, are combined as a single unified ensemble. Standard methods of statistics can then be applied to re-estimate the physical constants. A simple example is shown as a demonstration of the method. 1 figure
ZZ RFL-2-DTF, Group Constant Library of Reaction Cross-Section, Gas Production, Kerma, DPA
International Nuclear Information System (INIS)
1 - Description of program or function: Format: DTF format and the structure is adopted from the MACKLIB-IV library. Number of groups: group library of reaction cross sections, gas production, kerma and DPA. Materials: H-1, H-2, H-3, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-Nat, N-14, N-15, O-16, F-19, Na-23, Mg-Nat, Al-27, Si-28, P-31, S-Nat, Cl-Nat, Ar-36, Ar-38, Ar-40, K-Nat, Ca-Nat, Ti-Nat, V-Nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Zr-Nat, Nb-93, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-100, Ag-107, Ag-109, Cd-Nat, In-113, In-115, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-12,2 Sn-124, Ba-130, Ba-134, Ba-135, Ba-136, Ba-137, Ba-138, Hf-174, Hf-176, Hf-177, Hf- 178, Hf-179, Hf-180, Ta-181, W-Nat, W-182, W-183, W-184, W-186, Re-185, Re-187, Pb-Nat, Bi-209. Temperatures: T=293.6 K. Origin: GEFF-2 and GEPDL. RFL-2 is a group library of reaction cross sections, gas production, kerma and DPA based upon GEFF-2 and GEPDL - which are included in the package ZZ-GEFF-2-GENDF - and upon DECNET - included in the ZZ-DECNET-GENDF package (see below the description of these libraries). RFL-2 has been derived from them by the GENTORFL code (GENdf To RFL). Its primary use is to complete the neutron transport libraries in ANISN or FIDO format with data normally not present in the traditional files. It includes all GEFF-2 materials at T=293.6 K and σ0 = infinity; as qualifying point it gives 'delayed' kerma and 'delayed' gamma-ray production matrices, i.e. the energy release and the photons, respectively, generated by the decay of radioactive nuclei produced in the primary reactions; decay events that occur within 10000 seconds from the primary reaction are taken into account. The library includes many isotopes, since for each natural element included in GEFF-2 the decay of all component isotopes have been traced out. The library is in DTF format and the structure is
International Nuclear Information System (INIS)
A 75 group neutron-photon coupled cross-section library has been developed for 42 reactor nuclides utilizing the basic cross-section files - ENDF/B-IV for neutrons and DLC-7F for photons. 50 neutron energy groups and gamma energy groups are included in this library which should be well suited to carry out safety, shielding and core physics studies of nuclear reactors based on fission or fusion processes. This library is also adequate for oil logging and mineral exploration investigations. (author). 11 refs., 3 tabs
Energy Technology Data Exchange (ETDEWEB)
Sohail, Muhammad; Kim, Myunghyun [Kyung Hee Univ., Yongin (Korea, Republic of)
2013-05-15
It has the applicability for the cases of arbitrary geometry or direct whole-core transport calculation. Conventionally in subgroup method the subgroup data is generated without considering resonance interference and is therefore included at the use of subgroup data. A modification in subgroup method to consider resonance interference explicitly in more consistent way has been proposed in this study. Owing to the fact that these self-shielded cross-sections in interference term is also lethargy dependent, it can be converted to subgroup level dependent self-shielded cross-sections. The proposed method is implemented in 3-D whole core transport lattice code nTRACER. More consistent method of resonance interference interaction has shown relatively negligible error in self shielded cross-section. This new interference treatment method is investigated at various temperatures and has shown better results regardless of temperature changes of mixture of resonance isotopes mixture.
Manning, Shannon D.; Springman, A. Cody; Million, Amber D.; Milton, Nicole R.; McNamara, Sara E.; Somsel, Patricia A.; Bartlett, Paul; Davies, H. Dele
2010-01-01
Background While Group B Streptococcus (GBS) human colonization and infection has long been suspected as originating from cows, several investigators have suggested that ongoing interspecies GBS transmission is unlikely due to genotyping data demonstrating that human and bovine-derived GBS strains represent mostly distinct populations. The possibility of ongoing transmission between humans and their livestock has not been systematically examined. Methodology/Principal Findings To examine ongoing interspecies transmission, we conducted a prospective cross-sectional cohort study of 68 families and their livestock. Stool specimens were collected from 154 people and 115 livestock; GBS was detected in 19 (12.3%) humans and 2 (1.7%) animals (bovine and sheep). Application of multilocus sequence typing (MLST) identified 8 sequence types (STs or clones), with STs 1 and 23 predominating. There were 11 families in which two members submitted stools and at least one had GBS colonization. In 3 of these families, both members (consisting of couples) were colonized, yielding a co-colonization rate of 27% (95% CI: 7%–61%). Two of these couples had strains with identical MLST, capsule (cps) genotype, susceptibility, and RAPD profiles. One couple co-colonized with ST-1 (cps5) strains also had a bovine colonized with the identical strain type. On multivariate analysis of questionnaire data, cattle exposure was a predictor of GBS colonization, with each unit increase in days of cattle exposure increasing the odds of colonization by 20% (P = 0.02). These results support interspecies transmission with additional evidence for transmission provided by the epidemiological association with cattle exposure. Conclusions/Significance Although GBS uncommonly colonizes livestock stools, increased frequency of cattle exposure was significantly associated with human colonization and one couple shared the same GBS strains as their bovine suggesting intraspecies transmission. These results
Díez, C. J.; Cabellos, O.; Martínez, J. S.
2015-01-01
Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.
ZZ BARC-27GRP, 27-Group Infinitely Dilute and Bondarenko Cross-Section Library from ENDF/B
International Nuclear Information System (INIS)
1 - Description of problem or function: - BARC-27GRP: Format: 1-DX; Number of groups: 27; Nuclides: U-235, U-238, Pu-239, Pu-240, Pu-241, C, O, H, Al, Si, Na, Mg, Cr, Fe, Ni, Mo; Origin: ENDF/B-IV; Weighting spectrum: flux weighting proportional to 1/ΣT(u); fission weighting plus 1/E spectrum. - BARC-35-A: Format: SPHINX, Fx2-TH; Number of groups: 35; Nuclides: Al, He, Si, H, Fe, O, C, Na, Li, B, Be, N, Ca, Mn, V, Mo, Pb, Pu, Gd, K, Sm, Dy, Lu, Nb, U, Cr, Ni, Th, Np, Am, Zr, Cd, Eu, Mg, Ta, Cm, F, Ti, W. Origin: ENDF/B-IV; Weighting spectrum: fission - 1/E - thermal Maxwellian. - IAEA0856/01: 27-group resonance self-shielding factors and infinite diluted Cross sections for U-235, U-238, Pu-239, Pu-240, Pu-241, C, O, H, Al, Si, Na, Mg, Cr, Fe, Ni, Mo, generated by using the basic cross section and resonance parameter data from the ENDF/B-4 library. 2 - Method of solution: The 27-group constants were obtained by integrating the microscopic data over group intervals using a flux weighting proportional to 1/ΣT(u) and a fission plus 1/E spectrum. The standard ABBN group structure is used. The self-shielding factors were calculated for the following temperatures: 300, 900, 2100 (degrees Kelvin) and for potential scattering Cross sections of 10000, 100, 10, 1 barns. A thermal group is also included. For the 35-group library, resonance self-shielding factors are given at 300, 900, and 2100 K for a variety of dilution constants. Group Cross sections cover the energy range from 15 MeV to 0.005 eV and have been derived using Bondarenko flux approximation with a fission-1/E-thermal Maxwellian spectrum. The scattering Cross sections have been represented by a P3 Legendre expansion
Directory of Open Access Journals (Sweden)
Mitsuya Yamakita
Full Text Available Participation in a sports group is key for the prevention of incident functional disability. Little is known about the correlates of older adults' participation in sports groups, although this could assist with the development of effective health strategies. The purpose of this study was to identify the demographic and biological, psychosocial, behavioral, social and cultural, and environmental correlates of sports group participation among Japanese older adults.Data were obtained from the Japan Gerontological Evaluation study, which was a population-based cohort of people aged ≥65 years without disability enrolled from 31 municipalities across Japan (n = 78,002. Poisson regression analysis was used to determine the associations between the factors and participation in sports groups.Non-regular participation in sports groups was associated with lower educational level, being employed, and working the longest in the agricultural/forestry/fishery industry among the demographic and biological factors and poor self-rated health and depression among the psychosocial factors. Of the behavioral factors, current smoking was negatively associated and current drinking was positively associated with regular participation in sports groups. Among the social and cultural factors, having emotional social support and participating in hobby clubs, senior citizen clubs, or volunteer groups were associated with a high prevalence of participation in sports groups. Perceptions of the presence of parks or sidewalks, good access to shops, and good accessibility to facilities were positively associated with participation in sports groups among the environmental factors.Our study suggests that the promotion of activities that could increase older adults' participation in sports groups should consider a broad range of demographic and biological, psychosocial, behavioral, social and cultural, and environmental factors. Although future longitudinal studies to elucidate
Standard reference and other important nuclear data by the Cross Section Evaluation Working Group
International Nuclear Information System (INIS)
This report is a current review of the status of standard reference and other important nuclear data pointing out data discrepancies, recommending new measurements, and comparing the current version of ENDF/B with data. Neutron reactions with 1H, 3H, 6Li, 10B, 12C, 59Co, 80Kr, 136Xe, 153Eu, and 197Au are included, along with reaction and other data for the actinide nuclei. Cross sections, spectra, etc., are given for some of the nuclides considered
International Nuclear Information System (INIS)
This paper presents the quantification of resonance interference effect for multi-group effective cross-section in lattice physics calculation. In the resonance self-shielding method based on the equivalence theory, the resonance interference effect among multiple nuclides cannot be treated directly to the multi-group effective cross-section. The continuous energy or the ultra-fine-group treatment can directly consider the effect, but the application to the fuel assembly geometry is not realistic with practical computation time. In the present study, the resonance interference effect to the multi-group effective cross-section is simply quantified by the resonance interference factor (RIF) in order to confirm the benefit for considering the effect. The RIF is generated for the typical pin-cell geometry of water moderated system. The multi-group effective cross-sections with and without RIFs are compared with the continuous energy Monte-Carlo result. As a result, the significant impact for considering the resonance interference effect is confirmed to the limited nuclide, reaction type and energy group. Fortunately, these have small effect on k-infinity because the resonance interference effect is mainly induced by the wide resonances of 238U to the other minor nuclides (e.g., 235U, 239Pu) in the limited resonance energy ranges. The results also show that the effect is small to the absorption cross-section of 238U, which is the dominant resonance nuclide in the fuel. The quantification results in the present study indicate a useful material to investigate the more advanced resonance treatment for the next generation lattice physics code. (author)
International Nuclear Information System (INIS)
We have developed a group of computer codes to realize the accurate transport calculation by using the multi-group double-differential form cross section. This type of cross section can correctly take account of the energy-angle correlated reaction kinematics. Accordingly, the transport phenomena in materials with highly anisotropic scattering are accurately calculated by using this cross section. They include the following four codes or code systems: PROF-DD : a code system to generate the multi-group double-differential form cross section library by processing basic nuclear data file compiled in the ENDF / B-IV or -V format, ANISN-DD : a one-dimensional transport code based on the discrete ordinate method, DOT-DD : a two-dimensional transport code based on the discrete ordinate method, MORSE-DD : a three-dimensional transport code based on the Monte Carlo method. In addition to these codes, several auxiliary codes have been developed to process calculated results. This report describes the calculation algorithm employed in these codes and how to use them. (author)
ZZ DLC-16 COBB, 123 Neutron-Group Cross-Section Library from ENDF/B for XSDRN Calculation
International Nuclear Information System (INIS)
1 - Nature of physical problem solved: Format: XSDRN; Number of groups: 123; Nuclides: H, D, He, Be-9, B-10, C-12, O-16, Na-23, Mg, Al-27, Ti, Cr, Mn-55, Fe, Ni, Cu, Cu-63, Cu-65, Nb-93, Mo, Xe-135, Sm-149, Eu-151, Eu-153, Gd, Dy-164, Lu-175, Lu-176, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-243, and Cm-244. Origin: Mainly ENDF/B; Weighting spectrum: Fast cross sections → 1/E (14 MeV to .414 eV) Thermal cross sections → 1/E (1.86 eV to 0.125 eV) → Maxwell-Boltzmann (0.125 eV to 0.0047 eV). The library is intended to be a source of evaluated data for the cross section preparation code XSDRN. It supplements, rather than replaces, the existing XSDRN master library which is distributed with the code package. The library contains data for H, D, He, 9-Be, 10-B, 12-C, 16-O, 23-Na, Mg, 27-Al, Ti, Cr, 55-Mn, Fe, Ni, Cu, 63-Cu, 65-Cu, 93-Nb, Mo, 135-Xe, 149-Sm, 151-Eu, 153-Eu, Gd, 164-Dy, 175-Lu, 176-Lu, 182-W, 183-W, 184-W, 186-W, 185-Re, 187-Re, 197-Au, 233-U, 234-U, 235-U, 236-U, 238-U, 237-Np, 238-Pu, 239-Pu, 240-Pu, 241-Pu, 242-Pu, 241-Am, 243-Am, and 244-Cm. 2 - Method of solution: The library contains ENDF/B version 2 cross sections processed through several steps (primarily by SUPERTOG) into the standard XSDRN 123-group energy structure. These steps are - (a) process fast cross sections with SUPERTOG into standard GAM-2 energy structure (14 MeV to 0.414 eV), using a 1/E weighting function, and produce a GAM-2 tape. (This step was performed by R. Q. Wright, Math Div., ORNL). (b) Process thermal cross sections with SUPERTOG into standard 30-group THERMOS energy group structure (1.86 eV to 0.0047 eV), using a Maxwell-Boltzmann distribution with temperature 293 deg.K as a weighting function for E < 0.125 eV coupled to a 1/E weighting function for E from 0.125 eV to 1.86 eV. (c) Compute room temperature free-gas kernels, using THERMOS tape-making program, and
Zupancic, Maja; Kavcic, Tina; Slobodskaya, Helena R.; Akhmetova, Olga A.
2016-01-01
Incremental predictive value of 5 broad and 13 narrow personality traits for academic achievement over and beyond age, gender, parental education, and country was examined in Russian and Slovene 8- to 15-year-olds. Personality data were collected from mothers (Russia: N = 994, Slovenia: N = 624) and adolescents (Russia: N = 481, Slovenia: N = 310)…
Jamali, Safieh; Javadpour, Shohreh; Mosalanejad, Leili; Parnian, Razieh
2016-01-01
Background: Sexual function is affected by personal and interpersonal factors, familial and social traditions, culture, religion, menopause, and aging. So, ethnicity is a determining factor in sexual function. The present study aimed to investigate the prevalence of sexual dysfunction and attitudes towards sexuality in postmenopausal women among three different ethnic groups in Iran. Methods: This cross-sectional study was conducted on 746 postmenopausal women between 50 and 89 years who refe...
International Nuclear Information System (INIS)
The authors analyse experimental data on the transmission and fission self-indication functions for 239Pu in the unresolved resonance region. Use is made of the method of generating a cross-section structure based on the multi-level R-matrix formalism (stochastic K-matrix method). Evaluations of the average resonance parameters and group constants for 239Pu are made. (author)
Too LaySan; Abraham Mathew; Subramaniam Shamini; Thomas Susan; Beh LooSee
2011-01-01
Abstract Background This is a pilot cross sectional study using both quantitative and qualitative approach towards tutors teaching large classes in private universities in the Klang Valley (comprising Kuala Lumpur, its suburbs, adjoining towns in the State of Selangor) and the State of Negeri Sembilan, Malaysia. The general aim of this study is to determine the difficulties faced by tutors when teaching large group of students and to outline appropriate recommendations in overcoming them. Fin...
International Nuclear Information System (INIS)
1 - Description of problem or function: - ZZ-IRDF-82: ENDF-5 Format; 620 group (SAND II) Dosimetry Library. Nuclides: Li, B, F, Na, Mg, Al, P, S, Sc, Ti, Mn, Fe, Co, Ni, Cu, Zn, Zr, Nb, Rh, In, I, Au, Th, U, Np, Pu, Am. - ZZ-IRDF-90: ENDF-6 Format; 640 groups extended SAND II structure. Nuclides: Li, B, F, Mg, Al, P, S, Sc, Ti, Cr, Mn, Fe, Co, Ni, Cu, Zn, Zr, Nb, Rh, Cd, Ir, Gd, Au, Th, U, Np, Pu, V. Damage cross section for Fe, Cr, Ni. Weighting spectrum: Maxwell spectrum, 1/E spectrum and Watt fission spectrum. - ZZ-IRDF-2002: ENDF-6 Format (pointwise cross-section data). SAND II 640 energy group structure (multigroup data). Nuclides: Li, B, F, Na, Mg, Al, P, S, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, As, Y, Zr, Nb, Rh, Ag, In, I, La, Pr, Tm, Ta, W, Au, Hg, Pb, Th, U, Np, Pu, Am, Cd, Gd. Damage cross section for Fe, Cr, Ni, Si, GaAs displacement. Weighting spectrum: - Typical MTR spectrum used in the input of the cross-section uncertainty processing code. - Flat weighting spectrum used in converting the pointwise cross-section data to the extended SAND-II group structure. - ZZ-IRDF-2002-ACE: ACE Format (continuous energy cross-section data for Monte Carlo). Nuclides: Li, B, F, Na, Mg, Al, P, S, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, As, Y, Zr, Nb, Rh, Ag, In, I, La, Pr, Tm, Ta, W, Au, Hg, Pb, Th, U, Np, Pu, Am, Cd, Gd. Damage cross section for Fe, Cr, Ni, Si, GaAs displacement. - (A) ZZ-IRDF-82: The 1982 version of the International Reactor Dosimetry File is composed of two different parts. The first part is made up of a collection of dosimetry Cross sections and the second part contains a collection of benchmark spectra. For ease of use in dosimetry applications both Cross sections and spectra are distributed in multigroup form. Each of these two parts is in the ENDF/B-V Format as a separate computer file. I) The dosimetry cross section library contains the following data: (1) The entire ENDF/B-V Dosimetry Library (Mod. 1) in the form of 620 group averaged Cross
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Bretscher, M.M.
1993-12-31
The WIMS-D4 code has been modified (WIMS-D4M) to produce microscopic isotopic cross sections in ISOTXS format for use in diffusion and transport calculations. Beginning with 69-group libraries based on ENDF/B-V data, numerous cell calculations have been made to prepare a set of broad group cross sections for use in diffusion calculations. Global calculations have been made for two control rod states of the Romanian steady state TRIGA reactor with 29 fresh HEU fuel clusters. Detailed Monte Carlo calculations also have been performed for the same reactor configurations using data based on ENDF/B-V. Results from these global calculations are compared with each other and with the measured excess reactivities. Although region-averaged macroscopic principal cross sections obtained from WIMS-D4M are in good agreement with the corresponding Monte Carlo values, problems exist with the high energy (E > 10 keV) microscopic hydrogen transport cross sections.
International Nuclear Information System (INIS)
1 - Description of problem or function: - JFS-V2-1: Format: ABBN energy structure; Number of groups: 25 group constants; Nuclides: Be, B-10, B-11, C, O, Na, Al, Si, Cr, Mn, Fe, Ni, Cu, Mo, Th-232, U-233, U-234, U-235, U-236, U-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, fission products of U-235, and fission products of Pu-239. Origin: ENDF/B-IV; Weighting spectrum: The cross-section adjustment has been made by using an auxiliary equation for simultaneous evaluations for U-235, U-238, and Pu-239. - JFS-V2-2: Format: JFS energy structure; Number of groups: 70 group constants; Nuclides: Be, B-10, B-11, C, O, Na, Al, Si, Cr, Mn, Fe, Ni, Cu, Mo, Th-232, U-233, U-234, U-235, U-236, U-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, fission products of U-235, and fission products of Pu-239. Origin: ENDF/B-IV; Weighting spectrum: The cross-section adjustment has been made by using an auxiliary equation for simultaneous evaluations for U-235, U-238, and Pu-239. - JFS-3/J2: Number of groups: 70 group constants with quarter-lethargy width; Nuclides: H-1, He-4, Be-9, B-10, B-11, C-12, N-14, O-16, Na-23, Al-27, Si, Ar, Ti, V, Cr, Mn-55, Fe, Ni, Cu, Zr, Nb-93, Mo, Eu-151, Eu-153, Gd, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, Ta-181, W, Th-228, Th-230, Th-232, Th-233, Th-234, Pa-233, U-233, U-234, U-235, U-236, U-238, Np-237, Np-239, Pu-236, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-242m, Am-242g, Am-243, Cm-242, Cm-243, Cm-244, Cm-245 and 12 LFPs for 4 mother nuclides (U-235, U-238, Pu-239 and Pu-241) and 3 burnup days (180, 1080 and 1800). Origin: JENDL-2 and ENDF/B-V; Weighting spectrum: the collision density spectrum for a typical large LMFBR core spectrum is used as the weighting function. ZZ-JFS-3/J2: To improve fast reactor group constant set JFS-3-J2 to be applicable for high burnup reactor calculation, burnup depending lumped FP group cross sections for mother fission isotopes of U-235, U-238, Pu-239 and Pu-241 have been generated by using 155 FP nuclides of JENDL-2
Use of smokeless tobacco among groups of Pakistani medical students – a cross sectional study
Directory of Open Access Journals (Sweden)
Ilyas Mahwish
2007-09-01
Full Text Available Abstract Background Use of smokeless tobacco is common in South Asia. Tobacco is a major preventable cause of morbidity and mortality. Doctors make one of the best avenues to influence patients' tobacco use. However, medical students addicted to tobacco are likely to retain this habit as physicians and are unlikely to counsel patients against using tobacco. With this background, this study was conducted with the objective of determining the prevalence of smokeless tobacco among Pakistani medical students. Methods A cross sectional study was carried out in three medical colleges of Pakistan – one from the north and two from the southern region. 1025 students selected by convenient sampling completed a peer reviewed, pre-tested, self-administered questionnaire. Questions were asked regarding lifetime use (at least once or twice in their life, current use (at least once is the last 30 days, and established use (more than 100 times in their life of smokeless tobacco. Chi square and logistic regression analyses were used. Results Two hundred and twenty (21.5% students had used tobacco in some form (smoked or smokeless in their lifetime. Sixty six (6.4% students were lifetime users of smokeless tobacco. Thirteen (1.3% were daily users while 18 (1.8% fulfilled the criterion for established users. Niswar was the most commonly used form of smokeless tobacco followed by paan and nass. Most naswar users belonged to NWFP while most paan users studied in Karachi. On univariate analysis, lifetime use of smokeless tobacco showed significant associations with the use of cigarettes, student gender (M > F, student residence (boarders > day scholars and location of the College (NWFP > Karachi. Multivariate analysis showed independent association of lifetime use of smokeless tobacco with concomitant cigarette smoking, student gender and location of the medical college. Conclusion The use of smokeless tobacco among medical students cannot be ignored. The
Chronic low back pain patient groups in primary care - A cross sectional cluster analysis
Viniol, Annika; Jegan, Nikita; Hirsch, Oliver; Leonhardt, Corinna; Brugger, Markus; Strauch, Konstantin; Barth, Juergen; Baum, Erika; Becker, Annette
2013-01-01
Background Due to the heterogeneous nature of chronic low back pain (CLBP), it is necessary to identify patient groups and evaluate treatments within these groups. We aimed to identify groups of patients with CLBP in the primary care setting. Methods We performed a k-means cluster analysis on a large data set (n = 634) of primary care patients with CLBP. Variables of sociodemographic data, pain characteristics, psychological status (i.e., depression, anxiety, somatization), and the patient re...
