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Sample records for multigroup x-sect library

  1. FENDL multigroup libraries

    International Nuclear Information System (INIS)

    Ganesan, S.; Muir, D.W.

    1992-01-01

    Selected neutron reaction nuclear data libraries and photon-atomic interaction cross section libraries for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into MATXSR format using the NJOY system on the VAX4000 computer of the IAEA. This document lists the resulting multigroup data libraries. All the multigroup data generated are available cost-free upon request from the IAEA Nuclear Data Section. (author). 9 refs

  2. Generating and verification of ACE-multigroup library for MCNP

    International Nuclear Information System (INIS)

    Chen Chaobin; Hu Zehua; Chen Yixue; Wu Jun; Yang Shouhai

    2012-01-01

    The Monte Carlo code MCNP can handle multigroup calculations and a sample multigroup set based on ENDF/B-V, MGXSNP, is available for MCNP for coupled neutron-photon transport. However, this library is not suit- able for all problems, and there is a need for users to be able to generate multigroup libraries tailored to their specific applications. For these purposes CSPT (cross section processing tool) is created to generate multigroup library for MCNP from deterministic multigroup cross sections (GENDF or ANISN format at present). Several ACE-multigroup libraries based on ENDF/B-VII.0 converted and verified in this work, we drawn the conclusion that the CSPT code works correctly and the libraries produced are credible. (authors)

  3. NDS multigroup cross section libraries

    International Nuclear Information System (INIS)

    DayDay, N.

    1981-12-01

    A summary description and documentation of the multigroup cross section libraries which exist at the IAEA Nuclear Data Section are given in this report. The libraries listed are available either on tape or in printed form. (author)

  4. Procedure to Generate the MPACT Multigroup Library

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-17

    The CASL neutronics simulator MPACT is under development for the neutronics and T-H coupled simulation for the light water reactor. The objective of this document is focused on reviewing the current procedure to generate the MPACT multigroup library. Detailed methodologies and procedures are included in this document for further discussion to improve the MPACT multigroup library.

  5. Multigroup cross section library; WIMS library

    International Nuclear Information System (INIS)

    Kannan, Umasankari

    2000-01-01

    The WIMS library has been extensively used in thermal reactor calculations. This multigroup constants library was originally developed from the UKNDL in the late 60's and has been updated in 1986. This library has been distributed with the WIMS-D code by NEA data bank. The references to WIMS library in literature are the 'old' which is the original as developed by the AEA Winfrith and the 'new' which is the current 1986 WIMS library. IAEA has organised a CRP where a new and fully updated WIMS library will soon be available. This paper gives an overview of the definitions of the group constants that go into any basic nuclear data library used for reactor calculations. This paper also outlines the contents of the WIMS library and some of its shortcomings

  6. Research of the application of multi-group libraries based on ENDF/B-VII library in the reactor design

    International Nuclear Information System (INIS)

    Mi Aijun; Li Junjie

    2010-01-01

    In this paper the multi-group libraries were constructed by processing ENDF/B-VII neutron incident files into multi-group structure, and the application of the multi-group libraries in the pressurized-water reactor(PWR) design was studied. The construction of the multi-group library is realized by using the NJOY nuclear data processing system. The code can process the neutron cross section files form ENDF format to MATXS format which was required in SN code. Two dimension transport theory code of discrete ordinates DORT was used to verify the multi-group libraries and the method of the construction by comparing calculations for some representative benchmarks. We made the PWR shielding calculation by using the multi-group libraries and studied the influence of the parameters involved during the construction of the libraries such as group structure, temperatures and weight functions on the shielding design of the PWR. This work is the preparation for the construction of the multi-group library which will be used in PWR shielding design in engineering. (authors)

  7. MT71x: Multi-Temperature Library Based on ENDF/B-VII.1

    Energy Technology Data Exchange (ETDEWEB)

    Conlin, Jeremy Lloyd [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parsons, Donald Kent [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gray, Mark Girard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lee, Mary Beth [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); White, Morgan Curtis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-12-16

    The Nuclear Data Team has released a multitemperature transport library, MT71x, based upon ENDF/B-VII.1 with a few modifications as well as additional evaluations for a total of 427 isotope tables. The library was processed using NJOY2012.39 into 23 temperatures. MT71x consists of two sub-libraries; MT71xMG for multigroup energy representation data and MT71xCE for continuous energy representation data. These sub-libraries are suitable for deterministic transport and Monte Carlo transport applications, respectively. The SZAs used are the same for the two sub-libraries; that is, the same SZA can be used for both libraries. This makes comparisons between the two libraries and between deterministic and Monte Carlo codes straightforward. Both the multigroup energy and continuous energy libraries were verified and validated with our checking codes checkmg and checkace (multigroup and continuous energy, respectively) Then an expanded suite of tests was used for additional verification and, finally, verified using an extensive suite of critical benchmark models. We feel that this library is suitable for all calculations and is particularly useful for calculations sensitive to temperature effects.

  8. SERKON program for compiling a multigroup library to be used in BETTY calculation

    International Nuclear Information System (INIS)

    Nguyen Phuoc Lan.

    1982-11-01

    A SERKON-type program was written to compile data sets generated by FEDGROUP-3 into a multigroup library for BETTY calculation. A multigroup library was generated from the ENDF/B-IV data file and tested against the TRX-1 and TRX-2 lattices with good results. (author)

  9. Verification and validation of multi-group library MUSE1.0 created from ENDF/B-VII.0

    International Nuclear Information System (INIS)

    Chen Yixue; Wu Jun; Yang Shouhai; Zhang Bin; Lu Daogang; Chen Chaobin

    2010-01-01

    A multi-group library set named MUSE1.0 with 172-neutron group and 42-photon group is produced based on ENDF/B-VII.0 using NJOY code. Weight function of the multi-group library set is taken from the Vitanim-e library and the max legendre order of scattering matrix is six. All the nuclides have thermal scattering data created using free-gas scattering law and 10 Bondarenko background cross sections se lected to generate the self-shielded multi-group cross sections. The final libraries have GENDF-format, MATXS-format and ACE-multi-group sub-libraries and each sub-library generated under 4 temperatures(293 K,600 K,800 K and 900 K). This paper provides a summary of the procedure to produce the library set and a detail description of the validation of the multi-group library set by several critical benchmark devices and shielding benchmark devices using MCNP code. The ability to handle the thermal neutron transport and resonance self-shielding problems are investigated specially. In the end, we draw the conclusion that the multi-group libraries produced is credible and can be used in the R and D process of Supercritical Water Reactor Design. (authors)

  10. The LAW Library -- A multigroup cross-section library for use in radioactive waste analysis calculations

    International Nuclear Information System (INIS)

    Greene, N.M.; Arwood, J.W.; Wright, R.Q.; Parks, C.V.

    1994-08-01

    The 238-group LAW Library is a new multigroup neutron cross-section library based on ENDF/B-V data, with five sets of data taken from ENDF/B-VI ( 14 N 7 , 15 N 7 , 16 O 8 , 154Eu 63 , and 155 Eu 63 ). These five nuclides are included because the new evaluations are thought to be superior to those in Version 5. The LAW Library contains data for over 300 materials and will be distributed by the Radiation Shielding Information Center, located at Oak Ridge National Laboratory. It was generated for use in neutronics calculations required in radioactive waste analyses, although it has equal utility in any study requiring multigroup neutron cross sections

  11. Multi-Group Library Generation with Explicit Resonance Interference Using Continuous Energy Monte Carlo Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ho Jin; Cho, Jin Young [KAERI, Daejeon (Korea, Republic of); Kim, Kang Seog [Oak Ridge National Laboratory, Oak Ridge (United States); Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    In this study, multi-group cross section libraries for the DeCART code were generated using a new procedure. The new procedure includes generating the RI tables based on the MC calculations, correcting the effective fission product yield calculations, and considering most of the fission products as resonant nuclides. KAERI (Korea Atomic Energy Research Institute) has developed the transport lattice code KARMA (Kernel Analyzer by Ray-tracing Method for fuel Assembly) and DeCART (Deterministic Core Analysis based on Ray Tracing) for a multi-group neutron transport analysis of light water reactors (LWRs). These codes adopt the method of characteristics (MOC) to solve the multi-group transport equation and resonance fixed source problem, the subgroup and the direct iteration method with resonance integral tables for resonance treatment. With the development of the DeCART and KARMA code, KAERI has established its own library generation system for a multi-group transport calculation. In the KAERI library generation system, the multi-group average cross section and resonance integral (RI) table are generated and edited using PENDF (point-wise ENDF) and GENDF (group-wise ENDF) produced by the NJOY code. The new method does not need additional processing because the MC method can handle any geometry information and material composition. In this study, the new method is applied to the dominant resonance nuclide such as U{sup 235} and U{sup 238} and the conventional method is applied to the minor resonance nuclides. To examine the newly generated multi-group cross section libraries, various benchmark calculations such as pin-cell, FA, and core depletion problem are performed and the results are compared with the reference solutions. Overall, the results by the new method agree well with the reference solution. The new procedure based on the MC method were verified and provided the multi-group library that can be used in the SMR nuclear design analysis.

  12. Development and testing of multigroup library with correction of self-shielding effects in fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Zou Jun; He Zhaozhong; Zeng Qin; Qiu Yuefeng; Wang Minghuang

    2010-01-01

    A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the K eff , neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.

  13. AMPX: a modular code system for generating coupled multigroup neutron-gamma libraries from ENDF/B

    Energy Technology Data Exchange (ETDEWEB)

    Greene, N.M.; Lucius, J.L.; Petrie, L.M.; Ford, W.E. III; White, J.E.; Wright, R.Q.

    1976-03-01

    AMPX is a modular system for producing coupled multigroup neutron-gamma cross section sets. Basic neutron and gamma cross-section data for AMPX are obtained from ENDF/B libraries. Most commonly used operations required to generate and collapse multigroup cross-section sets are provided in the system. AMPX is flexibly dimensioned; neutron group structures, and gamma group structures, and expansion orders to represent anisotropic processes are all arbitrary and limited only by available computer core and budget. The basic processes provided will (1) generate multigroup neutron cross sections; (2) generate multigroup gamma cross sections; (3) generate gamma yields for gamma-producing neutron interactions; (4) combine neutron cross sections, gamma cross sections, and gamma yields into final ''coupled sets''; (5) perform one-dimensional discrete ordinates transport or diffusion theory calculations for neutrons and gammas and, on option, collapse the cross sections to a broad-group structure, using the one-dimensional results as weighting functions; (6) plot cross sections, on option, to facilitate the ''evaluation'' of a particular multigroup set of data; (7) update and maintain multigroup cross section libraries in such a manner as to make it not only easy to combine new data with previously processed data but also to do it in a single pass on the computer; and (8) output multigroup cross sections in convenient formats for other codes. (auth)

  14. MENDF71x. Multigroup Neutron Cross Section Data Tables Based upon ENDF/B-VII.1

    International Nuclear Information System (INIS)

    Conlin, Jeremy Lloyd; Parsons, Donald Kent; Gardiner, Steven J.; Gray, Mark Girard; Lee, Mary Beth; White, Morgan Curtis

    2015-01-01

    A new multi-group neutron cross section library has been released along with the release of NDI version 2.0.20. The library is named MENDF71x and is based upon the evaluations released in ENDF/B-VII.1 which was made publicly available in December 2011. ENDF/B-VII.1 consists of 423 evaluations of which ten are excited states evaluations and 413 are ground state evaluations. MENDF71x was created by processing the 423 evaluations into 618-group, downscatter only NDI data tables. The ENDF/B evaluation files were processed using NJOY version 99.393 with the exception of 35 Cl and 233 U. Those two isotopes had unique properties that required that we process the evaluation using NJOY version 2012. The MENDF71x library was only processed to room temperature, i.e., 293.6 K. In the future, we plan on producing a multi-temperature library based on ENDF/B-VII.1 and compatible with MENDF71x.

  15. Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

    International Nuclear Information System (INIS)

    Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Moses, D.L.; Petrie, L.M.; Primm, R.T. III; Slater, C.O.; Westfall, R.M.; Wright, R.Q.

    1990-09-01

    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

  16. AMZ, multigroup constant library for EXPANDA code, generated by NJOY code from ENDF/B-IV

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Moraes, Marisa de

    1985-01-01

    It is described a library of multigroup constants with 70 energy groups and 37 isotopes to fast reactor calculation. The cross sections, scattering matrices and self-shielding factors were generated by NJOY code and RGENDF interface program, from ENDF/B-IV'S evaluated data. The library is edited in adequated format to be used by EXPANDA code. (M.C.K.) [pt

  17. A computer program with graphical user interface to plot the multigroup cross sections of WIMS-D library

    International Nuclear Information System (INIS)

    Thiyagarajan, T.K.; Ganesan, S.; Jagannathan, V.; Karthikeyan, R.

    2002-01-01

    As a result of the IAEA Co-ordinated Research Programme entitled 'Final Stage of the WIMS Library Update Project', new and updated WIMS-D libraries based upon ENDF/B-VI.5, JENDL-3.2 and JEF-2.2 have become available. A project to prepare an exhaustive handbook of WIMS-D cross sections from old and new libraries has been taken up by the authors. As part of this project, we have developed a computer program XnWlup with user-friendly graphical interface to help the users of WIMS-D library to enable quick visualization of the plots of the energy dependence of the multigroup cross sections of any nuclide of interest. This software enables the user to generate and view the histogram of 69 multi-group cross sections as a function of neutron energy under Microsoft Windows environment. This software is designed using Microsoft Visual C++ and Microsoft Foundation Classes Library. The current features of the software, on-line help manual and future plans for further development are described in this paper

  18. XNWLUP, Graphical user interface to plot WIMS-D library multigroup cross sections

    International Nuclear Information System (INIS)

    Ganesan, S.; Jagannathan, V.; Thiyagarajan, T.K.

    2005-01-01

    1 - Description of program or function: XnWlup is a computer program with user-friendly graphical interface to help the users of WIMS-D library to enable quick visualisation of the plots of the energy dependence of the multigroup cross sections of any nuclide of interest. This software enables the user to generate and view the histogram of 69 multi-group cross sections as a function of neutron energy under Microsoft Windows environment. This software is designed using Microsoft Visual C++ and Microsoft Foundation Classes Library. IAEA1395/05: New features of version 3.0: - Plotting absorption and fission cross sections of resonant nuclide after applying the self-shielding cross section. - Plotting the data of Resonant Integral table, as a function of dilution cross section for a selected temperature and for a given energy group. - Plotting the data of Resonant Integral table, as a function of temperature for a selected background dilution cross section and for a given energy group. - Clearing all the graphs except one graph from the display screen is easily done by using a tool bar button. - Displaying the coordinate of the cursor point with appropriate units. 2 - Methods: XnWlup helps to obtain histogram plots of the values of cross section data of an element/isotope available as 69-group WIMS-D library as a function of energy bins. The software XnWlup is developed with this graphical user interface in order to help those users who frequently refer to the WIMS-D library cross section data of neutron-nuclear reactions. The software also helps to produce handbook of WIMS-D cross sections

  19. ERRFILS: a preliminary library of 30-group multigroup covariance data for use in CTR sensitivity studies

    International Nuclear Information System (INIS)

    LaBauve, R.J.; Muir, D.W.

    1978-01-01

    A library of 30-group multigroup covariance data was prepared from preliminary ENDF/B-V data with the NJOY code. Data for Fe, Cr, Ni, 10 B, C, Cu, H, and Pb are included in this library. Reactions include total cross sections, elastic and inelastic scattering cross sections, and the most important absorption cross sections. Typical data from the file are shown. 3 tables

  20. ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies

    International Nuclear Information System (INIS)

    Ford, W.E. III; Arwood, J.W.; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Waddell, M.W.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.

    1987-01-01

    Multigroup P3 neutron, P0-P3 secondary gamma ray production (SGRP), and P6 gamma ray interaction (GRI) cross section libraries have been generated to support design work on the Advanced Neutron Source (ANS) reactor. The libraries, designated ANSL-V (Advanced Neutron Source Cross-Section Libraries), are data bases in a format suitable for subsequent generation of problem dependent cross sections. The ANSL-V libraries are available on magnetic tape from the Radiation Shielding Information Center at Oak Ridge National Laboratory

  1. PROF-DD, Generator of Multigroup Cross-Sections Library DDX for MORSE-DD, ANISN-DD, DOT-DD

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Ishiguro, Yukio

    2002-01-01

    1 - Description of program or function: The code system PROF-DD generates a multi-group double-differential cross section library DDX from evaluated data in ENDF/B-IV or ENDF/B-V format. The system consists of the following five modules: PROF-DDX is the main module of the system. It calculates the multigroup DDX and stores them on a master PDS file. MCFILEF generates a control file for PROF-DDX, which contains energy group and angle bin structures. SPINPTF prepares an input data file for PROF-DDX by combining the control file with other input data. DDXLIBMK edits a DDX library from the master PDS file for transport calculations. RESENDD performs resonance cross section and Doppler broadening calculations. 2 - Restrictions on the complexity of the problem: The numbers of energy groups and angle bins are less than 150 and 40, respectively

  2. Development of multi-group xs libraries for the gfr 2400 reactor

    International Nuclear Information System (INIS)

    Cerba, Š.; Vrban, B.; Lüley, J.; Necas, V.

    2016-01-01

    GFR 2400 is considered as a conceptual design of the large scale GEN IV Gas-Cooled Fast Reactor. In general, the GEN IV technologies are seen as reliable but also very challenging reactor concepts. Since GFR 2400 lacks any experimental data, the questions on its safety are even more complex and the assessment of its performance could be made only based on computational experience. The paper deals with the development process of multi-group XS libraries based on a hybrid deterministic-Stochastic methodology, using the NJOY99, TRANSX, DIF3D, PARTISN and MCNP5 codes. A new optimized 25 group SBJ E 71 2 5G cross section library was developed based on ENDF/B-VII.1 evaluated data, ZZ-KAFAX-E70 background cross sections and GFR 2400 neutron spectrum. The created library was validated through integral experiments evaluated on the HEX-Z deterministic models in DIF3D. The results were also compared with MCNP5 calculations. (authors)

  3. Assessment and comparison of different multigroup neutron cross section libraries for dosimetry purposes

    International Nuclear Information System (INIS)

    Erradi, L.; Karouani, K.

    1994-01-01

    Many multigroup neutron cross section libraries have been processed from basic evaluated nuclear data for use in neutron dosimetry, reactor shielding calculation and in the development of fusion reactors. Most of these libraries have been tested only for fission spectra and were not validated for fusion spectra. Fifteen of these libraries such as DOSCROS84, IRDF85 and ENDFB5 have been used along with the neutron spectra unfolding code SAND II to evaluate about fifteen threshold detector saturated activities. The comparison between these computed activities and the measured ones of a set of foils placed in different places along the axis of a paraffin cylinder and irradiated by 14 MeV neutrons generated by a D-T source, hence giving rise to complex spectra, leads to different types of discrepancies. The analysis of these discrepancies allows to select from these libraries the ones that can be recommended. 1 fig., 4 refs. (author)

  4. Recent validation experience with multigroup cross-section libraries and scale

    International Nuclear Information System (INIS)

    Bowman, S.M.; Wright, R.Q.; DeHart, M.D.; Parks, C.V.; Petrie, L.M.

    1995-01-01

    This paper will discuss the results obtained and lessons learned from an extensive validation of new ENDF/B-V and ENDF/B-VI multigroup cross-section libraries using analyses of critical experiments. The KENO V. a Monte Carlo code in version 4.3 of the SCALE computer code system was used to perform the critical benchmark calculations via the automated SCALE sequence CSAS25. The cross-section data were processed by the SCALE automated problem-dependent resonance-processing procedure included in this sequence. Prior to calling KENO V.a, CSAS25 accesses BONAMI to perform resonance self-shielding for nuclides with Bondarenko factors and NITAWL-II to process nuclides with resonance parameter data via the Nordheim Integral Treatment

  5. Nuclear data processing and multigroup cross section generation

    International Nuclear Information System (INIS)

    Kim, Jeong Do; Kil, Chung Sub

    1996-01-01

    The multigroup constants for WIMS/CASMO were updated with ENDF/B-VI and were tested. The continuous energy MCNP library developed last year was validated against the LWR-simulated critical experiments. The MCNP library will be used to design and analyze nuclear and shielding facilities. The system for generation of MATXS multigroup library and TRANSX code, which is able to prepare the data for the discrete ordinates and diffusion codes from the MATXS library, was established. The MATXS libraries for analyses of thermal and fast critical experiments were generated and tested. The MATXS/TRANSX system for the discrete ordinates and diffusion codes will be useful for nuclear analyses. 10 tabs., 5 figs., 17 refs. (Author)

  6. AMZ, library of multigroup constants for EXPANDA computer codes, generated by NJOY computer code from ENDF/B-IV

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Moraes, M. de.

    1984-01-01

    A 70-group, 37-isotope library of multigroup constants for fast reactor nuclear design calculations is described. Nuclear cross sections, transfer matrices, and self-shielding factors were generated with NJOY code and an auxiliary program RGENDF using evaluated data from ENDF/B-IV. The output is being issued in a format suitable for EXPANDA code. Comparisons with JFS-2 library, as well as, test resuls for 14 CSEWG benchmark critical assemblies are presented. (Author) [pt

  7. Use of CPXSD for generation of effective fast multigroup libraries for pressure vessel fluence calculations

    International Nuclear Information System (INIS)

    Alpan, F. Arzu; Haghighat, Alireza

    2008-01-01

    Multigroup (i.e., broad-group) libraries play a significant role in the accuracy of transport calculations. There are several broad-group libraries available for particular applications. For example the 47-neutron (26 fast groups), 20-gamma-group BUGLE libraries are commonly used for light water reactor shielding and pressure vessel dosimetry problems. However, there is no publicly available methodology to construct group structures for a problem and objective of interest. Therefore, we have developed the Contribution and Point-wise Cross-Section Driven (CPXSD) methodology, which constructs effective fine-and broad-group structures. In this paper, we use the CPXSD methodology to construct broad-group structures for fast neutron dosimetry problems. It is demonstrated that the broad-group libraries generated from CPXSD constructed group structures, while only 14 groups (rather than 26 groups) in the fast energy range are in good agreement (similar to 1 %-2 %) with the fine-group library from which they were derived, in reaction rate calculations.

  8. Achievement and qualification of multigroup cross-section library for light water reactor calculation

    International Nuclear Information System (INIS)

    Gastaldi, B.

    1986-07-01

    This study intends to improve then to check on integral experiments, the calculation of the main neutronic parameters in light water moderated lattices: Uranium 238 capture and consequently Plutonium 239 build-up, multiplication factor, temperature coefficient. The first part of this work concerns the resonant reaction rate calculation method implemented in the APOLLO code, the so-called LIVOLANT and JEANPIERRE formalism. The errors introduced by the corresponding assumptions are quantified and we propose substitution methods which avoid large biases and supply satisfactory results. The second part is dedicated to the cross-section evaluation of uranium major isotopes and to the achievement of APOLLO multigroup cross-sections. This cross-section set takes into considerations on the one hand the recent differential information and the other hand the various integral information obtained in the French Atomic Energy Commission facilities. The nuclear data file (JEF abd ENDF/B5) processing, for multigroup and self-shielded cross-sections achieving enable us to check the new THEMIS computer code. In the last part, the experimental validation of the proposed procedure (accurate formalism mutuel shielding and new multigroup library) is presented. This qualification is based on the reinterpretation of critical experiments performed in the EOLE reactor at Cadarache and spent fuel analysis. The corresponding results demonstrate that our propositions provide improvements on the computation of the PWR neutronic parameters; calculation-experiment discrepancies are now consistent with experimental uncertainty margins. 46 refs; 31 figs; 23 tabl [fr

  9. Generation and performance of a multigroup coupled neutron-gamma cross-section library for deterministic and Monte Carlo borehole logging analysis

    International Nuclear Information System (INIS)

    Kodeli, I.; Aldama, D. L.; De Leege, P. F. A.; Legrady, D.; Hoogenboom, J. E.; Cowan, P.

    2004-01-01

    As part of the IRTMBA (Improved Radiation Transport Modelling for Borehole Applications) project of the EU community's 5. framework program a special purpose multigroup cross-section library was prepared for use in deterministic and Monte Carlo oil well logging particle transport calculations. This library is expected to improve the prediction of the neutron and gamma spectra at the detector positions of the logging tool, and their use for the interpretation of the neutron logging measurements was studied. Preparation and testing of this library is described. (authors)

  10. Kalpakkam multigroup cross section set for fast reactor applications - status and performance

    International Nuclear Information System (INIS)

    Ramanadhan, M.M.; Gopalakrishnan, M.M.

    1986-01-01

    This report documents the status of the presently created set of multigroup constants at Kalpakkam. The list of nuclides processed and the details of multigroup structure are given. Also included are the particulars of dilutions and temperatures for each nuclide in the multigroup cross section set for which self shielding factors have been calculated. Using this new multigroup cross section set, measured integral quantities such as K-eff, central reaction rate ratios, central reactivity worths etc. were calculated for a few fast critical benchmark assemblies and the calculated values of neutronic parameters obtained were compared with those obtained using the available Cadarache cross section library and those published in literature for ENDF/B-IV based set and Japanese evaluated nuclear data library (JENDL). The details of analyses are documented along with the conclusions. (author). 17 refs., 12 tabs

  11. BUGLE-96: A revised multigroup cross section library for LWR applications based on ENDF/B-VI Release 3

    International Nuclear Information System (INIS)

    White, J.E.; Ingersoll, D.T.; Slater, C.O.; Roussin, R.W.