International Nuclear Information System (INIS)
A 35 group cross-section set with P3-anisotropic scattering matrices and resonance self-shielding factors has been generated from the basic ENDF/B-IV cross-section Library for 57 reactor elements. This library, called BARC35, is considered to be well suited for the neutronics and safety analysis of fission, fusion and hybrid systems. (author)
ZZ WM-NRSM, Neutron and Gamma Group Cross-Section Library for Nuclear Rocket Shielding Calculations
International Nuclear Information System (INIS)
Description of problem or function: - Master Library 1: Format: ANISN-W, DOT-IIW and APPROPOS. Number of groups: 52; Nuclides: Al, Be, B, B-10, Cd, C, Cr, Co, Cu, Gd, Au, H, In-115, Fe, Pb, Li, Li-6, Li-7, Mg, Mn, Mo, Ni, Nb, N, O, Si, Ta, Ti, W, U-235, U-238, Zr. Origin: Westinghouse Astro-nuclear Laboratory. Weighting spectrum: 1/E, flux and current spectra. - Master Library 2: Format: ANISN-W, DOT-IIW and APPROPOS. Number of groups: 52; Nuclides: Al, Be, B, B-10, Cd, C, Cr, Co, Cu, Gd, Au, H, In-115, Fe, Pb, Li, Li-6, Li-7, Mg, Mn, Mo, Ni, Nb, N, O, Si, Ta, Ti, W, U-235, U-238, Zr. Origin: Westinghouse Astro-nuclear Laboratory. Weighting spectrum: 1/E, flux and current spectra. - Master Library 3: Format: APPROPOS. Number of groups: 52; Nuclides: Al, Be, B, B-10, Cd, C, Cr, Co, Cu, Gd, Au, H, In-115, Fe, Pb, Li, Li-6, Li-7, Mg, Mn, Mo, Ni, Nb, N, O, Si, Ta, Ti, W, U-235, U-238, Zr. Origin: Westinghouse Astro-nuclear Laboratory. Weighting spectrum: 1/E, flux and current spectra. - Master Library 5: Format: KAP-VI, GAMLEG-W, MAC and SCAP. Number of groups: energy points in the range of 0.01 MeV to 20.0 MeV; Nuclides: H, He, Li, Be, B, C, N, O, Na, Mg, Al, Si, P, S, K, Ca, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Y, Zr, Nb, Mo, Ag, Cd, In, Sn, Cs, Ba, Sm, Gd, Dy, Y, Hf, Ta, W, Au, Hg, Pb, Po, Th, Pa, U, Np, Pu. Origin: Westinghouse Astro-nuclear Laboratory. - Basic set of nuclear data (Library 6): Format: ANISN-W and DOT-IIW. Number of groups: 52; Nuclides: H, Be, B, C, U-235, U-238, N, O, Mg, Al, Si, Cr, Mn, Fe Co, Ni, Cu, Zr, Mo, Ag, In, Cd, Gd, Pb, Nb, Ti, Ta, Li-6, Li-7, B-10 W, S. Origin: Master Libraries. Weighting spectrum: decided by user. WANL-MSFC Nuclear Rocket Shielding Data Generators GAMLEG-W, APPROPOS, NAGS, SATURN and Neutron and photon Cross Section Libraries 1-6. Applications of the Data: Transport codes which use the data are ANISN-W, KAP-VI, DOT-IIW, MAC and SCAP. The transport codes, also available from RSIC, the cross section processing codes, and
ZZ DLC-2D/100G, 100 Neutron-Group Cross-Section Library by SUPERTOG Calculation for ANISN, DOT
International Nuclear Information System (INIS)
1 - Nature of physical problem solved: Format: ANISN, DOT or DTF-4; Number of groups: 100; Nuclides: H, D, He, He-3, Li-6, Li-7, Be-9, B-10, B-11, C-12, N-14, O-16, Na-23, Mg, Al-27, Si, Cl, K, Ca, V, Cr, Mn-55, Fe, Co-59, Ni, Cu, Cu-63, Cu-65, Nb, Mo, Ag-107, Xe-135, Cs-133, Sm-149, Eu-151, Eu-153, Gd, Dy-164, Lu-175, Lu-176, Ta-181, Ta-182, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, Pb, Th-232, Pa-233, U-234, U-235, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-243, and Cm-244. Origin: The nuclides in DLC-2 are those which have been released as category I ENDF/B by the National Neutron Cross Section Center, Brookhaven National Laboratory. Weighting spectrum: The explicit assumption was made that the flux has the shape of a fission spectrum joined at 0.0674 MeV by a 1/E tail. Neutron transport calculations can be performed with DLC-2 data. Since the data are intended for use in multigroup discrete-ordinates or Monte Carlo transport codes which treat anisotropic scattering, possible cross section angular expansion is limited only by the options available in the particular code used. Specifically, the retrieval program manipulates DLC-2 such that it conforms to input requirements of the ANISN, DOT, or DTF-4 codes, or any computer code using data in the ANISN or DTF-4 format. The nuclides in DLC-2 are those which have been released as category I ENDF/B by the National Neutron Cross Section Center, Brookhaven National Laboratory. The library contains data for H, D, He, 3-He, 6-Li, 7-Li, 9-Be, 10-B, 11-B, 12-C, 14-N, 16-O, 23-Na, Mg, 27-Al, Si, Cl, K, Ca, V, Cr, 55-Mn, Fe, 59-Co, Ni, Cu, 63-Cu, 65-Cu, Nb, Mo, 107-Ag, 135-Xe, 133-Cs, 149-Sm, 151-Eu, 153-Eu, Gd, 164-Dy, 175-Lu, 176-Lu, 181-Ta, 182-Ta, 182-W, 183-W, 184-W, 186-W, 185-Re, 187-Re, 197-Au, Pb, 232-Th, 233-Pa, 234-U, 235-U, 238-U, 238-Pu, 239-Pu, 240-Pu, 241-Pu, 242-Pu, 241-Am, 243-Am, and 244-Cm. 2 - Method of solution: DLC-2 was generated by SUPERTOG from nuclear data in either point
Verification of a Multi-group Cross Section Library for Burnup Calculation
Energy Technology Data Exchange (ETDEWEB)
Daing, Aung Tharn; Kim, Myung Hyun [Kyung Hee Univ., Yongin (Korea, Republic of); Joo, Hang Yu [Seoul National Univ., Seoul (Korea, Republic of)
2013-05-15
Despite satisfying the estimation of the neutronic parameters without depletion to some extent, it still requires detailed investigation of the behavior of a fuel with strong neutron absorber over its operating life time by nTRACER, the direct whole core calculation code with the conventional semi Predictor-Corrector method. This study is mainly focused on the verification of the newly generated multi-group library for burnup calculation by nTRACER through the analysis of its performance of depletion calculation of UO{sub 2} fuel with strong neutron absorbers such as Gadolinium. Firstly, the depletion calculation results of nTRACER are presented by comparing the evolution of k-inf and the inventories of commonly found important isotopes as a function of burnup in the cases of gadolinia(GAD)-bearing fuel pin and fuel assembly (FA) with those of MCNPX-version.2.6.0. The newly generated multi-group library for burnup calculation by nTRACER was verified through GAD-bearing fuel after the new approach of resonance treatment had been employed. Though very good agreement in the overall effect reflected on the multiplication factor of FA at BOC, the evolution of k-inf along fuel irradiation history was systematically well underestimated by nTRACER when compared to Monte Carlo results.
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Smith, D.L. [ed.] [comp.] [Argonne National Lab., IL (United States); McLane, V. [ed.] [comp.] [Brookhaven National Lab., Upton, NY (United States)
1997-10-01
The Cross-Section Evaluation Working Group (CSEWG) is a long-standing committee charged with responsibility for organizing and overseeing the US cross-section evaluation effort. It`s main product is the official US evaluated nuclear data file, ENDF. In 1992 CSEWG added the Measurements Committee to its list of standing committees and subcommittees. This action was based on a recognition of the importance of experimental data in the evaluation process as well as the realization that measurement activities in the US were declining at an alarming rate and needed considerable encouragement to avoid the loss of this resource. The mission of the Committee is to maintain contact with experimentalists in the Us and to encourage them to contribute to the national nuclear data effort. Improved communication and the facilitation of collaborative activities are among the tools employed in achieving this objective. In 1994 the Committee was given an additional mission, namely, to serve as an interface between the applied interests represented in CSEWG and the basic nuclear science community. Accordingly, its name was changed to the Measurement and Basic Physics Committee. The present annual report is the third such document issued by the Committee. It contains voluntary contributions from several laboratories in the US. Their contributions were submitted to the Chairman for compilation and editing.
Internet addiction in a group of medical students: a cross sectional study.
Pramanik, T; Sherpa, M T; Shrestha, R
2012-03-01
The use of Internet for education, recreation and communication is increasing day by day. Nevertheless, the possibility of exploitation and addiction leading to impairment in academic performance and emotional balance cannot be denied, especially among young population. The study was aimed to measure the degree of Internet addiction among a group of medical students. Internet addiction test questionnaire developed by Young was used to assess mild, moderate and severe addiction. Amongst the study population (n=130, age 19-23 years), 40% had mild addiction. Moderate and severe addiction was found in 41.53% and 3.07% of the participants respectively. The study revealed that 24% often and 19.2% always found themselves using Internet longer than they had planned or thought. Late night Internet surfing leading to sleep deprivation was found in 31.53% of the participants. Almost one fourth of them (25.38%) occasionally tried to cut down the time they spent on the Internet but failed and 31.53% sometimes experienced restlessness when deprived of Internet access. Results reflected that a significant number of participants suffered from mild to moderate addiction. The role of counseling and education should be emphasized for prevention of Internet addiction. PMID:23441494
ZZ SINEX, 100 Neutron-Group Neutron Reaction Cross-Section Library from ENDF/B by SUPERTOG for ANISN
International Nuclear Information System (INIS)
1 - Nature of physical problem solved: Format: ANISN; Number of groups: 100 group reaction cross sections for neutron interactions. Nuclides: H, D, He, He-3, Li-6, Li-7, Be-9, B-10, B-11, C-12, N-14, O-16, Na-23, Mg, Al-27, Si, Cl, K, Ca, V, Cr, Mn-55, Fe, Co-59, Ni, Cu, Cu-63, Cu-65, Nb, Mo, Ag-107, Ag-109, Xe-135, Cs-133, Sm-149, Eu-151, Eu-153, Gd, Dy-164, Lu-175, Lu-176, Ta-181, Ta-182, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, Pb, Th-232, Pa-233, U-234, U-235, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-243, and Cm-244. Origin: ENDF/B; Weighting spectrum: For the top 99 groups, the explicit assumption was made that the flux (weighting function) has the shape of a fission spectrum jointed at 0.0674 MeV by a 1/E tail. For the thermal group (group 100), values for all materials except hydrogen were taken from the Maxwellian average values derived from the ENDF/B data. The data can be used in combination with 100 group neutron transport calculations (using, e. g., the DLC-2 library) to determine the spatial distribution of individual reaction rates. In particular, the retrieval program allows the preparation of dummy materials based on DLC-24 which can be used in the activity calculation option in ANISN to calculate the desired reaction rates. The library consists of 100 group reaction cross sections for neutron interactions as follows - total, elastic, inelastic, (n,2n), fission, (n,n'α), (n,n'3α), (n,2nα), absorption, (n,n'p), capture, (n,γ), (n,p), (n,d), (n,t), (n,He3), (n,α), (n,2α), and ν-bar. The units are barns, except that ν-bar is the average number of neutrons per fission event. A table listing the reactions included for each material is found in ref.1. The nuclides in DLC-24 are those which have been released as category I ENDF/B by the National Neutron Cross Section Center, Brookhaven National Laboratory. The library contains data for H, D, He, 3-He, 6-Li, 7-Li, 9-Be, 10-B, 11-B, 12-C, 14-N, 16-O, 23-Na, Mg, 27-Al, Si
International Nuclear Information System (INIS)
The Cross-Section Evaluation Working Group (CSEWG) is a long-standing committee charged with responsibility for organizing and overseeing the US cross-section evaluation effort. Its main product is the official US evaluated nuclear data file, ENDF. The current version of this file is Version VI. All evaluations included in ENDF, as well as periodic modifications and updates to the file, are reviewed and approved by CSEWG and issued by the US Nuclear Data Center, Brookhaven National Laboratory. CSEWG is comprised of volunteers from the US nuclear data community who possess expertise in evaluation methodologies and who collectively have been responsible for producing most of the evaluations included in ENDF. In 1992 CSEWG added the Measurements Committee to its list of standing committees and subcommittees. This action was based on a recognition of the importance of experimental data in the evaluation process as well as the realization that measurement activities in the US were declining at an alarming rate and needed considerable encouragement to avoid the loss of this resource. The mission of the Committee is to maintain contact with experimentalists in the US and to encourage them to contribute to the national nuclear data effort. Improved communication and the facilitation of collaborative activities are among the tools employed in achieving this objective. In 1994 the Committee was given an additional mission, namely, to serve as an interface between the applied interests represented in CSEWG and the basic nuclear science community. Accordingly, its name was changed to the Measurement and Basic Physics Committee. The present annual report is the third such document issued by the Committee. It contains voluntary contributions from several laboratories in the US. Their contributions were submitted to the Chairman for compilation and editing
International Nuclear Information System (INIS)
The Cross-Section Evaluation Working Group (CSEWG) is a long-standing committee charged with the responsibility for organizing and overseeing the U.S. cross-section evaluation effort. It's main product is the official U.S. evaluated nuclear data file, ENDF. The current version of this file is Version VI. All evaluations included in ENDF are reviewed and approved by CSEWG and issued by the U.S. Nuclear Data Center, Brookhaven National Laboratory. CSEWG is comprised of volunteers from the U.S. nuclear data community who possess expertise in evaluation methodologies and who collectively have been responsible for producing most of the evaluations included in ENDF. In 1992 CSEWG added the Measurements Committee to its list of standing committees and subcommittees. This action was based on a recognition of the importance of experimental data in the evaluation process as well as the realization that measurement activities in the U.S. were declining at an alarming rate and needed all possible encouragement to avoid the loss of this resource. The mission of the Committee is to maintain a network of experimentalists in the U.S. that would provide needed encouragement to the national nuclear data measurement effort through improved communication and facilitation of collaborative activities. In 1994, an additional charge was added to the responsibilities of this Committee, namely, to serve as an interface between the more applied interests represented in CSEWG and the basic nuclear science community. This annual report is the second such document issued by the Committee. It contains voluntary contributions from eleven laboratories in the U.S. which have been prepared by members of the Committee and submitted to the Chairman for compilation and editing. It is hoped that the information provided here on the work that is going on at the reporting laboratories will prove interesting and stimulating to the readers
Energy Technology Data Exchange (ETDEWEB)
SMITH,D.L.; MCLANE,V.
1998-10-20
The Cross-Section Evaluation Working Group (CSEWG) is a long-standing committee charged with responsibility for organizing and overseeing the US cross-section evaluation effort. Its main product is the official US evaluated nuclear data file, ENDF. The current version of this file is Version VI. All evaluations included in ENDF, as well as periodic modifications and updates to the file, are reviewed and approved by CSEWG and issued by the US Nuclear Data Center, Brookhaven National Laboratory. CSEWG is comprised of volunteers from the US nuclear data community who possess expertise in evaluation methodologies and who collectively have been responsible for producing most of the evaluations included in ENDF. In 1992 CSEWG added the Measurements Committee to its list of standing committees and subcommittees. This action was based on a recognition of the importance of experimental data in the evaluation process as well as the realization that measurement activities in the US were declining at an alarming rate and needed considerable encouragement to avoid the loss of this resource. The mission of the Committee is to maintain contact with experimentalists in the US and to encourage them to contribute to the national nuclear data effort. Improved communication and the facilitation of collaborative activities are among the tools employed in achieving this objective. In 1994 the Committee was given an additional mission, namely, to serve as an interface between the applied interests represented in CSEWG and the basic nuclear science community. Accordingly, its name was changed to the Measurement and Basic Physics Committee. The present annual report is the third such document issued by the Committee. It contains voluntary contributions from several laboratories in the US. Their contributions were submitted to the Chairman for compilation and editing.
A Pebble Bed Reactor cross section methodology
International Nuclear Information System (INIS)
A method is presented for the evaluation of microscopic cross sections for the Pebble Bed Reactor (PBR) neutron diffusion computational models during convergence to an equilibrium (asymptotic) fuel cycle. This method considers the isotopics within a core spectral zone and the leakages from such a zone as they arise during reactor operation. The randomness of the spatial distribution of fuel grains within the fuel pebbles and that of the fuel and moderator pebbles within the core, the double heterogeneity of the fuel, and the indeterminate burnup of the spectral zones all pose a unique challenge for the computation of the local microscopic cross sections. As prior knowledge of the equilibrium composition and leakage is not available, it is necessary to repeatedly re-compute the group constants with updated zone information. A method is presented to account for local spectral zone composition and leakage effects without resorting to frequent spectrum code calls. Fine group data are pre-computed for a range of isotopic states. Microscopic cross sections and zone nuclide number densities are used to construct fine group macroscopic cross sections, which, together with fission spectra, flux modulation factors, and zone buckling, are used in the solution of the slowing down balance to generate a new or updated spectrum. The microscopic cross-sections are then re-collapsed with the new spectrum for the local spectral zone. This technique is named the Spectral History Correction (SHC) method. It is found that this method accurately recalculates local broad group microscopic cross sections. Significant improvement in the core eigenvalue, flux, and power peaking factor is observed when the local cross sections are corrected for the effects of the spectral zone composition and leakage in two-dimensional PBR test problems.
Directory of Open Access Journals (Sweden)
Too LaySan
2011-09-01
Full Text Available Abstract Background This is a pilot cross sectional study using both quantitative and qualitative approach towards tutors teaching large classes in private universities in the Klang Valley (comprising Kuala Lumpur, its suburbs, adjoining towns in the State of Selangor and the State of Negeri Sembilan, Malaysia. The general aim of this study is to determine the difficulties faced by tutors when teaching large group of students and to outline appropriate recommendations in overcoming them. Findings Thirty-two academics from six private universities from different faculties such as Medical Sciences, Business, Information Technology, and Engineering disciplines participated in this study. SPSS software was used to analyse the data. The results in general indicate that the conventional instructor-student approach has its shortcoming and requires changes. Interestingly, tutors from Medicine and IT less often faced difficulties and had positive experience in teaching large group of students. Conclusion However several suggestions were proposed to overcome these difficulties ranging from breaking into smaller classes, adopting innovative teaching, use of interactive learning methods incorporating interactive assessment and creative technology which enhanced students learning. Furthermore the study provides insights on the trials of large group teaching which are clearly identified to help tutors realise its impact on teaching. The suggestions to overcome these difficulties and to maximize student learning can serve as a guideline for tutors who face these challenges.
International Nuclear Information System (INIS)
FILLC is a utility subroutine for use in the KAPROS versions of multigrop SN codes. Within KAPROS a couple of modules create and modify nuclear cross sections and store them in a so-called SIGMN file. FILLC stores group cross sections in an array C used by some SN codes with data provided in a SIGMN file. This report documents this interface with release number 3.4. (orig.)
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Spittaels Heleen
2012-12-01
Full Text Available Abstract Background From a health perspective it is suggested to promote a positive balance between time spent in light intensity physical activity (LIPA and sedentary behaviour (SB (i.e. spending more time in LIPA than time spent in SB. However, no studies have reported prevalence rates of the LIPA-SB balance yet. The aim of this study was to objectively investigate the time spent in SB, in LIPA and moderate-to-vigorous intensity physical activity (MVPA in four Belgian age groups and to explore which proportion of the population had a favorable balance between LIPA and SB and combined this with recommended amount of MVPA. Methods Accelerometer data from 7 cross-sectional studies (N=2083 in four age groups (preschoolers, primary schoolchildren, secondary schoolchildren and adults were aggregated. Differences in SB and PA between age groups and between men and women were determined by two-way MANCOVA. LIPA-SB balance was calculated and participants were categorized into one of four groups: (1 positive LIPA-SB balance (LIPA> SB & sufficient MVPA (2 negative LIPA-SB balance & sufficient MVPA (3 positive LIPA-SB balance & insufficient MVPA (4 negative LIPA-SB balance & insufficient MVPA. Results For the total sample, 55% of the waking time was spent in SB, 39% in LIPA and 6% in MVPA. Differences in SB between age groups was dependent from gender (p Conclusion A high proportion of the Belgian population is at risk if taking into account both SB and PA levels. Secondary schoolgirls have the unhealthiest SB and PA profile and are therefore an important target group for interventions both increasing MVPA and decreasing SB. In men more attention should be given in promoting a positive LIPA-SB balance independently from their compliance with the MVPA guidelines.