    1996-01-01

    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

  12. Nuclear data and multigroup methods in fast reactor calculations

    International Nuclear Information System (INIS)

    Gur, Y.

    1975-03-01

    The work deals with fast reactor multigroup calculations, and the efficient treatment of basic nuclear data, which serves as raw material for the calculations. Its purpose is twofold: to build a computer code system that handles a large, detailed library of basic neutron cross section data, (such as ENDF/B-III) and yields a compact set of multigroup cross sections for reactor calculations; to use the code system for comparative analysis of different libraries, in order to discover basic uncertainties that still exist in the measurement of neutron cross sections, and to determine their influence upon uncertainties in nuclear calculations. A program named NANICK which was written in two versions is presented. The first handles the American basic data library, ENDF/B-III, while the second handles the German basic data library, KEDAK. The mathematical algorithm is identical in both versions, and only the file management is different. This program calculates infinitely diluted multigroup cross sections and scattering matrices. It is complemented by the program NASIF that calculates shielding factors from resonance parameters. Different versions of NASIF were written to handle ENDF/B-III or KEDAK. New methods for evaluating in reactor calculations the long term behavior of the neutron flux as well as its fine structure are described and an efficient calculation of the shielding factors from resonance parameters is offered. (B.G.)

  13. Review of multigroup nuclear cross-section processing

    Energy Technology Data Exchange (ETDEWEB)

    Trubey, D.K.; Hendrickson, H.R. (comps.)

    1978-10-01

    These proceedings consist of 18 papers given at a seminar--workshop on ''Multigroup Nuclear Cross-Section Processing'' held at Oak Ridge, Tennessee, March 14--16, 1978. The papers describe various computer code systems and computing algorithms for producing multigroup neutron and gamma-ray cross sections from evaluated data, and experience with several reference data libraries. Separate abstracts were prepared for 13 of the papers. The remaining five have already been cited in ERA, and may be located by referring to the entry CONF-780334-- in the Report Number Index. (RWR)

  14. MINX: a multigroup interpretation of nuclear X-sections from ENDF/B

    International Nuclear Information System (INIS)

    Weisbin, C.R.; Soran, P.D.; MacFarlane, R.E.; Harris, D.R.; LaBauve, R.J.; Hendricks, J.S.; White, J.E.; Kidman, R.B.

    1976-09-01

    MINX calculates fine-group averaged infinitely dilute cross sections, self-shielding factors, and group-to-group transfer matrices from ENDF/B-IV data. Its primary purpose is to generate pseudo-composition independent multigroup libraries in the standard CCCC-III interface formats for use in the design and analysis of nuclear systems. MINX incorporates and improves upon the resonance capabilities of existing codes such as ETOX and ENDRUN and the high-Legendre-order transfer matrices of ETOG and SUPERTOG. Group structure, Legendre order, weight function, temperature, dilutions, and processing tolerances are all under user control. Paging and variable dimensioning allow very large problems to be run. Both CDC and IBM versions of MINX are available

  15. Calculation of multigroup constants in WIMS format with programs fedgroup and flange and comparison of the results obtained using different evaluated libraries

    International Nuclear Information System (INIS)

    Trkov, A.; Budnar, M.; Copic, M.; Perdan, A.; Ravnik, M.

    1982-01-01

    Multigroup constants for 1-H-1, 92-U-235, and 92-U-238 have been calculated. Averaged cross-sections and other constants have been prepared in the WIMS 69-group format. Comparison has been made between group constants obtained with several evaluated libraries (KEDAK-3 1975, 1979, ENDF/B-4, ENDF/B-5) and the WIMS-D library. Observed differences are most pronounced in the resonance and fast region. From test runs on fuel cell with the WIMS program it can be deduced that these differences affect the fewgroup constants significantly. (author)

  16. The LAW library

    International Nuclear Information System (INIS)

    Green, N.M.; Parks, C.V.; Arwood, J.W.

    1989-01-01

    The 238 group LAW library is a new multigroup library based on ENDF/B-V data. It contains data for 302 materials and will be distributed by the Radiation Shielding Information Center, located at Oak Ridge National Laboratory. It was generated for use in neutronics calculations required in radioactive waste analyses, though it has equal utility in any study requiring multigroup neutron cross sections

  17. Generation, Testing, and Validation of a WIMS-D/4 Multigroup Cross-Section Library Based on the JENDL-3.2 Nuclear Data

    International Nuclear Information System (INIS)

    Rahman, Mafizur; Takano, Hideki

    2001-01-01

    A new 69-group library of multigroup constants for the lattice code WIMS-D/4 has been generated with an improved resonance treatment, processing nuclear data from JENDL-3.2 by NJOY91.108. A parallel ENDF/B-VI based library has also been constructed for intercomparison of results. Benchmark calculations for a number of thermal reactor critical assemblies of both uranium and plutonium fuels have been performed with the code WIMS-D/4.1 with its three different libraries: the original WIMS library (NEA-0329/10) and the new ENDF/B-VI and JENDL-3.2 based libraries. The results calculated with both ENDF and JENDL based libraries show a similar tendency and are found in better agreement with the experimental values. Benchmark parameters are further calculated with the comprehensive lattice code SRAC95. The results from SRAC95 and WIMS-D/4.1 (both using JENDL-3.2 based libraries) agree well with each other. The new library is also verified for its applicability to mixed-oxide cores of varying plutonium contents

  18. FENDL/MG. Library of multigroup cross sections in GENDF and MATXS format for neutron-photon transport calculations. Version 1.1 of March 1995. Summary documentation

    International Nuclear Information System (INIS)

    Pashchenko, A.B.; Wienke, H.; Ganesan, S.

    1996-01-01

    Selected neutron reaction nuclear data evaluations and photon-atomic interaction cross section libraries for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into GENDF and MATXS format using the NJOY system by R.E. MacFarlane, in VITAMIN-J group structure with VITAMIN-E weighting spectrum. This document summarizes the resulting multigroup data library FENDL/MG version 1.1. The data are available costfree, upon request from the IAEA Nuclear Data Section, online or on magnetic tape. (author). 7 refs, 1 tab

  19. Status of multigroup cross-section data for shielding applications

    International Nuclear Information System (INIS)

    Roussin, R.W.; Maskewitz, B.F.; Trubey, D.K.

    1983-01-01

    Multigroup cross-section libraries for shielding applications in formats for direct use in discrete ordinates or Monte Carlo codes have long been a part of the Data Library Collection (DLC) of the Radiation Shielding Information Center (RSIC). In recent years libraries in more flexible and comprehensive formats, which allow the user to derive his own problem-dependent sets, have been added to the collection. The current status of both types is described, as well as projections for adding data libraries based on ENDF/B-V

  20. DySectAPI: Scalable Prescriptive Debugging

    DEFF Research Database (Denmark)

    Jensen, Nicklas Bo; Karlsson, Sven; Quarfot Nielsen, Niklas

    We present the DySectAPI, a tool that allow users to construct probe trees for automatic, event-driven debugging at scale. The traditional, interactive debugging model, whereby users manually step through and inspect their application, does not scale well even for current supercomputers. While...... lightweight debugging models scale well, they can currently only debug a subset of bug classes. DySectAPI fills the gap between these two approaches with a novel user-guided approach. Using both experimental results and analytical modeling we show how DySectAPI scales and can run with a low overhead...

  1. MPI version of NJOY and its application to multigroup cross-section generation

    Energy Technology Data Exchange (ETDEWEB)

    Alpan, A.; Haghighat, A.

    1999-07-01

    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

  2. MPI version of NJOY and its application to multigroup cross-section generation

    International Nuclear Information System (INIS)

    Alpan, A.; Haghighat, A.

    1999-01-01

    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

  3. ZZ ANSLV, Multigroup Cross Sections Library for ANS Reactor Design Studies

    International Nuclear Information System (INIS)

    2000-01-01

    A - Description of program or function: - Format: AMPX Master Interface Library format. Number of groups: Fine Group (99 energy groups) General Purpose Neutron Library. Materials: H, He, Be, B, Graphite, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Kr, Zr, Mo, Tc, Ru, Ag, Cd, Cs, Ce, Pr, Pm, Sm, Eu, Hf, Ta, U, C, F, Cu, Sn, Pb, Rh, I, Xe, Nd, Th, Np, Pu, Am, Cm, Bk, Cf, Es, MAFP, WAFP. Origin: ENDF/B-V. - Format: AMPX Master Interface Library format. Number of groups: Broad Group (39 energy groups) General Purpose Neutron Library. Materials: H, He, Be, B, Graphite, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Kr, Zr, Mo, Tc, Ru, Ag, Cd, Cs, Ce, Pr, Pm, Sm, Eu, Hf, Ta, U, C, F, Cu, Sn, Pb, Rh, I, Xe, Nd, Th, Np, Pu, Am, Cm, Bk, Cf, Es, MAFP, WAFP. Origin: ENDF/B-V. - Format: AMPX Master Interface Library format. Number of groups: Gamma-Ray Interaction (GRI) Library in 44-groups. Materials: H, He, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Xe, Sm, Eu, Hf, Ta, Ir, Pb, Th, U, Pu. Origin: ENDF/B-V; LENDL-V evaluations for 12 materials. - Format: AMPX Master Interface Library format. Number of groups: Coupled Library containing (CNG) 99-group neutron and 44-group gamma-ray data. Materials: H, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Eu, Hf, Ta, Pb, Th, U, Pu. Origin: ENDF/B-V. - Format: AMPX Master Interface Library format. Number of groups: Coupled neutron-gamma (CNG) Library containing 39-group, and 44-group gamma-ray data. Materials: H, Be, B, C, N, O, Na, Mg, Al, Si, K, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zr, Mo, Ag, Cd, Eu, Hf, Ta, Pb, Th, U, Pu. Origin: ENDF/B-V. Weighting spectrum: Maxwellian 300 K + 1/(E*sigma-total) + fission spectrum4 types of boundaries have been used depending isotope and library type (see report). Pseudo-problem-independent, multigroup cross section libraries were generated to support the Advanced Neutron source (ANS) reactor design studies. The ANS was

  4. Radiation Protection Ordinance 1989. Supplement with Radiation Protection Register Ordinance, general administration regulation pursuant to Sect. 45 Radiation Protection Ordinance, general administration regulation pursuant to Sect. 62 sub-sect. radiation passport

    International Nuclear Information System (INIS)

    Veith, H.M.

    1990-01-01

    The addendum contains regulations issued supplementary to the Radiation Protection Ordinance: The Radiation Protection Register as of April 3, 1990 including the law on the setting up of a Federal Office on Radiation Protection; the general administration regulation pursuant to Sect. 45 Radiation Protection Ordinance as of February 21, 1990; the general administration regulation pursuant to Sect. 62 sub-sect. 2 Radiation Protection Ordinance as of May 3, 1990 (AVV Radiation passport). The volume contains, apart from the legal texts, the appropriate decision by the Bundesrat, the official explanation from the Bundestag Publications as well as a comprehensive introduction into the new legal matter. (orig.) [de

  5. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Chiang, Min-Han; Wang, Jui-Yu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Sheu, Rong-Jiun, E-mail: rjsheu@mx.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Liu, Yen-Wan Hsueh [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China); Department of Engineering System and Science, National Tsing Hua University, 101, Section 2, Kung-Fu Road, Hsinchu 30013, Taiwan (China)

    2014-05-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects.

  6. Evaluation of the HTTR criticality and burnup calculations with continuous-energy and multigroup cross sections

    International Nuclear Information System (INIS)

    Chiang, Min-Han; Wang, Jui-Yu; Sheu, Rong-Jiun; Liu, Yen-Wan Hsueh

    2014-01-01

    The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several benchmark problems were repeated first to validate the calculation models. Careful code-to-code comparisons were made to ensure that two calculation models are both correct and equivalent. Compared with experimental data, the two models show a consistent bias of approximately 20–30 mk overestimation in effective multiplication factor for a wide range of core states. Most of the bias could be related to the ENDF/B-VII.0 cross-section library or incomplete modeling of impurities in graphite. After that, a series of systematic analyses was performed to investigate the effects of cross sections on the HTTR criticality and burnup calculations, with special interest in the comparison between continuous-energy and multigroup results. Multigroup calculations in this study were carried out in 238-group structure and adopted the SCALE double-heterogeneity treatment for resonance self-shielding. The results show that multigroup calculations tend to underestimate the system eigenvalue by a constant amount of ∼5 mk compared to their continuous-energy counterparts. Further sensitivity studies suggest the differences between multigroup and continuous-energy results appear to be temperature independent and also insensitive to burnup effects

  7. CLUB - a multigroup integral transport theory code for lattice calculations of PHWR cells

    International Nuclear Information System (INIS)

    Krishnani, P.D.

    1992-01-01

    The computer code CLUB has been developed to calculate lattice parameters as a function of burnup for a pressurised heavy water reactor (PHWR) lattice cell containing fuel in the form of cluster. It solves the multigroup integral transport equation by the method based on combination of small scale collision probability (CP) method and large scale interface current technique. The calculations are performed by using WIMS 69 group cross section library or its condensed versions of 27 or 28 group libraries. It can also compute Keff from the given geometrical buckling in the input using multigroup diffusion theory in fundamental mode. The first order differential burnup equations can be solved by either Trapezoidal rule or Runge-Kutta method. (author). 17 refs., 2 figs

  8. Sensitivity of 238U resonance absorption to library multigroup structure as calculated by WIMS-AECL

    International Nuclear Information System (INIS)

    Laughton, P.J.; Donnelly, J.V.

    1995-01-01

    In simulations of the TRX-1 experimental lattice, WIMS-AECL overpredicts, relative to MCNP, resonance absorption in neutron-energy groups containing the three large, low-lying resonances of 238 U when a standard ENDF/B-V-based library is used. A total excess in these groups of 4.0 neutron captures by 238 U per thousand fission neutrons has been observed. Similar comparisons are made in this work for the MIT-4 experimental lattice and simplified CANDU lattice cells containing 37-element fuel, with and without heavy-water coolant. Eleven different 89-group cross-section libraries were constructed for WIMS-AECL from ENDF/B-V data: only the neutron-energy-group boundaries used in generating multigroup cross sections and the Goldstein-Cohen correction factors differ from one library to the next. The first library uses the original 89-group structure, and the other ten involve energy groups of varying widths centred on the three large, low-lying resonances of 238 U. For TRX-1, some reduction in total discrepancy in 238 U capture can be achieved by using a new structure, although the improvement is small. The discrepancies in 238 U capture are of the same order for the MIT-4 case as those observed for TRX-1 for both the original group structure and the ten new structures. The WIMS-AECL calculation of 238 U resonance absorption in the same ranges of energy for the simplified CANDU 37-element lattice are in better agreement with MCNP than they are for TRX-1 and MIT-4: when the original structure is used, WIMS-AECL underpredicts total capture rate by 238 U in the energy range of interest by only 0.56 per thousand fission neutrons (coolant present) and 0.88 per thousand fission neutrons (voided coolant channel). The discrepancies are reduced when some of the new structures are used. For almost all of the cases considered here-TRX-1, MIT-4 and CANDU with coolant-better group-by-group agreement of 238 U capture around the 6.67-eV resonance is achieved by using a new library

  9. Solution of multi-group diffusion equation in x-y-z geometry by finite Fourier transformation

    International Nuclear Information System (INIS)

    Kobayashi, Keisuke

    1975-01-01

    The multi-group diffusion equation in three-dimensional x-y-z geometry is solved by finite Fourier transformation. Applying the Fourier transformation to a finite region with constant nuclear cross sections, the fluxes and currents at the material boundaries are obtained in terms of the Fourier series. Truncating the series after the first term, and assuming that the source term is piecewise linear within each mesh box, a set of coupled equations is obtained in the form of three-point equations for each coordinate. These equations can be easily solved by the alternative direction implicit method. Thus a practical procedure is established that could be applied to replace the currently used difference equation. This equation is used to solve the multi-group diffusion equation by means of the source iteration method; and sample calculations for thermal and fast reactors show that the present method yields accurate results with a smaller number of mesh points than the usual finite difference equations. (auth.)

  10. Program to solve the multigroup discrete ordinates transport equation in (x,y,z) geometry

    International Nuclear Information System (INIS)

    Lathrop, K.D.

    1976-04-01

    Numerical formulations and programming algorithms are given for the THREETRAN computer program which solves the discrete ordinates, multigroup transport equation in (x,y,z) geometry. An efficient, flexible, and general data-handling strategy is derived to make use of three hierarchies of storage: small core memory, large core memory, and disk file. Data management, input instructions, and sample problem output are described. A six-group, S 4 , 18 502 mesh point, 2 800 zone, k/sub eff/ calculation of the ZPPR-4 critical assembly required 144 min of CDC-7600 time to execute to a convergence tolerance of 5 x 10 -4 and gave results in good qualitative agreement with experiment and other calculations. 6 references

  11. RGENDF - An interface program between the NJOY code and codes using multigroup cross-sections

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Anaf, J.

    1988-02-01

    An interface program for reformatting multigroup cross-section libraries generated by NJOY into ENDF/B-V format and the EXPANDA, PFCOND and COMPAR input formats is presented. (author). 7 refs, 1 fig., 1 tab

  12. UFMGLIB: a multigroup library for the WIMS code

    International Nuclear Information System (INIS)

    Aboustta, Mohamed A.; Mello, Jair Carlos

    1995-01-01

    The British code WIMS is distributed by the NEA in its D4 version. It has a proper data library (Standard Library) in 69 energy groups covering, in its 1981 version, more than 100 materials with sufficient details. The Standard library was generated, basically, from the UKNDL data files and has been submitted to many alterations since the 1960's. Completely new versions such as the WIMKAL 88, generated from the basic library ENDF/B-V, were introduced and are available from the IAEA. The library UFMGLIB was generated from the ENDF/B-VI basic data library with some of the Standard data being maintained. The library contains evaluations for 131 different materials including the most common in thermal reactors. Results obtained from running the WIMS-D4 code with this library compare very well with results from the other libraries when running benchmark cases. This paper shows a general description of the library and some of the steps taken to generate it. (author). 12 refs, 2 tabs

  13. CTD: a computer program to solve the three dimensional multi-group diffusion equation in X, Y, Z, and triangular Z geometries

    Energy Technology Data Exchange (ETDEWEB)

    Fletcher, J K

    1973-05-01

    CTD is a computer program written in Fortran 4 to solve the multi-group diffusion theory equations in X, Y, Z and triangular Z geometries. A power print- out neutron balance and breeding gain are also produced. 4 references. (auth)

  14. VARI-QUIR-3, 2-D Multigroup Steady-State Neutron Diffusion in X-Y R-Z or R-Theta Geometry

    International Nuclear Information System (INIS)

    Collier, George

    1984-01-01

    1 - Nature of physical problem solved: The steady-state, multigroup, two-dimensional neutron diffusion equations are solved in x-y, r-z, and r-theta geometry. 2 - Method of solution: A Gauss-Seidel type of solution with inner and outer iterations is used. The source is held constant during the inner iterations

  15. MCFT: a program for calculating fast and thermal neutron multigroup constants

    International Nuclear Information System (INIS)

    Yang Shunhai; Sang Xinzeng

    1993-01-01

    MCFT is a program for calculating the fast and thermal neutron multigroup constants, which is redesigned from some codes for generation of thermal neutron multigroup constants and for fast neutron multigroup constants adapted on CYBER 825 computer. It uses indifferently as basic input with the evaluated nuclear data contained in the ENDF/B (US), KEDAK (Germany) and UK (United Kingdom) libraries. The code includes a section devoted to the generation of resonant Doppler broadened cross section in the framework of single-or multi-level Breit-Wigner formalism. The program can compute the thermal neutron scattering law S (α, β, T) as the input data in tabular, free gas or diffusion motion form. It can treat up to 200 energy groups and Legendre moments up to P 5 . The output consists of various reaction multigroup constants in all neutron energy range desired in the nuclear reactor design and calculation. Three options in input file can be used by the user. The output format is arbitrary and defined by user with a minimum of program modification. The program includes about 15,000 cards and 184 subroutines. FORTRAN 5 computer language is used. The operation system is under NOS 2 on computer CYBER 825

  16. The X-Shooter spectral library

    NARCIS (Netherlands)

    Chen, Y. P.; Trager, S. C.; Peletier, R. F.; Lançon, A.; Prugniel, Ph.; Koleva, M.

    2012-01-01

    We are building a new spectral library with the X-Shooter instrument on ESO's VLT: XSL, the X-Shooter Spectral Library. We present our progress in building XSL, which covers the wavelength range from the near-UV to the near-IR with a resolution of R˜10000. As of now we have collected spectra for

  17. Analysis of benchmark experiments for testing the IKE multigroup cross-section libraries based on ENDF/B-III and IV

    International Nuclear Information System (INIS)

    Keinert, J.; Mattes, M.

    1975-01-01

    Benchmark experiments offer the most direct method for validation of nuclear cross-section sets and calculational methods. For 16 fast and thermal critical assemblies containing uranium and/or plutonium of different compositions we compared our calculational results with measured integral quantities, such as ksub(eff), central reaction rate ratios or fast and thermal activation (dis)advantage factors. Cause of the simple calculational modelling of these assemblies the calculations proved as a good test for the IKE multigroup cross-section libraries essentially based on ENDF/B-IV. In general, our calculational results are in excellent agreement with the measured values. Only with some critical systems the basic ENDF/B-IV data proved to be insufficient in calculating ksub(eff), probably due to Pu neutron data and U 238 fast capture cross-sections. (orig.) [de

  18. Neutron cross section libraries for analysis of fusion neutronics experiments

    International Nuclear Information System (INIS)

    Kosako, Kazuaki; Oyama, Yukio; Maekawa, Hiroshi; Nakamura, Tomoo

    1988-03-01

    We have prepared two computer code systems producing neutron cross section libraries to analyse fusion neutronics experiments. First system produces the neutron cross section library in ANISN format, i.e., the multi-group constants in group independent format. This library can be obtained by using the multi-group constant processing code system MACS-N and the ANISN format cross section compiling code CROKAS. Second system is for the continuous energy cross section library for the MCNP code. This library can be obtained by the nuclear data processing system NJOY which generates pointwise energy cross sections and the cross section compiling code MACROS for the MCNP library. In this report, we describe the production procedures for both types of the cross section libraries, and show six libraries with different conditions in ANISN format and a library for the MCNP code. (author)

  19. Notes on Cantharellus sect. Leptocantharellus

    NARCIS (Netherlands)

    Donk, M.A.

    1969-01-01

    Cantharellus sect. Leptocantharellus Peck is an earlier name for Cantharellus subgen. Phaeocantharellus Corner. The European species fall apart in two groups (Lepto-Plicati and Lepto-Phlebini) on the basis of the hymenophoral configuration. Most of the older names provided in profusion for the few

  20. Multigroup and coupled forward-adjoint Monte Carlo calculation efficiencies for secondary neutron doses from proton beams

    International Nuclear Information System (INIS)

    Kelsey IV, Charles T.; Prinja, Anil K.

    2011-01-01

    We evaluate the Monte Carlo calculation efficiency for multigroup transport relative to continuous energy transport using the MCNPX code system to evaluate secondary neutron doses from a proton beam. We consider both fully forward simulation and application of a midway forward adjoint coupling method to the problem. Previously we developed tools for building coupled multigroup proton/neutron cross section libraries and showed consistent results for continuous energy and multigroup proton/neutron transport calculations. We observed that forward multigroup transport could be more efficient than continuous energy. Here we quantify solution efficiency differences for a secondary radiation dose problem characteristic of proton beam therapy problems. We begin by comparing figures of merit for forward multigroup and continuous energy MCNPX transport and find that multigroup is 30 times more efficient. Next we evaluate efficiency gains for coupling out-of-beam adjoint solutions with forward in-beam solutions. We use a variation of a midway forward-adjoint coupling method developed by others for neutral particle transport. Our implementation makes use of the surface source feature in MCNPX and we use spherical harmonic expansions for coupling in angle rather than solid angle binning. The adjoint out-of-beam transport for organs of concern in a phantom or patient can be coupled with numerous forward, continuous energy or multigroup, in-beam perturbations of a therapy beam line configuration. Out-of-beam dose solutions are provided without repeating out-of-beam transport. (author)

  1. JSD1000: multi-group cross section sets for shielding materials

    International Nuclear Information System (INIS)

    Yamano, Naoki

    1984-03-01

    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)

  2. MO-FG-CAMPUS-JeP1-04: Evaluating DECT Vs SECT Range Differences in Proton Therapy Using Clinical Data

    Energy Technology Data Exchange (ETDEWEB)

    Hudobivnik, N; Dedes, G; Parodi, K; Landry, G [Ludwig-Maximilians-Universitaet Muenchen, Garching (Germany); Schwarz, F; Johnson, T; Thieke, C; Belka, C; Sommer, W [LMU Munich, Munich (Germany); Tessonnier, T [Ludwig-Maximilians-Universitaet Muenchen, Garching (Germany); Universitaetsklinikum Heidelberg, Heidelberg (Germany); Verhaegen, F [Maastro Clinic, Maastricht (Netherlands)

    2016-06-15

    Purpose: Range uncertainty from X-ray CT number conversion to stopping power ratio (SPR) is one of the key factors limiting the potential of proton therapy. The large margins required for deep seated tumors degrade the organ sparing achievable with the technology. Of interest is the application of dual energy CT (DECT) to SPR estimation. In this planning study proton range differences between SECT and DECT have been quantified for brain cases. Methods: A last generation dual source DECT scanner was used to acquire SECT (150 kVp with Sn filtration) and DECT (additionally 90 kVp) scans of phantoms and 5 head trauma patients, acting as surrogate cancer patients. Phantom materials were characterized in terms of SPR in a particle beam to obtain reference values. IMPT treatment plans were generated on the basis of SECT and DECT SPR images for hypothetical brain tumors using a short and a long beam path. Range differences between SECT and DECT from plan recalculations were evaluated in beam-eye-view (BEV) by comparing the 80% isodose. Results: For the 18 phantom materials the SECT RMS SPR errors were 2.6% compared to 1.1% for DECT. Group median relative range differences between SECT and DECT plans were −1.0% for the short beam path over the 5 patients investigated in this study. For the long beam path the median difference was −1.4%. These relative range differences corresponded to −0.5 mm and −1.4 mm shifts respectively. Conclusion: This is the first study performing proton therapy treatment planning on DECT patient images. Important range differences of more than 1 mm were observed between SECT and DECT treatment plans, and DECT SPR accuracy was found superior on the basis of phantom measurements. While the patients investigated in this study did not have brain tumors, the findings we observed should apply to cancer patients. Deutsche Forschungsgemeinschaft (MAP); Bundesministerium fur Bildung und Forschung (01IB13001)

  3. ESELEM 4: a code for calculating fine neutron spectrum and multi-group cross sections in plate lattice

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Katsuragi, Satoru; Narita, Hideo.

    1976-07-01

    The multi-group treatment has been used in the design study of fast reactors and analysis of experiments at fast critical assemblies. The accuracy of the multi-group cross sections therefore affects strongly the results of these analyses. The ESELEM 4 code has been developed to produce multi-group cross sections with an advanced method from the nuclear data libraries used in the JAERI Fast set. ESELEM 4 solves integral transport equation by the collision probability method in plate lattice geometry to obtain the fine neutron spectrum. A typical fine group mesh width is 0.008 in lethargy unit. The multi-group cross sections are calculated by weighting the point data with the fine structure neutron flux. Some devices are applied to reduce computation time and computer core storage required for the calculation. The slowing down sources are calculated with the use of a recurrence formula derived for elastic and inelastic scattering. The broad group treatment is adopted above 2 MeV for dealing with both light any heavy elements. Also the resonance cross sections of heavy elements are represented in a broad group structure, for which we use the values of the JAERI Fast set. The library data are prepared by the PRESM code from ENDF/A type nuclear data files. The cross section data can be compactly stored in the fast computer core memory for saving the core storage and data processing time. The programme uses the variable dimensions to increase its flexibility. The users' guide for ESELEM 4 and PRESM is also presented in this report. (auth.)

  4. Revision of Dorstenia sect. Nothodorstenia (Moraceae)

    NARCIS (Netherlands)

    Berg, C.C.

    1978-01-01

    The African genus Craterogyne Lanjouw has to be united with Dorstenia L. Most species of Craterogyne can be included in Dorstenia sect. Nothodorstenia Engl. A revision of this section is presented. 5 species have been accepted, viz. D. elliptica Bureau, D. djettii J. L. Guillaumet, D. oligogyna

  5. Multigroup cross section collapsing optimization of a He-3 detector assembly model using deterministic transport techniques

    International Nuclear Information System (INIS)

    Huang, Mi; Yi, Ce; Manalo, Kevin L.; Sjoden, Glenn E.

    2011-01-01

    Multigroup optimization is performed on a neutron detector assembly to examine the validity of transport response in forward and adjoint modes. For SN transport simulations, we discuss the multigroup collapse of an 80 group library to 40, 30, and 16 groups, constructed from using the 3-D parallel PENTRAN and macroscopic cross section collapsing with YGROUP contribution weighting. The difference in using P_1 and P_3 Legendre order in scattering cross sections is investigated; also, associated forward and adjoint transport responses are calculated. We conclude that for the block analyzed, a 30 group cross section optimizes both computation time and accuracy relative to the 80 group transport calculations. (author)

  6. Correction of multigroup cross sections for resolved resonance interference in mixed absorbers

    International Nuclear Information System (INIS)

    Williams, M.L.

    1982-07-01

    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

  7. MARS-ORNL, Processing Program Collection for AMPX, CCCC, ANISN, DOT, MORSE Format Library. LINX, MINX Library Utility, Data Merge. BINX, MINX Utility and SPHINX Utility, BCD to BIN Library Conversion. CINX, MINX Utility and SPHINX Utility, Library Data Collapsing

    International Nuclear Information System (INIS)

    2001-01-01

    Description of problem or function: MARS-ORNL is a selection of computer codes for the generation of problem-dependent multigroup cross section libraries. They are selected modules from the AMPX-2 system for AMPX interface format libraries, LASL codes for CCCC interfaces, and processing codes for libraries to be used by ANISN, DOT, or MORSE codes. The codes in the collection are used in connection with the following DLC data libraries: ZZ-LIB-IV (DLC-0040), ZZ-VITAMIN-C (DLC-0041), VITAMIN-4C (DLC-0053), ZZ-CLEAR/42B (DLC-0042), ZZ-CSRL/43B (DLC-0043), and EPRMASTER (DLC-0052). The functions of these processing codes are briefly described: A. AMPX Modules: AIM: Converts AMPX Master Interface Files from EBCDIC to binary form and back. AJAX: Merges, collects, assembles, re-orders, joins, and copies selected nuclides from AMPX Master Interfaces. BONAMI: Accesses Bondarenko factors from an AMPX Master Library and performs resonance self-shielding calculations. CHOX: Produces a coupled interface library in AMPX format by combining neutron libraries (generated by module XLACS), gamma libraries (generated by module SMUG), and photon production libraries (generated by module LAPHNGAS). CHOXM: Combines self-shielding factors as generated by the code SPHINX (PSR-0129) and an infinite dilution neutron master interface (generated by XLACS) to generate a self-shielded neutron AMPX Interface File. The interface produced by CHOXM is an input to the NITAWL module of AMPX. CHOXM is a modified version of CHOX. COMAND: Collapses ANISN cross section libraries. DIAL: Produces edits from AMPX Master Interfaces. ICE-II: Accepts cross sections from an AMPX working library and produces mixed cross sections in four formats: (1) AMPX working library format; (2) ANISN format; (3) group-independent ANISN format; (4) Monte Carlo processed cross section library format. NITAWL: Produces self-shielded and working cross section libraries in the formats required by the ANISN, DOT, or MORSE codes

  8. Conception and development of an adaptive energy mesher for multigroup library generation of the transport codes

    International Nuclear Information System (INIS)

    Mosca, P.

    2009-12-01

    The deterministic transport codes solve the stationary Boltzmann equation in a discretized energy formalism called multigroup. The transformation of continuous data in a multigroup form is obtained by averaging the highly variable cross sections of the resonant isotopes with the solution of the self-shielding models and the remaining ones with the coarse energy spectrum of the reactor type. So far the error of such an approach could only be evaluated retrospectively. To remedy this, we studied in this thesis a set of methods to control a priori the accuracy and the cost of the multigroup transport computation. The energy mesh optimisation is achieved using a two step process: the creation of a reference mesh and its optimized condensation. In the first stage, by refining locally and globally the energy mesh, we seek, on a fine energy mesh with subgroup self-shielding, a solution equivalent to a reference solver (Monte Carlo or pointwise deterministic solver). In the second step, once fixed the number of groups, depending on the acceptable computational cost, and chosen the most appropriate self-shielding models to the reactor type, we look for the best bounds of the reference mesh minimizing reaction rate errors by the particle swarm optimization algorithm. This new approach allows us to define new meshes for fast reactors as accurate as the currently used ones, but with fewer groups. (author)

  9. Approximate albedo boundary conditions for energy multigroup X,Y-geometry discrete ordinates nuclear global calculations

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Davi J.M.; Nunes, Carlos E.A.; Alves Filho, Hermes; Barros, Ricardo C., E-mail: davijmsilva@yahoo.com.br, E-mail: ceanunes@yahoo.com.br, E-mail: rcbarros@pq.cnpq.br [Secretaria Municipal de Educacao de Itaborai, RJ (Brazil); Universidade Estacio de Sa (UNESA), Rio de Janeiro, RJ (Brazil); Universidade do Estado do Rio de Janeiro (UERJ), Novra Friburgo, RJ (Brazil). Instituto Politecnico. Departamento de Modelagem Computacional

    2017-11-01

    Discussed here is the accuracy of approximate albedo boundary conditions for energy multigroup discrete ordinates (S{sub N}) eigenvalue problems in two-dimensional rectangular geometry for criticality calculations in neutron fission reacting systems, such as nuclear reactors. The multigroup (S{sub N}) albedo matrix substitutes approximately the non-multiplying media around the core, e.g., baffle and reflector, as we neglect the transverse leakage terms within these non-multiplying regions. Numerical results to a typical model problem are given to illustrate the accuracy versus the computer running time. (author)

  10. Status of multigroup sensitivity profiles and covariance matrices available from the radiation shielding information center

    International Nuclear Information System (INIS)

    Roussin, R.W.; Drischler, J.D.; Marable, J.H.

    1980-01-01

    In recent years multigroup sensitivity profiles and covariance matrices have been added to the Radiation Shielding Information Center's Data Library Collection (DLC). Sensitivity profiles are available in a single package. DLC-45/SENPRO, and covariance matrices are found in two packages, DLC-44/COVERX and DLC-77/COVERV. The contents of these packages are described and their availability is discussed

  11. MC2-3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Yang, W. S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2013-11-08

    The MC2-3 code is a Multigroup Cross section generation Code for fast reactor analysis, developed by improving the resonance self-shielding and spectrum calculation methods of MC2-2 and integrating the one-dimensional cell calculation capabilities of SDX. The code solves the consistent P1 multigroup transport equation using basic neutron data from ENDF/B data files to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (~2000) or hyperfine (~400,000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified isotopic temperatures. The pointwise cross sections are directly used in the hyperfine group calculation whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for two-dimensional whole-core problems to generate region-dependent broad-group cross sections. Multigroup cross sections are written in the ISOTXS format for a user-specified group structure. The code is executable on UNIX, Linux, and PC Windows systems, and its library includes all isotopes of the ENDF/BVII. 0 data.

  12. Nuclear libraries for SCALE5.1 system

    International Nuclear Information System (INIS)

    Vertes, P.

    2009-01-01

    Codes for preparing master and working AMPX libraries and point-wise nuclear libraries for SCALE5.1 system have been created. Master and working libraries are constructed from multigroup library in matxs form which are produced by means of the NJOY code. The point-wise cross-section library is derived from pend files obtained also by NJOY. The AMPX libraries may contain neutron, gamma production and gamma transport data, as well. The produced master libraries can be used either with stand-alone functional modules or with control modules. An assistant package of programs also has been developed in order to facilitate the usage of NJOY. (Authors)

  13. Nuclear libraries for SCALE5.1 system

    International Nuclear Information System (INIS)

    Vertes, P.

    2009-01-01

    Codes for preparing master and working AMPX libraries and point-wise (PW) nuclear libraries for SCALE5.1 system have been created. Master and working libraries are constructed from multigroup library in matxs form which are produced by means of the NJOY code. The PW cross-section library is derived from pend files obtained also by NJOY. The AMPX libraries may contain neutron, gamma production and gamma transport data, as well. The produced master libraries can be used either with stand-alone functional modules or with control modules. An assistant package of programs also has been developed in order to facilitate the usage of NJOY. (author)

  14. Solution of multigroup transport equation in x-y-z geometry by the spherical harmonics method using finite Fourier transformation

    International Nuclear Information System (INIS)

    Kobayashi, Keisuke; Kikuchi, Hirohiko; Tsutsuguchi, Ken

    1993-01-01

    A neutron multigroup transport equation in x-y-z geometry is solved by the spherical harmonics method using finite Fourier transformation. Using the first term of the Fourier series for the space variables of spherical harmonics moments, three-point finite difference like equations are derived for x-, y- and z-axis directions, which are more consistent and accurate than those derived using the usual finite difference approximation, and these equations are solved by the iteration method in each axis direction alternatively. A method to find an optimum acceleration factor for this inner iteration is described. It is shown in the numerical examples that the present method gives higher accuracy with less mesh points that the usual finite difference method. (author)

  15. Requests on domestic nuclear data library from BWR design

    International Nuclear Information System (INIS)

    Maruyama, Hiromi

    2003-01-01

    Requests on the domestic nuclear data library JENDL and activities of the Nuclear Data Center have been presented from the perspective of BWR design and design code development. The requests include a standard multi-group cross section library, technical supports, and clarification of advantage of JENDL as well as requests from physical aspects. (author)

  16. WIMS-D library update

    International Nuclear Information System (INIS)

    2007-05-01

    WIMS-D (Winfrith Improved Multigroup Scheme-D) is the name of a family of software packages for reactor lattice calculations and is one of the few reactor lattice codes in the public domain and available on noncommercial terms. WIMSD-5B has recently been released from the OECD Nuclear Energy Agency Data Bank, and features major improvements in machine portability, as well as incorporating a few minor corrections. This version supersedes WIMS-D/4, which was released by the Winfrith Technology Centre in the United Kingdom for IBM machines and has been adapted for various other computer platforms in different laboratories. The main weakness of the WIMS-D package is the multigroup constants library, which is based on very old data. The relatively good performance of WIMS-D is attributed to a series of empirical adjustments to the multigroup data. However, the adjustments are not always justified on the basis of more accurate and recent experimental measurements. Following the release of new and revised evaluated nuclear data files, it was felt that the performance of WIMS-D could be improved by updating the associated library. The WIMS-D Library Update Project (WLUP) was initiated in the early 1990s with the support of the IAEA. This project consisted of voluntary contributions from a large number of participants. Several benchmarks for testing the library were identified and analysed, the WIMSR module of the NJOY code system was upgraded and the author of NJOY accepted the proposed updates for the official code system distribution. A detailed parametric study was performed to investigate the effects of various data processing input options on the integral results. In addition, the data processing methods for the main reactor materials were optimized. Several partially updated libraries were produced for testing purposes. The final stage of the WLUP was organized as a coordinated research project (CRP) in order to speed up completion of the fully updated library

  17. Cassandre : a two-dimensional multigroup diffusion code for reactor transient analysis

    International Nuclear Information System (INIS)

    Arien, B.; Daniels, J.

    1986-12-01

    CASSANDRE is a two-dimensional (x-y or r-z) finite element neutronics code with thermohydraulics feedback for reactor dynamics prior to the disassembly phase. It uses the multigroup neutron diffusion theory. Its main characteristics are the use of a generalized quasistatic model, the use of a flexible multigroup point-kinetics algorithm allowing for spectral matching and the use of a finite element description. The code was conceived in order to be coupled with any thermohydraulics module, although thermohydraulics feedback is only considered in r-z geometry. In steady state criticality search is possible either by control rod insertion or by homogeneous poisoning of the coolant. This report describes the main characterstics of the code structure and provides all the information needed to use the code. (Author)

  18. Prioritizing Identities: Cross Categorization of Ethnicity and Religious Sects in Turkey Identities: Ethnicity and/or Religion

    Directory of Open Access Journals (Sweden)

    Goklem Tekdemir Yurtdas

    2015-12-01

    Full Text Available This study aims to investigate how different ethnic groups (Turkish/Kurdish and religious sects (Alevi/Sunni are perceived in Turkey. These groups have a long history of conflicts. In order to examine the perception of these conflicting group identities, we adopted the theoretical frameworks of simple and cross categorization developed by the Social Identity Theory and the Category Differentiation Model. Both theories converge on the idea of in-group favoritism in the case of simple categorization while they offer different explanations for cross categorization condition. In order to test these differing theoretical propositions, we asked our participants to evaluate simple and cross categorization conditions based on variables of ethnicity and sect.  Our sample consisted of 106 individuals from two ethnic groups (Kurdish/Turkish and two religious sects (Alevi/Sunni. The participants completed a questionnaire based on Zavalloni’s focused introspection technique. Within subjects repeated measures ANOVA analysis were carried out for both simple and cross categorization. The results of the analysis revealed that the participants emphasized in-group similarities and out-group differences both for ethnicity and religious sect in the simple categorization condition. Moreover, in-group similarities based on religious sects rather than ethnicity were prioritized in the cross categorization condition. The results confirmed the Social Identity Theory’s assumptions generally. Results were discussed in terms of relevant literature, and in relation to historical and political issues regarding ethnicity and sects in Turkey.

  19. SU-E-J-136: Investigation Into Robustness of Stopping Power Calculated by DECT and SECT for Proton Therapy Treatment Planning

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, J [University of Adelaide, Adelaide, SA (Australia); Penfold, S [University of Adelaide, Adelaide, SA (Australia); Royal Adelaide Hospital, Adelaide, SA (Australia)

    2015-06-15

    Purpose: To investigate the robustness of dual energy CT (DECT) and single energy CT (SECT) proton stopping power calibration techniques and quantify the associated errors when imaging a phantom differing in chemical composition to that used during stopping power calibration. Methods: The CIRS tissue substitute phantom was scanned in a CT-simulator at 90kV and 140kV. This image set was used to generate a DECT proton SPR calibration based on a relationship between effective atomic number and mean excitation energy. A SECT proton SPR calibration based only on Hounsfield units (HUs) was also generated. DECT and SECT scans of a second phantom of known density and chemical composition were performed. The SPR of the second phantom was calculated with the DECT approach (SPR-DECT),the SECT approach (SPR-SECT) and finally the known density and chemical composition of the phantom (SPR-ref). The DECT and SECT image sets were imported into the Pinnacle{sup 3} research release of proton therapy treatment planning. The difference in dose when exposed to a common pencil beam distribution was investigated. Results: SPR-DECT was found to be in better agreement with SPR-ref than SPR- SECT. The mean difference in SPR for all materials was 0.51% for DECT and 6.89% for SECT. With the exception of Teflon, SPR-DECT was found to agree with SPR-ref to within 1%. Significant differences in calculated dose were found when using the DECT image set or the SECT image set. Conclusion: The DECT calibration technique was found to be more robust to situations in which the physical properties of the test materials differed from the materials used during SPR calibration. Furthermore, it was demonstrated that the DECT and SECT SPR calibration techniques can Result in significantly different calculated dose distributions.

  20. ZZ COVFILS, 30-Group Covariance Library from ENDF/B-5 for Sensitivity Studies

    International Nuclear Information System (INIS)

    Muir, D.W.

    1997-01-01

    1 - Description of program or function: Format: ENDB/F; Number of groups: 30-Group Covariance Library; Nuclides: H-1, B-10, C, O-16, Cr, Fe, Ni, Cu, Pb. Origin: ENDF/B-V. COVFILS is a 30-Group Covariance Library. It contains neutron cross sections, and their uncertainties and correlation in multigroup form. These data can be used, in conjunction with sensitivity information, to estimate the data-related uncertainty in calculated integral quantities such as radiation-damage or heating. 2 - Method of solution: COVFILS was obtained by processing evaluations from ENDF/B-V with ERRORR module of the NJOY nuclear data processing system (LA-9303-M, Vols. 1).The group structure is the Los Alamos 30-group structure which is listed in 'File 1' of each multigroup data set in the library

  1. Production and testing of the ENEA-Bologna VITJEFF32.BOLIB (JEFF-3.2) multi-group (199 n + 42 γ) cross section library in AMPX format for nuclear fission applications

    Science.gov (United States)

    Pescarini, Massimo; Orsi, Roberto; Frisoni, Manuela

    2017-09-01

    The ENEA-Bologna Nuclear Data Group produced the VITJEFF32.BOLIB multi-group coupled neutron/photon (199 n + 42 γ) cross section library in AMPX format, based on the OECD-NEA Data Bank JEFF-3.2 evaluated nuclear data library. VITJEFF32.BOLIB was conceived for nuclear fission applications as European counterpart of the ORNL VITAMIN-B7 similar library (ENDF/B-VII.0 data). VITJEFF32.BOLIB has the same neutron and photon energy group structure as the former ORNL VITAMIN-B6 reference library (ENDF/B-VI.3 data) and was produced using similar data processing methodologies, based on the LANL NJOY-2012.53 nuclear data processing system for the generation of the nuclide cross section data files in GENDF format. Then the ENEA-Bologna 2007 Revision of the ORNL SCAMPI nuclear data processing system was used for the conversion into the AMPX format. VITJEFF32.BOLIB contains processed cross section data files for 190 nuclides, obtained through the Bondarenko (f-factor) method for the treatment of neutron resonance self-shielding and temperature effects. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the ORNL DOORS system, can be generated from VITJEFF32.BOLIB through the cited SCAMPI version. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF32.BOLIB validation.

  2. Production and testing of the ENEA-Bologna VITJEFF32.BOLIB (JEFF-3.2 multi-group (199 n + 42 γ cross section library in AMPX format for nuclear fission applications

    Directory of Open Access Journals (Sweden)

    Pescarini Massimo

    2017-01-01

    Full Text Available The ENEA-Bologna Nuclear Data Group produced the VITJEFF32.BOLIB multi-group coupled neutron/photon (199 n + 42 γ cross section library in AMPX format, based on the OECD-NEA Data Bank JEFF-3.2 evaluated nuclear data library. VITJEFF32.BOLIB was conceived for nuclear fission applications as European counterpart of the ORNL VITAMIN-B7 similar library (ENDF/B-VII.0 data. VITJEFF32.BOLIB has the same neutron and photon energy group structure as the former ORNL VITAMIN-B6 reference library (ENDF/B-VI.3 data and was produced using similar data processing methodologies, based on the LANL NJOY-2012.53 nuclear data processing system for the generation of the nuclide cross section data files in GENDF format. Then the ENEA-Bologna 2007 Revision of the ORNL SCAMPI nuclear data processing system was used for the conversion into the AMPX format. VITJEFF32.BOLIB contains processed cross section data files for 190 nuclides, obtained through the Bondarenko (f-factor method for the treatment of neutron resonance self-shielding and temperature effects. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the ORNL DOORS system, can be generated from VITJEFF32.BOLIB through the cited SCAMPI version. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF32.BOLIB validation.

  3. Three-dimensional h-adaptivity for the multigroup neutron diffusion equations

    KAUST Repository

    Wang, Yaqi

    2009-04-01

    Adaptive mesh refinement (AMR) has been shown to allow solving partial differential equations to significantly higher accuracy at reduced numerical cost. This paper presents a state-of-the-art AMR algorithm applied to the multigroup neutron diffusion equation for reactor applications. In order to follow the physics closely, energy group-dependent meshes are employed. We present a novel algorithm for assembling the terms coupling shape functions from different meshes and show how it can be made efficient by deriving all meshes from a common coarse mesh by hierarchic refinement. Our methods are formulated using conforming finite elements of any order, for any number of energy groups. The spatial error distribution is assessed with a generalization of an error estimator originally derived for the Poisson equation. Our implementation of this algorithm is based on the widely used Open Source adaptive finite element library deal.II and is made available as part of this library\\'s extensively documented tutorial. We illustrate our methods with results for 2-D and 3-D reactor simulations using 2 and 7 energy groups, and using conforming finite elements of polynomial degree up to 6. © 2008 Elsevier Ltd. All rights reserved.

  4. On the convergence of multigroup discrete-ordinates approximations

    International Nuclear Information System (INIS)

    Victory, H.D. Jr.; Allen, E.J.; Ganguly, K.

    1987-01-01

    Our analysis is divided into two distinct parts which we label for convenience as Part A and Part B. In Part A, we demonstrate that the multigroup discrete-ordinates approximations are well-defined and converge to the exact transport solution in any subcritical setting. For the most part, we focus on transport in two-dimensional Cartesian geometry. A Nystroem technique is used to extend the discrete ordinates multigroup approximates to all values of the angular and energy variables. Such an extension enables us to employ collectively compact operator theory to deduce stability and convergence of the approximates. In Part B, we perform a thorough convergence analysis for the multigroup discrete-ordinates method for an anisotropically-scattering subcritical medium in slab geometry. The diamond-difference and step-characteristic spatial approximation methods are each studied. The multigroup neutron fluxes are shown to converge in a Banach space setting under realistic smoothness conditions on the solution. This is the first thorough convergence analysis for the fully-discretized multigroup neutron transport equations

  5. Introduction of corrections taking into account interdependence of multigroup constants to the results of multigroup perturbation theory calculations

    International Nuclear Information System (INIS)

    Raskach, K. F.

    2012-01-01

    In multigroup calculations of reactivity and sensitivity coefficients, methodical errors can appear if the interdependence of multigroup constants is not taken into account. For this effect to be taken into account, so-called implicit components of the aforementioned values are introduced. A simple technique for computing these values is proposed. It is based on the use of subgroup parameters.