International Nuclear Information System (INIS)
1 - Description of problem or function: Format: AMPX, CCCC format. Number of groups: 171 neutron, 36 gamma-ray group cross section; Nuclides: H, He, Li, Be, B, C, N, O, F, Na, Mg, Al, Si, P, S, K, Ca, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Nb, Mo, Ag, Cd, Sn, Eu, Ta, W, Pb, Th, U, Pu, Am Origin: ENDF/B; Weighting spectrum: - Maxwellian thermal spectrum (300 K) up to 0.125 eV; - 1/E slowing down spectrum to 820.8 KeV; - fission spectrum to 10.0 MeV; - 1/E spectrum to 12.57 MeV; - velocity exponential fusion peak to 15.57 MeV - 1/E spectrum to 17.333 MeV. Gamma cross sections are flat-weighted. ZZ-VITAMIN-C/B is a data library containing 171 neutron, 36 gamma-ray group cross section data for fusion and LMFBR neutronics. The weighting function for the neutron data consists of six regions: a Maxwellian thermal spectrum (300 K) up to 0.125 eV, a 1/E slowing down spectrum to 820.8 KeV, a fission spectrum to 10.0 MeV, a 1/E spectrum to 12.57 MeV, a velocity exponential fusion peak to 15.57 MeV, and a 1/E spectrum to 17.333 MeV. The gamma cross sections are flat-weighted. The Legendre expansion order is P3. Data are available in the form of a 171 neutron group AMPX interface, a 171 neutron 36 gamma- ray production AMPX interface, and a 36 group gamma-ray interaction AMPX interface. The neutron data are also available in CCCC format as ISOTXS neutron cross section and BRKOXS self-shielding factor standard interface files. Data exists for 60 nuclides used in proposed fusion or fission/ fusion hybrid systems. 2 - Method of solution: The data has been computed from ENDF/B files using the program MINX. The Bondarenko method is employed to create group dependent resonance self-shielding factors to account for temperature and dilution effects. The fractional error tolerances were set to 0.5 percent, and 0.1 percent for integration
Energy Technology Data Exchange (ETDEWEB)
Lee, B.L. Jr. [Battelle, Columbus, OH (United States); D`Aquila, D.M. [Lockheed Martin Utility Services, Inc., Oak Ridge, TN (United States)
1996-01-01
The original validation report, POEF-T-3636, was documented in August 1994. The document was based on calculations that were executed during June through August 1992. The statistical analyses in Appendix C and Appendix D were completed in October 1993. This revision is written to clarify the margin of safety being used at Portsmouth for nuclear criticality safety calculations. This validation gives Portsmouth NCS personnel a basis for performing computerized KENO V.a calculations using the Lockheed Martin Nuclear Criticality Safety Software. The first portion of the document outlines basic information in regard to validation of NCSS using ENDF/B-IV 27-group cross sections on the IBM3090 at ORNL. A basic discussion of the NCSS system is provided, some discussion on the validation database and validation in general. Then follows a detailed description of the statistical analysis which was applied. The results of this validation indicate that the NCSS software may be used with confidence for criticality calculations at the Portsmouth Gaseous Diffusion Plant. For calculations of Portsmouth systems using the specified codes and systems covered by this validation, a maximum k{sub eff} including 2{sigma} of 0.9605 or lower shall be considered as subcritical to ensure a calculational margin of safety of 0.02. The validation of NCSS on the IBM 3090 at ORNL was extended to include NCSS on the IBM 3090 at K-25.
International Nuclear Information System (INIS)
The original validation report, POEF-T-3636, was documented in August 1994. The document was based on calculations that were executed during June through August 1992. The statistical analyses in Appendix C and Appendix D were completed in October 1993. This revision is written to clarify the margin of safety being used at Portsmouth for nuclear criticality safety calculations. This validation gives Portsmouth NCS personnel a basis for performing computerized KENO V.a calculations using the Lockheed Martin Nuclear Criticality Safety Software. The first portion of the document outlines basic information in regard to validation of NCSS using ENDF/B-IV 27-group cross sections on the IBM3090 at ORNL. A basic discussion of the NCSS system is provided, some discussion on the validation database and validation in general. Then follows a detailed description of the statistical analysis which was applied. The results of this validation indicate that the NCSS software may be used with confidence for criticality calculations at the Portsmouth Gaseous Diffusion Plant. For calculations of Portsmouth systems using the specified codes and systems covered by this validation, a maximum keff including 2σ of 0.9605 or lower shall be considered as subcritical to ensure a calculational margin of safety of 0.02. The validation of NCSS on the IBM 3090 at ORNL was extended to include NCSS on the IBM 3090 at K-25
Energy Technology Data Exchange (ETDEWEB)
DUNFORD, C.; HOLDEN, N.; PEARLSTEIN, S.
2001-11-05
This publication has been prepared to record some of the history of the Cross Section Evaluation Working Group (CSEWG). CSEWG is responsible for creating the evaluated nuclear data file (ENDF/B) which is widely used by scientists and engineers who are involved in the development and maintenance of applied nuclear technologies. This organization has become the model for the development of nuclear data libraries throughout the world. The data format (ENDF) has been adopted as the international standard. On November 5, 2001, a symposium was held at Brookhaven National Laboratory to celebrate the 50 th meeting of the CSEWG organization and the 35 th anniversary of its first meeting in November 1966. The papers presented in this volume were prepared by present and former CSEWG members for presentation at the November 2001 symposium. All but two of the presentations are included. I have included an appendix to list all of the CSEWG members and their affiliations, which has been compiled from the minutes of each of the CSEWG meetings. Minutes exist for all meetings except the 4 th meeting held in January 1968. The list includes 348 individuals from 71 organizations. The dates for each of the 50 CSEWG meetings are listed. The committee structure and chairmen of all committees and subcommittees are also included in the appendix. This volume is dedicated to three individuals whose foresight and talents made CSEWG possible and successful. They are Henry Honeck who lead the effort to develop the ENDF format and the CSEWG system, Ira Zartman, the Atomic Energy Commission program manager who provided the programmatic direction and support, and Sol Pearlstein who led the development of the CESWG organization and the ENDF/B evaluated nuclear data library.
Minnesota Department of Natural Resources — FEMA Cross Sections are required for any Digital Flood Insurance Rate Map database where cross sections are shown on the Flood Insurance Rate Map (FIRM). Normally...
Orsi R.; Sinitsa V.; Pescarini M.; Frisoni M.
2013-01-01
This paper presents a synthesis of the ENEA-Bologna Nuclear Data Group programme dedicated to generate and validate group-wise cross section libraries for shielding and radiation damage deterministic calculations in nuclear fission reactors, following the data processing methodology recommended in the ANSI/ANS-6.1.2-1999 (R2009) American Standard. The VITJEFF311.BOLIB and VITENDF70.BOLIB finegroup coupled n-γ (199 n + 42 γ – VITAMIN-B6 structure) multi-purpose cross section libraries, based o...
Energy Technology Data Exchange (ETDEWEB)
Kim, Jong Woon; Kim, Sang Ji; Gil, Choong-Sup; Lee, Young-Ouk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2014-10-15
The unresolved resonance region (URR) begins at an energy where it is difficult to measure individual resonances and extends to an energy where the effects of fluctuations in the resonance cross sections become unimportant for practical calculations. In ENDF-format evaluations, this 'unresolved range' is handled by giving average values for the resonance spacing and the various partial widths, together with their probability distributions. These unresolved resonance parameters are used two ways in view of transport solver. For a deterministic method, the self-shielded multi-group cross sections are generated by UNRESR and GROUPR modules of NJOY code which use Bondarenko method. For a Monte Carlo method, so-called Bondarenko method is not very useful for continuous-energy Monte Carlo codes like MCNP. The natural approach for treating unresolved-resonance self-shielding for Monte Carlo codes is the 'Probability Table' method. The PURR module produces probability tables that can be used in versions of MCNP from 4B on to treat unresolved-resonance self-shielding. We present a method to generate self-shielded multi-group cross sections in URR for easy numerical integration and tested on the total cross section of {sup 239}Pu. This is the first phase of study and the effects of statistical resonances in URR are identified by comparing generated multi-group cross sections. Test will be performed on several other nuclides and this method might be used as a one of items for developing multi-group cross section generation code for fast reactor analysis.
International Nuclear Information System (INIS)
The unresolved resonance region (URR) begins at an energy where it is difficult to measure individual resonances and extends to an energy where the effects of fluctuations in the resonance cross sections become unimportant for practical calculations. In ENDF-format evaluations, this 'unresolved range' is handled by giving average values for the resonance spacing and the various partial widths, together with their probability distributions. These unresolved resonance parameters are used two ways in view of transport solver. For a deterministic method, the self-shielded multi-group cross sections are generated by UNRESR and GROUPR modules of NJOY code which use Bondarenko method. For a Monte Carlo method, so-called Bondarenko method is not very useful for continuous-energy Monte Carlo codes like MCNP. The natural approach for treating unresolved-resonance self-shielding for Monte Carlo codes is the 'Probability Table' method. The PURR module produces probability tables that can be used in versions of MCNP from 4B on to treat unresolved-resonance self-shielding. We present a method to generate self-shielded multi-group cross sections in URR for easy numerical integration and tested on the total cross section of 239Pu. This is the first phase of study and the effects of statistical resonances in URR are identified by comparing generated multi-group cross sections. Test will be performed on several other nuclides and this method might be used as a one of items for developing multi-group cross section generation code for fast reactor analysis
Energy Technology Data Exchange (ETDEWEB)
Voi, Dante Luiz Voi [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil); Rocha, Helio Fenandes da [Universidade Federal, Rio de Janeiro, RJ (Brazil). Inst. de Puericultura e Pediatria Martagao Gesteira
2002-07-01
Amino acids used in parenteral administration in hospital patients with special importance in nutritional applications were analyzed to compare with the manufactory data. Individual amino acid samples of phenylalanine, cysteine, methionine, tyrosine and threonine were measured with the neutron crystal spectrometer installed at the J-9 irradiation channel of the 1 kW Argonaut Reactor of the Instituto de Engenharia Nuclear (IEN). Gold and D{sub 2}O high purity samples were used for the experimental system calibration. Neutron cross section values were calculated from chemical composition, conformation and molecular structure analysis of the materials. Literature data were manipulated by parceling and grouping neutron cross sections. (author)
International Nuclear Information System (INIS)
Characteristics and contents of the CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data) 227-neutron-group AMPX master and pointwise cross-section libraries are described. Results obtained in using CSRL-V to calculate performance parameters of selected thermal reactor and criticality safety benchmarks are discussed
International Nuclear Information System (INIS)
The ENEA-Bologna Nuclear Data Group produced the JEFF-3.1 VITJEFF31.BOLIB and MATJEFF31. BOLIB fine-group coupled neutron and photon (199 n + 42 γ) cross section libraries for nuclear fission applications, respectively in AMPX and MATXS format, with the same specifications and energy group structure of the Endf/B-VI-3 VITAMIN-B6 American library. Each library, containing 181 nuclide cross section files, was generated from the same set of cross section data files in GENDF format, obtained through the Bondarenko (f-factor) method, with an ENEA-Bologna revised version of the GROUPR module of the NJOY-99.160 system. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the DANTSYS and DOORS systems, can be generated from VITJEFF31.BOLIB and MATJEFF31.BOLIB through, respectively, further data processing with an ENEA-Bologna revised version of the SCAMPI system and with the TRANSX code. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF31.BOLIB validation. (authors)
International Nuclear Information System (INIS)
The procedure to calculate the effective cross section in the criticality safety evaluation code system JACS was to interpolate the Multigroup Cross Section Library MGCL with respect to the background cross section. For a reference calculation to the calculation following the Bondarenko method, a computational module RABTH has been developed to obtain the eigen ultra-fine (64,194 group) neutron flux with the collision probability method and the effective cross section by weighting with the flux. In the RABTH code module, the neutron source has an energy spectrum of fission neutrons from 235U. The module utilizes the RABBLE code to solve the equations for one-dimensional cells in the fast energy groups, higher than about 1.9 eV, and the THERMOS code in the thermal energy groups, less than this energy. In this way the neutron flux distribution that covers the whole energy range is obtained. Both codes have been extended to treat not only slab and cylindrical cells but also a spherical cell in both complete reflective and vacuum boundary conditions, and the THERMOS code has been further revised for higher precision. This report includes practical information to treat RABTH module and basic equations for the extension and revision made to RABBLE and THERMOS codes. (author)
Pandey, Madhu; Singh, Jaideep; Mangal, Garima; Yadav, Pramod
2014-01-01
Objective: This study was carried out to know the level of awareness regarding orthodontic procedures among preadolescents as there is very high prevalence of malocclusion. Methods: This cross-sectional study was conducted among a sample of 1010 subjects with a mean age of (in years) was 13.02 ± 2.146. A self-administered structured questionnaire proforma was used. Pilot study was done to validate the questionnaire, which was constituted of nine items. The Student's t-test and ANOVA test alon...
International Nuclear Information System (INIS)
The one-dimension SN method code ANISN and specific cross section library ZPR-22 have been used to perform the design calculation of dose rate distribution along the radial and axial direction of HWZPR shielding. Through multi-case calculations and optimization analysis works, a double slab cover structure is adopted. It is combined with the feasibility of structure and the possibility of boron concentration to be merged in paraffin for design case. The calculation results of axial direction: the core lattice distance is 18 cm; core radius R = 113 cm; reflector saving of radial direction is 25 cm; transfer leakage Dy = Dz = 244.6 cm. The calculation results of radial direction; the core lattice distance is 18 cm; critical water level 138.5 cm; reflector saving of axial direction is 20 cm; transfer leakage correction parameter Dy = 160 cm
International Nuclear Information System (INIS)
The ENEA-Bologna Nuclear Data Group produced the VITJEF22.BOLIB (NEA-1699/01 ZZ VITJEF22.BOLIB) and MATJEF22.BOLIB (NEA-1740/01 ZZ MATJEF22.BOLIB) fine-group coupled neutron and photon (199 n + 42 γ) cross section libraries for nuclear fission applications, respectively in AMPX and MATXS format and based on the JEF-2.2 European nuclear data file. Both the libraries were produced from the same set of cross section files in GENDF format, generated with the NJOY-94.66 nuclear data processing system. The present libraries can be considered as European counterparts of the VITAMIN-B6 (DLC-0184 ZZ VITAMIN-B6) American library in AMPX format, based on the ENDF/B-VI Release 3 American nuclear data file. In fact they have the same general features and the same neutron and photon energy group structures as VITAMIN-B6. In particular, all these libraries are pseudo-problem-independent and based on the Bondarenko (f-factor) method for the treatment of neutron resonance self-shielding and temperature effects. Each ENEA-Bologna library contains a set of 133 nuclide cross section files processed at 4 temperatures (300 K, 600 K, 1000 K and 2100 K) and obtained for the most part with 6 to 8 values of the background cross section σ0. Thermal scattering cross sections were processed at all the temperatures available in the JEF-2.2 thermal scattering law data file for 5 additional bound nuclides: H-1 in light water, H-1 in polyethylene, H-2 in heavy water, C in graphite and Be in beryllium metal. Collapsed working libraries of self-shielded cross sections in the formats used by the deterministic transport codes of the DANTSYS and DOORS systems can be generated from VITJEF22.BOLIB and MATJEF22.BOLIB through, respectively, further problem-dependent data processing with the AMPX or SCAMPI nuclear data processing systems and with the TRANSX code. (authors)
Energy Technology Data Exchange (ETDEWEB)
Pescarini, M.; Orsi, R.; Martinelli, T.; Sinitsa, V. [ENEA - Centro Ricerche - Ezio Clementel - Bologna (Italy); Blokhin, A.I. [Institute of Physics and Power Engineering (IPPE), Kaluga Region (Russian Federation)
2005-07-01
The ENEA-Bologna Nuclear Data Group produced the VITJEF22.BOLIB (NEA-1699/01 ZZ VITJEF22.BOLIB) and MATJEF22.BOLIB (NEA-1740/01 ZZ MATJEF22.BOLIB) fine-group coupled neutron and photon (199 n + 42 {gamma}) cross section libraries for nuclear fission applications, respectively in AMPX and MATXS format and based on the JEF-2.2 European nuclear data file. Both the libraries were produced from the same set of cross section files in GENDF format, generated with the NJOY-94.66 nuclear data processing system. The present libraries can be considered as European counterparts of the VITAMIN-B6 (DLC-0184 ZZ VITAMIN-B6) American library in AMPX format, based on the ENDF/B-VI Release 3 American nuclear data file. In fact they have the same general features and the same neutron and photon energy group structures as VITAMIN-B6. In particular, all these libraries are pseudo-problem-independent and based on the Bondarenko (f-factor) method for the treatment of neutron resonance self-shielding and temperature effects. Each ENEA-Bologna library contains a set of 133 nuclide cross section files processed at 4 temperatures (300 K, 600 K, 1000 K and 2100 K) and obtained for the most part with 6 to 8 values of the background cross section {sigma}{sub 0}. Thermal scattering cross sections were processed at all the temperatures available in the JEF-2.2 thermal scattering law data file for 5 additional bound nuclides: H-1 in light water, H-1 in polyethylene, H-2 in heavy water, C in graphite and Be in beryllium metal. Collapsed working libraries of self-shielded cross sections in the formats used by the deterministic transport codes of the DANTSYS and DOORS systems can be generated from VITJEF22.BOLIB and MATJEF22.BOLIB through, respectively, further problem-dependent data processing with the AMPX or SCAMPI nuclear data processing systems and with the TRANSX code. (authors)
Pescarini Massimo; Orsi Roberto; Frisoni Manuela
2016-01-01
The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the ORNL TORT-3.2 3D SN code. PCA-Replica, specifically conceived to test the accuracy of nuclear data and transport codes employed in LWR shielding and radiation damage calculations, reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a PWR pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with ...
2009-01-01
Objective: Even with an increasing immigrant population in Norway, there are still a limited number of studies among the group. Chronic musculoskeletal and psychiatric disorders frequently occur and there is a need to establish the magnitude of prevalence and the strength of association between the two chronic disorders in a local context. Methods: Cross-sectional data from the Oslo Immigrant Health Study in 2002 were analyzed. Questionnaires were sent to age cohorts, between 20 and 60 ...
Energy Technology Data Exchange (ETDEWEB)
Goluoglu, S.
2003-12-01
A review of the degree of applicability of benchmarks containing gadolinium using the computer code KENO V.a and the gadolinium cross sections from the 238-group SCALE cross-section library has been performed for a system that contains {sup 239}Pu, H{sub 2}O, and Gd{sub 2}O{sub 3}. The system (practical problem) is a water-reflected spherical mixture that represents a dry-out condition on the bottom of a sludge receipt and adjustment tank around steam coils. Due to variability of the mixture volume and the H/{sup 239}Pu ratio, approximations to the practical problem, referred to as applications, have been made to envelop possible ranges of mixture volumes and H/{sup 239}Pu ratios. A newly developed methodology has been applied to determine the degree of applicability of benchmarks as well as the penalty that should be added to the safety margin due to insufficient benchmarks.
Impact of the ENDF/B-VI Cross Section on the RPV Fluence Determination
International Nuclear Information System (INIS)
The calculations with the broad-group cross-section library Bugle-96, and atom displacement (dpa) cross sections for iron, both derived from ENDF/B-VI data, result in higher calculated fast neutron fluxes, better agreement of calculations with radiometric dosimeter measurements, and significantly slower dpa rate attenuation through pressure vessel walls relative to the results with their predecessors: the Sailor library and ASTM iron dpa cross sections
EJ2-XMAS. Contents of the JEF2.2 based neutron cross-section library in the XMAS group structure
International Nuclear Information System (INIS)
This report describes the contents of the EJ2-XMAS library. The EJ2-XMAS library is a JEF2.2 based 172-group AMPX-Master library in the XMAS group structure for reactor calculations with the SCALE-4 system, as implemented at ECN-Petten. The group cross section data were generated with NJOY89/NSLINK4 and NJOY91/NSLINK4. The data on the EJ2-XMAS library allow resolved-resonance treatment by NITAWL and unresolved resonance self-shielding by BONAMI. These codes are based upon the Nordheim and Bondarenko methods, respectively. (orig.)
Pescarini, Massimo; Sinitsa, Valentin; Orsi, Roberto; Frisoni, Manuela
2016-02-01
Two broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format, dedicated to LWR shielding and pressure vessel dosimetry applications, were generated following the methodology recommended by the US ANSI/ANS-6.1.2-1999 (R2009) standard. These libraries, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, are respectively based on JEFF-3.1.1 and ENDF/B-VII.0 nuclear data and adopt the same broad-group energy structure (47 n + 20 γ) of the ORNL BUGLE-96 similar library. They were respectively obtained from the ENEA-Bologna VITJEFF311.BOLIB and VITENDF70.BOLIB libraries in AMPX format for nuclear fission applications through problem-dependent cross section collapsing with the ENEA-Bologna 2007 revision of the ORNL SCAMPI nuclear data processing system. Both previous libraries are based on the Bondarenko self-shielding factor method and have the same AMPX format and fine-group energy structure (199 n + 42 γ) as the ORNL VITAMIN-B6 similar library from which BUGLE-96 was obtained at ORNL. A synthesis of a preliminary validation of the cited BUGLE-type libraries, performed through 3D fixed source transport calculations with the ORNL TORT-3.2 SN code, is included. The calculations were dedicated to the PCA-Replica 12/13 and VENUS-3 engineering neutron shielding benchmark experiments, specifically conceived to test the accuracy of nuclear data and transport codes in LWR shielding and radiation damage analyses.