  6. Multigroup P8 - elastic scattering matrices of main reactor elements

    International Nuclear Information System (INIS)

    Garg, S.B.; Shukla, V.K.

    1979-01-01

    To study the effect of anisotropic scattering phenomenon on shielding and neutronics of nuclear reactors multigroup P8-elastic scattering matrices have been generated for H, D, He, 6 Li, 7 Li, 10 B, C, N, O, Na, Cr, Fe, Ni, 233 U, 235 U, 238 U, 239 Pu, 240 Pu, 241 Pu and 242 Pu using their angular distribution, Legendre coefficient and elastic scattering cross-section data from the basic ENDF/B library. Two computer codes HSCAT and TRANS have been developed to complete this task for BESM-6 and CDC-3600 computers. These scattering matrices can be directly used as input to the transport theory codes ANISN and DOT. (auth.)

  7. MCNP and MATXS cross section libraries based on JENDL-3.3

    International Nuclear Information System (INIS)

    Kosako, Kazuaki; Konno, Chikara; Fukahori, Tokio; Shibata, Keiichi

    2003-01-01

    The continuous energy cross section library for the Monte Carlo transport code MCNP-4C, FSXLIB-J33, has been generated from the latest version of JENDL-3.3. The multigroup cross section library with the MATXS format, MATXS-J33, has been generated also from JENDL-3.3. Both libraries contain all nuclides in JENDL-3.3 and are processed at 300 K by the nuclear data processing system NJOY99. (author)

  8. Preparation of multigroup lumped fission product cross-sections from ENDF/B-VI for FBRs

    International Nuclear Information System (INIS)

    Devan, K.; Gopalakrishnan, V.; Mohanakrishnan, P.; Sridharan, M.S.

    1997-01-01

    Multigroup pseudo fission product cross-sections were computed from the American evaluated nuclear data library ENDF/B-VI, corresponding to various burnups of the proposed 500 MWe prototype fast breeder reactor (PFBR), in India. The data were derived from the cross-sections of 111 selected fission products that account for almost complete capture of fission products in an FBR. The dependence of burnup on the pseudo fission product cross-sections, and comparison with other data sets, viz. JNDC, ENDF/B-IV and ABBN, are discussed. (author)

  9. FCXSEC: multigroup cross-section libraries for nuclear fuel cycle shielding calculations

    International Nuclear Information System (INIS)

    Ford, W.E. III; Webster, C.C.; Diggs, B.R.; Pevey, R.E.; Croff, A.G.

    1980-05-01

    Starting with the pseudo-composition-independent VITAMIN-C cross-sectin library, composition-dependent fine-(171n-36γ) and broad-group (22n-21γ) self-shielded AMPX master, broad-group microscopic ANISN-formatted, and broad-group macroscopic ANISN-formatted cross-section libraries were generated to be used for nuclear fuel cycle shielding calculations. The specifications for the data and the procedure used to prepare the libraries are described

  10. KAFAX-F22 : development and benchmark of multi-group library for fast reactor using JEF-2.2. Neutron 80 group and Photon 24 group

    International Nuclear Information System (INIS)

    Kim, Jung Do; Gil, Choong Sup.

    1997-03-01

    The KAFAX-F22 was developed from JEF-2.2, which is a MATXS format, multigroup library of fast reactor. The KAFAX-F22 has 80 and 24 energy group structures for neutron and photon, respectively. It includes 89 nuclide data processed by NJOY94.38. The TRANSX/TWODANT system was used for benchmark calculations of fast reactor and one- and two-dimensional calculations of ONEDANT and TWODANT were carried out with 80 group, P 3 S 16 and with 25 group, P 3 S 8 , respectively. The average values of multiplication factors are 0.99652 for MOX cores, 1.00538 for uranium cores and 1.00032 for total cores. Various central reaction rate ratios also give good agreements with the experimental values considering experimental uncertainties except for VERA-11A, VERA-1B, ZPR-6-7 and ZPR-6-6A cores of which experimental values seem to involve some problems. (author). 13 refs., 18 tabs., 2 figs

  11. The significant event compilation tree-sect: Theory and application

    International Nuclear Information System (INIS)

    Ishack, G.A.

    1990-01-01

    The Significant Event Compilation Tree (SECT) is a computer programme that was developed by staff of the Canadian Atomic Energy Control Board during the period 1984-86. Its primary purpose is to link seemingly unrelated events, or parts of events, that could have occurred at different points in time at various nuclear power plants. Using such a software tool aids in the identification of potential paths and/or scenarios that: a. may not have been foreseen in the accident analysis (including fault tree verification), b. could lead to a certain failure; or c. could have been caused by a certain initiating event (which may have ended or been terminated at an earlier stage). This paper describes: a. the basic idea of SECT; b. the criteria whereby events are selected and coded; c. the options available to the user; d. an example of the programme's application in Canada; and e. a demonstration of its possible use in conjunction with the NEA-IRS

  12. A study of the morphoanatomical characters of the leaves of Chamaecrista (L. Moench sect. Apoucouita (Leguminosae-Caesalpinioideae

    Directory of Open Access Journals (Sweden)

    Ítalo Antônio Cotta Coutinho

    2016-01-01

    Full Text Available ABSTRACT Little attention has been paid to species of Chamaecrista sect. Apoucouita (Leguminosae-Caesalpinioideae, especially regarding anatomical studies. When only vegetative material is available, the identification of such species may be difficult. Additionally, vegetative material of some species of C. sect. Apoucouita may be even harder to identify because they can resemble species of Inga Mill. (Leguminosae-Mimosoideae. The present study focused on recognizing morphoanatomical characters of leaves that are taxonomically useful for the species of C. sect. Apoucouita by employing standard anatomical techniques. The arrangement of the vascular system in the petiole/rachis, dorsiventral mesophyll, mucilage idioblasts in the epidermis of leaflets and hypostomatic leaves were some of the characters shared by all species studied. Length of the petiole, position and type of extrafloral nectaries, leaflet venation, presence and type of papillae on the epidermis of the leaflet blades and sclereids in the mesophyll were some of the characters useful in the distinction of taxa. The vascular arrangement of the petiole/rachis is a promising character in the distinction of species of C. sect. Apoucouita and Inga. Based on morphoanatomical data, the taxonomic revision of some species and varieties ascribed to C. sect. Apoucouita is suggested.

  13. A comparison of the BUGLE-80, SAILOR, and ELXSIR neutron cross-section libraries for PWR pressure vessels surveillance dosimetry and shielding applications

    International Nuclear Information System (INIS)

    Basha, H.S.; Manahan, M.P.

    1992-01-01

    In this paper three multigroup neutron cross-section libraries are used in synthesized three-dimensional discrete ordinates transport analyses to investigate their similarities, differences, and results for pressurized water reactor (PWR) pressure vessel surveillance dosimetry and shielding applications. The calculated-to-experimental (C/E) rations and the calculated reaction rates of several fast reactions are compared for the BUGLE-80, SAILOR, and ELXSIR cross-section libraries at the 97-deg surveillance capsule of the San Onofre Nuclear Generation Station Unit 2 (SONGS-2) and at the 90- and 97-deg (C/E ratios only) cavity dosimetry locations for another PWR (referred to as Reactor X)

  14. The Chandra X-ray Observatory PSF Library

    Science.gov (United States)

    Karovska, M.; Beikman, S. J.; Elvis, M. S.; Flanagan, J. M.; Gaetz, T.; Glotfelty, K. J.; Jerius, D.; McDowell, J. C.; Rots, A. H.

    Pre-flight and on-orbit calibration of the Chandra X-Ray Observatory provided a unique base for developing detailed models of the optics and detectors. Using these models we have produced a set of simulations of the Chandra point spread function (PSF) which is available to the users via PSF library files. We describe here how the PSF models are generated and the design and content of the Chandra PSF library files.

  15. Nuclear cross section library for oil well logging analysis

    International Nuclear Information System (INIS)

    Kodeli, I.; Kitsos, S.; Aldama, D.L.; Zefran, B.

    2003-01-01

    As part of the IRTMBA (Improved Radiation Transport Modelling for Borehole Applications) Project of the EU Community's 5 th Programme a special purpose multigroup cross section library to be used in the deterministic (as well as Monte Carlo) oil well logging particle transport calculations was prepared. This library is expected to improve the prediction of the neutron and gamma spectra at the detector positions of the logging tool, and their use for the interpretation of the neutron logging measurements was studied. Preparation and testing of this library is described. (author)

  16. Review of uncertainty files and improved multigroup cross section files for FENDL

    International Nuclear Information System (INIS)

    Ganesan, S.

    1994-03-01

    The IAEA Nuclear Data Section, in co-operation with several national nuclear data centers and research groups, is creating an internationally available Fusion Evaluated Nuclear Data Library (FENDL), which will serve as a comprehensive source of processed and tested nuclear data tailored to the requirements of the Engineering and Development Activities (EDA) of the International Thermonuclear Experimental Reactor (ITER) Project and other fusion-related development projects. The FENDL project of the International Atomic Energy Agency has the task of coordination with the goal of assembling, processing and testing a comprehensive, fusion-relevant Fusion Evaluated Nuclear Data Library with unrestricted international distribution. The present report contains the summary of the IAEA Advisory Group Meeting on ''Review of Uncertainty Files and Improved Multigroup Cross Section Files for FENDL'', held during 8-12 November 1993 at the Tokai Research Establishment, JAERI, Japan, organized in cooperation with the Japan Atomic Energy Research Institute. The report presents the current status of the FENDL activity and the future work plans in the form of conclusions and recommendations of the four Working Groups of the Advisory Group Meeting on (1) experimental and calculational benchmarks, (2) preparation processed libraries for FENDL/ITER, (3) specifying procedures for improving FENDL and (4) selection of activation libraries for FENDL. (author). 1 tab

  17. Description of WIMS Library Update Project (WLUP)

    International Nuclear Information System (INIS)

    Leszczynski, Francisco

    2002-01-01

    WIMS-D is one of the few reactor lattice codes that are in the public domain and therefore are available on non-commercial terms, for research and power nuclear reactor calculations. The main weakness of the WIMS-D package is its multi-group constants library, which is based on very old data. Relatively good performance of WIMS-D is attributed to a series of empirical adjustments to the multi-group data. However, the adjustments are not always justified by more accurate and recent experimental measurements. In view of the recently available new, or revised, evaluated nuclear data files it was felt that the performance of WIMS-D could be improved by updating its library. The WIMS-D Library Update Project (WLUP) was initiated in the early 1990's and finished in 2001. The International Atomic Energy Agency (IAEA) supported its co-ordination, but the project itself consisted of voluntary contributions from a large number of participants. In due course, several benchmarks for testing the library were identified and analyzed, the WIMSR module of the NJOY code system was upgraded, a detailed parametric study was performed to investigate the effects of various data processing input options on integral results and, the data processing methods for the main reactor materials were optimized. The final product, available on CD-ROM from NDS-IAEA includes: 69 and 172 group WIMSD libraries prepared from the selected evaluated data files, IAEA-TECDOC with detailed documentation, Processing inputs, Benchmark inputs and, the system of auxiliary codes developed under the project. (author)

  18. ETOA, ABBN Multigroup Constants from ENDF/B for Fast Reactors

    International Nuclear Information System (INIS)

    Nishimura, Hideo

    1977-01-01

    1 - Nature of physical problem solved: Production of ABBN type group constants up to 70 groups for fast reactor calculations, reading ENDF/B library as input. 2 - Method of solution: The multigroup method of Bondarenko et al. is used for processing basic nuclear data. Calculational algorithms for an unresolved resonance region are the same as those in the MC 2 code. For a resolved resonance region, an ultrafine energy structure dependent on a level scheme is adopted. 3 - Restrictions on the complexity of the problem: Maximum number of: energy groups: 70; sigma 0 values: 6; temperatures: 5. Self-shielding factors for an unrealistically low value of sigma 0 are not guaranteed because of the approximations used in the unresolved resonance region

  19. Generation of the WIMS code library from the ENDF/B-VI basic library

    International Nuclear Information System (INIS)

    Aboustta, Mohamed Ali Bashir.

    1994-01-01

    The WIMS code is being presently used in many research centers and educational institutions in the world. It has proven to be versatile, reliable and diverse as it is used to calculate different reactor systems. Its data library is rich of useful information that can even be condensed to serve other codes, but the copy distributed with the code is not updated. Some of its data has never been changed, others had changed many times to accommodate certain experimental setups and some data is, simply, not included. This work is an attempt to dominate the techniques used in generating a multigroup library as being applied to the WIMS data library. This new library is called UFMGLIB. A new set of consistent data was generated from the basic ENDF/B-VI library, including complete data for the fission product nuclides and more elaborated burnup chains. The performance of the library is comparable to that of the Standard library accompanying the code and a later library, WIMKAL 88, generated by a group of the Korean Research Institute of Atomic Energy. (author). 38 refs., 40 figs., 30 tabs

  20. Multi-Group Covariance Data Generation from Continuous-Energy Monte Carlo Transport Calculations

    International Nuclear Information System (INIS)

    Lee, Dong Hyuk; Shim, Hyung Jin

    2015-01-01

    The sensitivity and uncertainty (S/U) methodology in deterministic tools has been utilized for quantifying uncertainties of nuclear design parameters induced by those of nuclear data. The S/U analyses which are based on multi-group cross sections can be conducted by an simple error propagation formula with the sensitivities of nuclear design parameters to multi-group cross sections and the covariance of multi-group cross section. The multi-group covariance data required for S/U analysis have been produced by nuclear data processing codes such as ERRORJ or PUFF from the covariance data in evaluated nuclear data files. However in the existing nuclear data processing codes, an asymptotic neutron flux energy spectrum, not the exact one, has been applied to the multi-group covariance generation since the flux spectrum is unknown before the neutron transport calculation. It can cause an inconsistency between the sensitivity profiles and the covariance data of multi-group cross section especially in resolved resonance energy region, because the sensitivities we usually use are resonance self-shielded while the multi-group cross sections produced from an asymptotic flux spectrum are infinitely-diluted. In order to calculate the multi-group covariance estimation in the ongoing MC simulation, mathematical derivations for converting the double integration equation into a single one by utilizing sampling method have been introduced along with the procedure of multi-group covariance tally

  1. Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh; Calculs de reference avec un maillage multigroupe fin sur des assemblages critiques par Apollo2

    Energy Technology Data Exchange (ETDEWEB)

    Aggery, A

    1999-12-01

    The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)

  2. One-, two- and three-dimensional transport codes using multi-group double-differential form cross sections

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Sasaki, Makoto.

    1988-11-01

    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)

  3. Multigroup Moderation Test in Generalized Structured Component Analysis

    Directory of Open Access Journals (Sweden)

    Angga Dwi Mulyanto

    2016-05-01

    Full Text Available Generalized Structured Component Analysis (GSCA is an alternative method in structural modeling using alternating least squares. GSCA can be used for the complex analysis including multigroup. GSCA can be run with a free software called GeSCA, but in GeSCA there is no multigroup moderation test to compare the effect between groups. In this research we propose to use the T test in PLS for testing moderation Multigroup on GSCA. T test only requires sample size, estimate path coefficient, and standard error of each group that are already available on the output of GeSCA and the formula is simple so the user does not need a long time for analysis.

  4. CHARTB multigroup transport package

    International Nuclear Information System (INIS)

    Baker, L.

    1979-03-01

    The physics and numerical implementation of the radiation transport routine used in the CHARTB MHD code are discussed. It is a one-dimensional (Cartesian, cylindrical, and spherical symmetry), multigroup,, diffusion approximation. Tests and applications will be discussed as well

  5. 12G: code for conversion of isotope-ordered cross-section libraries into group-ordered cross-section libraries

    International Nuclear Information System (INIS)

    Resnik, W.M. II; Bosler, G.E.

    1977-09-01

    Many current reactor physics codes accept cross-section libraries in an isotope-ordered form, convert them with internal preprocessing routines to a group-ordered form, and then perform calculations using these group-ordered data. Occasionally, because of storage and time limitations, the preprocessing routines in these codes cannot convert very large multigroup isotope-ordered libraries. For this reason, the I2G code, i.e., ISOTXS to GRUPXS, was written to convert externally isotope-ordered cross section libraries in the standard file format called ISOTXS to group-ordered libraries in the standard format called GRUPXS. This code uses standardized multilevel data management routines which establish a strategy for the efficient conversion of large libraries. The I2G code is exportable contingent on access to, and an intimate familiarization with, the multilevel routines. These routines are machine dependent, and therefore must be provided by the importing facility. 6 figures, 3 tables

  6. Range calculations using multigroup transport methods

    International Nuclear Information System (INIS)

    Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

    1979-01-01

    Several aspects of radiation damage effects in fusion reactor neutron and ion irradiation environments are amenable to treatment by transport theory methods. In this paper, multigroup transport techniques are developed for the calculation of particle range distributions. These techniques are illustrated by analysis of Au-196 atoms recoiling from (n,2n) reactions with gold. The results of these calculations agree very well with range calculations performed with the atomistic code MARLOWE. Although some detail of the atomistic model is lost in the multigroup transport calculations, the improved computational speed should prove useful in the solution of fusion material design problems

  7. Creation and validation of a neutron-gamma coupled multigroup cross section library

    International Nuclear Information System (INIS)

    Devan, K.; Gopalakrishnan, V.; Lee, S.M.

    1995-01-01

    The task of creating our own neutron-gamma coupled library was taken up. By using 1985 version of NJOY code system, a coupled set called IGC-DE4-S1 in ANISN format for 25 nuclides has been arrived at based on ENDF/B-IV neutron library and DLC-99 gamma library, with Legendre order of up to 5. The flow chart for the creation of coupled set is given. 5 refs, 1 fig., 3 tabs

  8. Generation and testing of the shielding data library EURLIB for fission and fusion technology

    International Nuclear Information System (INIS)

    Caglioti, E.; Hehn, G.; Herrnberger, V.; Mattes, M.; Nicks, R.; Penkuhn, H.

    1977-01-01

    For the common field of core physics and shielding, the CSEWG group structure of 239 fast neutron groups had been proposed, of which the 100 neutron groups of the EURLIB Library is a sub-set for shielding. This standard group Library EURLIB had been initiated by the NEA-specialist group on shielding benchmarks in 1974. The wide acceptance of the Library for interpretation of benchmarks in the NEA program represents an important step forward in the standardization of group data which is the basic requirement for a useful collaboration. On the other side the interpretation of a series of different benchmark experiments with the EURLIB Library provides the best check of the cross section data for neutron and gamma-rays showing the needs for further improvements. The paper describes the joint work of IKE, Stuttgart and EURATOM, Ispra in generating multigroup libraries for neutron and gamma-rays. Special effort has been devoted to improve the flux weighting for both types of radiation and proper treatment of thermal neutrons. The coupled multigroup Library of 100 neutron and 20 gamma groups is collapsed into few group structures for typical designs of LWR, LMFBR, gas cooled and thermonuclear reactors. The work for optimal few group representation is done in cooperation with EIR, Wurenlingen. The testing of the EURLIB Library is a common effort of several institutions participating in the NEA shielding benchmark program

  9. WIMSD5, Deterministic Multigroup Reactor Lattice Calculations

    International Nuclear Information System (INIS)

    2004-01-01

    1 - Description of program or function: The Winfrith improved multigroup scheme (WIMS) is a general code for reactor lattice cell calculation on a wide range of reactor systems. In particular, the code will accept rod or plate fuel geometries in either regular arrays or in clusters and the energy group structure has been chosen primarily for thermal calculations. The basic library has been compiled with 14 fast groups, 13 resonance groups and 42 thermal groups, but the user is offered the choice of accurate solutions in many groups or rapid calculations in few groups. Temperature dependent thermal scattering matrices for a variety of scattering laws are included in the library for the principal moderators which include hydrogen, deuterium, graphite, beryllium and oxygen. WIMSD5 is a successor version of WIMS-D/4. 2 - Method of solution: The treatment of resonances is based on the use of equivalence theorems with a library of accurately evaluated resonance integrals for equivalent homogeneous systems at a variety of temperatures. The collision theory procedure gives accurate spectrum computations in the 69 groups of the library for the principal regions of the lattice using a simplified geometric representation of complicated lattice cells. The computed spectra are then used for the condensation of cross-sections to the number of groups selected for solution of the transport equation in detailed geometry. Solution of the transport equation is provided either by use of the Carlson DSN method or by collision probability methods. Leakage calculations including an allowance for streaming asymmetries may be made using either diffusion theory or the more elaborate B1-method. The output of the code provides Eigenvalues for the cases where a simple buckling mode is applicable or cell-averaged parameters for use in overall reactor calculations. Various reaction rate edits are provided for direct comparison with experimental measurements. 3 - Restrictions on the complexity of

  10. BUGLE-93 (ENDF/B-VI) cross-section library data testing using shielding benchmarks

    International Nuclear Information System (INIS)

    Hunter, H.T.; Slater, C.O.; White, J.E.

    1994-01-01

    Several integral shielding benchmarks were selected to perform data testing for new multigroup cross-section libraries compiled from the ENDF/B-VI data for light water reactor (LWR) shielding and dosimetry. The new multigroup libraries, BUGLE-93 and VITAMIN-B6, were studied to establish their reliability and response to the benchmark measurements by use of radiation transport codes, ANISN and DORT. Also, direct comparisons of BUGLE-93 and VITAMIN-B6 to BUGLE-80 (ENDF/B-IV) and VITAMIN-E (ENDF/B-V) were performed. Some benchmarks involved the nuclides used in LWR shielding and dosimetry applications, and some were sensitive specific nuclear data, i.e. iron due to its dominant use in nuclear reactor systems and complex set of cross-section resonances. Five shielding benchmarks (four experimental and one calculational) are described and results are presented

  11. Integral tests of coupled multigroup neutron and gamma cross sections with fission and fusion sources

    International Nuclear Information System (INIS)

    Schriewer, J.; Hehn, G.; Mattes, M.; Pfister, G.; Keinert, J.

    1978-01-01

    Calculations were made for different benchmark experiments in order to test the coupled multigroup neutron and gamma library EURLIB-3 with 100 neutron groups and 20 gamma groups. In cooperation with EURATOM, Ispra, we produced this shielding library recently from ENDF/B-IV data for application in fission and fusion technology. Integral checks were performed for natural lithium, carbon, oxygen, and iron. Since iron is the most important structural material in nuclear technology, we started with calculations of iron benchmark experiments. Most of them are integral experiments of INR, Karlsruhe, but comparisons were also done with benchmark experiments from USA and Japan. For the experiments with fission sources we got satisfying results. All details of the resonances cannot be checked with flux measurements and multigroup cross sections used. But some averaged resonance behaviour of the measured and calculated fluxes can be compared and checked within the error limits given. We get greater differences in the calculations of benchmark experiments with 14 MeV neutron sources. For iron the group cross sections of EURLIB-3 produce an underestimation of the neutron flux in a broad energy region below the source energy. The conclusion is that the energy degradation by inelastic scattering is too strong. For fusion application the anisotropy of the inelastic scatter process must be taken into account, which isn't done by the processing codes at present. If this effect isn't enough, additional corrections have to be applied to the inelastic cross sections of iron in ENDF/B-IV. (author)

  12. Generation of broad-group neutron/photon cross-section libraries for shielding applications

    International Nuclear Information System (INIS)

    Ingersoll, D.T.; Roussin, R.W.; Fu, C.Y.; White, J.E.

    1989-01-01

    The generation and use of multigroup cross-section libraries with broad energy group structures is primarily for the economy of computer resources. Also, the establishment of reference broad-group libraries is desirable in order to avoid duplication of effort, both in terms of the data generation and verification, and to assure a common data base for all participants in a specific project. Uncertainties are inevitably introduced into the broad-group cross sections due to approximations in the grouping procedure. The dominant uncertainty is generally with regard to the energy weighting function used to average the pointwise or fine-group data within a single broad group. Intelligent choice of the weighting functions can reduce such uncertainties. Also, judicious selection of the energy group structure can help to reduce the sensitivity of the computed responses to the weighting function, at least for a selected set of problems. Two new multigroup cross section libraries have been recently generated from ENDF/B-V data for two specific shielding applications. The first library was prepared for use in sodium-cooled reactor systems and is available in both broad-group structures. The second library, just recently completed, was prepared for use in air-over-ground environments and is available in a broad-group (46-neutron, 23-photon) energy structure. The selection of the specific group structures and weighting functions was an important part of the generation of both libraries

  13. STUDY ON THE HISTORY AND ARCHITECTURE OF THE RINZAI ZEN SECT BUDDHIST MONASTERIES IN KYÔTO

    Directory of Open Access Journals (Sweden)

    Antariksa Antariksa

    2000-01-01

    Full Text Available In the middle of the 14th century, the Five Temples or Five Mountains (Gozan of Rinzai Zen sect were developed rapidly in the city of Kyôto. The ranking system of officially sponsored by Zen Buddhist monasteries were created by the Kamakura and Muromachi government. The Gozan system originally included three monasteries in Kyôto and three in Kamakura, but immediately spread to five monasteries in several cities. While the two great monasteries, are not members of the Gozan group, which belonged to the Rinka group of Rinzai Zen sect temples. The system, which corresponds to a Chinese hierarchical model, was established in Japan during the Kamakura period. The purpose of this study is to attempt to clarify the history and architecture of Rinzai Zen sect Buddhist monastery. The central functioning building of the Zen Buddhist monastery reflects the repetition, consistency, persistence, and order of the monastic ritual.

  14. Development of the CPXSD Methodology for Generation of Fine-Group Libraries for Shielding Applications

    International Nuclear Information System (INIS)

    Alpan, F. Arzu; Haghighat, Alireza

    2005-01-01

    Multigroup cross sections are one of the major factors that cause uncertainties in the results of deterministic transport calculations. Thus, it is important to prepare effective cross-section libraries that include an appropriate group structure and are based on an appropriate spectrum. There are several multigroup cross-section libraries available for particular applications. For example, the 47-neutron, 20-gamma group BUGLE library that is derived from the 199-neutron, 42-gamma group VITAMIN-B6 library is widely used for light water reactor (LWR) shielding and pressure vessel dosimetry applications. However, there is no publicly available methodology that can construct problem-dependent libraries. Thus, the authors have developed the Contributon and Point-wise Cross Section Driven (CPXSD) methodology for constructing effective fine- and broad-group structures. In this paper, new fine-group structures were constructed using the CPXSD, and new fine-group cross-section libraries were generated. The 450-group LIB450 and 589-group LIB589 libraries were developed for neutrons sensitive to the fast and thermal energy ranges, respectively, for LWR shielding problems. As compared to a VITAMIN-B6-like library, the new fine-group library developed for fast neutron dosimetry calculations resulted in closer agreement to the continuous-energy predictions. For example, for the fast neutron cavity dosimetry, ∼4% improvement was observed for the 237 Np(n,f) reaction rate. For the thermal neutron 1 H(n, γ) reaction, a maximum improvement of ∼14% was observed in the reaction rate at the middowncomer position

  15. Self-shielding phenomenon modelling in multigroup transport code Apollo-2; Modelisation du phenomene d'autoprotection dans le code de transport multigroupe Apollo 2

    Energy Technology Data Exchange (ETDEWEB)

    Coste-Delclaux, M

    2006-03-15

    This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)

  16. Neutron-photon multigroup cross sections for neutron energies up to 400 MeV: HILO86R

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Nakane, Yoshihiro; Hasegawa, Akira; Tanaka, Shun-ichi

    1993-02-01

    A macroscopic multigroup cross section library of 66 neutron and 22 photon groups for neutron energies up to 400 MeV: HILO86R is prepared for 10 typical shielding materials; water, concrete, iron, air, graphite, polyethylene, heavy concrete, lead, aluminum and soil. The library is a revision of the DLC-119/HILO86, in which only the cross sections below 19.6 MeV have been exchanged with a group cross section processed from the JENDL-3 microscopic cross section library. In the HILO86R library, self shielding factors are used to produce effective cross sections for neutrons less than 19.6 MeV considering rather coarse energy meshes. Energy spectra and dose attenuation in water, concrete and iron have been compared among the HILO, HILO86 and HILO86R libraries for different energy neutron sources. Significant discrepancy has been observed in the energy spectra less than a couple of MeV energy in iron among the libraries, resulting large difference in the dose attenuation. The difference was attributed to the effect of self-shielding factor, namely to the difference between infinite dilution and effective cross sections. Even for 400 MeV neutron source the influence of the self-shielding factor is significant, nevertheless only the cross sections below 19.6 MeV are exchanged. (author)

  17. NJOY processed multigroup library for fast reactor applications and point data library for MCNP - Experience and validation

    International Nuclear Information System (INIS)

    Kim Jung-Do; Gil Choong-Sup

    1996-01-01

    JEF-1-based 50-group cross section library for fast reactor applications and point data library for continuous-energy Monte Carlo code MCNP have been generated using NJOY91.38 system. They have been examined by analyzing measured integral quantities such as criticality and central reaction rate ratios for 8 small fast critical assemblies. (author). 9 refs, 2 figs, 10 tabs

  18. COVFILS: 30-group covariance library based on ENDF/B-V

    International Nuclear Information System (INIS)

    Muir, D.W.; LaBauve, R.J.

    1981-03-01

    A library of 30-group cross sections and covariances called COVFILS has been prepared from ENDF/B-V data using the NJOY code system. COVFILS includes data on the total cross section, scattering cross sections, and the most important absorption cross sections for 1 H, 10 B, C, 16 O, Cr, Fe, Ni, Cu, and Pb. This report contains detailed descriptions of various features of the library, a listing of a FORTRAN retrieval program, and 143 plots of the multigroup cross-section uncertainties and their correlations

  19. Generation of Matxs-formated nuclear data libraries

    International Nuclear Information System (INIS)

    Vontobel, P.

    1989-01-01

    Using the NJOY nuclear data processing system, three multigroup MATXS-formated nuclear data libraries were generated based on the European data files JEF-1 and EFF-1. After processing with TRAMIX, TRANSX, or TRANSX-CTR these libraries can be red into most transport and diffusion codes. For the neutron analysis of gas-cooled or water moderated thermal reactor systems (including high converter PWR's) a 70-group WIMS-BOXER structured library was generated. A general purpose fine group library in 308 groups is provided for thermal as well as for fast reactor systems. A coupled 175 neutron/42 photon-group library in VITAMIN-J structure was created for the analysis of shielding problems and fusion blanket design. A problem found when using CRAY's CFT77 compiler to implement NJOY87 is discussed. The problem of irregular selfshielding factors from UNRESR for some isotopes and (σ 0 , material temperature)-combinations in the unresolved resonance range is addressed

  20. WIMS library up-date project: first stage results

    International Nuclear Information System (INIS)

    Prati, A.; Claro, L.H.

    1990-01-01

    The following benchmarks: TRX1, TRX2, BAPL-UO sub(2)-1, BAPL-UO sub (2)-2, BAPL-UO sub(2)-3 have been calculated with the WIMSD/4 code, as a contribution of CTA/IEAv, to the first stage of the WIMS Library Update Project, coordinated by the International Atomic Energy Agency. The card image input for each benchmark has been attached and the major input options/parameters are commented. The version of the WIMSD/4 code and its multigroup cross section library used to run the benchmarks are specified. Results from the major integral parameters are presented and discussed. (author)

  1. Library analytics and metrics using data to drive decisions and services

    CERN Document Server

    2015-01-01

    This book will enable libraries to make informed decisions, develop new services and improve user experience by collecting, analysing and utilising data. With the wealth of data available to library and information services, analytics are the key to understanding your users and your field of operations better and improving the services that you offer. This book sets out the opportunities that analytics present to libraries, and provides inspiration for how they can use the data within their systems to help inform decisions and drive services. Using case studies to provide real-life examples of current developments and services, and packed full of practical advice and guidance for libraries looking to realise the value of their data, this will be an essential guide for librarians and information professionals. This volume will bring together a group of internationally recognised experts to explore some of the key issues in the exploitation of data analytics and metrics in the library and cultural heritage sect...

  2. ZZ BOREHOLE-EB6.8-MG, multi group cross-section library for deterministic and Monte Carlo codes

    International Nuclear Information System (INIS)

    Kodeli, Ivo; Aldama, Daniel L.; Leege, Piet F.A. de; Legrady, David; Hoogenboom, J. Eduard

    2007-01-01

    1 - Description: Format: MATXS and ACE; Number of groups: 175 neutron, 45 gamma-ray; Nuclides: H-1, C-12, O-16, Na-23, Mg-nat, Al-27, Si-28, -29, -30, S-nat, Cl-35, -37, K-nat, Ca-nat, Mn-55, Fe-54, -56, -57, -58, I-127, W-nat. Origin: ENDF/B-VI.8; Weighting spectrum: Fission and fusion peak at high energies and a 1/E + thermal Maxwellian extension at low energies. The following materials/nuclides are included in the library: H-1, C-12, O-16, Na-23, Mg-nat, Al-27, Si-28, -29, -30, S-nat, Cl-35, -37, K-nat, Ca-nat, Fe-54, -56, -57, -58, Mn-55, I-127, W-nat. ZZ-BOREHOLE-EB6.8-MG is a multigroup cross section library for deterministic (DOORS, DANTSYS) and Monte Carlo (MCNP) transport codes developed for the oil well logging applications. The library is based on the ENDF/B-VI.8 evaluation and was processed by the NJOY-99 code. The cross sections are given in the 175 neutron and 45 gamma ray group structure. The MATXS format library can be directly used in TRANSX code to prepare the multigroup self-shielded cross sections for deterministic discrete ordinates codes like DOORS and DANTSYS. The data provided in the GROUPR and GAMINR format were converted to the MCNP ACE format by the NSLINK, SCALE and CRSRD codes. IAEA1398/03: Multigroup cross section data for Mn-55 were added in TRANSX format

  3. Update of PHOENIX-P 42 group library from CENDL-2

    International Nuclear Information System (INIS)

    Zhang Baocheng

    1998-01-01

    PHOENIX-P is a lattice physics code system, developed by the Westinghouse Electric Corporation (WEC), which was transplanted and used at Dayabay Nuclear Power Plant (DNPJVC). The associated multi-group (42-group) library was derived from the evaluated nuclear data of ENDF/B-5. Since the original library is from the old evaluated nuclear data, it can not meet all the requirements of reactor physics calculations of the nuclear power plant. So it is necessary to update the library with the latest version of evaluated nuclear data. To do so, based on the investigation of the old library and the information about the library, some programs were developed at China Nuclear Data Center (CNDC) to produce PHOENIX-P format data sets mainly from CENDL-2 and the new data were used to supersede the old ones of the PHOENIX-P library

  4. Self-shielding phenomenon modelling in multigroup transport code Apollo-2; Modelisation du phenomene d'autoprotection dans le code de transport multigroupe Apollo 2

    Energy Technology Data Exchange (ETDEWEB)

    Coste-Delclaux, M

    2006-03-15

    This document describes the improvements carried out for modelling the self-shielding phenomenon in the multigroup transport code APOLLO2. They concern the space and energy treatment of the slowing-down equation, the setting up of quadrature formulas to calculate reaction rates, the setting-up of a method that treats directly a resonant mixture and the development of a sub-group method. We validate these improvements either in an elementary or in a global way. Now, we obtain, more accurate multigroup reaction rates and we are able to carry out a reference self-shielding calculation on a very fine multigroup mesh. To end, we draw a conclusion and give some prospects on the remaining work. (author)

  5. Reference calculations on critical assemblies with Apollo2 code working with a fine multigroup mesh

    International Nuclear Information System (INIS)

    Aggery, A.

    1999-12-01

    The objective of this thesis is to add to the multigroup transport code APOLLO2 the capability to perform deterministic reference calculations, for any type of reactor, using a very fine energy mesh of several thousand groups. This new reference tool allows us to validate the self-shielding model used in industrial applications, to perform depletion calculations, differential effects calculations, critical buckling calculations or to evaluate precisely data required by the self shielding model. At its origin, APOLLO2 was designed to perform routine calculations with energy meshes around one hundred groups. That is why, in the current format of cross sections libraries, almost each value of the multigroup energy transfer matrix is stored. As this format is not convenient for a high number of groups (concerning memory size), we had to search out a new format for removal matrices and consequently to modify the code. In the new format we found, only some values of removal matrices are kept (these values depend on a reconstruction precision choice), the other ones being reconstructed by a linear interpolation, what reduces the size of these matrices. Then we had to show that APOLLO2 working with a fine multigroup mesh had the capability to perform reference calculations on any assembly geometry. For that, we successfully carried out the validation with several calculations for which we compared APOLLO2 results (obtained with the universal mesh of 11276 groups) to results obtained with Monte Carlo codes (MCNP, TRIPOLI4). Physical analysis led with this new tool have been very fruitful and show a great potential for such an R and D tool. (author)

  6. A revision of photon interaction data in the UKAEA nuclear data library

    International Nuclear Information System (INIS)

    Knipe, A.D.

    1975-10-01

    Photon interaction data in the UKAEA Nuclear Data Library have been updated and extended to cover all elements up to Atomic Number 94. Cross-sections for the photoelectric effect, Compton scattering, pair-production, and the total cross-section, are stored at 40 energy points in the range 0.01 MeV to 20 MeV. The angular distribution for Compton scattering is also included in the library. This report describes the derivation and accuracy of the data, and tabulates the cross-sections and angular distribution in the appendices. The preparation of multigroup cross-sections from the library's data is also discussed. (author)

  7. MUXS: a code to generate multigroup cross sections for sputtering calculations

    International Nuclear Information System (INIS)

    Hoffman, T.J.; Robinson, M.T.; Dodds, H.L. Jr.

    1982-10-01

    This report documents MUXS, a computer code to generate multigroup cross sections for charged particle transport problems. Cross sections generated by MUXS can be used in many multigroup transport codes, with minor modifications to these codes, to calculate sputtering yields, reflection coefficients, penetration distances, etc

  8. Eccolib-Jeff-3.1 libraries; Les bibliotheques Eccolib-Jeff-3.1

    Energy Technology Data Exchange (ETDEWEB)

    Sublet, J Ch; Dean, Ch; Plisson-Rieunier, D

    2006-07-01

    The multi-group, multi-temperature libraries ECCOLIB-JEFF-3.1, derived from the JEFF-3.1 evaluations, have been produced using the NJOY-99 and CALENDF-2005 processing codes and the MERGE-3.8 and GECCO-1.4 interface tools. The processing steps are explained along side the Quality Assurance processes linked to such libraries. Final results encompass five main libraries; two fine 1968 group with probability tables for 104 major isotopes allowing detailed criticality studies, another in the same groups structure for the thermal compounds, and two others in coarser 172 XMAS and 175 Vitamin-J group structures for other criticality and shielding, transport studies. (authors)

  9. Development and validation of library MUSE-F1.0

    International Nuclear Information System (INIS)

    Tang Haibo; Chen Yixue; Wu Jun

    2013-01-01

    The multi-group transport library MUSE-F1.0 based on ENDF/B-VII.0 with 175-group neutron and 42-group photon was developed by NJOY99. Weighting function is thermal--l/e--fast reactor-fission + fusion, and Legendre order is six. The library was validated by a series of critical and shielding benchmarks. The shielding test involves nuclear data including fission reactor, fusion reactor and accelerator. The result shows that MUSE-F1.0 is suitable for critical and shielding calculation. And it is competent for the application of fast neutron reactor design. (authors)

  10. Use of OsiriX in developing a digital radiology teaching library

    International Nuclear Information System (INIS)

    Shamshuddin, S.; Matthews, H.R.

    2014-01-01

    Widespread adoption of digital imaging in clinical practice and for the image-based examinations of the Royal College of Radiologists has created a desire to provide a digital radiology teaching library in many hospital departments around the UK. This article describes our experience of using OsiriX software in developing digital radiology teaching libraries

  11. Development of Calcomp compatible interface library 'piflib' on X Window System

    International Nuclear Information System (INIS)

    Tanabe, Hidenobu; Yokokawa, Mitsuo; Onuma, Yoshio.

    1993-05-01

    Graphics processing at JAERI mainly has been executed on mainframe computers with Calcomp compatible graphics libraries. With spread of engineering workstations (EWS), it is important that those enormous graphics softwares be able to be carried out on EWS. The Calcomp compatible interface library 'piflib' has been developed on the X Window System, which is the most popular window environments on EWS. In this report, specifications of the library 'piflib' and its usages are presented. The cooperative processing with mainframe computers is also described. (author)

  12. Validation of multigroup neutron cross sections and calculational methods for the advanced neutron source against the FOEHN critical experiments measurements

    International Nuclear Information System (INIS)

    Smith, L.A.; Gallmeier, F.X.; Gehin, J.C.

    1995-05-01

    The FOEHN critical experiment was analyzed to validate the use of multigroup cross sections and Oak Ridge National Laboratory neutronics computer codes in the design of the Advanced Neutron Source. The ANSL-V 99-group master cross section library was used for all the calculations. Three different critical configurations were evaluated using the multigroup KENO Monte Carlo transport code, the multigroup DORT discrete ordinates transport code, and the multigroup diffusion theory code VENTURE. The simple configuration consists of only the fuel and control elements with the heavy water reflector. The intermediate configuration includes boron endplates at the upper and lower edges of the fuel element. The complex configuration includes both the boron endplates and components in the reflector. Cross sections were processed using modules from the AMPX system. Both 99-group and 20-group cross sections were created and used in two-dimensional models of the FOEHN experiment. KENO calculations were performed using both 99-group and 20-group cross sections. The DORT and VENTURE calculations were performed using 20-group cross sections. Because the simple and intermediate configurations are azimuthally symmetric, these configurations can be explicitly modeled in R-Z geometry. Since the reflector components cannot be modeled explicitly using the current versions of these codes, three reflector component homogenization schemes were developed and evaluated for the complex configuration. Power density distributions were calculated with KENO using 99-group cross sections and with DORT and VENTURE using 20-group cross sections. The average differences between the measured values and the values calculated with the different computer codes range from 2.45 to 5.74%. The maximum differences between the measured and calculated thermal flux values for the simple and intermediate configurations are ∼ 13%, while the average differences are < 8%

  13. Multigroup calculations of low-energy neutral transport in tokamak plasmas

    International Nuclear Information System (INIS)

    Gilligan, J.G.; Gralnick, S.L.; Price, W.G. Jr.; Kammash, T.

    1978-01-01

    Multigroup discrete ordinates methods avoid many of the approximations that have been used in previous neutral transport analyses. Of particular interest are the neutral profiles generated as an integral part of larger plasma system simulation codes. To determine the appropriateness of utilizing a particular multigroup code, ANISN, for this purpose, results are compared with the neutral transport module of the Duechs code. For a typical TFTR plasma, predicted neutral densities differ by a maximum factor of three on axis and outfluxes at the plasma boundary by approximately 40%. This is found to be significant for a neutral transport module. Possible sources of the observed discrepancies are indicated from an analysis of the approximations used in the Duechs model. Recommendations are made concerning the future application of the multigroup method. (author)

  14. The Suppression of Energy Discretization Errors in Multigroup Transport Calculations

    International Nuclear Information System (INIS)

    Larsen, Edward

    2013-01-01

    The Objective of this project is to develop, implement, and test new deterministric methods to solve, as efficiently as possible, multigroup neutron transport problems having an extremely large number of groups. Our approach was to (i) use the standard CMFD method to 'coarsen' the space-angle grid, yielding a multigroup diffusion equation, and (ii) use a new multigrid-in-space-and-energy technique to efficiently solve the multigroup diffusion problem. The overall strategy of (i) how to coarsen the spatial an energy grids, and (ii) how to navigate through the various grids, has the goal of minimizing the overall computational effort. This approach yields not only the fine-grid solution, but also coarse-group flux-weighted cross sections that can be used for other related problems.

  15. Optimization of multi-group cross sections for fast reactor analysis

    International Nuclear Information System (INIS)

    Chin, M. R.; Manalo, K. L.; Edgar, C. A.; Paul, J. N.; Molinar, M. P.; Redd, E. M.; Yi, C.; Sjoden, G. E.

    2013-01-01

    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% PUO 2 -UO 2 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)

  16. ZZ MCNPDATA, Standard Neutron, Photon and Electron Data Libraries for MCNP-4C and MCB1C

    International Nuclear Information System (INIS)

    2002-01-01

    of data: A wide variety of continuous-energy, discrete, multigroup, thermal and dosimetry neutron data libraries are available in this release. The continuous-energy neutron data libraries available include: ENDF60, RMCCS, RMCCSA, ENDF5U, ENDF5P, NEWXS, ENDF5MT*, MISC5XS**, ENDL85, KIDMAN, 100XS, URES***, ENDF6DN***, ENDF62MT***, and ENDL92***. The discrete neutron data libraries include: NEWXSD, DRMCCS, and DRE5. The multigroup neutron data library is MGXSNP, and the thermal S(alpha, beta) libraries are TMCCS and THERXS. The neutron dosimetry libraries are 531DOS, 532DOS, and LLLDOS. The photon transport libraries are MCPLIB and MCPLIB02, and the electron libraries are EL and EL03***. The photon and electron data libraries contain data for elements having Z<95. The data libraries, as distributed, are in ASCII, or type 1, format. - The data library ENDF5MT contains data previously available in the library EPRIXS, along with the U600K data library. - The data library MISC5XS contains corrected data for ENDF/B-V based Zr as described below, and the libraries previously known as IRNAT, MISCXS, ARKRC, TM169, GDT2GP, and T2DDC. The ENDF/B-V Zr data has been corrected for five ZAID's from the libraries RMCCS, DRMCCS, ENDF5P, DRE5, and EPRIXS. Below is a summary of the changes that have been made for Zr (Previous → Corrected): RMCCS 40000.51c → 40000.57c MISC5XS 300K; DRMCCS 40000.51d → 40000.57d MISC5XS 300K; ENDF5P 40000.50c → 40000.56c MISC5XS 300K; DRE5 40000.50d → 40000.56d MISC5X 300K; EPRIXS 40000.53c → 40000.58c MISC5XS 600K. - These five data libraries (URES, ENDF6DN, ENDF62MT, ENDL92, and EL03) are newly released in DLC-200. In the February 2001 update, URES was replaced with URESA. Please consult the README file and the more detailed documentation provided for descriptions of these libraries

  17. Begonia hosensis (sect. Reichenheimia, Begoniaceae, a new species from Sarawak, Malaysia

    Directory of Open Access Journals (Sweden)

    Che-Wei Lin

    2014-12-01

    Full Text Available Begonia hosensis, a new species of sect. Reichenheimia from Sarawak, Malaysia is here described and illustrated. Begonia hosensis resembles B. andersonii, differing in the stipule prominently keeled and female flower 5- (vs. 4- tepaled. Ecologically, the two species are also separable: B. hosensis occurs on granitic cliffs ca. 400 km east of Kuching, the only known locality of B. andersonii, a limestone species.

  18. THEMIS-4: a coherent punctual and multigroup cross section library for Monte Carlo and SN codes from ENDF/B4

    International Nuclear Information System (INIS)

    Dejonghe, G.; Gonnord, J.; Monnier, A.; Nimal, J.C.

    1983-05-01

    The THEMIS cross section processing system has been developped to produce punctual data for MONTE CARLO and coherent multigroup data for SN codes from ENDF/B. The THEMIS-4 data base has been generated from ENDF/B4 using the system and can be accessed by the 3-D Monte Carlo system TRIPOLI-2 and by the SN codes ANISN and DOT. An interpretation of ORNL fusion shielding benchmark is presented

  19. Generation of multigroup cross-sections from micro-group ones in code system SUHAM-U used for VVER-1000 reactor core calculations with MOX loading

    Energy Technology Data Exchange (ETDEWEB)

    Boyarinov, V.F.; Davidenko, V.D.; Polismakov, A.A.; Tsybulsky, V.F. [RRC Kurchatov Institute, Moscow (Russian Federation)

    2005-07-01

    At the present time, the new code system SUHAM-U for calculation of the neutron-physical processes in nuclear reactor core with triangular and square lattices based both on the modern micro-group (about 7000 groups) cross-sections library of code system UNK and on solving the multigroup (up to 89 groups) neutron transport equation by Surface Harmonics Method is elaborated. In this paper the procedure for generation of multigroup cross-sections from micro-group ones for calculation of VVER-1000 reactor core with MOX loading is described. The validation has consisted in computing VVER-1000 fuel assemblies with uranium and MOX fuel and has shown enough high accuracy under corresponding selection of the number and boundaries of the energy groups. This work has been fulfilled in the frame of ISTC project 'System Analyses of Nuclear Safety for VVER Reactors with MOX Fuels'.

  20. Taxonomic revision of the tropical African group of Carex subsect. Elatae (sect. Spirostachyae, Cyperaceae

    Directory of Open Access Journals (Sweden)

    Escudero, Marcial

    2011-12-01

    Full Text Available The tropical African monophyletic group of Carex subsect. Elatae (sect. Spirostachyae is distributed in continental tropical Africa, Madagascar, the Mascarene archipelago, and Bioko Island (32 km off the coast of West Africa, in the Gulf of Guinea. The first monographic treatment of this Carex group, as well as of the tribe Cariceae, was published by Kükenthal (as sect. Elatae Kük.. Recently, the first molecular (nrDNA, cpDNA phylogeny of Carex sect. Elatae has been published, which also included the species of sect. Spirostachyae. In the resulting consensus trees, most species of sect. Elatae were embedded within core Spirostachyae and so this section was joined with sect. Spirostachyae as subsect. Elatae. Within subsect. Elatae, several groups were described, one of which was termed the “tropical African group”. Here we present a taxonomic revision of this group, based on more than 280 vouchers from 29 herbaria as well as in field trips in Tropical Africa. In the revision, we recognise 12 species (16 taxa within the tropical African group, and so have somewhat modified our previous view, in which 10 species, 12 taxa were listed. One new species from Tanzania is included in this treatment, C. uluguruensis Luceño & M. Escudero. Several combinations are made, C. cyrtosaccus is treated as a synonym of C. vallis-rosetto and, finally, the binomial C. greenwayi has been recognised.Las especies de la subsección Elatae (sección Spirostachyae del género Carex que se distribuyen por África tropical continental, Madagascar, el archipiélago de las Mascareñas y la isla de Bioko (a 32 km del litoral de África occidental, en el golfo de Guinea forman un grupo monofilético. El primer tratamiento taxonómico de este grupo de cárices, así como de la tribu Cariceae en su conjunto, fue elaborado por Kükenthal (sección Elatae Kük.; recientemente, se ha publicado el primer estudio de filogenia molecular (nrDNA, cpDNA de los táxones de este grupo

  1. Cyclotron radiation by a multi-group method

    International Nuclear Information System (INIS)

    Chu, T.C.