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Thermal reactor design calculations are being performed in India using the WIMS/D-4 multi group cross section library, obtained in late 60's, reflecting the status of the basic nuclear data and processing technology then available. Significant improvements in basic evaluated data files such as ENDF/B-IV to VI and JEF data files etc. have been made in the past four decades and the multigroup libraries have been updated world over using improved and comprehensive nuclear data processing code systems. A few of such updated multigroup cross sections in WIMS/D-4 format are available from KAERI and NEA data bank sources. This paper presents the analysis of a set of enriched UO2 and U-metal uniform critical lattice experiments. These include TRX(4), BAPL (3) and B and W (17) lattice, 64 enriched UO2 lattices complied in NEACRP-U-190 report, 56 enriched UO2 lattices and 61 U-metal lattices which were used for validating the WIMKAL-1988 library. Calculated reaction rate values from the participants of WIMS library update project (WLUP) are available for TRX, BAPL lattices. Integral data measured in the lattices of TRX, BAPL, B and W and NEACRP compilations are available in the open literature. Different calculational methods like J± and Pij, and resonance interpolation schemes were examined in the theoretical analysis. Possible shortcomings of the WIMS-D/4 multigroup cross section library currently being used are also identified. (author)
International Nuclear Information System (INIS)
Conventionally the data preparation of the neutron cross sections for reactor-core calculations pursues with 2D cell codes. Aim of this thesis was, to develop a 3D cell code, to study with this code 3D effects, and to evaluate the necessarity of a 3D data preparation of the neutron cross sections. For the calculation of the neutron transport the method of the first-collision probabilities, which are calculated with the ray-tracing method, was chosen. The mathematical algorithms were implemented in the 2D/3D cell code TransRay. For the geometry part of the program the geometry module of a Monte Carlo code was used.The ray tracing in 3D was parallelized because of the high computational time. The program TransRay was verified on 2D test problems. For a reference pressured-water reactor following 3D problems were studied: A partly immersed control rod and void (vacuum or steam) around a fuel rod as model of a steam void. All problems were for comparison calculated also with the programs HELIOS(2D) and MCNP(3D). The dependence of the multiplication factor and the averaged two-group cross section on the immersion depth of the control rod respectively of the height of the steam void were studied. The 3D-calculated two-group cross sections were compared with three conventional approximations: Linear interpolation, interpolation with flux weighting, and homogenization, At the 3D problem of the control rod it was shown that the interpolation with flux weighting is a good approximation. Therefore here a 3D data preparation is not necessary. At the test case of the single control rod, which is surrounded by the void, the three approximation for the two-group cross sections were proved as unsufficient. Therefore a 3D data preparation is necessary. The single fuel-rod cell with void can be considered as the limiting case of a reactor, in which a phase interface has been formed
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A new executable, identified as NJOY99.0 has been created to generate the 69-group cross-section library for the reactor lattice transport code WIMS. The new code incorporates modifications in the WIMSR module of NJOY to generate the 69-group library, which will be used for TRIGA reactor calculations. The basic evaluated nuclear data file JEF-2.2 was used to generate the 69-group cross-section library in WIMS format. The results for TRX-1, TRX-2, BAPL-1, BAPL-2, and BAPL-3 benchmarks obtained by using the generated 69-group cross-section library from JEF-2.2 were analyzed. The following integral parameters were considered for the validation of the 69-group library: finite medium effective multiplication factor (keff), Ratio of epithermal to thermal 238U captures (ρ28), Ratio of epithermal to thermal 235U fission (δ25), Ratio of 238U fission to 235U fission (δ28) and Ratio of 238U captures to 235U fissions (C*). The TRX and BAPL benchmark lattices were modeled with optimized inputs, which were suggested in the final report of the WIMS Library Update Project (WLUP) Stage-I by Ravnik. The calculated results of the integral parameters of TRX and BAPL Benchmark Lattices obtained by using the new version of code WIMSD-5B were found to be in good agreement with the experimental values. Besides, The TRX and BAPL calculation results showed that JEF-2.2 is reliable for thermal reactor calculations and validated the 69-group library, which will be used for the neutronic calculation of the TRIGA Mark-II research reactor at AERE, Savar, Dhaka, Bangladesh. (authors)
MPI version of NJOY and its application to multigroup cross-section generation
International Nuclear Information System (INIS)
Multigroup cross-section libraries are needed in performing neutronics calculations. These libraries are referred to as broad-group libraries. The number of energy groups and group structure are highly dependent on the application and/or user's objectives. For example, for shielding calculations, broad-group libraries such as SAILOR and BUGLE with 47-neutron and 20-gamma energy groups are used. The common procedure to obtain a broad-group library is a three-step process: (1) processing pointwise ENDF (PENDF) format cross sections; (2) generating fine-group cross sections; and (3) collapsing fine-group cross sections to broad-group. The NJOY code is used to prepare fine-group cross sections by processing pointwise ENDF data. The code has several modules, each one performing a specific task. For instance, the module RECONR performs linearization and reconstruction of the cross sections, and the module GROUPR generates multigroup self-shielded cross sections. After fine-group, i.e., groupwise ENDF (GENDF), cross sections are produced, cross sections are self-shielded, and a one-dimensional transport calculation is performed to obtain flux spectra at specific regions in the model. These fluxes are then used as weighting functions to collapse the fine-group cross sections to obtain a broad-group cross-section library. The third step described is commonly performed by the AMPX code system. SMILER converts NJOY GENDF filed to AMPX master libraries, AJAX collects the master libraries. BONAMI performs self-shielding calculations, NITAWL converts the AMPX master library to a working library, XSDRNPM performs one-dimensional transport calculations, and MALOCS collapses fine-group cross sections to broad-group. Finally, ALPO is used to generate ANISN format libraries. In this three-step procedure, generally NJOY requires the largest amount of CPU time. This time varies depending on the user's specified parameters for each module, such as reconstruction tolerances, temperatures
International Nuclear Information System (INIS)
To improve the accuracy of the neutron analyses for subcritical systems with thermal fission blanket, a coupled neutron and photon (315 n + 42γ) fine-group cross section library HENDL3.0/FG based on ENDF/B-Ⅶ. 0 has been produced by FDS team. In order to test the availability and reliability of the HENDL3.0/FG data library, shielding and critical safety benchmarks were performed with VisualBUS code. The testing results indicated that the discrepancy between calculation and experimental values of nuclear parameters fell in a reasonable range. (authors)
G. GiacomelliBologna University and INFN
2014-01-01
The measurements of the hadron-hadron total cross sections are the first measurements performed when a new hadron accelerator opens up a new energy region; the measurements were made as function of the incoming beam momentum or c.m. energy and have often been repeated with improved accuracy and finer energy spacing.
International Nuclear Information System (INIS)
1 - Description: Format: MATXS, 142 nuclides processed with NJOY99.245. Number of groups: 150 neutron-, 12 photon-groups. 142 nuclides: H-1, H-2, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-nat, N-14, N-15, O-16, F-19, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, P-31, Cl-35, Cl-37, Ar-40, K-39, K-40, K-41, Ca-40, Ca-42, Ca-43, Ca-44, Ca-46, Ca-48, Ti-46, Ti-47, Ti-48, Ti-49, Ti-50, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Ga-nat, Y-89, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Nb-93, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Ag-107, Ag-109, Cd-106, Cd-108, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-115m, Cd-116, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-125, Sn-126, Eu-151, Eu-153, Gd-152, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, Pb-206, Pb-207, Pb-208, Bi-209, Th-232, Pa-233, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-242, Am-242m, Am-243, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246. Origin: JEFF-3.1. Weighting spectrum: 300, 600, 900, 1200 K. The KAFAX-F31 is a MATXS-format, 150-group neutron and 12-group photon cross section library for fast reactors based on JEFF-3.1. This library was originally generated for the KALIMER (Korea Advanced LIquid Metal Reactor) core analyses. It includes 142 nuclide data (Table 1) processed by the NJOY99.245 code patched with NEA020. The library can be utilized to generate the problem-dependent group constants for neutron and/or photon transport calculations through the DANTSYS, DOORS, or PARTISN code systems. 2 - Methods: The KAFAX-F31 was generated at 300, 600, 900, and 1200 K. It contains the self-shielded cross sections for 5 to 10 background cross sections depending on the nuclides. The neutron group structure consists of one-eighth lethargy widths in almost
International Nuclear Information System (INIS)
1 - Description: Format: MATXS, 144 nuclides processed with NJOY99.245. Number of groups: 150 neutron-, 12 photon-groups. 144 nuclides: H-1, H-2, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-nat, N-14, N-15, O-16, F-19, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, P-31, Cl-35, Cl-37, Ar-40, K-39, K-40, K-41, Ca-40, Ca-42, Ca-43, Ca-44, Ca-46, Ca-48, Ti-46, Ti-47, Ti-48, Ti-49, Ti-50, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Ga-69, Ga-71, Y-89, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Nb-93, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Ag-107, Ag-109, Cd-106, Cd-108, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-115m, Cd-116, Sn-112, Sn-113, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-125, Sn-126, Eu-151, Eu-153, Gd-152, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, W-182, W-183, W-184, W-186, Re-185, Re-187, Au-197, Pb-206, Pb-207, Pb-208, Bi-209, Th-232, Pa-233, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-242, Am-242m, Am-243, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246. Origin: ENDF/B-VII.0. Weighting spectrum: 300, 600, 900, 1200 k. The ZZ-KAFAX-E70 is a MATXS-format, 150-group neutron and 12-group photon cross section library for fast reactors based on ENDF/B-VII.0. This library was originally generated for the KALIMER (Korea Advanced Liquid Metal Reactor) core analyses. It includes 144 nuclide data processed with the NJOY99.245 code patched with NEA020. The library can be used to generate the problem-dependent group constants for neutron and/or photon transport calculations through the DANTSYS, DOORS, or PARTISN code systems. 2 - Methods: The KAFAX-E70 was generated at 300, 600, 900, and 1200 K. It contains the self-shielded cross sections for 5 to 10 background cross sections depending on the nuclides. The neutron group structure consists of one-eighth lethargy
International Nuclear Information System (INIS)
1 - Description: Format: MATXS, 136 nuclides processed with NJOY99.245. Number of groups: 150 neutron-, 12 photon-groups. 136 Nuclides: H-1, H-2, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-nat, N-14, N-15, O-16, F-19, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, P-31, Cl-35, Cl-37, Ar-40, K-39, K-40, K-41, Ca-40, Ca-42, Ca-43, Ca-44, Ca-46, Ca-48, Ti-46, Ti-47, Ti-48, Ti-49, Ti-50, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Ga-69, Ga-71, Y-89, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Ag-107, Ag-109, Cd-106, Cd-108, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-116, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-126, Eu-151, Eu-153, Gd-152, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, W-182, W-183, W-184, W-186, Pb-206, Pb-208, Bi-209, Th-232, Pa-233, U-233, U-234, U-235, U-236, U-238, Np-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Am-242, Am-242m, Am-243, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246. Origin: JENDL-3.3. Weighting spectrum: 300, 600, 900, 1200 K. The KAFAX-J33 is a MATXS-format, 150-group neutron and 12-group photon cross section library for fast reactors based on JENDL-3.3. This library was originally generated for the KALIMER (Korea Advanced LIquid Metal Reactor) core analyses. It includes 136 nuclide data processed by the NJOY99.245 code patched with NEA020. The library can be utilized to generate the problem-dependent group constants for neutron and/or photon transport calculations through the DANTSYS, DOORS, or PARTISN code systems. 2 - Methods: The KAFAX-J33 was generated at 300, 600, 900, and 1200 K. It contains the self-shielded cross sections for 5 to 10 background cross sections depending on the nuclides. The neutron group structure consists of one-eighth lethargy widths in almost all the energy ranges, except between 1 and 10 keV in
Directory of Open Access Journals (Sweden)
Devillé Walter L
2009-04-01
Full Text Available Abstract Background To determine gender differences in health and health care utilisation within and between various ethnic groups in the Netherlands. Methods Data from the second Dutch National Survey of General Practice (2000–2002 were used. A total of 7,789 persons from the indigenous population and 1,512 persons from the four largest migrant groups in the Netherlands – Morocco, Netherlands Antilles, Turkey and Surinam – aged 18 years and older were interviewed. Self-reported health outcomes studied were general health status and the presence of acute (past 14 days and chronic conditions (past 12 months. And self-reported utilisation of the following health care services was analysed: having contacted a general practitioner (past 2 months, a medical specialist, physiotherapist or ambulatory mental health service (past 12 months, hospitalisation (past 12 months and use of medication (past 14 days. Gender differences in these outcomes were examined within and between the ethnic groups, using logistic regression analyses. Results In general, women showed poorer health than men; the largest differences were found for the Turkish respondents, followed by Moroccans, and Surinamese. Furthermore, women from Morocco and the Netherlands Antilles more often contacted a general practitioner than men from these countries. Women from Turkey were more hospitalised than Turkish men. Women from Morocco more often contacted ambulatory mental health care than men from this country, and women with an indigenous background more often used over the counter medication than men with an indigenous background. Conclusion In general the self-reported health of women is worse compared to that of men, although the size of the gender differences may vary according to the particular health outcome and among the ethnic groups. This information might be helpful to develop policy to improve the health status of specific groups according to gender and ethnicity. In
Respirator fit of a medium mask on a group of South Africans: a cross-sectional study
Directory of Open Access Journals (Sweden)
Wilson Kerry S
2011-03-01
Full Text Available Abstract Background In South Africa, respiratory protective equipment is often the primary control method used to protect workers. This preliminary study investigated how well a common disposable P2 respirator fitted persons with a range of facial dimensions. Methods Quantitative respirator fit tests were performed on 29 volunteers from different racial, gender and face size groups. Two facial dimensions width (bizygomatic and length (menton-sellion were measured for all participants. Results In this study 13.8% of the participants demonstrated a successful fit with the medium sized mask. These included participants from three different racial and both gender groups. The large percentage of failed fit tests (86% indicates that reliance on off-the-shelf respirators could be problematic in South Africa. Conclusions The limitations of this preliminary study notwithstanding, respirator fit appear to be associated with individual facial characteristics and are not specific to racial/ethnic or gender characteristics.
Respirator fit of a medium mask on a group of South Africans: a cross-sectional study
Wilson Kerry S; Spies Adri; Ferrie Robert
2011-01-01
Abstract Background In South Africa, respiratory protective equipment is often the primary control method used to protect workers. This preliminary study investigated how well a common disposable P2 respirator fitted persons with a range of facial dimensions. Methods Quantitative respirator fit tests were performed on 29 volunteers from different racial, gender and face size groups. Two facial dimensions width (bizygomatic) and length (menton-sellion) were measured for all participants. Resul...
Background-cross-section-dependent subgroup parameters
International Nuclear Information System (INIS)
A new set of subgroup parameters was derived that can reproduce the self-shielded cross section against a wide range of background cross sections. The subgroup parameters are expressed with a rational equation which numerator and denominator are expressed as the expansion series of background cross section, so that the background cross section dependence is exactly taken into account in the parameters. The advantage of the new subgroup parameters is that they can reproduce the self-shielded effect not only by group basis but also by subgroup basis. Then an adaptive method is also proposed which uses fitting procedure to evaluate the background-cross-section-dependence of the parameters. One of the simple fitting formula was able to reproduce the self-shielded subgroup cross section by less than 1% error from the precise evaluation. (author)
Pescarini, Massimo; Orsi, Roberto; Frisoni, Manuela
2016-02-01
The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the ORNL TORT-3.2 3D SN code. PCA-Replica, specifically conceived to test the accuracy of nuclear data and transport codes employed in LWR shielding and radiation damage calculations, reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a PWR pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with the same energy group structure (47 n + 20 γ) and based on different nuclear data were alternatively used: the ENEA BUGJEFF311.BOLIB (JEFF-3.1.1) and BUGENDF70.BOLIB (ENDF/B-VII.0) libraries and the ORNL BUGLE-96 (ENDF/B-VI.3) library. Dosimeter cross sections derived from the IAEA IRDF-2002 dosimetry file were employed. The calculated reaction rates for the Rh-103(n,n')Rh-103 m, In-115(n,n')In-115m and S-32(n,p)P-32 threshold activation dosimeters and the calculated neutron spectra are compared with the corresponding experimental results.
Directory of Open Access Journals (Sweden)
Pescarini Massimo
2016-01-01
Full Text Available The PCA-Replica 12/13 (H2O/Fe neutron shielding benchmark experiment was analysed using the ORNL TORT-3.2 3D SN code. PCA-Replica, specifically conceived to test the accuracy of nuclear data and transport codes employed in LWR shielding and radiation damage calculations, reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a PWR pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with the same energy group structure (47 n + 20 γ and based on different nuclear data were alternatively used: the ENEA BUGJEFF311.BOLIB (JEFF-3.1.1 and BUGENDF70.BOLIB (ENDF/B-VII.0 libraries and the ORNL BUGLE-96 (ENDF/B-VI.3 library. Dosimeter cross sections derived from the IAEA IRDF-2002 dosimetry file were employed. The calculated reaction rates for the Rh-103(n,n′Rh-103 m, In-115(n,n′In-115m and S-32(n,pP-32 threshold activation dosimeters and the calculated neutron spectra are compared with the corresponding experimental results.
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S Noorbakhsh
2011-06-01
Full Text Available Background and Objective: Group A beta-hemolytic streptococcus (GABHS is an important pharyngotonsillitis etiologic agent in children. The objective of this study was diagnosis of streptococcal pharyngitis based on rapid antigen detection test and conventional pharyngeal culture.Materials and Methods: The rapid GABHS antigen detection test was compared to culture on blood agar, the gold standard for the diagnosis of this etiologic agent.Results: Streptococcal antigen was detected in pharyngeal specimens of 34.5% of cases by rapid strip test. We detected group A Streptococcus in 17.2% of pharyngeal culture. There was no agreement between two methods ( PV < 0.1. The negative pharyngeal culture results are probably due to antibiotic usage in 43.2 % of patients. Positive rapid test results in pharyngeal swab was age dependent ( P < 0.05. There was good correlation between observing the "petechia in pharynx of patients" and positive rapid test in pharyngeal swab (P < 0.004. Throat culture results were relatated to previous antibiotic usage ( P < 0.03.Conclusion: The rapid test in pharyngeal swab is helpful for rapid diagnosis and treatment of GABHS pharyngitis. Diagnosis of GABHS pharyngitis based on soley clinical findings is misleading in the majority of cases. Petechia observed in pharynx of the cases was highly predictive of streptococcal pharyngitis.
New activation cross section data
International Nuclear Information System (INIS)
New nuclear cross section libraries (known as USACT92) have been created for activation calculations. A point-wise file was created from merging the previous version of the activation library, the U.S. Nuclear Data Library (ENDF/B-VI), and the European Activation File (EAF-2). 175 and 99 multi-group versions were also created. All the data are available at the National Energy Research Supercomputer Center
Pellerone, Monica; Tolini, Giacomo; Polopoli, Caterina
2016-01-01
Background Literature has demonstrated the adaptive function of identity development and parenting toward manifestation of problem behaviors in adolescence. These dimensions act on both internalizing and externalizing symptoms. Methods The objective is to investigate the relationship between identity status, parenting, and adolescent problems, which may manifest through internalized (phobias, obsessions, depression, eating disorders, entropy) and externalized modes (alcohol use and school discomfort). The research involved 198 Italian students (104 males and 94 females) in the 4th year (mean =16.94 years, standard deviation =0.35) and 5th year (mean =17.94 years, standard deviation =0.43) of senior secondary schools, who live in Caltanissetta, a town located in Sicily, Italy. The research lasted for 1 school year. The general group consisted of 225 students with a mortality rate of 12%. They completed an anamnestic questionnaire to provide 1) basic information, 2) alcohol consumption attitude in the past 30 days, and 3) their beliefs about alcohol; the “Ego Identity Process Questionnaire” to investigate identity development; the “Parental Bonding Instrument” to measure the perception of parenting during childhood; and the “Constraints of Mind” to value the presence of internalizing symptoms. Results Data show that identity status influences alcohol consumption. Low-profile identity and excessive maternal control affect the relational dependence and the tendency to perfectionism in adolescents. Among the predictors of alcohol use, there are socioeconomic status, parental control, and the presence of internalizing symptoms. Conclusion Family is the favored context of learning beliefs, patterns, and values that affect the broader regulatory social environment, and for this reason, it is considered the privileged context on which to intervene to reduce the adolescents’ behavior problems. This deviance could be an external manifestation of the difficulty
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Tiwari Ranjana
2009-01-01
Full Text Available Background: Obesity has become a major chronic disorder affecting the larger population more than any other disease in the world. Objectives: 1 To determine the prevalence of obesity in both sexes in persons aged 30 years and above. 2 To determine the relationship of epidemiological determinants on the obesity status in the study subjects. Materials and Methods: The present study had been undertaken in literate high income group colonies of Gwalior city in which persons aged 30 years and above, in a family, were interviewed. A house-to-house survey method on a pre-designed, pre- tested structured questionnaire was used. Information regarding socio-demographic profile, eating habits and current health status were recorded. Anthropometric data regarding height, weight and blood pressure was also taken. The data was collected and analyzed using statistical software and chi square and proportional statistical test were applied. Results: The study showed that 34.4% of males and 31.3 % of females, both aged 30years and above were either obese or over weight. There was a statistically significant difference noted in the likening of fried food and fast food between obese and overweight persons and persons with normal body mass index. Conclusions: It can be concluded from the present study that obesity is a chronic illness. Early detection of it can prevent various complications associated with it. BMI plays a crucial role in its early detection as it is simple to calculate and can even detect the pre-obesity stage in time.
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Pellerone M
2016-07-01
Full Text Available Monica Pellerone, Giacomo Tolini, Caterina Polopoli Faculty of Human and Social Sciences, “Kore” University of Enna, Enna, Italy Background: Literature has demonstrated the adaptive function of identity development and parenting toward manifestation of problem behaviors in adolescence. These dimensions act on both internalizing and externalizing symptoms.Methods: The objective is to investigate the relationship between identity status, parenting, and adolescent problems, which may manifest through internalized (phobias, obsessions, depression, eating disorders, entropy and externalized modes (alcohol use and school discomfort. The research involved 198 Italian students (104 males and 94 females in the 4th year (mean =16.94 years, standard deviation =0.35 and 5th year (mean =17.94 years, standard deviation =0.43 of senior secondary schools, who live in Caltanissetta, a town located in Sicily, Italy. The research lasted for 1 school year. The general group consisted of 225 students with a mortality rate of 12%. They completed an anamnestic questionnaire to provide 1 basic information, 2 alcohol consumption attitude in the past 30 days, and 3 their beliefs about alcohol; the “Ego Identity Process Questionnaire” to investigate identity development; the “Parental Bonding Instrument” to measure the perception of parenting during childhood; and the “Constraints of Mind” to value the presence of internalizing symptoms.Results: Data show that identity status influences alcohol consumption. Low-profile identity and excessive maternal control affect the relational dependence and the tendency to perfectionism in adolescents. Among the predictors of alcohol use, there are socioeconomic status, parental control, and the presence of internalizing symptoms.Conclusion: Family is the favored context of learning beliefs, patterns, and values that affect the broader regulatory social environment, and for this reason, it is considered the privileged
Directory of Open Access Journals (Sweden)
Mohammadi G
2011-11-01
well as emotionally and sexually. The violence was reported to be exerted by husband (42.6%, parents (38.4%, or both (19.0%. Among 39 participants who ran away from home, 38 participants reported to be inflicted by violence. Unwanted pregnancy was reported by 64.6% of the participants. Abortion was reported in 50.0% of participants. Contraception was completely ignored in 44.6% of participants. Among eligible women, 53.3% never participated in cervical cancer screening examination. Mean sexual performance scale score was 21.9 (5.5 and 75 (83.3% participants scored less than 28.Conclusion: A high prevalence of poor reproductive health was documented among a group of Middle Eastern socially damaged women.Keywords: sexual behavior, domestic violence, pregnancy, drop-in center, abortion, contraception, cervical cancer screening
International Nuclear Information System (INIS)
1 - Description: Format: MATXS, 204 nuclides processed with NJOY99.245. Number of groups: 199 neutron-, 42 photon-groups. 204 Nuclides including 8 thermal scattering law data: H-1, H-2, H-3, He-3, He-4, Li-6, Li-7, Be-9, Be-9, Be-9, B-10, B-11, C-nat, C-nat, N-14, N-15, O-16, O-17, F-19, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, P-31, S-32, S-33, S-34, S-36, Cl-35, Cl-37, K-39, K-40, K-41, Ca-40, Ca-42, Ca-43, Ca-44, Ca-46, Ca-48, Sc-45, Ti-46, Ti-47, Ti-48, Ti-49, Ti-50, V-nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Ga-69, Ga-71, Y-89, Zr-90, Zr-91, Zr-92, Zr-94, Zr-96, Nb-93, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-100, Pd-102, Pd-104, Pd-105, Pd-106, Pd-108, Pd-110, Ag-107, Ag-109, Cd-106, Cd-108, Cd-110, Cd-112, Cd-113, Cd-114, Cd-116, In-113, In-115, I-127, Xe-124, Xe-126, Xe-128, Xe-129, Xe-130, Xe-131, Xe-132, Xe-134, Xe-136, Cs-133, Ba-138, Pr-141, Nd-143, Nd-145, Nd-146, Nd-148, Nd-150, Pm-147, Sm-147, Sm-151, Sm-152, Eu-151, Eu-152, Eu-153, Eu-154, Eu-155, Gd-152, Gd-154, Gd-155, Gd-156, Gd-157, Gd-158, Gd-160, Dy-164, Ho-165, Lu-175, Lu-176, Hf-174, Hf-176, Hf-177, Hf-178, Hf-179, Hf-180, Ta-181, Ta-182, W-182, W-183, W-184, W-186, Re-185, Re-187, Ir-191, Ir-193, Au-197, Pb-206, Pb-207, Pb-208, Bi-209, Th-230, Th-232, Pa-231, Pa-233, U-232, U-233, U-234, U-235, U-236, U-237, U-238, Np-237, Np-238, Np-239, Pu-236, Pu-237, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Pu-243, Pu-244, Am-241, Am-242, Am-242m, Am-243, Cm-241, Cm-242, Cm-243, Cm-244, Cm-245, Cm-246, Cm-247, Cm-248, Bk-249, Cf-249, Cf-250, Cf-251, Cf-252, Cf-253, Es-253. Origin: ENDF/B-VII.0. Weighting spectrum: 300, 600, 1000, 2100 K. The KASHIL-E70 is a MATXS-format, 199-group neutron and 42-group photon cross section library for shielding applications based on ENDF/B-VII.0. The library contains 204 nuclide data including 8 thermal scattering law data processed by the NJOY99.259 code patched with NEA
Gollapinni, Sowjanya
2016-01-01
The study of neutrino-nucleus interactions has recently received renewed attention due to their importance in interpreting the neutrino oscillation data. Over the past few years, there has been continuous disagreement between neutrino cross section data and predictions due to lack of accurate nuclear models suitable for modern experiments which use heavier nuclear targets. Also, the current short and long-baseline neutrino oscillation experiments focus in the few GeV region where several distinct neutrino processes come into play resulting in complex nuclear effects. Despite recent efforts, more experimental input is needed to improve nuclear models and reduce neutrino interaction systematics which are currently dominating oscillation searches together with neutrino flux uncertainties. A number of new detector concepts with diverse neutrino beams and nuclear targets are currently being developed to provide necessary inputs required for next generation oscillation experiments. This paper summarizes these effor...