    1980-01-01

    A multi-energy group technique is developed to study conditions under which cyclotron radiation emission can shift a Maxwellian electron distribution into a non-Maxwellian; and if the electron distribution is non-Maxwellian, to study the rate of cyclotron radiation emission as compared to that emitted by a Maxwellian having the same mean electron density and energy. The assumptions in this study are: the electrons should be in an isotropic medium and the magnetic field should be uniform. The multi-group technique is coupled into a multi-group Fokker-Planck computer code to study electron behavior under the influence of cyclotron radiation emission in a self-consistent fashion. Several non-Maxwellian distributions were simulated to compare their cyclotron emissions with the corresponding energy and number density equivalent Maxwellian distribtions

  2. Use of OsiriX in developing a digital radiology teaching library.

    Science.gov (United States)

    Shamshuddin, S; Matthews, H R

    2014-10-01

    Widespread adoption of digital imaging in clinical practice and for the image-based examinations of the Royal College of Radiologists has created a desire to provide a digital radiology teaching library in many hospital departments around the UK. This article describes our experience of using OsiriX software in developing digital radiology teaching libraries. Copyright © 2014 The Royal College of Radiologists. Published by Elsevier Ltd. All rights reserved.

  3. Cross-section libraries and kerma factors

    International Nuclear Information System (INIS)

    Little, R.C.; MacFarlane, R.E.; Seamon, R.E.

    1991-01-01

    A large amount of data is required in order to accurately simulate various aspects of Cold Neutron Sources using radiation transport codes such as MCNP and TWODANT. In particular, the following types of data are needed: couple neutron/photon transport libraries, neutron thermal S(α,β) data, response function data (including energy deposition), and proton interaction data. This paper concentrates on the coupled neutron/photon transport libraries and energy deposition. Data libraries available to radiation transport codes are obtained as a result of efforts in many areas, including differential and integral measurements, theoretical model codes, data evaluations, data processing, and data testing. A wide variety of data libraries are available to users of radiation transport codes, including pointwise and multigroup libraries. At Los Alamos, the authors generally recommend the use of data libraries derived from ENDF/B-V. It is often important to know how much energy is deposited in various regions of a device. This problem is typically modeled in radiation transport codes by folding the calculated fluences with an energy-dependent 'heating number'. The heating number represents the average energy deposited locally per collision. Calculation of these heating numbers from evaluated data libraries is fraught with difficulty. Many past difficulties related to energy deposition should be resolved by the release of ENDF/B-VI

  4. ZZ TEMPEST/MUFT, Thermal Neutron and Fast Neutron Multigroup Cross-Section Library for Program LEOPARD

    International Nuclear Information System (INIS)

    Kim, Jung-Do; Lee, Jong Tai

    1986-01-01

    Description of problem or function: Format: TEMPEST and MUFT; Number of groups: 246 thermal groups in TEMPEST Format and 54 fast groups in MUFT Format. From this library, the program SPOTS4 generates a 172-54 group library as input to the code LEOPARD. Nuclides: H, O, Zr, C, Fe, Ni, Al, Cr, Mn, U, Pu, Th, Pa, Xe, Sm, B and D. Origin: ENDF/B-4; Weighting spectrum: 1/E + U 235 fission spectrum. Data library of thermal and fast neutron group Cross sections to generate input to the program LEOPARD. The data is based on ENDF/B-4 and consists of two parts: (1) 246 thermal groups in TEMPEST Format. (2) 54 fast groups in MUFT Format. From this library, the program SPOTS4 generates a 172-54 group library as input to the code LEOPARD (NESC0279)

  5. A library for X-ray-matter interaction cross sections for X-ray fluorescence applications

    Energy Technology Data Exchange (ETDEWEB)

    Brunetti, A. [Istituto di Matematica e Fisica, Universita di Sassari, via Vienna 2, 07100 Sassari (Italy) and INFN, Sezione di Cagliari (Italy)]. E-mail: brunetti@uniss.it; Sanchez del Rio, M. [European Synchrotron Radiation Facility, 6 rue Jules Horowitz, 38043 Grenoble Cedex (France); Golosio, B. [INFN, Sezione di Cagliari (Italy); European Synchrotron Radiation Facility, 6 rue Jules Horowitz, 38043 Grenoble Cedex (France); Simionovici, A. [European Synchrotron Radiation Facility, 6 rue Jules Horowitz, 38043 Grenoble Cedex (France); Laboratoire de Sciences de la Terre, Ecole Normale Superieure, Lyon, F-69364 (France); Somogyi, A. [European Synchrotron Radiation Facility, 6 rue Jules Horowitz, 38043 Grenoble Cedex (France)

    2004-10-08

    Quantitative estimate of elemental composition by spectroscopic and imaging techniques using X-ray fluorescence requires the availability of accurate data of X-ray interaction with matter. Although a wide number of computer codes and data sets are reported in literature, none of them is presented in the form of freely available library functions which can be easily included in software applications for X-ray fluorescence. This work presents a compilation of data sets from different published works and an xraylib interface in the form of callable functions. Although the target applications are on X-ray fluorescence, cross sections of interactions like photoionization, coherent scattering and Compton scattering, as well as form factors and anomalous scattering functions, are also available.

  6. A numerical method for multigroup slab-geometry discrete ordinates problems with no spatial truncation error

    International Nuclear Information System (INIS)

    Barros, R.C. de; Larsen, E.W.

    1991-01-01

    A generalization of the one-group Spectral Green's Function (SGF) method is developed for multigroup, slab-geometry discrete ordinates (S N ) problems. The multigroup SGF method is free from spatial truncation errors; it generated numerical values for the cell-edge and cell-average angular fluxes that agree with the analytic solution of the multigroup S N equations. Numerical results are given to illustrate the method's accuracy

  7. [Polymorphism of KPI-A genes from plants of the subgenus Potatoe (sect. Petota, Estolonifera and Lycopersicum) and subgenus Solanum].

    Science.gov (United States)

    Krinitsyna, A A; Mel'nikova, N V; Belenikin, M S; Poltronieri, P; Santino, A; Kudriavtseva, A V; Savilova, A M; Speranskaia, A S

    2013-01-01

    Kunitz-type proteinase inhibitor proteins of group A (KPI-A) are involved in the protection of potato plants from pathogens and pests. Although sequences of large number of the KPI-A genes from different species of cultivated potato (Solanum tuberosum subsp. tuberosum) and a few genes from tomato (Solanum lycopersicum) are known to date, information about the allelic diversity of these genes in other species of the genus Solanum is lacking. In our work, the consensus sequences of the KPI-A genes were established in two species of subgenus Potatoe sect. Petota (Solanum tuberosum subsp. andigenum--5 genes and Solanum stoloniferum--2 genes) and in the subgenus Solanum (Solanum nigrum--5 genes) by amplification, cloning, sequencing and subsequent analysis. The determined sequences of KPI-A genes were 97-100% identical to known sequences of the cultivated potato of sect. Petota (cultivated potato Solanum tuberosum subsp. tuberosum) and sect. Etuberosum (S. palustre). The interspecific variability of these genes did not exceed the intraspecific variability for all studied species except Solanum lycopersicum. The distribution of highly variable and conserved sequences in the mature protein-encoding regions was uniform for all investigated KPI-A genes. However, our attempts to amplify the homologous genes using the same primers and the genomes of Solanum dulcamarum, Solanum lycopersicum and Mandragora officinarum resulted in no product formation. Phylogenetic analysis of KPI-A diversity showed that the sequences of the S. lycopersicum form independent cluster, whereas KPI-A of S. nigrum and species of sect. Etuberosum and sect. Petota are closely related and do not form species-specific subclasters. Although Solanum nigrum is resistant to all known races of economically one of the most important diseases of solanaceous plants oomycete Phytophthora infestans aminoacid sequences encoding by KPI-A genes from its genome have nearly or absolutely no differences to the same from

  8. ADS-Lib/V1.0. A test library for Accelerator Driven Systems. Summary documentation

    International Nuclear Information System (INIS)

    Lopez Aldama, D.; Trkov, A.

    2005-08-01

    The report describes the generation of a test library for a number of code systems used in the analysis of Accelerator Driven Systems (ADS). The generation of the ADS library was undertaken by IAEA-NDS and the data files are available to users at http://wwwnds. iaea.org/ads/ and also as CD-ROM (upon request).The source of the evaluated nuclear data was the JEFF-3.1 library. Processing was carried out using NJOY-99.90 with the local updates at IAEA-NDS. The resulting processed files are available in ACE format for MCNP and in MATXS format for multi-group transport calculations. (author)

  9. 3-DB, 3-D Multigroup Diffusion, X-Y-Z, R-Theta-Z, Triangular-Z Geometry, Fast Reactor Burnup

    International Nuclear Information System (INIS)

    Hardie, R.W.; Little, W.W. Jr.; Mroz, W.

    1974-01-01

    1 - Description of problem or function: 3DB is a three-dimensional (x-y-z, r-theta-z, triangular-z) multigroup diffusion code for use in detailed fast-reactor criticality and burnup analysis. The code can be used to - (a) compute k eff and perform criticality searches on time absorption, reactor composition, and reactor dimensions by means of either a flux or an adjoint model, (b) compute material burnup using a flexible material shuffling scheme, and (c) compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Eigenvalues are computed by standard source- iteration techniques. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Adjoint solutions are obtained by inverting the input data and redefining the source terms. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy-averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes are formed by the user. The code does not contain built- in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated

  10. CASTRO: A NEW COMPRESSIBLE ASTROPHYSICAL SOLVER. III. MULTIGROUP RADIATION HYDRODYNAMICS

    International Nuclear Information System (INIS)

    Zhang, W.; Almgren, A.; Bell, J.; Howell, L.; Burrows, A.; Dolence, J.

    2013-01-01

    We present a formulation for multigroup radiation hydrodynamics that is correct to order O(v/c) using the comoving-frame approach and the flux-limited diffusion approximation. We describe a numerical algorithm for solving the system, implemented in the compressible astrophysics code, CASTRO. CASTRO uses a Eulerian grid with block-structured adaptive mesh refinement based on a nested hierarchy of logically rectangular variable-sized grids with simultaneous refinement in both space and time. In our multigroup radiation solver, the system is split into three parts: one part that couples the radiation and fluid in a hyperbolic subsystem, another part that advects the radiation in frequency space, and a parabolic part that evolves radiation diffusion and source-sink terms. The hyperbolic subsystem and the frequency space advection are solved explicitly with high-order Godunov schemes, whereas the parabolic part is solved implicitly with a first-order backward Euler method. Our multigroup radiation solver works for both neutrino and photon radiation.

  11. Are white evangelical Protestants lower class? A partial test of church-sect theory.

    Science.gov (United States)

    Schwadel, Philip

    2014-07-01

    Testing hypotheses derived from church-sect theory and contemporary research about changes in evangelical Protestants' social status, I use repeated cross-sectional survey data spanning almost four decades to examine changes in the social-class hierarchy of American religious traditions. While there is little change in the social-class position of white evangelical Protestants from the early 1970s to 2010, there is considerable change across birth cohorts. Results from hierarchical age-period-cohort models show: (1) robust, across-cohort declines in social-class differences between white evangelical Protestants and liberal Protestants, affiliates of "other" religions, and the unaffiliated, (2) stability in social-class differences between white evangelical Protestants and moderate, Pentecostal, and nondenominational Protestants, (3) moderate across-cohort growth in social-class differences between white evangelical Protestants and Catholics, and (4) these patterns vary across indicators of social class. The findings in this article provide partial support for church-sect theory as well as other theories of social change that emphasize the pivotal role of generations. Copyright © 2014 Elsevier Inc. All rights reserved.

  12. A library of neutron data for calculating group constants

    International Nuclear Information System (INIS)

    Koshcheev, V.N.; Nikolaev, M.N.

    1987-01-01

    This paper describes the first version of a computerized library evaluated neutron data files (FOND) which includes data on the 67 most important nuclear reactor and radiation shielding materials. The data are represented in the ENDF/B format. The sources of data were the results of evaluations of data from differential neutron physics experiments conducted both in the USSR and abroad. The first version of the FOND library is not intended for use in reactor and shielding design calculations, but rather to serve as the basis for developing a corrected version which will guarantee adequate description of the results of a representative set of macroscopic experiments, and for generating multigroup constant systems in methodological research. (author)

  13. Elaboration of a nodal method to solve the steady state multigroup diffusion equation. Study and use of the multigroup diffusion code DAHRA

    International Nuclear Information System (INIS)

    Halilou, A.; Lounici, A.

    1981-01-01

    The subject is divided in two parts: In the first part a nodal method has been worked out to solve the steady state multigroup diffusion equation. This method belongs to the same set of nodal methods currently used to calculate the exact fission powers and neutron fluxes in a very short computing time. It has been tested on a two dimensional idealized reactors. The effective multiplication factor and the fission powers for each fuel element have been calculated. The second part consists in studying and mastering the multigroup diffusion code DAHRA - a reduced version of DIANE - a two dimensional code using finite difference method

  14. Methodological study of the adjustment of multigroup cross sections for fast reactors, by means of integral data

    International Nuclear Information System (INIS)

    Bastos, H.F.B.N.

    1979-01-01

    In this work a study of the methodology of the adjustment of multigroup cross sections by means of integral data is presented. A synthesis of the principal methods existent and the mathematical development of the adaptation of one of them are made. A calculational system is built from this reference method, with the basic conditions for the operation of the process of adjustment. In order to test the system developed and analyze several problems related to the adjustment, a series of trial adjustments was made with the value of the U 235 fission cross section from the infinite dilution library used in the calculational system for fast reactors of the Instituto de Engenharia Nuclear. (author)

  15. Cross sections in 25 groups obtained from ENDF/B-IV and ENDL/78 libraries, processed with GALAXY and NJOY computer codes

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Corcuera, R.P.

    1982-01-01

    The discrepancies existing between ENDF/B-IV and ENDL/78 libraries, in diferent energy regions are identified, and the order of the differences in multigroup sections are determined, when GALAXY or NJOY computer codes are used. (E.G.) [pt

  16. The Multigroup Neutron Diffusion Equations/1 Space Dimension

    Energy Technology Data Exchange (ETDEWEB)

    Linde, Sven

    1960-06-15

    A description is given of a program for the Ferranti Mercury computer which solves the one-dimensional multigroup diffusion equations in plane, cylindrical or spherical geometry, and also approximates automatically a two-dimensional solution by separating the space variables. In section A the method of calculation is outlined and the preparation of data for two group problems is described. The spatial separation of two-dimensional equations is considered in section B. Section C covers the multigroup equations. These parts are self contained and include all information required for the use of the program. Details of the numerical methods are given in section D. Three sample problems are solved in section E. Punching and operating instructions are given in an appendix.

  17. The Multigroup Neutron Diffusion Equations/1 Space Dimension

    International Nuclear Information System (INIS)

    Linde, Sven

    1960-06-01

    A description is given of a program for the Ferranti Mercury computer which solves the one-dimensional multigroup diffusion equations in plane, cylindrical or spherical geometry, and also approximates automatically a two-dimensional solution by separating the space variables. In section A the method of calculation is outlined and the preparation of data for two group problems is described. The spatial separation of two-dimensional equations is considered in section B. Section C covers the multigroup equations. These parts are self contained and include all information required for the use of the program. Details of the numerical methods are given in section D. Three sample problems are solved in section E. Punching and operating instructions are given in an appendix

  18. Proposal to extend CSEWG neutron and photon multigroup structures for wider applications

    International Nuclear Information System (INIS)

    LaBauve, R.J.; Wilson, W.B.

    1976-02-01

    The 239-group neutron multigroup structure recommended by the Codes and Formats Subcommittee of the cross section evaluation working group (CSEWG) for use in LMFBR design is not well suited for application in certain other areas, particularly thermal reactor design. This report describes a proposal for a neutron group structure consisting of 347 groups, which is an extension of the CSEWG group structure into the thermal range, and also includes more detail in other energy ranges important in LWR, HTGR, GCFR, and CTR design. Similarly, a proposed extension of the CSEWG 94-group photon multigroup structure to 103 groups is described. A subset of the neutron multigroup structure, consisting of 154 groups and for use in power reactor studies, is also presented

  19. Application of direct discrete method (DDM) to multigroup neutron transport problems

    International Nuclear Information System (INIS)

    Vosoughi, Naser; Salehi, Ali Akbar; Shahriari, Majid

    2003-01-01

    The Direct Discrete Method (DDM), which produced excellent results for one-group neutron transport problems, has been developed for multigroup energy. A multigroup neutron transport discrete equation has been produced for a cylindrical shape fuel element with and without associated coolant regions with two boundary conditions. The calculations are illustrated for two-group energy by graphs showing the fast and thermal fluxes. The validity of the results are tested against the results obtained by the ANISN code. (author)

  20. Report from AmSECT's International Consortium for Evidence-Based Perfusion: American Society of Extracorporeal Technology Standards and Guidelines for Perfusion Practice: 2013.

    Science.gov (United States)

    Baker, Robert A; Bronson, Shahna L; Dickinson, Timothy A; Fitzgerald, David C; Likosky, Donald S; Mellas, Nicholas B; Shann, Kenneth G

    2013-09-01

    One of the roles of a professional society is to develop standards and guidelines of practice as an instrument to guide safe and effective patient care. The American Society of Extracorporeal Technology (AmSECT) first published its Essentials for Perfusion Practice, Clinical Function: Conduct of Extracorporeal Circulation in 1993. The International Consortium for Evidence-Based Perfusion (ICEBP), a committee within AmSECT, was tasked with updating this document in 2010. The aim of this report is to describe the method of development and content of AmSECT's new professional standards and guidelines. The ICEBP committee independently evaluated and provided input regarding the current "Essentials and Guidelines." Structural changes were made to the entire document, and a draft document was developed, presented, and circulated to the AmSECT Board of Directors and broader membership for comment. Informed by these reviews, a revised document was then presented to the Society for a membership vote. The final document consists of 15 areas of practice covered by 50 Standards and 38 Guidelines (see Appendix 1) with the first standard focusing on the development of institutional protocols to support their implementation and use. A majority of the membership voted to accept the document (81.2% of the voting membership accepting, 18.8% rejecting). After an audit of the balloting process by AmSECT's Ethics Committee, the results were reported to the membership and the document was officially adopted on July 24, 2013. The Standards and Guidelines will serve as a useful guide for cardiac surgical teams that wish to develop institution-specific standards and guidelines to improve the reliability, safety, and effectiveness of adult cardiopulmonary bypass. The ICEBP recognizes that the development of a Standards and Guidelines statement alone will not change care. Safe, reliable, and effective care will be best served through the development and implementation of institutional

  1. Phylogenetic analyses of Begonia sect. Coelocentrum and allied limestone species of China shed light on the evolution of Sino-Vietnamese karst flora.

    Science.gov (United States)

    Chung, Kuo-Fang; Leong, Wai-Chao; Rubite, Rosario Rivera; Repin, Rimi; Kiew, Ruth; Liu, Yan; Peng, Ching-I

    2014-12-01

    The picturesque limestone karsts across the Sino-Vietnamese border are renowned biodiversity hotspot, distinguished for extremely high endemism of calciphilous plants restricted to caves and cave-like microhabitats that have functioned as biological refugia on the otherwise harsh habitats. To understand evolutionary mechanisms underlying the splendid limestone flora, dated phylogeny is reconstructed for Asian Begonia, a species-rich genus on limestone substrates represented by no less than 60 species in southern China, using DNA sequences of nrITS and chloroplast rpL16 intron. The sampling includes 94 Begonia species encompassing most major Asian clades with a special emphasized on Chinese species. Except for two tuberous deciduous species and a species with upright stems, a majority of Sino-Vietnamese limestone Begonia (SVLB), including sect. Coelocentrum (19 species sampled) and five species of sect. Diploclinium, Leprosae, and Petermannia, are rhizomatous and grouped in a strongly supported and yet internally poorly resolved clade (Clade SVLB), suggesting a single evolutionary origin of the adaptation to limestone substrates by rhizomatous species, subsequent species radiation, and a strong tendency to retain their ancestral niche. Divergence-time estimates indicate a late Miocene diversification of Clade SVLB, coinciding with the onset of the East Asian monsoon and the period of extensive karstification in the area. Based on our phylogenetic study, Begonia sect. Coelocentrum is recircumscribed and expanded to include other members of the Clade SVLB (sect. Diploclinium: B. cavaleriei, B. pulvinifera, and B. wangii; sect. Leprosae: B. cylindrica and B. leprosa; sect. Petermannia: B. sinofloribunda). Because species of Clade SVLB have strong niche conservatism to retain in their ancestral habitats in cave-like microhabitats and Begonia are generally poor dispersers prone to diversify allopatrically, we propose that extensive and continuous karstification of the

  2. MVP/GMVP II, MC Codes for Neutron and Photon Transport Calc. based on Continuous Energy and Multigroup Methods

    International Nuclear Information System (INIS)

    2005-01-01

    A - Description of program or function: (1) Problems to be solved: MVP/GMVP can solve eigenvalue and fixed-source problems. The multigroup code GMVP can solve forward and adjoint problems for neutron, photon and neutron-photon coupled transport. The continuous-energy code MVP can solve only the forward problems. Both codes can also perform time-dependent calculations. (2) Geometry description: MVP/GMVP employs combinatorial geometry to describe the calculation geometry. It describes spatial regions by the combination of the 3-dimensional objects (BODIes). Currently, the following objects (BODIes) can be used. - BODIes with linear surfaces: half space, parallelepiped, right parallelepiped, wedge, right hexagonal prism; - BODIes with quadratic surface and linear surfaces: cylinder, sphere, truncated right cone, truncated elliptic cone, ellipsoid by rotation, general ellipsoid; - Arbitrary quadratic surface and torus. The rectangular and hexagonal lattice geometry can be used to describe the repeated geometry. Furthermore, the statistical geometry model is available to treat coated fuel particles or pebbles for high temperature reactors. (3) Particle sources: The various forms of energy-, angle-, space- and time-dependent distribution functions can be specified. (4) Cross sections: The ANISN-type PL cross sections or the double-differential cross sections can be used in the multigroup code GMVP. On the other hand, the specific cross section libraries are used in the continuous-energy code MVP. The libraries are generated from the evaluated nuclear data (JENDL-3.3, ENDF/B-VI, JEF-3.0 etc.) by using the LICEM code. The neutron cross sections in the unresolved resonance region are described by the probability table method. The neutron cross sections at arbitrary temperatures are available for MVP by just specifying the temperatures in the input data. (5) Boundary conditions: Vacuum, perfect reflective, isotropic reflective (white), periodic boundary conditions can be

  3. Problem Oriented Neutron-Gamma Cross Sections Libraries for WWER-440 and WWER-1000 Shielding and Reactor Vessel Dosimetry Application

    International Nuclear Information System (INIS)

    Belousov, S.; Antonov, S.; Ilieva, K.

    1997-01-01

    The 47 neutron and 20 gamma group libraries BGL-440 and BGL-1000 for the shielding and reactor vessel dosimetry application have been generated for WWER-440 and WWER-1000 by collapsing the VITAMIN-B6 library (199 neutron and 42 gamma groups on the base of ENDF/B-6). The first parts of the libraries for neutron-gamma transport calculation, BGL-440-1 (150 nuclides) and BGL-1000-1 (140 nuclides), have been generated by a modified version of SAS1X control module of the SCALE system. The appropriate zone-average neutron flux had been used for these sub-libraries collapsing. The BGL-440-2 and BGL-1000-2 sub-libraries consist of cross sections for all 120 nuclides of VITAMIN-B6, for calculation of the transport through non-reactor materials of dosimeters, capsules, specimens which may be placed in the cavity behind the reactor vessel. The neutron spectrum just beyond the RPV had been used for this collapsing. As the first test the comparative calculations of the neutron flux on/behind the WWER-1000 reactor vessel have been realised using the libraries BGL-1000 and BUGLE, intended for the American PWR reactors. The integral neutron flux values by BGL-1000 and BUGLE differ by 3% onto the vessel, and 5% behind the vessel. This result shows that the calculations of the neutron flux responses for the WWER vessel surveillance, especially in locations behind the WWER vessel have to be done by the appropriate BGL library. Key words: neutron transport, multigroup neutron cross section libraries

  4. Multigroup or multipoint thermal neutron data preparation. Programme SIGMA

    International Nuclear Information System (INIS)

    Matausek, M.V.; Kunc, M.

    1974-01-01

    When calculating the space energy distribution of thermal neutrons in reactor lattices, in either the multigroup or the multipoint approximation, it is convenient to divide the problem into two independent parts. Firstly, for all material regions of the given reactor lattice cell, the group or the point values of cross sections, scattering kernel and the outer source of thermal neutrons are calculated by a data preparation programme. These quantities are then used as input, by the programme which solves multigroup or multipoint transport equations, to generate the space energy neutron spectra in the cell considered and to determine the related integral quantities, namely the different reaction rates. The present report deals with the first part of the problem. An algorithm for constructing a set of thermal neutron input data, to be used with the multigroup or multipoint version of the code MULTI /1,2,3/, is presented and the new version of the programme SIGMA /4/, written in FORTRAN IV for the CDC-3600 computer, is described. For a given reactor cell material, composed of a number of different isotopes, this programme calculates the group or the point values of the scattering macroscopic absorption cross section, macroscopic scattering cross section, kernel and the outer source of thermal neutrons. Numerous options are foreseen in the programme, concerning the energy variation of cross sections and a scattering kernel, concerning the weighting spectrum in multigroup scheme or the procedure for constructing the scattering matrix in the multipoint scheme and, finally, concerning the organization of output. The details of the calculational algorithm are presented in Section 2 of the paper. Section 3 contains the description of the programme and the instructions for its use (author)

  5. Comparison of WIMS results using libraries based on new evaluated data files

    International Nuclear Information System (INIS)

    Trkov, A.; Ganesan, S.; Zidi, T.

    1996-01-01

    A number of selected benchmark experiments have been modelled with the WIMS-D/4 lattice code. Calculations were performed using multigroup libraries generated from a number of newly released evaluated data files. Data processing was done with the NJOY91.38 code. Since the data processing methods were the same in all cases, the results may serve to determine the impact on integral parameters due to differences in the basic data. The calculated integral parameters were also compared to the measured values. Observed differences were small, which means that there are no essential differences between the evaluated data libraries. The results of the analysis cannot serve to discriminate in terms of quality of the data between the evaluated data libraries considered. For the test cases considered the results with the new, unadjusted libraries are at least as good as those obtained with the old, adjusted WIMS library which is supplied with the code. (author). 16 refs, 3 tabs

  6. Final report [on solving the multigroup diffusion equations

    International Nuclear Information System (INIS)

    Birkhoff, G.

    1975-01-01

    Progress achieved in the development of variational methods for solving the multigroup neutron diffusion equations is described. An appraisal is made of the extent to which improved variational methods could advantageously replace difference methods currently used

  7. Numerical analysis for multi-group neutron-diffusion equation using Radial Point Interpolation Method (RPIM)

    International Nuclear Information System (INIS)

    Kim, Kyung-O; Jeong, Hae Sun; Jo, Daeseong

    2017-01-01

    Highlights: • Employing the Radial Point Interpolation Method (RPIM) in numerical analysis of multi-group neutron-diffusion equation. • Establishing mathematical formation of modified multi-group neutron-diffusion equation by RPIM. • Performing the numerical analysis for 2D critical problem. - Abstract: A mesh-free method is introduced to overcome the drawbacks (e.g., mesh generation and connectivity definition between the meshes) of mesh-based (nodal) methods such as the finite-element method and finite-difference method. In particular, the Point Interpolation Method (PIM) using a radial basis function is employed in the numerical analysis for the multi-group neutron-diffusion equation. The benchmark calculations are performed for the 2D homogeneous and heterogeneous problems, and the Multiquadrics (MQ) and Gaussian (EXP) functions are employed to analyze the effect of the radial basis function on the numerical solution. Additionally, the effect of the dimensionless shape parameter in those functions on the calculation accuracy is evaluated. According to the results, the radial PIM (RPIM) can provide a highly accurate solution for the multiplication eigenvalue and the neutron flux distribution, and the numerical solution with the MQ radial basis function exhibits the stable accuracy with respect to the reference solutions compared with the other solution. The dimensionless shape parameter directly affects the calculation accuracy and computing time. Values between 1.87 and 3.0 for the benchmark problems considered in this study lead to the most accurate solution. The difference between the analytical and numerical results for the neutron flux is significantly increased in the edge of the problem geometry, even though the maximum difference is lower than 4%. This phenomenon seems to arise from the derivative boundary condition at (x,0) and (0,y) positions, and it may be necessary to introduce additional strategy (e.g., the method using fictitious points and

  8. TWOTRAN-2, 2-D Multigroup Transport in X-Y, R-Z, R-Theta Geometry with Anisotropic Scattering

    International Nuclear Information System (INIS)

    Lathrop, K.D.; Brinkley, F.W.

    1995-01-01

    1 - Description of problem or function: TWOTRAN2 solves the two-dimensional multigroup transport equation in (x,y), (r,theta), and (r,z) geometries. Both regular and adjoint, inhomogeneous and homogeneous (k eff and eigenvalue searches) problems subject to vacuum, reflective, periodic, white or input-specified boundary flux conditions are solved. General anisotropic scattering is allowed and anisotropic inhomogeneous sources are permitted. 2 - Method of solution: The discrete ordinates approximation for the angular variable is used in finite difference form which is solved with the central (diamond) difference approximation. Negative fluxes are eliminated by a local set-to zero and correct algorithm. Standard inner (within-group) and outer iterative cycles are accelerated by a coarse-mesh re-balancing on a coarse mesh which may be independent of the material mesh. 3 - Restrictions on the complexity of the problem: Variable dimensioning is used so that any combination of problem parameters leading to a container array less than MAXLEN can be accommodated. On IBM machines, TWOTRAN2 will execute in the 4-byte mode so that any combination of problem parameters leading to a container array less than MAXLEN can be accommodated. MAXLEN can be several hundred thousand and most problems can be core-contained. On the CDC machines MAXLEN can be slightly greater than 40,000 words and peripheral storage is used for most group-dependent data

  9. Generation of multigroup cross sections from ENDF/B-IV nuclear data library

    International Nuclear Information System (INIS)

    Chapot, J.L.C.; Thome Filho, Z.D.

    1980-04-01

    The generation of nuclear data compacted in energy groups is made. The nuclear data library ENDF/B-IV, Evaluated Nuclear Data File, and the new version of the codes ETOG-3 and ETOT-3 are utilized. The data obtained are compared with data from other sources. (L.F.) [pt

  10. Proposal to extend CSEWG neutron and photon multigroup structures for wider applications. [Tables

    Energy Technology Data Exchange (ETDEWEB)

    LaBauve, R.J.; Wilson, W.B.

    1976-02-01

    The 239-group neutron multigroup structure recommended by the Codes and Formats Subcommittee of the cross section evaluation working group (CSEWG) for use in LMFBR design is not well suited for application in certain other areas, particularly thermal reactor design. This report describes a proposal for a neutron group structure consisting of 347 groups, which is an extension of the CSEWG group structure into the thermal range, and also includes more detail in other energy ranges important in LWR, HTGR, GCFR, and CTR design. Similarly, a proposed extension of the CSEWG 94-group photon multigroup structure to 103 groups is described. A subset of the neutron multigroup structure, consisting of 154 groups and for use in power reactor studies, is also presented.

  11. Corolla morphology influences diversification rates in bifid toadflaxes (Linaria sect. Versicolores)

    Science.gov (United States)

    Fernández-Mazuecos, Mario; Blanco-Pastor, José Luis; Gómez, José M.; Vargas, Pablo

    2013-01-01

    Background and Aims The role of flower specialization in plant speciation and evolution remains controversial. In this study the evolution of flower traits restricting access to pollinators was analysed in the bifid toadflaxes (Linaria sect. Versicolores), a monophyletic group of ∼30 species and subspecies with highly specialized corollas. Methods A time-calibrated phylogeny based on both nuclear and plastid DNA sequences was obtained using a coalescent-based method, and flower morphology was characterized by means of morphometric analyses. Directional trends in flower shape evolution and trait-dependent diversification rates were jointly analysed using recently developed methods, and morphological shifts were reconstructed along the phylogeny. Pollinator surveys were conducted for a representative sample of species. Key Results A restrictive character state (narrow corolla tube) was reconstructed in the most recent common ancestor of Linaria sect. Versicolores. After its early loss in the most species-rich clade, this character state has been convergently reacquired in multiple lineages of this clade in recent times, yet it seems to have exerted a negative influence on diversification rates. Comparative analyses and pollinator surveys suggest that the narrow- and broad-tubed flowers are evolutionary optima representing divergent strategies of pollen placement on nectar-feeding insects. Conclusions The results confirm that different forms of floral specialization can lead to dissimilar evolutionary success in terms of diversification. It is additionally suggested that opposing individual-level and species-level selection pressures may have driven the evolution of pollinator-restrictive traits in bifid toadflaxes. PMID:24142920

  12. Space synthesis: an application of synthesis method to two and three dimensional multigroup neutron diffusion equations; Synthese spatiale: une application de la methode de synthese aux equations de diffusion neutronique multigroupe a deux et trois dimensions

    Energy Technology Data Exchange (ETDEWEB)

    Nguyen-Ngoc, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1969-07-01

    In order to reduce computing time, two and three-dimensional multigroup neutron diffusion equations in cylindrical, rectangular (X, Y), (X, Y, Z) and hexagonal geometries are solved by the method of synthesis using an appropriate variational principle (stationary principle). The basic idea is to reduce the number of independent variables by constructing two or three-dimensional solutions from solutions of fewer variables, hence the name 'synthesis method'. Whatever the geometry, we are led to solve a system of ordinary differential equations with matrix coefficients to which one can apply well-known numerical methods: CHEBYSHEV's polynomial method, Gaussian elimination. Numerical results furnished by synthesis programs written for the IBM 7094, the IBM 360-75 and the CDC 6600 computers, are confronted with those which are given by programs employing the classical finite difference method. [French] En vue de reduire le-temps de calcul, les equations de diffusion neutronique, multigroupe, a deux et trois dimensions d'espace dans les geometries cylindrique, rectangulaire (X, Y), (X, Y, Z) et hexagonale sont resolues par la methode de synthese utilisant un principe variationnel approprie (principe stationnaire). L'idee consiste a reduire le nombre de variables independantes par construction d'une solution bi ou tridimensionnelle au moyen de solutions dependant d'un nombre inferieur de variables, d'ou le nom de la methode. Dans tous les cas de geometrie, nous sommes conduits a resoudre un systeme d'equations differentielles a coefficients matriciels auquel peuvent s'appliquer les methodes numeriques courantes; methode polynomiale de TCHEBYCHEFF et methode d'elimination de GAUSS. Les resultats numeriques obtenus par nos codes de synthese programmes sur IBM 7094, IBM 360-75 et CDC 6600, sont confrontes avec ceux que fournissent les programmes adoptant la methode classique des differences finies. (auteur)

  13. Status of standard cross section library and future plan

    International Nuclear Information System (INIS)

    Zukeran, Atsushi

    2001-01-01

    JSSTDL-300 multi-group cross section library with 300 neutron energy groups coupled with 104 group γ-ray cross sections was developed for general users in nuclear reactor physics and/or design, whose source data is the evaluated nuclear data library JENDL-3.2. For the purpose of a standard or common use, several famous cross section libraries worldwide used, i.e., ABBN-25, GAM-123, VITAMIN-C/J(E+C), MGCL-137, BERMUDA-12 and FNS-125 for neutron, and LANL-12, -24-, -48, and CSEWG-94 for γ-ray, are consulted about setting the common energy group structure. Furthermore, in order to expand the applicability, the top energy is set on 20 MeV and the lowest energy is 10 -5 eV. In the thermal neutron energy region, the JSSTDL-300 has about 20 energy groups. Besides, many utility codes for group collapsing and for data format transformation are provided for general users. (author)

  14. Pollen morphology of the Onobrychis sect. Heliobrychis in Iran

    Directory of Open Access Journals (Sweden)

    Roya Karamian

    2010-01-01

    Full Text Available Pollen morphology of 20 species belonging to Onobrychis sect. Heliobrychis was studied in Iran. The pollen grains from the herbarium specimens were prepared by acetolysis method and then 8 quantitative characters were studied by light microscopy. Data analysis was carried out by cluster analysis and principal component analysis methods. Pollen grains of the sections were 3-colpate with reticulate ornamentation. Results of the study showed that although the pollen grains in the section were comparatively homogenous, but had a considerable variation in the studied quantitative characters. However among the studied characters, equatorial and polar diameters of pollen grains were the most significant characters. The members of the section can be divided into 3 groups based on these characters.

  15. An energy recondensation method using the discrete generalized multigroup energy expansion theory

    International Nuclear Information System (INIS)

    Zhu Lei; Forget, Benoit

    2011-01-01

    Highlights: → Discrete-generalized multigroup method was implemented as a recondensation scheme. → Coarse group cross-sections were recondensed from core-level solution. → Neighboring effect of reflector and MOX bundle was improved. → Methodology was shown to be fully consistent when a flat angular flux approximation is used. - Abstract: In this paper, the discrete generalized multigroup (DGM) method was used to recondense the coarse group cross-sections using the core level solution, thus providing a correction for neighboring effect found at the core level. This approach was tested using a discrete ordinates implementation in both 1-D and 2-D. Results indicate that 2 or 3 iterations can substantially improve the flux and fission density errors associated with strong interfacial spectral changes as found in the presence of strong absorbers, reflector of mixed-oxide fuel. The methodology is also proven to be fully consistent with the multigroup methodology as long as a flat-flux approximation is used spatially.

  16. Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system

    International Nuclear Information System (INIS)

    Yang, W.S.; Lee, C.H.

    2008-01-01

    Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC 2 -2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC 2 -2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC 2 -2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC 2 -2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC 2 -2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC 2 -2, VIM, and NJOY. For almost all nuclides considered, MC 2 -2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC 2 -2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC 2 -2/TWODANT calculations were in good agreement with MCNP solutions within ∼0.25% Δρ, except a few small LANL fast assemblies. Relative to the MCNP solution, the MC 2 -2/TWODANT

  17. Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

    Energy Technology Data Exchange (ETDEWEB)

    Yang, W. S.; Lee, C. H. (Nuclear Engineering Division)

    2008-05-16

    Under the fast reactor simulation program launched in April 2007, development of an advanced multigroup cross section generation code was initiated in July 2007, in conjunction with the development of the high-fidelity deterministic neutron transport code UNIC. The general objectives are to simplify the existing multi-step schemes and to improve the resolved and unresolved resonance treatments. Based on the review results of current methods and the fact that they have been applied successfully to fast critical experiment analyses and fast reactor designs for last three decades, the methodologies of the ETOE-2/MC{sup 2}-2/SDX code system were selected as the starting set of methodologies for multigroup cross section generation for fast reactor analysis. As the first step for coupling with the UNIC code and use in a parallel computing environment, the MC{sup 2}-2 code was updated by modernizing the memory structure and replacing old data management package subroutines and functions with FORTRAN 90 based routines. Various modifications were also made in the ETOE-2 and MC{sup 2}-2 codes to process the ENDF/B-VII.0 data properly. Using the updated ETOE-2/MC{sup 2}-2 code system, the ENDF/B-VII.0 data was successfully processed for major heavy and intermediate nuclides employed in sodium-cooled fast reactors. Initial verification tests of the MC{sup 2}-2 libraries generated from ENDF/B-VII.0 data were performed by inter-comparison of twenty-one group infinite dilute total cross sections obtained from MC{sup 2}-2, VIM, and NJOY. For almost all nuclides considered, MC{sup 2}-2 cross sections agreed very well with those from VIM and NJOY. Preliminary validation tests of the ENDF/B-VII.0 libraries of MC{sup 2}-2 were also performed using a set of sixteen fast critical benchmark problems. The deterministic results based on MC{sup 2}-2/TWODANT calculations were in good agreement with MCNP solutions within {approx}0.25% {Delta}{rho}, except a few small LANL fast assemblies

  18. FENDL-3.0: Processing the Evaluated Nuclear Data Library for Fusion Applications

    International Nuclear Information System (INIS)

    Lopez Aldama, D.; Noy, R. Capote

    2011-12-01

    A description of the work undertaken towards the development of a new version of the neutron-induced part of the Fusion Evaluated Nuclear Data Library (FENDL) for applications is summarized. The main issues related to the selection and processing of evaluated nuclear data files using the NJOY-99 and PREPRO-2010 processing systems are described. The new version of FENDL for applications, termed FENDL-3.0, includes the evaluated nuclear data files in ENDF-6 format, the continuous-energy cross section files in ACE format for the MCNP family of Monte Carlo codes and the multi-group data library in MATXS format for deterministic transport calculations up to 55 MeV for 180 isotopes. Further, additional data are supplied in GENDF format for sensitivity studies. The library is freely available from the Nuclear Data Section at the International Atomic Energy Agency. (author)

  19. A Validated MCNP(X) Cross Section Library based on JEFF 3.1

    International Nuclear Information System (INIS)

    Haeck, W.; Verboomen, B.

    2006-01-01

    ALEPH-LIB is a multi-temperature neutron transport library for standard use by MCNP(X) and ALEPH generated with ALEPH-DLG. This is an auxiliary computer code to ALEPH, the Monte Carlo burn-up code under development at SCK-CEN in collaboration with Ghent university. ALEPH-DLG automates the entire process of generating library files with NJOY and takes care of the first requirement of a validated application library: verify the processing. It produces tailor made NJOY input files using data from the original ENDF file (initial temperature, the fact if the nuclide is fissile or if it has unresolved resonances, etc.) When the library files have been generated, ALEPH-DLG will also process the output from NJOY by extracting all messages and warnings. If ALEPH-DLG finds anything out of the ordinary, it will either warn the user or perform corrective actions. The temperatures included in the ALEPH-LIB library are 300, 600, 900, 1200, 1500 and 1800 K. Library files were produced for the JEF 2.2, JEFF 3.0, JEFF 3.1, JENDL 3.3 and ENDF/B-VI.8 nuclear data libraries. This will be extended with ENDF/B-VII when it becomes available. This report deals with the JEFF 3.1 files included in ALEPH-LIB that are now released by the NEA-OECD.

  20. A Validated MCNP(X) Cross Section Library based on JEFF 3.1

    Energy Technology Data Exchange (ETDEWEB)

    Haeck, W; Verboomen, B

    2006-10-15

    ALEPH-LIB is a multi-temperature neutron transport library for standard use by MCNP(X) and ALEPH generated with ALEPH-DLG. This is an auxiliary computer code to ALEPH, the Monte Carlo burn-up code under development at SCK-CEN in collaboration with Ghent university. ALEPH-DLG automates the entire process of generating library files with NJOY and takes care of the first requirement of a validated application library: verify the processing. It produces tailor made NJOY input files using data from the original ENDF file (initial temperature, the fact if the nuclide is fissile or if it has unresolved resonances, etc.) When the library files have been generated, ALEPH-DLG will also process the output from NJOY by extracting all messages and warnings. If ALEPH-DLG finds anything out of the ordinary, it will either warn the user or perform corrective actions. The temperatures included in the ALEPH-LIB library are 300, 600, 900, 1200, 1500 and 1800 K. Library files were produced for the JEF 2.2, JEFF 3.0, JEFF 3.1, JENDL 3.3 and ENDF/B-VI.8 nuclear data libraries. This will be extended with ENDF/B-VII when it becomes available. This report deals with the JEFF 3.1 files included in ALEPH-LIB that are now released by the NEA-OECD.

  1. Nuclear data libraries for Tripoli-3.5 code

    International Nuclear Information System (INIS)

    Vergnaud, Th.

    2001-01-01

    The TRIPOLI-3 code uses multigroup nuclear data libraries generated using the NJOY-THEMIS suite of modules: for neutrons, they are produced from the ENDF/B-VI evaluations and cover the range between 20 MeV and 10 -5 eV, either in 315 groups and for one temperature, or in 3209 groups and for five temperatures; for gamma-rays, they are from JEF2 and are processed in groups between 14 MeV and keV. The probability tables used for the neutron transport calculations have been derived from the ENDF/B-VI evaluations using the CALENDF code. Cross sections for gamma production by neutron interaction (fission, capture or inelastic scattering) have been derived from ENDF/B-VI in 315 neutron groups and 75 gamma groups. The code also uses two response function libraries: for neutrons; based on several sources, in particular the dosimetry libraries IRDF/85 and IRDF/90; for gamma-rays it is based on the JEF2 evaluation and contains the kerma factors for all the elements and cross sections for all interactions. (author)

  2. Multigroup neutron transport equation in the diffusion and P{sub 1} approximation

    Energy Technology Data Exchange (ETDEWEB)

    Obradovic, D [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1970-07-01

    Investigations of the properties of the multigroup transport operator, width and without delayed neutrons in the diffusion and P{sub 1} approximation, is performed using Keldis's theory of operator families as well as a technique . recently used for investigations into the properties of the general linearized Boltzmann operator. It is shown that in the case without delayed neutrons, multigroup transport operator in the diffusion and P{sub 1} approximation possesses a complete set of generalized eigenvectors. A formal solution to the initial value problem is also given. (author)

  3. Complex of two-dimensional multigroup programs for neutron-physical computations of nuclear reactor

    International Nuclear Information System (INIS)

    Karpov, V.A.; Protsenko, A.N.

    1975-01-01

    Briefly stated mathematical aspects of the two-dimensional multigroup method of neutron-physical computation of nuclear reactor. Problems of algorithmization and BESM-6 computer realisation of multigroup diffuse approximations in hexagonal and rectangular calculated lattices are analysed. The results of computation of fast critical assembly having complicated composition of the core are given. The estimation of computation accuracy of criticality, neutron fields distribution and efficiency of absorbing rods by means of computer programs developed is done. (author)

  4. Neutronic calculations in heavy water moderated multiplying media using GGC-3 library nuclear data

    International Nuclear Information System (INIS)

    Boado, H.J.; Gho, C.J.; Abbate, M.J.

    1981-01-01

    Differences in obtaining transference matrices between GGC-3 code and the system to produce multigroup cross sections using GGC-3 library, recently implemented at the Neutrons and Reactors Division, have been analized. Neutronic calculations in multiplicative systems containing heavy water have been made using both methods. From the obtained results, it is concluded that the new method is more appropriate to deal with systems including moderators other than light water. (author) [es

  5. Nigeria and the Boko Haram Sect: Adopting a Better Strategy for Resolving the Crisis

    Science.gov (United States)

    2013-06-01

    ed. Thomas A. Imobighe and Agatha N. T. Eguavoen (Ibadan Nigeria: Heinemann Education Books PLC, 2006), 7; Benjamin Maiangwa and Ufo O. Uzodike, “The...sect were lined up and shot in the back. 111 Hakeem Onapajo and Ufo O. Uzodike, “Boko Haram Terrorism in Nigeria: Man, the State, and the...of Management and Technology, 2008. Maiangwa, Benjamin and Ufo O. Uzodike. “The Changing Dynamics of Boko Haram Terrorism.” Aljazeera Centre for

  6. Multi-level nonlinear diffusion acceleration method for multigroup transport k-Eigenvalue problems

    International Nuclear Information System (INIS)

    Anistratov, Dmitriy Y.

    2011-01-01

    The nonlinear diffusion acceleration (NDA) method is an efficient and flexible transport iterative scheme for solving reactor-physics problems. This paper presents a fast iterative algorithm for solving multigroup neutron transport eigenvalue problems in 1D slab geometry. The proposed method is defined by a multi-level system of equations that includes multigroup and effective one-group low-order NDA equations. The Eigenvalue is evaluated in the exact projected solution space of smallest dimensionality, namely, by solving the effective one- group eigenvalue transport problem. Numerical results that illustrate performance of the new algorithm are demonstrated. (author)

  7. International benchmark tests of the FENDL-1 Nuclear Data Library

    International Nuclear Information System (INIS)

    Fischer, U.

    1997-01-01

    An international benchmark validation task has been conducted to validate the fusion evaluated nuclear data library FENDL-1 through data tests against integral 14 MeV neutron experiments. The main objective of this task was to qualify the FENDL-1 working libraries for fusion applications and to elaborate recommendations for further data improvements. Several laboratories and institutions from the European Union, Japan, the Russian Federation and US have contributed to the benchmark task. A large variety of existing integral 14 MeV benchmark experiments was analysed with the FENDL-1 working libraries for continuous energy Monte Carlo and multigroup discrete ordinate calculations. Results of the benchmark analyses have been collected, discussed and evaluated. The major findings, conclusions and recommendations are presented in this paper. With regard to the data quality, it is summarised that fusion nuclear data have reached a high confidence level with the available FENDL-1 data library. With few exceptions this holds for the materials of highest importance for fusion reactor applications. As a result of the performed benchmark analyses, some existing deficiencies and discrepancies have been identified that are recommended for removal in theforthcoming FENDL-2 data file. (orig.)