Methods for calculating anisotropic transfer cross sections
International Nuclear Information System (INIS)
The Legendre moments of the group transfer cross section, which are widely used in the numerical solution of the transport calculation can be efficiently and accurately constructed from low-order (K = 1--2) successive partial range moments. This is convenient for the generation of group constants. In addition, a technique to obtain group-angle correlation transfer cross section without Legendre expansion is presented. (author)
International Nuclear Information System (INIS)
1 - Description: Format: MATXS. Number of groups: 80 neutron-, 24 photon-groups. 97 Nuclides: 1-H-1, 1-H-2, 2-He-3, 2-He-4, 3-Li-6, 3-Li-7, 4-Be-9, 5-B-10, 5-B-11, 6-C- nat., 7-N-14, 7-N-15, 8-O-16, 9-F-19, 11-Na-23, 12-Mg-nat., 13-Al-27, 14-Si-nat., 15-P-31, 17-Cl-nat., 18-Ar-40, 19-K-nat., 20-Ca-nat., 22-Ti-nat., 23-V-nat., 24-Cr-50, 24-Cr-52, 24-Cr-53, 24-Cr-54, 25-Mn-25, 26-Fe-54, 26-Fe-56, 26-Fe-57, 26-Fe-58, 27-Co-59, 28-Ni-58, 28-Ni-60, 28-Ni-61, 28-Ni-62, 28-Ni-64, 29-Cu-nat., 31-Ga-nat., 39-Y-89, 40-Zr-nat., 41-Nb-93, 42-Mo-nat., 47-Ag-107, 47-Ag-109, 48-Cd-nat., 50-Sn-nat., 63-Eu-151, 63-Eu-153, 64-Gd-152, 64-Gd-154, 64-Gd-155, 64-Gd-156, 64-Gd-157, 64-Gd-158, 64-Gd-160, 73-Ta-181, 74-W-182, 74-W-183, 74-W-184, 74-W-186, 75-Re-185, 75-Re-187, 79-Au-197, 82-Pb-nat., 83-Bi-209, 90-Th-232, 91-Pa-233, 92-U-232, 92-U-233, 92-U-234, 92-U-235, 92-U-236, 92-U-237, 92-U-238, 93-Np-237, 93-Np-238, 94-Pu-238, 94-Pu-239, 94-Pu-240, 94-Pu-241, 94-Pu-242, 95-Am-241, 95-Am-242, 95-Am-242m, 95-Am-243, 96-Cm-242, 96-Cm-243, 96-Cm-244, 96-Cm-245, 96-Cm-246, 96-Cm-247, 96-Cm-248, 98-Cf-252 Origin: JEF-2.2; Weighting spectrum: Thermal + 1/E + fast reactor + fusion. The library is focused on the fast reactor analyses. It has 80 and 24 energy group structures for neutron and photon, respectively. It includes 97 nuclide data based on JEF-2.2 and has a Format of MATXS processed by the NJOY94 code. It can be used to calculate the problem dependant group constants with the TRANSX code for neutron and gamma transport. 2 - Methods: The data were generated at 300 ∼ 2500 Kelvin degrees and at 4∼7 background cross sections for the self shielding considerations. The weighting function used for group averaged neutron cross sections from the pointwise data is 'thermal + 1/E + fast reactor + fusion'. The library has been validated through the CSEWG benchmark analyses such as VERA-11A, ZPR-3-12, SNEAK-7B, ZPPR-2, ZPR-6-7, etc. 3 - Related or auxiliary programs: - BBC: Program to convert
Diffractive and rising cross sections
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The energy dependence of the diffractive component of the proton-proton cross section is discussed and its contribution to the rise of the total cross section at high energies is examined. 17 refs., 9 figs
[Fast neutron cross section measurements
International Nuclear Information System (INIS)
This paper discusses the following topics: 14 MeV pulsed neutron facility; detection and measurement system; 238U capture cross sections at 23 and 964 keV using photon neutron sources; capture cross sections of Au-197 at 23 and 964 keV; and yttrium nuclear cross section measurement
Estimation of multi-group cross section covariances for 235,238U, 239Pu, 241Am, 56Fe, 23Na and 27Al
International Nuclear Information System (INIS)
This paper presents the methodology used to estimate multi-group covariances for some major isotopes used in reactor physics. The starting point of this evaluation is the modelling of the neutron induced reactions based on nuclear reaction models with parameters. These latest are the vectors of uncertainties as they are absorbing uncertainties and correlation arising from the confrontation of nuclear reaction model to microscopic experiment. These uncertainties are then propagated towards multi-group cross sections. As major breakthroughs were then asked by nuclear reactor physicists to assess proper uncertainties to be used in applications, a solution is proposed by the use of integral experiment information at two different stages in the covariance estimation. In this paper, we will explain briefly the treatment of all type of uncertainties, including experimental ones (statistical and systematic) as well as those coming from validation of nuclear data on dedicated integral experiment (nuclear data oriented). We will illustrate the use of this methodology with various isotopes such as 235,238U, 239Pu, 241Am, 56Fe, 23Na and 27Al. (authors)
Spectral history correction of microscopic cross sections for the PBR using the slowing down balance
International Nuclear Information System (INIS)
A method has been formulated to account for depletion effects on microscopic cross sections within a Pebble Bed Reactor (PBR) spectral zone without resorting to calls to the spectrum (cross section generation) code or relying upon table interpolation between data at different values of burnup. In this method, infinite medium microscopic cross sections, fine group fission spectra, and modulation factors are pre-computed at selected isotopic states. This fine group information is used with the local spectral zone nuclide densities to generate new cross sections for each spectral zone. The local spectrum used to generate these microscopic cross sections is estimated through the solution to the cell-homogenized, infinite medium slowing down balance equation during the flux calculation. This technique is known as Spectral History Correction (SHC), and it is formulated to specifically account for burnup within a spectral zone. It was found that the SHC technique accurately calculates local broad group microscopic cross sections with local burnup information. Good agreement is obtained with cross sections generated directly by the cross section generator. Encouraging results include improvement in the converged fuel cycle eigenvalue, the power peaking factor, and the flux. It was also found that the method compared favorably to the benchmark problem in terms of the computational speed. (authors)
A nuclear cross section data handbook
Energy Technology Data Exchange (ETDEWEB)
Fisher, H.O.M.
1989-12-01
Isotopic information, reaction data, data availability, heating numbers, and evaluation information are given for 129 neutron cross-section evaluations, which are the source of the default cross sections for the Monte Carlo code MCNP. Additionally, pie diagrams for each nuclide displaying the percent contribution of a given reaction to the total cross section are given at 14 MeV, 1 MeV, and thermal energy. Other information about the evaluations and their availability in continuous-energy, discrete-reaction, and multigroup forms is provided. The evaluations come from ENDF/B-V, ENDL85, and the Los Alamos Applied Nuclear Science Group T-2. Graphs of all neutron and photon production cross-section reactions for these nuclides have been categorized and plotted. 21 refs., 5 tabs.
Directory of Open Access Journals (Sweden)
Yang Fang
2009-08-01
Full Text Available Abstract Background 1 To report site-specific normative values by age, sex and educational level for four components of the 10/66 Dementia Research Group cognitive test battery; 2 to estimate the main and interactive effects of age, sex, and educational level by site; and 3 to investigate the effect of site by region and by rural or urban location. Methods Population-based cross-sectional one phase catchment area surveys were conducted in Cuba, Dominican Republic, Venezuela, Peru, Mexico, China and India. The protocol included the administration of the Community Screening Instrument for Dementia (CSI 'D', generating the COGSCORE measure of global function, and the Consortium to Establish a Registry for Alzheimer's Disease (CERAD verbal fluency (VF, word list memory (WLM, immediate recall and recall (WLR, delayed recall tests. Only those free of dementia were included in the analysis. Results Older people, and those with less education performed worse on all four tests. The effect of sex was much smaller and less consistent. There was a considerable effect of site after accounting for compositional differences in age, education and sex. Much of this was accounted for by the effect of region with Chinese participants performing better, and Indian participants worse, than those from Latin America. The effect of region was more prominent for VF and WLM than for COGSCORE and WLR. Conclusion Cognitive assessment is a basic element for dementia diagnosis. Age- and education-specific norms are required for this purpose, while the effect of gender can probably be ignored. The basis of cultural effects is poorly understood, but our findings serve to emphasise that normative data may not be safely generalised from one population to another with quite different characteristics. The minimal effects of region on COGSCORE and WLR are reassuring with respect to the cross-cultural validity of the 10/66 dementia diagnosis, which uses only these elements of the 10
Beauchamp, Alison; Buchbinder, Rachelle; Dodson, Sarity; Batterham, Roy W.; Elsworth, Gerald R.; McPhee, Crystal; Sparkes, Louise; Hawkins, Melanie; Richard H. Osborne
2015-01-01
Background Recent advances in the measurement of health literacy allow description of a broad range of personal and social dimensions of the concept. Identifying differences in patterns of health literacy between population sub-groups will increase understanding of how health literacy contributes to health inequities and inform intervention development. The aim of this study was to use a multi-dimensional measurement tool to describe the health literacy of adults in urban and rural Victoria, ...
Correction of multigroup cross sections for resolved resonance interference in mixed absorbers
International Nuclear Information System (INIS)
The effect that interference between resolved resonances has on averaging multigroup cross sections is examined for thermal reactor-type problems. A simple and efficient numerical scheme is presented to correct a preprocessed multigroup library for interference effects. The procedure is implemented in a design oriented lattice physics computer code and compared with rigorous numerical calculations. The approximate method for computing resonance interference correction factors is applied to obtaining fine-group cross sections for a homogeneous uranium-plutonium mixture and a uranium oxide lattice. It was found that some fine group cross sections are changed by more than 40% due to resonance interference. The change in resonance interference correction factors due to burnup of a PWR fuel pin is examined and found to be small. The effect of resolved resonance interference on collapsed broad-group cross sections for thermal reactor calculations is discussed
Prospects for Precision Neutrino Cross Section Measurements
Energy Technology Data Exchange (ETDEWEB)
Harris, Deborah A. [Fermilab
2016-01-28
The need for precision cross section measurements is more urgent now than ever before, given the central role neutrino oscillation measurements play in the field of particle physics. The definition of precision is something worth considering, however. In order to build the best model for an oscillation experiment, cross section measurements should span a broad range of energies, neutrino interaction channels, and target nuclei. Precision might better be defined not in the final uncertainty associated with any one measurement but rather with the breadth of measurements that are available to constrain models. Current experience shows that models are better constrained by 10 measurements across different processes and energies with 10% uncertainties than by one measurement of one process on one nucleus with a 1% uncertainty. This article describes the current status of and future prospects for the field of precision cross section measurements considering the metric of how many processes, energies, and nuclei have been studied.
International Nuclear Information System (INIS)
Most of the fission products and a few of the actinides in ENDF/B-V do not have (n,2n) cross sections. A complete set of these cross sections is presented in the multigroup structure defined. These were constructed for future use in the DANDE Code System
XCOM: Photon Cross Sections Database
SRD 8 XCOM: Photon Cross Sections Database (Web, free access) A web database is provided which can be used to calculate photon cross sections for scattering, photoelectric absorption and pair production, as well as total attenuation coefficients, for any element, compound or mixture (Z <= 100) at energies from 1 keV to 100 GeV.
Cross Sections and Lorentz Violation
Colladay, Don; Kostelecky, Alan
2001-01-01
The derivation of cross sections and decay rates in the Lorentz-violating standard-model extension is discussed. General features of the physics are described, and some conceptual and calculational issues are addressed. As an illustrative example, the cross section for the specific process of electron-positron pair annihilation into two photons is obtained.
Directory of Open Access Journals (Sweden)
Piet Cools
Full Text Available One million neonates die each year in low- and middle-income countries because of neonatal sepsis; group B Streptococcus (GBS and Escherichia coli are the leading causes. In sub-Saharan Africa, epidemiological data on vaginal GBS and E. coli carriage, a prerequisite for GBS and E. coli neonatal sepsis, respectively, are scarce but necessary to design and implement prevention strategies. Therefore, we assessed vaginal GBS and E. coli carriage rates and risk factors and the GBS serotype distribution in three sub-Saharan countries.A total of 430 women from Kenya, Rwanda and South Africa were studied cross-sectionally. Vaginal carriage of GBS and E. coli, and GBS serotype were assessed using molecular techniques. Risk factors for carriage were identified using multivariable logistic regression analysis.Vaginal carriage rates in reference groups from Kenya and South Africa were 20.2% (95% CI, 13.7-28.7% and 23.1% (95% CI, 16.2-31.9%, respectively for GBS; and 25.0% (95% CI, 17.8-33.9% and 27.1% (95% CI, 19.6-36.2%, respectively for E. coli. GBS serotypes Ia (36.8%, V (26.3% and III (14.0% were most prevalent. Factors independently associated with GBS and E. coli carriage were Candida albicans, an intermediate vaginal microbiome, bacterial vaginosis, recent vaginal intercourse, vaginal washing, cervical ectopy and working as a sex worker. GBS and E. coli carriage were positively associated.Reduced vaginal GBS carriage rates might be accomplished by advocating behavioral changes such as abstinence from sexual intercourse and by avoidance of vaginal washing during late pregnancy. It might be advisable to explore the inclusion of vaginal carriage of C. albicans, GBS, E. coli and of the presence of cervical ectopy in a risk- and/or screening-based administration of antibiotic prophylaxis. Current phase II GBS vaccines (a trivalent vaccine targeting serotypes Ia, Ib, and III, and a conjugate vaccine targeting serotype III would not protect the majority of
International Nuclear Information System (INIS)
JENDL-3.2 and ENDL-84 data; Weighting spectrum: Maxwellian + 1/E + fission spectrum. The library was generated for the analyses of neutron and gamma shielding. The library includes 176 Nuclides based on ENDF/B-VI.5 and has a Format of MATXS processed by the NJOY97 code. It can be used to calculate the problem dependant group constants with the TRANSX code for neutron and gamma transport. 2 - Methods: The data were generated at 300 ∼ 2100 Kelvin degrees and at 6∼8 background cross sections for the self shielding considerations. The infinite diluted data of H-1, H-2, C and Li are included at the Temperatures of the thermal scattering law data. The scattering matrices were expanded to P5∼P7 of Legendre polynomial. The weighting function used for group averaged neutron cross sections from the pointwise data is 'Maxwellian + 1/E + fission spectrum'. The library has been validated through the shielding benchmarks such as PCA-REPLICA, NESDIP2, Winfrith Iron, Winfrith Iron88, Winfrith Graphite experiments, etc
Ferrand, Jean-François; Verret, Catherine; Trichereau, Julie; Rondier, Jean-Philippe; Viance, Patrice; Migliani, René
2012-01-01
Objectives To investigate the associations between psychosocial risk factors and self-reported health, taking into account other occupational risk factors. Design Cross-sectional survey using a self-administered questionnaire. Setting The three military hospitals in Paris, France. Participants Surveys were distributed to 3173 employees (1807 military and 1336 civilian), a total of 1728 employees completed surveys. Missing data prohibited the use of 26 surveys. Primary and secondary outcome me...
SCAMPI: A code package for cross-section processing
Energy Technology Data Exchange (ETDEWEB)
Parks, C.V.; Petrie, L.M.; Bowman, S.M.; Broadhead, B.L.; Greene, N.M.; White, J.E.
1996-04-01
The SCAMPI code package consists of a set of SCALE and AMPX modules that have been assembled to facilitate user needs for preparation of problem-specific, multigroup cross-section libraries. The function of each module contained in the SCANTI code package is discussed, along with illustrations of their use in practical analyses. Ideas are presented for future work that can enable one-step processing from a fine-group, problem-independent library to a broad-group, problem-specific library ready for a shielding analysis.
Measurement of fission cross sections
International Nuclear Information System (INIS)
A review is presented on the recent progress in the experiment of fission cross section measurement, including recent activity in Japan being carried out under the project of nuclear data measurement. (author)
R. Vogt
2007-01-01
We assess the theoretical uncertainties on the total charm cross section. We discuss the importance of the quark mass, the scale choice and the parton densities on the estimate of the uncertainty. We conclude that due to the small charm quark mass, which amplifies the effect of the other parameters in the calculation, the uncertainty on the total charm cross section is difficult to quantify.
International Nuclear Information System (INIS)
A project to prepare an exhaustive handbook of WIMS-D cross section libraries for thermal reactor applications comparing different WIMS-D compatible nuclear data libraries originating from various countries has been successfully designed. To meet the objectives of this project, a computer software package with graphical user interface for MS Windows has been developed at BARC, India. This article summarizes the salient features of this new software and presents significant improvements and extensions in relation to its first version [Ann Nucl Energ 29 (2002) 1735
Transport model based on three-dimensional cross-section generation for TRIGA core analysis
Kriangchaiporn, Nateekool
This dissertation addresses the development of a reactor core physics model based on 3-D transport methodology utilizing 3-D multigroup fuel lattice cross-section generation and core calculation for PSBR. The proposed 3-D transport calculation scheme for reactor core simulations is based on the TORT code. The methodology includes development of algorithms for 2-D and 3-D cross-section generation. The fine- and broad-group structures for the TRIGA cross-section generation problems were developed based on the CPXSD (Contributon and Point-wise Cross-Section Driven) methodology that selects effective group structure. Along with the study of cross section generation, the parametric studies for SN calculations were performed to evaluate the impact of the spatial meshing, angular, and scattering order variables and to obtain the suitable values for cross-section collapsing of the TRIGA cell problem. The TRIGA core loading 2 is used to verify and validate the selected effective group structures. Finally, the 13 group structure was selected to use for core calculations. The results agree with continuous energy for eigenvalues and normalized pin power distribution. The Monte Carlo solutions are used as the references.
International Nuclear Information System (INIS)
The Generation IV [1] International forum identified six advanced reactor concepts and related fuel cycles along with the R and D programs necessary to achieve the four key goals: (1) sustainability, (2) safety and reliability, (3) economics, (4) proliferation resistance and physical protection. Among these six promising reactor concepts, the lead-cooled fast reactor (LFR) has been selected for development by EURATOM, which in 2006 decided to finance the European Lead Cooled System (ELSY) project. The aim of the project is to demonstrate the possibility to design a safe and competitive lead-cooled fast power reactor using simple engineering solutions. This paper demonstrates the use of the code package SCALE5.1 and its NEWT/TRITON modules [3] for preliminary neutronic core analysis of a LFR within Generation IV Nuclear Energy systems program. More specifically, the analysis of the reference design of the ELSY-600 open square fuel assembly is presented. In particular, the use of ENDF/B-V and ENDF/B-VI.7 and multigroup energy structure was investigated. The homogenized cross sections calculated for the ELSY fuel assembly 2D model have been evaluated and compared to the results obtained with calculations performed with the deterministic code ERANOS/ECCO using JEFF2.2 cross section library. A good agreement has been observed in the energy range of interests, and generally for energy above 1 eV. (authors)
Zizin, M. N.; Zimin, V. G.; Zizina, S. N.; Kryakvin, L. V.; Pitilimov, V. A.; Tereshonok, V. A.
2010-12-01
The ShIPR intellectual code system for mathematical simulation of nuclear reactors includes a set of computing modules implementing the preparation of macro cross sections on the basis of the two-group library of neutron-physics cross sections obtained for the SKETCH-N nodal code. This library is created by using the UNK code for 3D diffusion computation of first VVER-1000 fuel loadings. Computation of neutron fields in the ShIPR system is performed using the DP3 code in the two-group diffusion approximation in 3D triangular geometry. The efficiency of all groups of control rods for the first fuel loading of the third unit of the Kalinin Nuclear Power Plant is computed. The temperature, barometric, and density effects of reactivity as well as the reactivity coefficient due to the concentration of boric acid in the reactor were computed additionally. Results of computations are compared with the experiment.
ZZ-SCALE5.1/COVA-44G, 44-group cross section covariance matrix library extracted from SCALE5.1
International Nuclear Information System (INIS)
1 - Description: ZZ-SCALE5.1/COVA-44G is a 44-group cross section covariance matrix library retrieved from the SCALE-5.1 package. The package includes the following 4 covariance libraries in COVERX format: - 44GROUPV5COV, Basic ENDF/B-V Covariance Library - 44GROUPV5REC, Recommended ENDF/B-V Covariance Library - 44GROUPV6COV, Basic ENDF/B-VI Covariance Library - 44GROUPV6REC, Recommended ENDF/B-VI Covariance Library The files contain the covariance data for the following reactions or parameters: total, elastic, inelastic, (n,2n), fission, chi, (n,gamma), (n,p), (n,d), (n,t), (n,3He), (n,α), and ν-bar. The nuclides or materials (in ZA order) for which covariance data are provided. In parentheses the total number of the different relative covariance matrices in the four libraries for each nuclide is specified. H-1(10),H-2(3),H-3(2),He-3(2),He-4,Li-6(2),Li-7(3),Be-9(2), B-10(3),B-11(2),C-0(6),N-14(2),N-15,O-16(3),O-17,F-19(3), Na-23(3),Mg-0,Al-27(2),Si-0(3),Si-28,Si-29,Si-29,Si-30, P-31,S-0,S-32,Cl-0,K-0,Ca-0,Sc-45(2),Ti-0, V-0(2),Cr-0(2),Cr-50,Cr-52,Cr-53,Cr-54,Mn-55(3),Fe-0(2), Fe-54,Fe-56,Fe-57,Fe-58,Co-59(3),Ni-0(2),Ni-58,Ni-60, Ni-61,Ni-62,Ni-64,Cu-0,Cu-63,Cu-65,Ga-0,Ge-72, Ge-73,Ge-74,Ge-76,As-75,Se-74,Se-76,Se-77,Se-78, Se-80,Se-82,Br-79,Br-81,Kr-78,Kr-80,Kr-82,Kr-83, Kr-84,Kr-85,Kr-86,Rb-85,Rb-87,Sr-84,Sr-86,Sr-87, Sr-88,Sr-89,Sr-90,Y-89,Y-89,Y-90,Y-91,Zr-0, Zr-90,Zr-91,Zr-92,Zr-93,Zr-94,Zr-96,Nb-93,Nb-93, Nb-94,Nb-95,Mo-0,Mo-94,Mo-95,Mo-96,Mo-97,Tc-99, Ru-96,Ru-99,Ru-100,Ru-101,Ru-102,Ru-104,Ru-105,Ru-106, Rh-103,Rh-105,Pd-102,Pd-104,Pd-105,Pd-106,Pd-107,Pd-108, Pd-110,Ag-107,Ag-109,Ag-111,Cd-0,Cd-106,Cd-108,Cd-110, Cd-111,Cd-112,Cd-113,Cd-114,Cd-116,In-0,In-113,In-115, Sn-112,Sn-114,Sn-115,Sn-116,Sn-117,Sn-118,Sn-119,Sn-120, Sn-122,Sn-124,Sb-121,Sb-123,Sb-124,Te-120,Te-122,Te-123, Te-124,Te-125,Te-126,Te-127(m),Te-128,Te-130,I-127,I-129, I-130,I-131,Xe-124,Xe-126,Xe-128,Xe-129,Xe-130,Xe-131, Xe-132,Xe-133,Xe-134,Xe-135,Xe-136,Cs-133,Cs-134,Cs-135, Cs-137
International Nuclear Information System (INIS)
Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations
Revolutionizing Cross-sectional Imaging
Fan, Yifang; Luo, Liangping; Lin, Wentao; Li, Zhiyu; Zhong, Xin; Shi, Changzheng; Newman, Tony; Zhou, Yi; Lv, Changsheng; Fan, Yuzhou
2014-01-01
Cross-sectional imaging is so important that, six Nobel Prizes have been awarded to the field of nuclear magnetic resonance alone because it revolutionized clinical diagnosis. The BigBrain project supported by up to 1 billion euro each over a time period of 10 years predicts to "revolutionize our ability to understand internal brain organization" (Evan 2013). If we claim that cross-sectional imaging diagnosis is only semi-quantitative, some may believe because no doctor would ever tell their patient that we can observe the changes of this cross-sectional image next time. If we claim that BigBrain will make no difference in clinical medicine, then few would believe because no doctor would ever tell their patient to scan this part of the image and compare it with that from the BigBrain. If we claim that the BigBrain Project and the Human Brain Project have defects in their key method, one might believe it. But this is true. The key lies in the reconstruction of any cross-sectional image along any axis. Using Ga...
Terahertz radar cross section measurements
DEFF Research Database (Denmark)
Iwaszczuk, Krzysztof; Heiselberg, Henning; Jepsen, Peter Uhd
2010-01-01
We perform angle- and frequency-resolved radar cross section (RCS) measurements on objects at terahertz frequencies. Our RCS measurements are performed on a scale model aircraft of size 5-10 cm in polar and azimuthal configurations, and correspond closely to RCS measurements with conventional radar...
Cross sections for nuclear astrophysics
International Nuclear Information System (INIS)
General properties of low-energy cross sections and of reaction rates are presented. We describe different models used in nuclear astrophysics: microscopic models, the potential model, and the R-matrix method. Two important reactions, 7Be(p,γ)8B and 12C(α,γ)16O, are then briefly discussed. (author)
Measurement cross sections for radioisotopes production
International Nuclear Information System (INIS)
New radioactive isotopes for nuclear medicine can be produced using particle accelerators. This is one goal of Arronax, a high energy - 70 MeV - high intensity - 2*350 μA - cyclotron set up in Nantes. A priority list was established containing β- - 47Sc, 67Cu - β+ - 44Sc, 64Cu, 82Sr/82Rb, 68Ge/68Ga - and α emitters - 211At. Among these radioisotopes, the Scandium 47 and the Copper 67 have a strong interest in targeted therapy. The optimization of their productions required a good knowledge of their cross-sections but also of all the contaminants created during irradiation. We launched on Arronax a program to measure these production cross-sections using the Stacked-Foils' technique. It consists in irradiating several groups of foils - target, monitor and degrader foils - and in measuring the produced isotopes by γ-spectrometry. The monitor - natCu or natNi - is used to correct beam loss whereas degrader foils are used to lower beam energy. We chose to study the natTi(p,X)47Sc and 68Zn(p,2p)67Cu reactions. Targets are respectively natural Titanium foil - bought from Goodfellow - and enriched Zinc 68 deposited on Silver. In the latter case, Zn targets were prepared in-house - electroplating of 68Zn - and a chemical separation between Copper and Gallium isotopes has to be made before γ counting. Cross-section values for more than 40 different reactions cross-sections have been obtained from 18 MeV to 68 MeV. A comparison with the Talys code is systematically done. Several parameters of theoretical models have been studied and we found that is not possible to reproduce faithfully all the cross-sections with a given set of parameters. (author)
Neutron cross section standards and instrumentation
1992-09-01
This report from the National Institute of Standards and Technology contains a summary of the accomplishments of the Neutron Cross Section Standards and Instrumentation Project during the second year of a three-year interagency agreement. This program includes a broad range of data measurements and evaluations. An emphasis has been focused on the (sup 10)B cross sections where serious discrepancies in the nuclear data base remain. In particular, there are important problems with the interpretation of the helium gas production associated with diagnostic measurements of interest in nuclear technology. The enhanced use of this isotope for medical treatment is also of significance. New measurements of neutron reaction cross sections for (sup 10)B are in progress in collaboration with scientists at the Oak Ridge National Laboratory. New experiments are in progress on the important dosimetry standards (sup 237)Np(n,f) and (sup 239)Pu(n,f) below 1 MeV neutron energy. In addition, new measurements of charged-particle production in basic biological elements for medical applications are underway. Further measurements are planned or in progress in collaborations which include fission fragment energy and angular distributions, and neutron energy spectra and angular distributions from neutron-induced fission. Also measurements of angular distributions of neutrons from scattering on protons, and determinations of capture cross section of gold are planned for a later time. Data evaluation will shift to include a unified international effort to motivate new measurements and evaluations. In response to the requests of the measurement community, NIST is beginning the formation of a national depository for fissionable isotope mass standards. This action will preserve for future measurements the valuable and irreplaceable critical samples whose masses and composition have been carefully determined and documented over the past 30 years of the nuclear program.
Neutron cross section standards and instrumentation
International Nuclear Information System (INIS)
This report from the National Institute of Standards and Technology contains a summary of the accomplishments of the Neutron Cross Section Standards and Instrumentation Project during the second year of a three-year interagency agreement. This program includes a broad range of data measurements and evaluations. An emphasis has been focused on the 10B cross sections where serious discrepancies in the nuclear data base remain. In particular, there are important problems with the interpretation of the helium gas production associated with diagnostic measurements of interest in nuclear technology. The enhanced use of this isotope for medical treatment is also of significance. New measurements of neutron reaction cross sections for 10B are in progress in collaboration with scientists at the Oak Ridge National Laboratory. New experiments are in progress on the important dosimetry standards 237Np(n,f) and 239Pu(n,f) below 1 MeV neutron energy. In addition, new measurements of charged-particle production in basic biological elements for medical applications are underway. Further measurements are planned or in progress in collaborations which include fission fragment energy and angular distributions, and neutron energy spectra and angular distributions from neutron-induced fission. Also measurements of angular distributions of neutrons from scattering on protons, and determinations of capture cross section of gold are planned for a later time. Data evaluation will shift to include a unified international effort to motivate new measurements and evaluations. In response to the requests of the measurement community, NIST is beginning the formation of a national depository for fissionable isotope mass standards. This action will preserve for future measurements the valuable and irreplaceable critical samples whose masses and composition have been carefully determined and documented over the past 30 years of the nuclear program
Metonymy and Cross Section Demand
Evstigneev, Igor V.; Hildenbrand, Werner; Jerison, Michael
1996-01-01
Cross section consumer expenditure data are frequently used to make conclusions about consumer demand behavior. Such conclusions, however, can only be justified under certain assumptions, which are often left unstated in the empirical demand literature. An assumption of this type, the metonymy hypothesis, was stated rigorously and then exploited by Hardle, Hildenbrand and Jerison when analyzing the monotonicity property of aggregate demand functions. The purpose of the present paper is to exa...
Wind Turbine Radar Cross Section
David Jenn; Cuong Ton
2012-01-01
The radar cross section (RCS) of a wind turbine is a figure of merit for assessing its effect on the performance of electronic systems. In this paper, the fundamental equations for estimating the wind turbine clutter signal in radar and communication systems are presented. Methods of RCS prediction are summarized, citing their advantages and disadvantages. Bistatic and monostatic RCS patterns for two wind turbine configurations, a horizontal axis three-blade design and a vertical axi...
Polynomial parameterized representation of macroscopic cross section for PWR reactor
Energy Technology Data Exchange (ETDEWEB)
Fiel, Joao Claudio B., E-mail: fiel@ime.eb.br [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Departamento de Engenharia Nuclear
2015-07-01
The purpose of this work is to describe, by means of Tchebychev polynomial, a parameterized representation of the homogenized macroscopic cross section for PWR fuel element as a function of soluble boron concentration, moderator temperature, fuel temperature, moderator density and {sup 235} U {sub 92} enrichment. Analyzed cross sections are: fission, scattering, total, transport, absorption and capture. This parameterization enables a quick and easy determination of the problem-dependent cross-sections to be used in few groups calculations. The methodology presented here will enable to provide cross-sections values to perform PWR core calculations without the need to generate them based on computer code calculations using standard steps. The results obtained by parameterized cross-sections functions, when compared with the cross-section generated by SCALE code calculations, or when compared with K{sub inf}, generated by MCNPX code calculations, show a difference of less than 0.7 percent. (author)
Microscopic cross sections: An utopia?
Energy Technology Data Exchange (ETDEWEB)
Hilaire, S. [CEA Bruyeres-le-Chatel, DIF 91 (France); Koning, A.J. [Nuclear Research and Consultancy Group, PO Box 25, 1755 ZG Petten (Netherlands); Goriely, S. [Institut d' Astronomie et d' Astrophysique, Universite Libre de Bruxelles, Campus de la Plaine, CP 226, 1050 Brussels (Belgium)
2010-07-01
The increasing need for cross sections far from the valley of stability poses a challenge for nuclear reaction models. So far, predictions of cross sections have relied on more or less phenomenological approaches, depending on parameters adjusted to available experimental data or deduced from systematical relations. While such predictions are expected to be reliable for nuclei not too far from the experimentally known regions, it is clearly preferable to use more fundamental approaches, based on sound physical bases, when dealing with very exotic nuclei. Thanks to the high computer power available today, all major ingredients required to model a nuclear reaction can now be (and have been) microscopically (or semi-microscopically) determined starting from the information provided by a nucleon-nucleon effective interaction. We have implemented all these microscopic ingredients in the TALYS nuclear reaction code, and we are now almost able to perform fully microscopic cross section calculations. The quality of these ingredients and the impact of using them instead of the usually adopted phenomenological parameters will be discussed. (authors)
Atomic-process cross section data, 1
International Nuclear Information System (INIS)
Compiled by the Data Study Group, the data are intended for fusion plasma physics research. Cross sections of the latest experimental and theoretic studies cover the processes involving H,D,T as principal plasma materials as well as photons and electrons: emission and absorption of electromagnetic wave, electron collision, ion collision, recombination, neutral atom mutual collision, etc. Edition is so made to enable the future renewal by users. (J.P.N.)
Directory of Open Access Journals (Sweden)
Sousa Renata
2009-10-01
Full Text Available Abstract Background Demographic ageing is occurring at an unprecedented rate in China. Chronic diseases and their disabling consequences will become much more common. Public policy has a strong urban bias, and older people living in rural areas may be especially vulnerable due to limited access to good quality healthcare, and low pension coverage. We aim to compare the sociodemographic and health characteristics, health service utilization, needs for care and informal care arrangements of representative samples of older people in two Beijing communities, urban Xicheng and rural Daxing. Methods A one-phase cross-sectional survey of all those aged 65 years and over was conducted in urban and rural catchment areas in Beijing, China. Assessments included questionnaires, a clinical interview, physical examination, and an informant interview. Prevalence of chronic diseases, self-reported impairments and risk behaviours was calculated adjusting for household clustering. Poisson working models were used to estimate the independent effect of rural versus urban residence, and to explore the predictors of health services utilization. Results We interviewed 1002 participants in rural Daxing, and 1160 in urban Xicheng. Those in Daxing were more likely to be younger, widowed, less educated, not receiving a pension, and reliant on family transfers. Chronic diseases were more common in Xicheng, when based on self-report rather than clinical assessment. Risk exposures were more common in Daxing. Rural older people were much less likely to access health services, controlling for age and health. Community health services were ineffective, particularly in Daxing, where fewer than 3% of those with hypertension were adequately controlled. In Daxing, care was provided by family, who had often given up work to do so. In Xicheng, 45% of those needing care were supported by paid caregivers. Caregiver strain was higher in Xicheng. Dementia was strongly associated with
Neutron total scattering cross sections of elemental antimony
International Nuclear Information System (INIS)
Neutron total cross sections are measured from 0.8 to 4.5 MeV with broad resolutions. Differential-neutron-elastic-scattering cross sections are measured from 1.5 to 4.0 MeV at intervals of 50 to 200 keV and at scattering angles distributed between 20 and 160 degrees. Lumped-level neutron-inelastic-scattering cross sections are measured over the same angular and energy range. The exPerimental results are discussed in terms of an optical-statistical model and are compared with respective values given in ENDF/B-V
Electron capture cross sections for stellar nucleosynthesis
Giannaka, P G
2015-01-01
In the first stage of this work, we perform detailed calculations for the cross sections of the electron capture on nuclei under laboratory conditions. Towards this aim we exploit the advantages of a refined version of the proton-neutron quasi-particle random-phase approximation (pn-QRPA) and carry out state-by-state evaluations of the rates of exclusive processes that lead to any of the accessible transitions within the chosen model space. In the second stage of our present study, we translate the above mentioned $e^-$-capture cross sections to the stellar environment ones by inserting the temperature dependence through a Maxwell-Boltzmann distribution describing the stellar electron gas. As a concrete nuclear target we use the $^{66}Zn$ isotope, which belongs to the iron group nuclei and plays prominent role in stellar nucleosynthesis at core collapse supernovae environment.
Surrogate reaction methods for neutron induced cross-sections
International Nuclear Information System (INIS)
A brief discussion on surrogate reaction methods and some of the recent results on neutron induced fission cross-section measurements carried out by our group and the possibility of extending the measurements for determining (n,g), (n,2n) and (n,p) reaction cross-sections by surrogate reaction method are presented
[Fast neutron cross section measurements
International Nuclear Information System (INIS)
In this report, we outline the progress achieved in two distinct under the DOE-sponsored cross section project: the initial results obtained from the pulsed 14 MeV neutron facility, and a cooperative effort with Argonne National Laboratory in the measurement of fast neutron cross sections in yttrium. In the 14 MeV neutron laboratory, this year has seen the maturation of the project into one in which initial scattering measurements are now underway. We have improved the accelerator and ion source in several significant ways, so that neutron intensities have now been proven to be adequate for our series of elastic scattering angular distribution measurements outlined in our initial proposal of two years ago. We have successfully tested all components of the time-of-flight spectrometer and recorded initial neutron spectra from the ring targets that we have obtained for our first angular distribution measurements. Examples of the time-of-flight spectra that have been obtained are given later in this report. At the present time, the accelerator is operating with the highest degree of reliability that we have experienced since installing the pulsing system. Improvements made over the past year have not only increased the available neutron intensity, but also increased our capability to deal with inevitable component failures that require repair or replacement. The measurements carried out in conjunction with Argonne have contributed significantly to the available database on fast neutron interactions in yttrium. Results indicate that the cross section for the 89 Y(n,p)89Sr reaction is substantially higher than represented in ENDF/B-VI
Wind Turbine Radar Cross Section
Directory of Open Access Journals (Sweden)
David Jenn
2012-01-01
Full Text Available The radar cross section (RCS of a wind turbine is a figure of merit for assessing its effect on the performance of electronic systems. In this paper, the fundamental equations for estimating the wind turbine clutter signal in radar and communication systems are presented. Methods of RCS prediction are summarized, citing their advantages and disadvantages. Bistatic and monostatic RCS patterns for two wind turbine configurations, a horizontal axis three-blade design and a vertical axis helical design, are shown. The unique electromagnetic scattering features, the effect of materials, and methods of mitigating wind turbine clutter are also discussed.
International Nuclear Information System (INIS)
In the paper the formulae for perturbation theory functionals calculation are given and equations are based on improved coarse mesh discretization of diffusion problem in 3-dimensional geometry (Hex-Z). Expressions for the reactivity effect components and reactivity coefficients, written in the framework of the first order perturbation theory, are presented. On this basis the formulae for estimation of the sensitivity coefficients of different reactivity effects group cross-sections were derived. Expressions for the reactivity effect and its components obtained in the framework of the strict perturbation theory, are also presented in the paper. (author)
Ardila, Carlos M; Posada-López, Adriana; Agudelo-Suárez, Andrés A
2016-02-01
Regional contextual factors and dental caries using multilevel modeling related to adults in minority ethnic groups have been scantily explored. The influence of the socioeconomic context on self-reported dental caries (SRDC) in individuals of minority ethnic groups (IEG) in Colombia was studied. Data from the 2007 National Public Health Survey were collected in 34,843 participants of the population. The influence of different factors on SRDC in IEG was investigated with logistic and multilevel regression analyses. A total of 6440 individuals belonged to an ethnic group. Multilevel analysis showed a significant variance in SRDC that was smaller in IEG level than between states. Multilevel multivariate analysis also associated SRDC with increasing age, lower education level, last dental visit >1 year, unmet dental need and low Gross Domestic Product (GDP). Minority ethnic groups were at risk to report higher dental caries, where low GDP was an important variable to be considered. PMID:25963050
Parametric equations for calculation of macroscopic cross sections
Energy Technology Data Exchange (ETDEWEB)
Botelho, Mario Hugo; Carvalho, Fernando, E-mail: mariobotelho@poli.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear
2015-07-01
Neutronic calculations of the core of a nuclear reactor is one thing necessary and important for the design and management of a nuclear reactor in order to prevent accidents and control the reactor efficiently as possible. To perform these calculations a library of nuclear data, including cross sections is required. Currently, to obtain a cross section computer codes are used, which require a large amount of processing time and computer memory. This paper proposes the calculation of macroscopic cross section through the development of parametric equations. The paper illustrates the proposal for the case of macroscopic cross sections of absorption (Σa), which was chosen due to its greater complexity among other cross sections. Parametric equations created enable, quick and dynamic way, the determination of absorption cross sections, enabling the use of them in calculations of reactors. The results show efficient when compared with the absorption cross sections obtained by the ALPHA 8.8.1 code. The differences between the cross sections are less than 2% for group 2 and less than 0.60% for group 1. (author)
Energy Technology Data Exchange (ETDEWEB)
Jordan, W.C.
1993-02-01
A version of KENO V.a and the 27-group library in SCALE-4.0 were validated for use in evaluating the nuclear criticality safety of low-enriched uranium systems. A total of 59 critical systems were analyzed. A statistical analysis of the results was performed, and subcritical acceptanced criteria are established.
Energy Technology Data Exchange (ETDEWEB)
Jordan, W.C.
1993-02-01
A version of KENO V.a and the 27-group library in SCALE-4.0 were validated for use in evaluating the nuclear criticality safety of low-enriched uranium systems. A total of 59 critical systems were analyzed. A statistical analysis of the results was performed, and subcritical acceptanced criteria are established.
Electron-Impact Ionization Cross Section Database
SRD 107 Electron-Impact Ionization Cross Section Database (Web, free access) This is a database primarily of total ionization cross sections of molecules by electron impact. The database also includes cross sections for a small number of atoms and energy distributions of ejected electrons for H, He, and H2. The cross sections were calculated using the Binary-Encounter-Bethe (BEB) model, which combines the Mott cross section with the high-incident energy behavior of the Bethe cross section. Selected experimental data are included.
[Fast neutron cross section measurements
International Nuclear Information System (INIS)
In the 14 MeV Neutron Laboratory, we have continued the development of a facility that is now the only one of its kind in operation in the United States. We have refined the klystron bunching system described in last year's report to the point that 1.2 nanosecond pulses have been directly measured. We have tested the pulse shape discrimination capability of our primary NE 213 neutron detector. We have converted the RF sweeper section of the beamline to a frequency of 1 MHz to replace the function of the high voltage pulser described in last year's report which proved to be difficult to maintain and unreliable in its operation. We have also overcome several other significant experimental difficulties, including a major problem with a vacuum leak in the main accelerator column. We have completed additional testing to prove the remainder of the generation and measurement systems, but overcoming some of these experimental difficulties has delayed the start of actual data taking. We are now in a position to begin our first series of ring geometry elastic scattering measurements, and these will be underway before the end of the current contract year. As part of our longer term planning, we are continuing the conceptual analysis of several schemes to improve the intensity of our current pulsed beam. These include the provision of a duoplasmatron ion source and/or the provision of preacceleration bunching. Additional details are given later in this report. A series of measurements were carried out at the Tandem Dynamatron Facility involving the irradiation of a series of yttrium foils and the determination of activation cross sections using absolute counting techniques. The experimental work has been completed, and final analysis of the cross section data will be completed within several months
Directory of Open Access Journals (Sweden)
Nadine Correia Santos
2014-02-01
Full Text Available It is relevant to unravel the factors that may mediate the cognitive decline observed during aging. Previous reports indicate that education has a positive influence on cognitive performance, while age, female gender and, especially, depressed mood were associated with poorer performances across multiple cognitive dimensions (memory and general executive function. Herein, the present study aimed to characterize the cognitive performance of community-dwelling individuals within distinct educational groups categorized by the number of completed formal school years: less than 4, 4, completed primary education, and more than 4. Participants (n = 1051 were randomly selected from local health registries and representative of the Portuguese population for age and gender. Neurocognitive and clinical assessments were conducted in local health care centers. Structural equation modeling was used to derive a cognitive score, and hierarchical linear regressions were conducted for each educational group. Education, age and depressed mood were significant variables in directly explaining the obtained cognitive score, while gender was found to be an indirect variable. In all educational groups, mood was the most significant factor with effect on cognitive performance. Specifically, a depressed mood led to lower cognitive performance. The clinical disease indices cardiac and stroke associated with a more negative mood, while moderate increases in BMI, alcohol consumption and physical activity associated positively with improved mood and thus benefitted cognitive performance. Results warrant further research on the cause-effect (longitudinal relationship between clinical indices of disease and risk factors and mood and cognition throughout aging.
Cross Sections for Electron Impact Excitation of Ions Relevant to Planetary Atmospheres Observation
Tayal, Swaraj S.
1998-01-01
The goal of this research grant was to calculate accurate oscillator strengths and electron collisional excitation strengths for inelastic transitions in atomic species of relevance to Planetary Atmospheres. Large scale configuration-interaction atomic structure calculations have been performed to obtain oscillator strengths and transition probabilities for transitions among the fine-structure levels and R-matrix method has been used in the calculations of electron-ion collision cross sections of C II, S I, S II, S III, and Ar II. A number of strong features due to ions of sulfur have been detected in the spectra of Jupiter satellite Io. The electron excitation cross sections for the C II and S II transitions are studied in collaboration with the experimental atomic physics group at the Jet Propulsion Laboratory. There is excellent agreement between experiment and theory which provide an accurate and broad-base test of the ability of theoretical methods used in the calculation of atomic processes. Specifically, research problems have been investigated for: electron impact excitation cross sections of C II: electron impact excitation cross sections of S III; energy levels and oscillator strengths for transitions in S III; collision strengths for electron collisional excitation of S II; electron impact excitation of inelastic transitions in Ar II; oscillator strengths of fine-structure transitions in neutral sulfur; cross sections for inelastic scattering of electrons from atomic nitrogen; and excitation of atomic ions by electron impact.
Evaluation of cross section for 103Rh
International Nuclear Information System (INIS)
A completely new evaluation for the neutron cross sections is presented. The experimental data mainly referred to EXFOR, and the recommended cross sections are compared with ENDF/B-6, BROND-2, JENDL-3.2 and JEF-2
Directory of Open Access Journals (Sweden)
Keegan Theresa HM
2007-10-01
Full Text Available Abstract Background Breast cancer is the most commonly diagnosed cancer among the rapidly growing population of Asian Americans; it is also the most common cause of cancer mortality among Filipinas. Asian women continue to have lower rates of mammographic screening than women of most other racial/ethnic groups. While prior studies have described the effects of sociodemographic and other characteristics of women on non-adherence to screening guidelines, they have not identified the distinct segments of the population who remain at highest risk of not being screened. Methods To better describe characteristics of Asian women associated with not having a mammogram in the last two years, we applied recursive partitioning to population-based data (N = 1521 from the 2001 California Health Interview Survey (CHIS, for seven racial/ethnic groups of interest: Chinese, Japanese, Filipino, Korean, South Asian, Vietnamese, and all Asians combined. Results We identified two major subgroups of Asian women who reported not having a mammogram in the past two years and therefore, did not follow mammography screening recommendations: 1 women who have never had a pap exam to screen for cervical cancer (68% had no mammogram, and 2 women who have had a pap exam, but have no women's health issues (osteoporosis, using menopausal hormone therapies, and/or hysterectomy nor a usual source of care (62% had no mammogram. Only 19% of Asian women who have had pap screening and have women's health issues did not have a mammogram in the past two years. In virtually all ethnic subgroups, having had pap or colorectal screening were the strongest delineators of mammography usage. Other characteristics of women least likely to have had a mammogram included: Chinese non-U.S. citizens or citizens without usual source of health care, Filipinas with no health insurance, Koreans without women's health issues and public or no health insurance, South Asians less than age 50 who were
2010-01-01
Background Primary prevention should be targeted at individuals with high global cardiovascular risk, but research is lacking on how best to identify such individuals in the general population. Family history is a good proxy measure of global risk and may provide an efficient mechanism for identifying high risk individuals. The aim was to test the feasibility of using patients with premature cardiovascular disease to recruit family members as a means of identifying and screening high-risk individuals. Findings We recruited family members of 50 patients attending a cardiology clinic for premature coronary heart disease (CHD). We compared their cardiovascular risk with a general population control group, and determined their perception of their risk and current level of screening. 103 (36%) family members attended screening (27 siblings, 48 adult offspring and 28 partners). Five (5%) had prevalent CHD. A significantly higher percentage had an ASSIGN risk score >20% compared with the general population (13% versus 2%, p < 0.001). Only 37% of family members were aware they were at increased risk and only 50% had had their blood pressure and serum cholesterol level checked in the previous three years. Conclusions Patients attending hospital for premature CHD provide a mechanism to contact family members and this can identify individuals with a high global risk who are not currently screened. PMID:20459771
Photoproduction total cross section and shower development
Cornet, F.; García Canal, C. A.; Grau, A.; Pancheri, G.; Sciutto, S. J.
2015-12-01
The total photoproduction cross section at ultrahigh energies is obtained using a model based on QCD minijets and soft-gluon resummation and the ansatz that infrared gluons limit the rise of total cross sections. This cross section is introduced into the Monte Carlo system AIRES to simulate extended air showers initiated by cosmic ray photons. The impact of the new photoproduction cross section on common shower observables, especially those related to muon production, is compared with previous results.
Photoproduction total cross section and shower development
Cornet, F; Grau, A; Pancheri, G; Sciutto, S J
2015-01-01
The total photoproduction cross section at ultra-high energies is obtained using a model based on QCD minijets and soft-gluon resummation and the ansatz that infrared gluons limit the rise of total cross sections. This cross section is introduced into the Monte Carlo system AIRES to simulate extended air-showers initiated by cosmic ray photons. The impact of the new photoproduction cross section on common shower observables, especially those related to muon production, is compared with previous results.
Directory of Open Access Journals (Sweden)
Pellerone M
2015-08-01
Full Text Available Monica Pellerone,1 Alessia Passanisi,1 Mario Filippo Paolo Bellomo2 1Faculty of Human and Social Science, “Kore” University of Enna, Enna, 2Credito Emiliano Bank, Piazza Armerina, Italy Background: Forming one’s identity is thought to be the key developmental task of adolescence, but profound changes in personality traits also occur in this period. The negotiation of complex social settings, the creation of an integrated identity, and career choice are major tasks of adolescence. The adolescent, having to make choices for his or her future, has not only to consider his or her own aspirations and interests but also to possess a capacity for exploration and commitment; in fact, career commitments can be considered as a fit between the study or career that is chosen and personal values, skills, and preferences. Methods: The objective of the study reported here was to investigate the role of identity on profile of interests; the relation between identity and decisional style; the correlation between identity, aptitudes, interests, and school performance; and the predictive variables to school success. The research involved 417 Italian students who live in Enna, a small city located in Sicily, Italy, aged 16–19 years (197 males and 220 females in the fourth year (mean =17.2, standard deviation =0.52 and the fifth year (mean =18.2, standard deviation =0.64 of senior secondary school. The research lasted for one school year; the general group of participants consisted of 470 students, and although all participants agreed to be part of the research, there was a dropout rate of 11.28%. They completed the Ego Identity Process Questionnaire to measure their identity development, the Intelligence Structure Test to investigate aptitudes, the Self-Directed Search to value interests, and General Decision Making Style questionnaire to describe their individual decisional style. Results: The data showed that high-school performance was positively
JENDL gas-production cross section file
International Nuclear Information System (INIS)
The JENDL gas-production cross section file was compiled by taking cross-section data from JENDL-3 and by using the ENDF-5 format. The data were given to 23 nuclei or elements in light nuclei and structural materials. Graphs of the cross sections and brief description on their evaluation methods are given in this report. (author)
Status of pseudo-fission-product cross-sections for fast reactors
International Nuclear Information System (INIS)
Within the framework of the Subgroup 17 (SG17) benchmark organized by a Working Party of the Nuclear Science Committee of the Nuclear Energy Agency (FR), a comparison of lumped or pseudo-fission-product cross-sections for fast reactors has been made. Several parameters have been compared: the one- group cross-sections and reactivity worths of the lumped nuclide for several partial absorption and scattering cross-sections, and the one-group cross sections of individual fission products. Graphs of the multi-group cross-sections and those of capture cross-sections for 27 nuclides have also been compared. (R.P.)
[Fast neutron cross section measurements
International Nuclear Information System (INIS)
From its inception, the Nuclear Data Project at the University of Michigan has concentrated on two major objectives: (1) to carry out carefully controlled nuclear measurements of the highest possible reliability in support of the national nuclear data program, and (2) to provide an educational opportunity for students with interests in experimental nuclear science. The project has undergone a successful transition from a primary dependence on our photoneutron laboratory to one in which our current research is entirely based on a unique pulsed 14 MeV fast neutron facility. The new experimental facility is unique in its ability to provide nanosecond bursts of 14 MeV neutrons under conditions that are ''clean'' and as scatter-free as possible, and is the only one of its type currently in operation in the United States. It has been designed and put into operation primarily by graduate students, and has met or exceeded all of its important initial performance goals. We have reached the point of its routine operation, and most of the data are now in hand that will serve as the basis for the first two doctoral dissertations to be written by participating graduate students. Our initial results on double differential neutron cross sections will be presented at the May 1993 Fusion Reactor Technology Workshop. We are pleased to report that, after investing several years in equipment assembly and optimization, the project has now entered its ''data production'' phase
SNL RML recommended dosimetry cross section compendium
Energy Technology Data Exchange (ETDEWEB)
Griffin, P.J.; Kelly, J.G.; Luera, T.F. [Sandia National Labs., Albuquerque, NM (United States); VanDenburg, J. [Science and Engineering Associates, Inc., Albuquerque, NM (United States)
1993-11-01
A compendium of dosimetry cross sections is presented for use in the characterization of fission reactor spectrum and fluence. The contents of this cross section library are based upon the ENDF/B-VI and IRDF-90 cross section libraries and are recommended as a replacement for the DOSCROS84 multigroup library that is widely used by the dosimetry community. Documentation is provided on the rationale for the choice of the cross sections selected for inclusion in this library and on the uncertainty and variation in cross sections presented by state-of-the-art evaluations.
Recent fission cross section standards measurements
Energy Technology Data Exchange (ETDEWEB)
Wasson, O.A.
1985-01-01
The /sup 235/U(n,f) reaction is the standard by which most neutron induced fission cross sections are determined. Most of these cross sections are derived from relatively easy ratio measurements to /sup 235/U. However, the more difficult /sup 235/U(n,f) cross section measurements require the use of advanced neutron detectors for the determination of the incident neutron fluence. Examples of recent standard cross section measurements are discussed, various neutron detectors are described, and the status of the /sup 235/U(n,f) cross section standard is assessed. 23 refs., 8 figs., 4 tabs.
Recent fission cross section standards measurements
International Nuclear Information System (INIS)
The 235U(n,f) reaction is the standard by which most neutron induced fission cross sections are determined. Most of these cross sections are derived from relatively easy ratio measurements to 235U. However, the more difficult 235U(n,f) cross section measurements require the use of advanced neutron detectors for the determination of the incident neutron fluence. Examples of recent standard cross section measurements are discussed, various neutron detectors are described, and the status of the 235U(n,f) cross section standard is assessed. 23 refs., 8 figs., 4 tabs
The Elusive p-air Cross Section
Block, Martin M
2006-01-01
For the $\\pbar p$ and $pp$ systems, we have used all of the extensive data of the Particle Data Group[K. Hagiwara {\\em et al.} (Particle Data Group), Phys. Rev. D 66, 010001 (2002).]. We then subject these data to a screening process, the ``Sieve'' algorithm[M. M. Block, physics/0506010.], in order to eliminate ``outliers'' that can skew a $\\chi^2$ fit. With the ``Sieve'' algorithm, a robust fit using a Lorentzian distribution is first made to all of the data to sieve out abnormally high $\\delchi$, the individual i$^{\\rm th}$ point's contribution to the total $\\chi^2$. The $\\chi^2$ fits are then made to the sieved data. We demonstrate that we cleanly discriminate between asymptotic $\\ln s$ and $\\ln^2s$ behavior of total hadronic cross sections when we require that these amplitudes {\\em also} describe, on average, low energy data dominated by resonances. We simultaneously fit real analytic amplitudes to the ``sieved'' high energy measurements of $\\bar p p$ and $pp$ total cross sections and $\\rho$-values for $\\...
Vertically stabilized elongated cross-section tokamak
Sheffield, George V.
1977-01-01
This invention provides a vertically stabilized, non-circular (minor) cross-section, toroidal plasma column characterized by an external separatrix. To this end, a specific poloidal coil means is added outside a toroidal plasma column containing an endless plasma current in a tokamak to produce a rectangular cross-section plasma column along the equilibrium axis of the plasma column. By elongating the spacing between the poloidal coil means the plasma cross-section is vertically elongated, while maintaining vertical stability, efficiently to increase the poloidal flux in linear proportion to the plasma cross-section height to achieve a much greater plasma volume than could be achieved with the heretofore known round cross-section plasma columns. Also, vertical stability is enhanced over an elliptical cross-section plasma column, and poloidal magnetic divertors are achieved.
Measurements of neutron capture cross sections
International Nuclear Information System (INIS)
A review of measurement techniques for the neutron capture cross sections is presented. Sell transmission method, activation method, and prompt gamma-ray detection method are described using examples of capture cross section measurements. The capture cross section of 238U measured by three different prompt gamma-ray detection methods (large liquid scintillator, Moxon-Rae detector, and pulse height weighting method) are compared and their discrepancies are resolved. A method how to derive the covariance is described. (author)
Compilation of cross-sections. Pt. 2
International Nuclear Information System (INIS)
A compilation of integrated cross-sections for hadronic reactions is presented. This is an updated version of CERN/HERA 79-1, 79-2, 79-3. It contains all data published up to the beginning of 1982, but some more recent data have also been included. Plots of the cross sections versus incident laboratory momentum are also given. This volume II contains cross-sections for K+ and K- induced reactions. (orig.)
Cross Sections for Electron Collisions with Methane
Energy Technology Data Exchange (ETDEWEB)
Song, Mi-Young, E-mail: mysong@nfri.re.kr; Yoon, Jung-Sik [Plasma Technology Research Center, National Fusion Research Institute, 814-2 Osikdo-dong, Gunsan, Jeollabuk-do 573-540 (Korea, Republic of); Cho, Hyuck [Department of Physics, Chungnam National University, Daejeon 305-764 (Korea, Republic of); Itikawa, Yukikazu [Institute of Space and Astronautical Science, Sagamihara 252-5210 (Japan); Karwasz, Grzegorz P. [Faculty of Physics, Astronomy and Applied Informatics, University Nicolaus Copernicus, Grudziadzka 5, 87100 Toruń (Poland); Kokoouline, Viatcheslav [Department of Physics, University of Central Florida, Orlando, Florida 32816 (United States); Nakamura, Yoshiharu [6-1-5-201 Miyazaki, Miyamae, Kawasaki 216-0033 (Japan); Tennyson, Jonathan [Department of Physics and Astronomy, University College London, Gower Street, London WC1E 6BT (United Kingdom)
2015-06-15
Cross section data are compiled from the literature for electron collisions with methane (CH{sub 4}) molecules. Cross sections are collected and reviewed for total scattering, elastic scattering, momentum transfer, excitations of rotational and vibrational states, dissociation, ionization, and dissociative attachment. The data derived from swarm experiments are also considered. For each of these processes, the recommended values of the cross sections are presented. The literature has been surveyed through early 2014.
Ion and electron impact ionization cross sections
International Nuclear Information System (INIS)
Several current projects are described in which cross sections of interest to radiation physics are being measured. These include total and multiple ionization cross sections for protons on several gases covering a wide energy range, the measurement of cross sections differential in the angle and energy of ejected electrons for several gases including water vapor, and a review of proton ionization data. The work on water vapor has also been extended to electron and neutral hydrogen impact. A brief discussion is also given of some systematics of ionization cross sections. 13 references
Improved Empirical Parametrization of Fragmentation Cross Sections
Sümmerer, Klaus
2012-01-01
A new version is proposed for the universal empirical formula, EPAX, which describes fragmentation cross sections in high-energy heavy-ion reactions. The new version, EPAX 3, can be shown to yield cross sections that are in better agreement with experimental data for the most neutron-rich fragments than the previous version. At the same time, the very good agreement of EPAX 2 with data on the neutron-deficient side has been largely maintained. Comparison with measured cross sections show that the bulk of the data is reproduced within a factor of about 2, for cross sections down to the pico-barn range.
Damage cross section library (DAMSIG77)
International Nuclear Information System (INIS)
The damage cross sections of various materials are converted to a data format, which can be used as library for the program SAND-II. The materials available in this library are graphite, stainless steel, aluminium, silicium, chromium, iron, nickel, copper, zirconium, molybdenum, tungsten, vanadium and niobium. A number of these materials have more than one cross section set, originating from different evaluations. Cross sections for some activation reactions, commonly used to determine thermal and fast neutron fluences have been included too. Moreover, also some artificial cross sections are introduced in this library which can be used to derive values for some physical quantities which may characterize neutron spectra
Burnup-dependent cross section data for research reactors
International Nuclear Information System (INIS)
Studies currently in progress consider research and test reactors which commonly have burnups of 50 atom percent in 235-U and may reach as high a 70 atom percent. At these levels of burnup changes in cross-section data with burnup become significant. Some preliminary studies of these effects lead to the development of a modified version of REBUS-2 which supports changes in cross-section data with burnup. This version of REBUS-2 allows for changes in the cross-section data only at each time sub-interval in the problem, and these cross-section changes for capture and fission are based on a least squares polynomial fit as a function of burnup. In this paper an attempt is made to evaluate the importance of burnup dependent data for the various isotopes and/or groups, and to assess the accuracy of this method by comparing the REBUS-2 results with results obtained from PDQ-7. The 10 MW IAEA benchmark problem has been selected for this study. A description of the reactor and the XY model can be found in the IAEA Guidebook. The EPRI-CELL4 code was used to generate burnup dependent cross section data for use with both REBUS-2 and PDQ-7. Cross-section data were generated at 10 time steps to a burnup of approximately 50 atom percent in 235-U. The agreement between the PDQ-7 results and the REBUS-2 results with fitted burnup dependent cross-section data are quite good. Burnup dependent cross sections are essential for accurate estimates of cycle lengths and reactivities, and low order polynomial fits of capture and fission data for selected isotopes and energy groups can provide this capability
Fast-neutron total and scattering cross sections of niobium
International Nuclear Information System (INIS)
Neutron total cross sections of niobium were measured from approx. = 0.7 to 4.5 MeV at intervals of less than or equal to 50 keV with broad resolution. Differential-elastic-scattering cross sections were measured from approx. = 1.5 to 4.0 MeV at intervals of 0.1 to 0.2 MeV and at 10 to 20 scattering angles distributed between approx. = 20 and 160 degrees. Inelastically-scattered neutrons, corresponding to the excitation of levels at: 788 +- 23, 982 +- 17, 1088 +- 27, 1335 +- 35, 1504 +- 30, 1697 +- 19, 1971 +- 22, 2176 +- 28, 2456 +- (.), and 2581 +- (.) keV, were observed. An optical-statistical model, giving a good description of the observables, was deduced from the measured differential-elastic-scattering cross sections. The experimental-results were compared with the respective evaluated quantities given in ENDF/B-V
International Nuclear Information System (INIS)
A revised multigroup cross-section library based ON ENDF/B-VI Release 3 has been produced for light water reactor shielding and reactor pressure vessel dosimetry applications. This new broad-group library, which is designated BUGLE-96, represents an improvement over the BUGLE-93 library released in February 1994 and is expected to replace te BUGLE-93 data. The cross-section processing methodology is the same as that used for producing BUGLE-93 and is consistent with ANSI/ANS 6.1.2. As an added feature, cross-section sets having upscatter data for four thermal neutron groups are included in the BUGLE-96 package available from the Radiation Shielding Information Center. The upscattering data should improve the application of this library to the calculation of more accurate thermal fluences, although more computer time will be required. The incorporation of feedback from users has resulted in a data library that addresses a wider spectrum of user needs
COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program
Energy Technology Data Exchange (ETDEWEB)
Woo Y. Yoon; David W. Nigg
2009-08-01
COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete
COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program
International Nuclear Information System (INIS)
COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete
Electron Swarm Parameters and Electron Collision Cross Sections
International Nuclear Information System (INIS)
Electron collision cross section data for atoms and molecules and electron swarm data in respective gases are important for quantitative modeling of related plasmas. This fact and wide application of plasmas in various fields boos data collection and evaluation activities worldwide. We have been measuring electron swarm parameters (drift velocity, longitudinal diffusion coefficient, ionization/attachment coefficients, and so on) over a wide E/N range (where E is the electric field and N the gas number density) in a number of gases. We also derived a set of electron collision cross sections for each gas so that the set was consistent with our experimental swarm data. Our speciality in studying molecular target is to measure swarm parameters not only in the pure molecular gas but also in dilute molecular gas-argon gas mixtures, the mix rations of the molecule are 0.5-5.0%. The swarm parameters in pure molecular gas depend primarily on the elastic momentum transfer cross section of the molecule and its vibrational excitation cross sections. Those in the mixtures, on the other hand, depend mainly on the elastic momentum transfer cross section of major argon atom and the vibrational cross sections of minor admixed molecule. Alternative use of swarm parameters in pure molecular gas and those in the mixtures enable us to derive the momentum transfer cross section and vibrational cross sections for the molecule separately. Combination of the Ramsauer-Townsend minimum of argon atom and sharp structures in vibrational cross sections of the molecule frequently gives rise prominent E/N dependences in swarm parameters, which can be used to determine the position and magnitude of resonances in the vibrational excitation cross sections. Detailed accounts of the procedure, including estimated uncertainty in our electron swarm data and also in the resultant set of electron collision cross sections, will be given in the presentation by referring to our recent results. Stress will be
Compilation of cross-sections. Pt. 4
International Nuclear Information System (INIS)
This is the fourth volume in our series of data compilations on integrated cross-sections for weak, electromagnetic, and strong interaction processes. This volume covers data on reactions induced by photons, neutrinos, hyperons, and KL0. It contains all data published up to June 1986. Plots of the cross-sections versus incident laboratory momentum are also given. (orig.)
Compilation of cross-sections. Pt. 1
International Nuclear Information System (INIS)
A compilation of integral cross-sections for hadronic reactions is presented. This is an updated version of CERN/HERA 79-1, 79-2, 79-3. It contains all data published up to the beginning of 1982, but some more recent data have also been included. Plots of the cross-sections versus incident laboratory momentum are also given. (orig.)
Nucleon-XcJ Dissociation Cross Sections
Institute of Scientific and Technical Information of China (English)
冯又层; 许晓明; 周代翠
2002-01-01
Nucleon-XcJ dissociation cross sections are calculated in a constituent interexchange model in which quark-quark potential is derived from the Buchmüller-Tye quark-anti-quark potential. These new cross sections for dominant reaction channels depend on the centre-of-mass energy of the nucleon and the charmonium.
Fission cross section calculations for Pa isotopes
International Nuclear Information System (INIS)
Based on the recently measured cross-section values for the neutron-induced fission of 231Pa and our experience gained with other isotopes, new self consistent neutron cross section calculations for n+231Pa have been performed up to 30 MeV. The results are quite different to the existing evaluations, especially above the first chance fission threshold. (authors)
COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program
Energy Technology Data Exchange (ETDEWEB)
Woo Y. Yoon; David W. Nigg
2008-09-01
COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete
COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program
International Nuclear Information System (INIS)
COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete
International Nuclear Information System (INIS)
The effect of self-shielding of resonance cross sections on the tritium breeding ratio was investigated for three promising fusion blanket designs with liquid lithium, lithium oxide and lithium-lead breeders. Calculations were performed using ANISN and MCNP transport codes with the ENDF/B-V based nuclear data libraries. It is found that the self-shielding effect cannot be neglected in the blanket design if the blanket is neutron leaky in the case when the blanket is thin or with lower Li-6 enrichment in Li. This may result in an underestimate of the tritium breeding ratio if the cross sections are infinitely diluted. This is due to the resonances in the structure materials in which the absorption cross sections are enhanced in the infinitely diluted case. Thus the effect of self-shielding of resonance cross sections should be considered in neutronics calculations of fusion reactors. It is shown that the MCNP results are better reproduced by those from the transport code with the infinitely diluted library. This is probably due to the weight function used to generate the library and to the number of groups considered. Thus for fusion applications it is recommanded to collapse broad group cross sections with the spectrum obtained from an accurate calculation based on many fine groups. (author)
Comparative analysis among several cross section sets
International Nuclear Information System (INIS)
Critical parameters were calculated using the one dimensional multigroup transport theory for several cross section sets. Calculations have been performed for water mixtures of uranium metal, plutonium metal and uranium-thorium oxide, and for metallics systems, to determine the critical dimensions of geometries (sphere and cylinder). For this aim, the following cross section sets were employed: 1) multigroup cross section sets obtained from the GAMTEC-II code; 2) the HANSEN-ROACH cross section sets; 3) cross section sets from the ENDF/B-IV, processed by the NJOY code. Finally, we have also calculated the corresponding critical radius using the one dimensional multigroup transport DTF-IV code. The numerical results agree within a few percent with the critical values obtained in the literature (where the greatest discrepancy occured in the critical dimensions of water mixtures calculated with the values generated by the NJOY code), a very good results in comparison with similar works. (Author)
Photoproton cross section for 17O
International Nuclear Information System (INIS)
The measurement of the 17O(γ,p)16N reaction from threshold to an excitation energy of 44 MeV is presented. These results have been summed with the previously measured total photoneutron cross section to provide an approximation to the total photoabsorption cross section of 17O. The magnitude of the 17O photoabsorption cross section at the peak of the Giant Dipole Resonance is considerably less than the equivalent value for the photoabsorption cross sections of 16O and 18O. In addition, the integrated total photoabsorption cross section for 17O (up to 40 MeV) exhausts only about 58% of the sum rule; the values for the cases of 16O and 18O are significantly larger than this. The present data along with results from other reaction channels of this nucleus, were used to make spin, parity, and isospin assignments for several states in 17O. 48 refs., 4 tabs., 7 figs
Recommended evaluation procedure for photonuclear cross section
Energy Technology Data Exchange (ETDEWEB)
Lee, Young-Ouk; Chang, Jonghwa; Fukahori, Tokio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
In order to generate photonuclear cross section library for the necessary applications, data evaluation is combined with theoretical evaluation, since photonuclear cross sections measured cannot provide all necessary data. This report recommends a procedure consisting of four steps: (1) analysis of experimental data, (2) data evaluation, (3) theoretical evaluation and, if necessary, (4) modification of results. In the stage of analysis, data obtained by different measurements are reprocessed through the analysis of their discrepancies to a representative data set. In the data evaluation, photonuclear absorption cross sections are evaluated via giant dipole resonance and quasi-deutron mechanism. With photoabsorption cross sections from the data evaluation, theoretical evaluation is applied to determine various decay channel cross sections and emission spectra using equilibrium and preequilibrium mechanism. After this, the calculated results are compared with measured data, and in some cases the results are modified to better describe measurements. (author)
Photoneutron cross sections for the silicon isotopes
International Nuclear Information System (INIS)
The photoneutron cross sections for 28Si, 29Si, and 30Si have been measured up to 33 MeV with monoenergetic photons from the annihilation in flight of fast positrons, using neutron multiplicity counting. Average neutron energies were obtained simultaneously with the cross-section data by the ring-ratio technique. The giant dipole resonance for 28Si and 30Si exhibit appreciable fragmentation; that for 29Si does not. The (γ,2n) cross section for 30Si is large; that for 29Si is consistent with zero. The (γ,1n) cross section for 30Si decreases sharply with energy to values near zero as the (γ,2n) cross section grows, then increases to appreciable values as the (γ,2n) cross section diminishes; this extreme behavior, although never seen before, is attributable to the competition between the (γ,n), (γ,2n), and (γ,pn) decay channels. Some properties of the isospin components of the giant resonance are inferred. Other features of the data, including the integrated cross sections, are found to be similar in many respects to corresponding results for the oxygen and magnesium isotopes. The 28Si nucleus is found to be a better core for 29Si and 30Si than might have been expected from previous descriptions of its open-shell character
The 42Ca photoneutron cross section
International Nuclear Information System (INIS)
The measurement of the 42Ca(γ,nsub(t)) is reported here over the energy range 10.5 - 28 MeV. Bremsstrahlung radiation from the 35 MeV Betatron at this University was used to measure a yield curve of photoneutrons, from which the (γ,nsub(t)) cross section was derived. Since proton and neutron emission are the major decay modes of the giant dipole resonance, summing these cross sections approximates the photo-absorption cross section. With this information the theoretical predictions can be checked
Compilation of cross-sections. Pt. 3
International Nuclear Information System (INIS)
A compilation of integrated cross-sections for hadronic reactions is presented. This is an updated version of CERN/HERA 79-1, 79-2, 79-3. It contains all data published up to the beginning of 1982, but some more recent data, particularly those from the CERN Collider, have also been included. Plots of the cross-sections versus incident laboratory momentum are also given. This volume III contains cross-sections for p and anti p induced reactions. (orig.)
Screening corrections to the Rutherford cross section
International Nuclear Information System (INIS)
Differential cross sections for elastic p-Au scattering were measured in the energy range between 0.2 and 0.8 MeV for scattering angles from 300 to 1500 in order to determine corrections to the Rutherford cross section due to the screening of the nuclear charge by the atomic electrons. Furthermore, differential cross sections have been calculated in the weakly screening region using various screening functions. A simple analytical expression has been derived for the representation of both experimental and theoretical results. (orig.)
International evaluation cooperation Subgroup 7: Multigroup cross section processing
Energy Technology Data Exchange (ETDEWEB)
Roussin, R.W.; White, J.E. (Oak Ridge National Lab., TN (USA)); Sartori, E. (NEA Data Bank, 91 - Gif-sur-Yvette (France)); Panini, G. (ENEA, Bologna (Italy)); MacFarlane, R. (Los Alamos National Lab., NM (USA)); Muir, D. (International Atomic Energy Agency, Vienna (Austria). Nuclear Data Section); Mattes, M. (Stuttgart Univ. (Germany, F.R.). Inst. fuer Kernenergetik und Energiesysteme); Hasegawa, I
1991-01-01
The chairmen of the ENDF/B, JEF, EFF, and JENDL evaluated data files adopted a proposal to develop a fine-group processed cross section library based on the VITAMIN'' concept. The authors listed above, with support from others, are participating in this project. The end result will be a pseudo-problem-independent fine-group cross section library generated from the latest evaluated data in ENDF/B-VI, JEF-2, EFF-2, and JENDL-3. Initial applications of the library will be for shielding, fast reactor physics, and fusion neutronics. Progress made to date will be discussed. 8 refs.
Differential cross sections of positron hydrogen collisions
Institute of Scientific and Technical Information of China (English)
于荣梅; 濮春英; 黄晓玉; 殷复荣; 刘旭焱; 焦利光; 周雅君
2016-01-01
We make a detailed study on the angular differential cross sections of positron–hydrogen collisions by using the momentum-space coupled-channels optical (CCO) method for incident energies below the H ionization threshold. The target continuum and the positronium (Ps) formation channels are included in the coupled-channels calculations via a complex equivalent-local optical potential. The critical points, which show minima in the differential cross sections, as a function of the scattering angle and the incident energy are investigated. The resonances in the angular differential cross sections are reported for the first time in this energy range. The effects of the target continuum and the Ps formation channels on the different cross sections are discussed.
Systematics of (n,2n) Cross Sections
Institute of Scientific and Technical Information of China (English)
2008-01-01
<正>The experimental data of (n, 2n) cross sections were collected and evaluated as complete as possible. There are 640 sets of experimental data for 130 nuclei. The data were fitted to the expressions that describe the
Photoneutron cross section of 34S
International Nuclear Information System (INIS)
Using an enriched 34S target, the reaction 34S(γ,sn)33S has been measured from below threshold (10.4 MeV) to 28 MeV by directly counting the photoneutrons as a function of bremsstrahlung energy. The resultant cross section shows gross splitting in the GDR region. The integrated cross section is discussed in the light of the systematics of similar nuclei having two neutrons outside a doubly closed shell/sub-shell core
Photoneutron cross section of 34S
International Nuclear Information System (INIS)
Using an enriched 34S target, the reaction 34S(γ, sn) has been measured from below threshold (10.4 MeV) to 28 MeV by directly counting the photoneutrons as a function of bremsstrahlung energy. The resultant cross section shows gross splitting in the GDR region. The integrated cross section is discussed in the light of the systematics of similar nuclei having two neutrons outside a doubly closed shell/sub-shell core. (orig.)
Neutron capture cross sections from Surrogate measurements
Scielzo N.D.; Dietrich F.S.; Escher J.E.
2010-01-01
The prospects for determining cross sections for compound-nuclear neutron-capture reactions from Surrogate measurements are investigated. Calculations as well as experimental results are presented that test the Weisskopf-Ewing approximation, which is employed in most analyses of Surrogate data. It is concluded that, in general, one has to go beyond this approximation in order to obtain (n,γ) cross sections of sufficient accuracy for most astrophysical and nuclear-energy applications.
Neutron capture cross sections from Surrogate measurements
Directory of Open Access Journals (Sweden)
Scielzo N.D.
2010-03-01
Full Text Available The prospects for determining cross sections for compound-nuclear neutron-capture reactions from Surrogate measurements are investigated. Calculations as well as experimental results are presented that test the Weisskopf-Ewing approximation, which is employed in most analyses of Surrogate data. It is concluded that, in general, one has to go beyond this approximation in order to obtain (n,γ cross sections of sufficient accuracy for most astrophysical and nuclear-energy applications.
Evaluation methods for neutron cross section standards
International Nuclear Information System (INIS)
Methods used to evaluate the neutron cross section standards are reviewed and their relative merits, assessed. These include phase-shift analysis, R-matrix fit, and a number of other methods by Poenitz, Bhat, Kon'shin and the Bayesian or generalized least-squares procedures. The problems involved in adopting these methods for future cross section standards evaluations are considered, and the prospects for their use, discussed. 115 references, 5 figures, 3 tables
Photoproton cross section for 14C
International Nuclear Information System (INIS)
Using bremsstrahlung, the 14C(γ,p) reaction cross section has been measured from threshold to 29 MeV. The integrated cross section up to 30 MeV is 18±3 MeV mb. Above 23.5 MeV, the reported cross section includes a contribution, estimated at 3.5 MeV mb, due to the 14C(γ,d) and 14Cγ,pn) reactions. Essentially the entire 14C(γ,p) cross section results from decay of T> dipole states. From knowledge of other decay channels estimates of the cross section, integrated to 30 MeV for the T and T> components of the giant resonance (GDR) of 81 MeV mb and 43 MeV mb are obtained. The splitting of the mean energies of the GDR isospin components is 8.5 MeV. Comparisons with several shell-model calculations are made with the data, and general agreement is found. A comparison of photonuclear absorption cross sections for 12,1314C and 16,17,18 O shows dramatic redistribution of dipole strength as neutrons are added to the core nuclei. 41 refs., 1 tab., 7 figs
Reference solution for cross section parametrization
International Nuclear Information System (INIS)
Core calculations of nuclear reactors are usually performed by core physics codes (e.g. with NEM or FDM solvers) in diffusion or SP3 approximation of the transport equation. For each fuel type parameterized data libraries are prepared by means of a lattice code. The data libraries are burnup dependent, and the parameterization covers the hyperspace of admissible values of all operational parameters (fuel temperature, moderator density, boron concentration etc.) This approach has two weak spots. The first is, that it is difficult to make perfect parameterization of the data library because of relatively broad range of the parameter values and the fact that the parameters' effect on the macroscopic cross-sections are not mutually independent. The second is that even for perfect parameterizations with precise approximations of the data changes with respect to the feedback parameters the so-called history effects are neglected. It is generally difficult to assess the cumulative errors arising due to the approximative parameterization of the data libraries and due to the history effects. It is as well difficult to assess the efficiency of techniques developed in order to incorporate the history effect in the data library (such as time integration). In this paper we present a tool for reference core calculations in which the above stated approximations are eliminated. This paper presents the solution method, its implementation, as well as the results of a demonstration calculation showing the improvement of the calculation results over the traditional approach, assessing the magnitude of history and parameterization effects importance. The most important feature of the presented method is that it provides the perfect parameterization of macroscopic data, allowing the core physics code developers to understand sources of modeling uncertainties by completely removing the parameterization error (including, unlike other approaches, a complete representation of the
Penning ionization cross sections of excited rare gas atoms
International Nuclear Information System (INIS)
Electronic energy transfer processes involving excited rare gas atoms play one of the most important roles in ionized gas phenomena. Penning ionization is one of the well known electronic energy transfer processes and has been studied extensively both experimentally and theoretically. The present paper reports the deexcitation (Penning ionization) cross sections of metastable state helium He(23S) and radiative He(21P) atoms in collision with atoms and molecules, which have recently been obtained by the authors' group by using a pulse radiolysis method. Investigation is made of the selected deexcitation cross sections of He(23S) by atoms and molecules in the thermal collisional energy region. Results indicate that the cross sections are strongly dependent on the target molecule. The deexcitation probability of He(23S) per collision increases with the excess electronic energy of He(23S) above the ionization potential of the target atom or molecule. Another investigation, made on the deexcitation of He(21P), suggests that the deexcitation cross section for He(21P) by Ar is determined mainly by the Penning ionization cross section due to a dipole-dipole interaction. Penning ionization due to the dipole-dipole interaction is also important for deexcitation of He(21P) by the target molecules examined. (N.K.)
Elastic differential cross sections for electron collisions with polyatomic molecules
International Nuclear Information System (INIS)
Experimental data for electron-polyatomic molecule collisions are reviewed in connection with fusion and processing plasmas, as well as with the associated environmental issues. The electron scattering experiments for differential cross section (DCS) measurements for various processes, such as elastic scattering, have been performed across a broad range of energies (1-100 eV), mainly, at Sophia University since 1978, and some done under the collaborations with the Australian National University, Flinders University, and the Chungnam National University. As a benchmark cross section, elastic DCS are essential for the absolute scale conversion of inelastic DCS, as well as for testing computational methods. The need for cross-section data for a wide variety of molecular species is also discussed, because there is an urgent need to develop an international program to provide the scientific and technological communities with authoritative cross sections for electron-molecule interactions. Note that the detailed comparison with other data available is not given here. Ruther, other available data can be found in the references we cite. This course of action was adopted to keep this report to a sensible length, so that only our numerical data is provided here. (author)
P.J. van Genderen (P.); P.P.A.M. van Thiel (Pieter P. A.); P.G.H. Mulder (Paul); D. Overbosch (David)
2014-01-01
textabstractBackground Previous studies investigating the travellers' knowledge, attitudes and practices (KAP) profile indicated an important educational need among those travelling to risk destinations. Methods In the years 2002-2009 an annually repeated cross-sectional questionnaire-based survey w
Cross section inference based on PDE-constrained optimization
International Nuclear Information System (INIS)
The problem of inferring the material properties (cross section) in noninvasive inverse problems is formulated as a PDE-constrained optimization problem, where the governing laws of the chosen physics act as a constraint. A standard Lagrangian functional, containing the objective function to be minimized and the constraints to satisfy, is formed. The resolution of the optimality conditions lead to a nonlinear problem that is tackled with a Gauss-Newton procedure. Results of cross section inference are presented in the case of 1-group 2D neutron diffusion theory. (authors)
International Nuclear Information System (INIS)
It is well known that the temperature and background dependent neutron cross-sections are conventionally represented, in a problem-independent multigroup cross-section set, by specifying, for each group and reaction, the unshielded cross-section along with a set of self-shielding factors for various background cross-sections and temperatures. Usually the unshielded group cross-section is assumed to be independent of temperature. The observation presented in this paper, with examples, shows that the unshielded cross-section could significantly depend on temperature, depending on the group boundaries. (author)
Directory of Open Access Journals (Sweden)
Syed Hubbe
2012-04-01
Full Text Available ABSTRACT: CONTEXT : Obesity is increasing in the developed as well as developing countries. The prevalence of obesity is on the rise among the slum population. Increased incidence of vis ceral adiposity, hypertension, n on insulin dependent diabetes mellitus (NIDDM and coron ary heart disease often cluster in the same individual and there have been speculations that a common mechanism may be responsible for all these pathological conditions. This risk factor constellation, which is associated with an enhanced risk for cardiova scular disease, is referred to as “Syndrome X . AIMS : To assess the prevalence of diabetes and hypertension among obese and non obese in above 40 years age group in a slum area of Chennai. SETTINGS AND DESIGN : Urban slum in Chennai, Cross sectional study . MATERIALS AND METHODS : P r esent study was undertaken in a s lum in Chennai in persons above 4 0 years age group . One slum was selected randomly and the households in the slum were sampled by a systematic random sampling method. A pre - designed and pre - tested questionnaire was used to collect information regarding the socio - demographic profile, the diet pattern , the intake of non - vegetarian and oily foods , past history of hypertension and diabetes . Anthropometric data regarding height and weight was taken to assess body mass index (BMI , blood pressure was checked using mercury column sphygmomanometer and blood gluco se level b y G lucometer. STATISTICAL ANALYSIS : The prevalence was expressed in percentage and the Chi square test was used to find association with the factors. RESULTS : The prevalence of obesity was 13.66% and of overweight was 27.72%. The prevalence of Hy pertension among obese was 39.13%, pre obese 32.39% and non obese 24.93%. The prevalence of Dia betes among obese was 28.98%, pre obese 19.71% and non obese 15.34%. CONCLUSION : There is a rising prevalence of overweight and obesity among the urban slum dwellers. The prevalence of
Reduction Methods for Total Reaction Cross Sections
Gomes, P. R. S.; Mendes Junior, D. R.; Canto, L. F.; Lubian, J.; de Faria, P. N.
2016-03-01
The most frequently used methods to reduce fusion and total reaction excitation functions were investigated in a very recent paper Canto et al. (Phys Rev C 92:014626, 2015). These methods are widely used to eliminate the influence of masses and charges in comparisons of cross sections for weakly bound and tightly bound systems. This study reached two main conclusions. The first is that the fusion function method is the most successful procedure to reduce fusion cross sections. Applying this method to theoretical cross sections of single channel calculations, one obtains a system independent curve (the fusion function), that can be used as a benchmark to fusion data. The second conclusion was that none of the reduction methods available in the literature is able to provide a universal curve for total reaction cross sections. The reduced single channel cross sections keep a strong dependence of the atomic and mass numbers of the collision partners, except for systems in the same mass range. In the present work we pursue this problem further, applying the reduction methods to systems within a limited mass range. We show that, under these circumstances, the reduction of reaction data may be very useful.
Neutron cross section of methane hydrate
Energy Technology Data Exchange (ETDEWEB)
Kiyanagi, Y.; Date, S.; Horikawa, T.; Takamine, J.; Iwasa, H.; Kamiyama, T. [Graduate School of Eng., Hokkaido Univ., Sapporo (Japan); Uchida, T.; Ebinuma, T.; Narrita, H. [National Inst. of Advanced Industrial Science, Tsukisamu, Sapporo (Japan); Bennington, S.M. [ISIS Dept., Rutherford Appleton, Chilton, Didcot, Oxon (United Kingdom)
2004-03-01
To estimate the neutronic characteristics of methane hydrate and also to synthesize cross section data for simulation we need neutron scattering data ranging wide energy and momentum region. We performed inelastic neutron scattering experiments to get information about the neutron cross section on methane hydrate. It was found that at high momentum transfer region rotational mode as well as vibration mode showed recoil like behavior. On the other hand, at low momentum region, as well known, free rotation like energy levels were observed. The energy level of ice in methane hydrate was very similar to normal ice. The results suggest that the rough expression of the cross section of the methane hydrate is presented by linear combination of the methane and ice. (orig.)
Radiation pressure cross section for fluffy aggregates
International Nuclear Information System (INIS)
We apply the discrete dipole approximation (DDA) to estimate the radiation pressure cross section for fluffy aggregates by computing the asymmetry parameter and the cross sections for extinction and scattering. The ballistic particle-cluster aggregate and the ballistic cluster-cluster aggregate consisting of either dielectric or absorbing material are considered to represent naturally existing aggregates. We show that the asymmetry parameter perpendicular to the direction of wave propagation is maximized where the wavelength is comparable to the aggregate size, which may be characterized by the area-equivalent radius or the radius of gyration rather than the volume-equivalent radius. The asymmetry parameter for the aggregate depends on the morphology of the particle, but not on the constituent material. Therefore, the dependence of the radiation pressure cross section on the material composition arises mainly from that of the extinction and scattering cross sections, in other words, the single-scattering albedo. We find that aggregates consisting of high-albedo material show a large deviation of radiation pressure from the direction of incident radiation. When the aggregates are illuminated by blackbody radiation, the deviation of the radiation pressure increases with increasing temperature of the blackbody. Since the parallel component of the radiation pressure cross section for the aggregates is smaller than that for the volume-equivalent spheres at the size parameter close to unity, the Planck-mean radiation pressure cross section for the aggregates having radius comparable to the effective wavelength of radiation shows a lower value, compared with the volume-equivalent sphere. Consequently, the slope of the radiation pressure force per mass of the particle as a function of particle mass shows a lower maximum for the aggregates than for compact spherical particles. (Copyright (c) 1998 Elsevier Science B.V., Amsterdam. All rights reserved.)