  8. Fruit anatomy of species of Solanum sect. Torva (Solanaceae Anatomía del fruto en especies de Solanum sect. Torva (Solanaceae

    Directory of Open Access Journals (Sweden)

    Franco E. Chiarini

    2010-12-01

    Full Text Available The mature fruits of 10 South American species of Solanum sect. Torva were studied. Cross and longitudinal microtome sections, stained with astra blue/basic fuchsin, were made for microscopic examination. All species present an epidermis formed by a unistrate layer of small, isodiametric cells, with dense content and cellulosic walls. Immediately below, a hypodermis is always found, consisting of a well-defined layer of lignified cells with a single calcium oxalate crystal occupying the whole lumen of each cell. This is followed by one layer of cellulosic, isodiametric cells with dense cytoplasm and then several collenchymatous layers, sometimes with sclerified cell walls. The mesocarp comprises two zones histologically differentiated: an external one (formed by regular, vacuolated, medium-sized cells with small intercellular spaces, and an internal one, commonly juicy, and developing proliferations among the seeds. The fruits analyzed are alike, and despite some particularities, they can be classified as berries in the conventional sense. All the traits examined agree with the ornithochorous dispersal syndrome. The homogeneity in fruit traits may be due to shared habit, habitat and sexual system.Se estudiaron los frutos maduros de 10 especies sudamericanas de Solanum sect. Torva. Se examinaron en microscopio cortes microtómicos transversales y longitudinales teñidos con azul astral/fucsina básica. Todas las especies presentaron una epidermis unistrata de células pequeñas, isodiamétricas, de contenido denso y paredes celulósicas. Inmediatamente por debajo se encontró siempre una hipodermis, formada por una capa bien definida de células lignificadas con un cristal de oxalato de calcio en el lúmen de cada célula. A continuación se halló otra capa de celulas isodiamétricas, celulósicas, de contenido denso, y luego varias capas de colénquima, en ocasiones con paredes esclerificadas. El mesocarpo presentó dos zonas histologicamente

  9. FINELM: a multigroup finite element diffusion code. Part II

    International Nuclear Information System (INIS)

    Davierwalla, D.M.

    1981-05-01

    The author presents the axisymmetric case in cylindrical coordinates for the finite element multigroup neutron diffusion code, FINELM. The numerical acceleration schemes incorporated viz. the Lebedev extrapolations and the coarse mesh rebalancing, space collapsing, are discussed. A few benchmark computations are presented as validation of the code. (Auth.)

  10. Depletion Calculations for MTR Core Using MCNPX and Multi-Group Nodal Diffusion Methods

    International Nuclear Information System (INIS)

    Jaradata, Mustafa K.; Park, Chang Je; Lee, Byungchul

    2013-01-01

    In order to maintain a self-sustaining steady-state chain reaction, more fuel than is necessary in order to maintain a steady state chain reaction must be loaded. The introduction of this excess fuel increases the net multiplication capability of the system. In this paper MCNPX and multi-group nodal diffusion theory will be used for depletion calculations for MTR core. The eigenvalue and power distribution in the core will be compared for different burnup. Multi-group nodal diffusion theory with combination of NEWT-TRITON system was used to perform depletion calculations for 3Χ3 MTR core. 2G and 6G approximations were used and compared with MCNPX results for 2G approximation the maximum difference from MCNPX was 40 mk and for 6G approximation was 6 mk which is comparable to the MCNPX results. The calculated power using nodal code was almost the same MCNPX results. Finally the results of the multi-group nodal theory were acceptable and comparable to the calculated using MCNPX

  11. RZ calculations for self shielded multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z. [Commissariat a l' Energie Atomique CEA, Direction de l' Energie Nucleaire, DEN/DM2S/SERMA/LENR, 91191 Gif-sur-Yvette Cedex (France)

    2006-07-01

    A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

  12. RZ calculations for self shielded multigroup cross sections

    International Nuclear Information System (INIS)

    Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z.

    2006-01-01

    A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)

  13. Development and testing of the VITAMIN-B7/BUGLE-B7 coupled neutron-gamma multigroup cross-section libraries

    Energy Technology Data Exchange (ETDEWEB)

    Risner, J.M.; Wiarda, D.; Miller, T.M.; Peplow, D.E.; Patton, B.W.; Dunn, M.E. [Oak Ridge National Laboratory, MS 6170, P.O. Box 2008, Oak Ridge, TN 37831-6170 (United States); Parks, B.T. [U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Mail Stop O10-B3, 11555 Rockville Pike, Rockville, MD 20852 (United States)

    2011-07-01

    The U.S. Nuclear Regulatory Commission's Regulatory Guide 1.190 states that calculational methods used to estimate reactor pressure vessel (RPV) fluence should use the latest version of the evaluated nuclear data file (ENDF). The VITAMIN-B6 fine-group library and BUGLE-96 broad-group library, which are widely used for RPV fluence calculations, were generated using ENDF/B-VI.3 data, which was the most current data when Regulatory Guide 1.190 was issued. We have developed new fine-group (VITAMIN-B7) and broad-group (BUGLE-B7) libraries based on ENDF/B-VII.0. These new libraries, which were processed using the AMPX code system, maintain the same group structures as the VITAMIN-B6 and BUGLE-96 libraries. Verification and validation of the new libraries were accomplished using diagnostic checks in AMPX, 'unit tests' for each element in VITAMIN-B7, and a diverse set of benchmark experiments including critical evaluations for fast and thermal systems, a set of experimental benchmarks that are used for SCALE regression tests, and three RPV fluence benchmarks. The benchmark evaluation results demonstrate that VITAMIN-B7 and BUGLE-B7 are appropriate for use in RPV fluence calculations and meet the calculational uncertainty criterion in Regulatory Guide 1.190. (authors)

  14. Development and Testing of the VITAMIN-B7/BUGLE-B7 Coupled Neutron-Gamma Multigroup Cross-Section Libraries

    International Nuclear Information System (INIS)

    Risner, Joel M.; Wiarda, Dorothea; Miller, Thomas Martin; Peplow, Douglas E.; Patton, Bruce W.; Dunn, Michael E.; Parks, Benjamin T.

    2011-01-01

    The U.S. Nuclear Regulatory Commission's Regulatory Guide 1.190 states that calculational methods used to estimate reactor pressure vessel (RPV) fluence should use the latest version of the Evaluated Nuclear Data File (ENDF). The VITAMIN-B6 fine-group library and BUGLE-96 broad-group library, which are widely used for RPV fluence calculations, were generated using ENDF/B-VI data, which was the most current data when Regulatory Guide 1.190 was issued. We have developed new fine-group (VITAMIN-B7) and broad-group (BUGLE-B7) libraries based on ENDF/B-VII. These new libraries, which were processed using the AMPX code system, maintain the same group structures as the VITAMIN-B6 and BUGLE-96 libraries. Verification and validation of the new libraries was accomplished using diagnostic checks in AMPX, unit tests for each element in VITAMIN-B7, and a diverse set of benchmark experiments including critical evaluations for fast and thermal systems, a set of experimental benchmarks that are used for SCALE regression tests, and three RPV fluence benchmarks. The benchmark evaluation results demonstrate that VITAMIN-B7 and BUGLE-B7 are appropriate for use in LWR shielding applications, and meet the calculational uncertainty criterion in Regulatory Guide 1.190.

  15. PUFF-IV, Code System to Generate Multigroup Covariance Matrices from ENDF/B-VI Uncertainty Files

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of program or function: The PUFF-IV code system processes ENDF/B-VI formatted nuclear cross section covariance data into multigroup covariance matrices. PUFF-IV is the newest release in this series of codes used to process ENDF uncertainty information and to generate the desired multi-group correlation matrix for the evaluation of interest. This version includes corrections and enhancements over previous versions. It is written in Fortran 90 and allows for a more modular design, thus facilitating future upgrades. PUFF-IV enhances support for resonance parameter covariance formats described in the ENDF standard and now handles almost all resonance parameter covariance information in the resolved region, with the exception of the long range covariance sub-subsections. PUFF-IV is normally used in conjunction with an AMPX master library containing group averaged cross section data. Two utility modules are included in this package to facilitate the data interface. The module SMILER allows one to use NJOY generated GENDF files containing group averaged cross section data in conjunction with PUFF-IV. The module COVCOMP allows one to compare two files written in COVERX format. 2 - Methods: Cross section and flux values on a 'super energy grid,' consisting of the union of the required energy group structure and the energy data points in the ENDF/B-V file, are interpolated from the input cross sections and fluxes. Covariance matrices are calculated for this grid and then collapsed to the required group structure. 3 - Restrictions on the complexity of the problem: PUFF-IV cannot process covariance information for energy and angular distributions of secondary particles. PUFF-IV does not process covariance information in Files 34 and 35; nor does it process covariance information in File 40. These new formats will be addressed in a future version of PUFF

  16. Numerical method for multigroup one-dimensional SN eigenvalue problems with no spatial truncation error

    International Nuclear Information System (INIS)

    Abreu, M.P.; Filho, H.A.; Barros, R.C.

    1993-01-01

    The authors describe a new nodal method for multigroup slab-geometry discrete ordinates S N eigenvalue problems that is completely free from all spatial truncation errors. The unknowns in the method are the node-edge angular fluxes, the node-average angular fluxes, and the effective multiplication factor k eff . The numerical values obtained for these quantities are exactly those of the dominant analytic solution of the S N eigenvalue problem apart from finite arithmetic considerations. This method is based on the use of the standard balance equation and two nonstandard auxiliary equations. In the nonmultiplying regions, e.g., the reflector, we use the multigroup spectral Green's function (SGF) auxiliary equations. In the fuel regions, we use the multigroup spectral diamond (SD) auxiliary equations. The SD auxiliary equation is an extension of the conventional auxiliary equation used in the diamond difference (DD) method. This hybrid characteristic of the SD-SGF method improves both the numerical stability and the convergence rate

  17. FINELM: a multigroup finite element diffusion code. Part I

    International Nuclear Information System (INIS)

    Davierwalla, D.M.

    1980-12-01

    The author presents a two dimensional code for multigroup diffusion using the finite element method. It was realized that the extensive connectivity which contributes significantly to the accuracy, results in a matrix which, although symmetric and positive definite, is wide band and possesses an irregular profile. Hence, it was decided to introduce sparsity techniques into the code. The introduction of the R-Z geometry lead to a great deal of changes in the code since the rotational invariance of the removal matrices in X-Y geometry did not carry over in R-Z geometry. Rectangular elements were introduced to remedy the inability of the triangles to model essentially one dimensional problems such as slab geometry. The matter is discussed briefly in the text in the section on benchmark problems. This report is restricted to the general theory of the triangular elements and to the sparsity techniques viz. incomplete disections. The latter makes the size of the problem that can be handled independent of core memory and dependent only on disc storage capacity which is virtually unlimited. (Auth.)

  18. Calculation of multigroup reaction rates for the Ghana Research ...

    African Journals Online (AJOL)

    The discrete ordinate spatial model, which pro-vides solution to the differential form of the transport equation by the Carlson-SN (N=4) approach was adopted to solve the Ludwig-Boltzmann multigroup neutron transport equation for this analysis. The results show that for any fissile resonance absorber, the reaction rates ...

  19. MINX, Multigroup Cross-Sections and Self-Shielding Factors from ENDF/B for Program SPHINX

    International Nuclear Information System (INIS)

    Soran, P.D.; MacFarlane, R.E.; Harris, D.R.; LaBauve, R.J.; Hendricks, J.S.; Kidman, R.B.; Weisbin, C.R.; White, J.E.

    1977-01-01

    1 - Description of problem or function: MINX calculates fine-group averaged infinitely diluted cross sections and self-shielding factors from ENDF/B-IV data. Its primary purpose is to generate a pseudo-composition-independent multigroup library which is input to the SPHINX space-energy collapse program (2) (PSR-0129) through standard CCCC-III (8) interfaces. MINX incorporates and improves upon the resonance capabilities of existing codes such as ETOX (5) (NESC0388) and ENDRUN (9) and the high-order group-to-group transfer matrices of SUPERTOG (10) (PSR-0013) and ETOG (11). Fine group energy boundaries, Legendre expansion order, gross spectral shape component (in the Bondarenko flux model), temperatures and dilutions can all be used specifically. 2 - Method of solution: Infinitely dilute, un-broadened point cross sections are obtained from resolved resonance parameters using a modified version of the RESEND program (3) (NESC0465). The SIGMA1 (4) (IAEA0854) kernel-broadening method is used to Doppler broaden and thin the tabulated linearized pointwise cross sections at 0 K (outside of the unresolved energy region). Effective temperature- dependent self-shielded pointwise cross sections are derived from the formulation in the ETOX code. The primary modification to the ETOX algorithm is associated with the numerical quadrature scheme used to establish the mean values of the fluctuation intervals. The selection of energy mesh points, at which the effective cross sections are calculated, has been modified to include the energy points given in the ENDF/B file or, if the energy-independent formalism was employed, points at half-lethargy intervals. Infinitely dilute group cross sections and self-shielding factors are generated using the Bondarenko flux weighting model with the gross spectral shape under user control. The integral over energy for each group is divided into a set of panels defined by the union of the grid points describing the total cross section, the

  20. Multigroup computation of the temperature-dependent Resonance Scattering Model (RSM) and its implementation

    Energy Technology Data Exchange (ETDEWEB)

    Ghrayeb, S. Z. [Dept. of Mechanical and Nuclear Engineering, Pennsylvania State Univ., 230 Reber Building, Univ. Park, PA 16802 (United States); Ouisloumen, M. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Ougouag, A. M. [Idaho National Laboratory, MS-3860, PO Box 1625, Idaho Falls, ID 83415 (United States); Ivanov, K. N.

    2012-07-01

    A multi-group formulation for the exact neutron elastic scattering kernel is developed. This formulation is intended for implementation into a lattice physics code. The correct accounting for the crystal lattice effects influences the estimated values for the probability of neutron absorption and scattering, which in turn affect the estimation of core reactivity and burnup characteristics. A computer program has been written to test the formulation for various nuclides. Results of the multi-group code have been verified against the correct analytic scattering kernel. In both cases neutrons were started at various energies and temperatures and the corresponding scattering kernels were tallied. (authors)

  1. ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies. Supplement 1

    Energy Technology Data Exchange (ETDEWEB)

    Wright, R.Q.; Renier, J.P.; Bucholz, J.A.

    1995-08-01

    The original ANSL-V cross-section libraries (ORNL-6618) were developed over a period of several years for the physics analysis of the ANS reactor, with little thought toward including the materials commonly needed for shielding applications. Materials commonly used for shielding applications include calcium barium, sulfur, phosphorous, and bismuth. These materials, as well as {sup 6}Li, {sup 7}Li, and the naturally occurring isotopes of hafnium, have been added to the ANSL-V libraries. The gamma-ray production and gamma-ray interaction cross sections were completely regenerated for the ANSL-V 99n/44g library which did not exist previously. The MALOCS module was used to collapse the 99n/44g coupled library to the 39n/44g broad- group library. COMET was used to renormalize the two-dimensional (2- D) neutron matrix sums to agree with the one-dimensional (1-D) averaged values. The FRESH module was used to adjust the thermal scattering matrices on the 99n/44g and 39n/44g ANSL-V libraries. PERFUME was used to correct the original XLACS Legendre polynomial fits to produce acceptable distributions. The final ANSL-V 99n/44g and 39n/44g cross-section libraries were both checked by running RADE. The AIM module was used to convert the master cross-section libraries from binary coded decimal to binary format (or vice versa).

  2. Multi-group transport methods for high-resolution neutron activation analysis

    International Nuclear Information System (INIS)

    Burns, K. A.; Smith, L. E.; Gesh, C. J.; Shaver, M. W.

    2009-01-01

    The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of multi-group deterministic methods for the simulation of neutron activation problems. Central to this work is the development of a method for generating multi-group neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so that the key signatures in neutron activation analysis (i.e., the characteristic line energies) are preserved. The mechanics of the cross-section preparation method are described and contrasted with standard neutron-gamma cross-section sets. These custom cross-sections are then applied to several benchmark problems. Multi-group results for neutron and photon flux are compared to MCNP results. Finally, calculated responses of high-resolution spectrometers are compared. Preliminary findings show promising results when compared to MCNP. A detailed discussion of the potential benefits and shortcomings of the multi-group-based approach, in terms of accuracy, and computational efficiency, is provided. (authors)

  3. Multi-level methods for solving multigroup transport eigenvalue problems in 1D slab geometry

    International Nuclear Information System (INIS)

    Anistratov, D. Y.; Gol'din, V. Y.

    2009-01-01

    A methodology for solving eigenvalue problems for the multigroup neutron transport equation in 1D slab geometry is presented. In this paper we formulate and compare different variants of nonlinear multi-level iteration methods. They are defined by means of multigroup and effective one-group low-order quasi diffusion (LOQD) equations. We analyze the effects of utilization of the effective one-group LOQD problem for estimating the eigenvalue. We present numerical results to demonstrate the performance of the iteration algorithms in different types of reactor-physics problems. (authors)

  4. Multi-group neutron transport theory

    International Nuclear Information System (INIS)

    Zelazny, R.; Kuszell, A.

    1962-01-01

    Multi-group neutron transport theory. In the paper the general theory of the application of the K. M. Case method to N-group neutron transport theory in plane geometry is given. The eigenfunctions (distributions) for the system of Boltzmann equations have been derived and the completeness theorem has been proved. By means of general solution two examples important for reactor and shielding calculations are given: the solution of a critical and albedo problem for a slab. In both cases the system of singular integral equations for expansion coefficients into a full set of eigenfunction distributions has been reduced to the system of Fredholm-type integral equations. Some results can be applied also to some spherical problems. (author) [fr

  5. Chromosome numbers and karyotype evolution in holoparasitic Orobanche (Orobanchaceae) and related genera

    Science.gov (United States)

    Schneeweiss, G.M.; Palomeque, T.; Colwell, A.E.; Weiss-Schneeweiss, H.

    2004-01-01

    Chromosome numbers and karyotypes of species of Orobanche, Cistanche, and Diphelypaea (Orobanchaceae) were investigated, and 108 chromosome counts of 53 taxa, 19 counted for the first time, are presented with a thorough compilation of previously published data. Additionally, karyotypes of representatives of these genera, including Orobanche sects. Orobanche and Trionychon, are reported. Cistanche (x = 20) has large meta- to submetacentric chromosomes, while those of Diphelypaea (x = 19) are medium-sized submeta-to acrocentrics. Within three analyzed sections of Orobanche, sects. Myzorrhiza (x = 24) and Trionychon (x = 12) possess medium-sized submeta- to acrocentrics, while sect. Orobanche (x = 19) has small, mostly meta- to submetacentric, chromosomes. Polyploidy is unevenly distributed in Orobanche and restricted to a few lineages, e.g., O. sect. Myzorrhiza or Orobanche gracilis and its relatives (sect. Orobanche). The distribution of basic chromosome numbers supports the groups found by molecular phylogenetic analyses: Cistanche has x = 20, the Orobanche-group (Orobanche sect. Orobanche, Diphelypaea) has x = 19, and the Phelipanche-group (Orobanche sects. Gymnocaulis, Myzorrhiza, Trionychon) has x = 12, 24. A model of chromosome number evolution in Orobanche and related genera is presented: from two ancestral base numbers, xh = 5 and xh = 6, independent polyploidizations led to x = 20 (Cistanche) and (after dysploidization) x = 19 (Orobanche-group) and to x = 12 and x = 24 (Phelipanche-group), respectively.

  6. On the calculation of multi-group fission spectrum vectors

    International Nuclear Information System (INIS)

    Mueller, E.Z.

    1984-05-01

    In this report, the problem of calculating fission spectrum vectors in a consistent manner is formulated. The practical implications of using fission spectrum vectors in multi-group transport calculations are also addressed. The significance of the weighting spectra used for the calculation of fission spectrum vectors is illustrated for the case of a simple neutronic assembly

  7. Multi-group diffusion perturbation calculation code. PERKY (2002)

    Energy Technology Data Exchange (ETDEWEB)

    Iijima, Susumu; Okajima, Shigeaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Perturbation calculation code based on the diffusion theory ''PERKY'' is designed for nuclear characteristic analyses of fast reactor. The code calculates reactivity worth on the multi-group diffusion perturbation theory in two or three dimensional core model and kinetics parameters such as effective delayed neutron fraction, prompt neutron lifetime and absolute reactivity scale factor ({rho}{sub 0} {delta}k/k) for FCA experiments. (author)

  8. Resonance self-shielding method using resonance interference factor library for practical lattice physics computations of LWRs

    International Nuclear Information System (INIS)

    Choi, Sooyoung; Khassenov, Azamat; Lee, Deokjung

    2016-01-01

    This paper presents a new method of resonance interference effect treatment using resonance interference factor for high fidelity analysis of light water reactors (LWRs). Although there have been significant improvements in the lattice physics calculations over the several decades, there exist still relatively large errors in the resonance interference treatment, in the order of ∼300 pcm in the reactivity prediction of LWRs. In the newly developed method, the impact of resonance interference to the multi-group cross-sections has been quantified and tabulated in a library which can be used in lattice physics calculation as adjustment factors of multi-group cross-sections. The verification of the new method has been performed with Mosteller benchmark, UO_2 and MOX pin-cell depletion problems, and a 17×17 fuel assembly loaded with gadolinia burnable poison, and significant improvements were demonstrated in the accuracy of reactivity and pin power predictions, with reactivity errors down to the order of ∼100 pcm. (author)

  9. Spectral Green’s function nodal method for multigroup SN problems with anisotropic scattering in slab-geometry non-multiplying media

    International Nuclear Information System (INIS)

    Menezes, Welton A.; Filho, Hermes Alves; Barros, Ricardo C.

    2014-01-01

    Highlights: • Fixed-source S N transport problems. • Energy multigroup model. • Anisotropic scattering. • Slab-geometry spectral nodal method. - Abstract: A generalization of the spectral Green’s function (SGF) method is developed for multigroup, fixed-source, slab-geometry discrete ordinates (S N ) problems with anisotropic scattering. The offered SGF method with the one-node block inversion (NBI) iterative scheme converges numerical solutions that are completely free from spatial truncation errors for multigroup, slab-geometry S N problems with scattering anisotropy of order L, provided L < N. As a coarse-mesh numerical method, the SGF method generates numerical solutions that generally do not give detailed information on the problem solution profile, as the grid points can be located considerably away from each other. Therefore, we describe in this paper a technique for the spatial reconstruction of the coarse-mesh solution generated by the multigroup SGF method. Numerical results are given to illustrate the method’s accuracy

  10. LTFR-4, Library Generated for Fast Reactor Design Program from JAERI Fast-Set Multigroup Constant

    International Nuclear Information System (INIS)

    Suzuki, Tomoo

    1971-01-01

    Nature of physical problem solved: The program processes JAERI-Fast group constants sets of less than 30-group and prepares a binary library tape for efficient usage by a series of related fast reactor design calculation programmes

  11. ENEA-Bologna production and testing of JEF-2.2 multi-group cross section libraries for nuclear fission applications

    International Nuclear Information System (INIS)

    Pescarini, M.; Orsi, R.; Martinelli, T.; Sinitsa, V.; Blokhin, A.I.

    2005-01-01

    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 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 σ 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)

  12. The evolutionary history of Eugenia sect. Phyllocalyx (Myrtaceae) corroborates historically stable areas in the southern Atlantic forests.

    Science.gov (United States)

    de Oliveira Bünger, Mariana; Fernanda Mazine, Fiorella; Forest, Félix; Leandro Bueno, Marcelo; Renato Stehmann, João; Lucas, Eve J

    2016-12-01

    Eugenia sect. Phyllocalyx Nied. includes 14 species endemic to the Neotropics, mostly distributed in the Atlantic coastal forests of Brazil. Here the first comprehensive phylogenetic study of this group is presented, and this phylogeny is used as the basis to evaluate the recent infrageneric classification in Eugenia sensu lato (s.l.) to test the history of the evolution of traits in the group and test hypotheses associated with the history of this clade. A total of 42 taxa were sampled, of which 14 were Eugenia sect. Phyllocalyx for one nuclear (ribosomal internal transcribed spacer) and four plastid markers (psbA-trnH, rpl16, trnL-rpl32 and trnQ-rps16). The relationships were reconstructed based on Bayesian analysis and maximum likelihood. Additionally, ancestral area analysis and modelling methods were used to estimate species dispersal, comparing historically climatic stable (refuges) and unstable areas. Maximum likelihood and Bayesian inferences indicate that Eugenia sect. Phyllocalyx is paraphyletic and the two clades recovered are characterized by combinations of morphological characters. Phylogenetic relationships support a link between Cerrado and south-eastern species and a difference in the composition of species from north-eastern and south-eastern Atlantic forest. Refugia and stable areas identified within unstable areas suggest that these areas were important to maintain diversity in the Atlantic forest biodiversity hotspot. This study provides a robust phylogenetic framework to address important historical questions for Eugenia s.l. within an evolutionary context, supporting the need for better taxonomic study of one of the largest genera in the Neotropics. Furthermore, valuable insight is offered into diversification and biome shifts of plant species in the highly environmentally impacted Atlantic forest of South America. Evidence is presented that climate stability in the south-eastern Atlantic forest during the Quaternary contributed to the

  13. Development of a novel algorithm and production of new nuclear data libraries for the treatment of sequential (x,n) reactions in fusion material activation calculations

    International Nuclear Information System (INIS)

    Cierjacks, S.W.; Oblozinsky, P.; Kelzenberg, S.; Rzehorz, B.

    1993-01-01

    A new algorithm and three major nuclear data libraries were developed for the kinematically complete treatment of sequential (x,n) reactions in fusion material activation calculations. The new libraries include data for virtually all isotopes with Z ≤ 84 (A ≤ 210) and half-lives exceeding 1 day; primary neutron energies E n 3 He, and α with energies E x < 24 MeV. While production cross sections of charged particles for primary (n,x) reactions can be deduced from the European activation file, the KFKSPEC data file was created for the corresponding normalized charged-particle spectra. The second data file, KFKXN, contains cross sections for secondary (x,n) reactions. The third data file, KFKSTOP, has a complete set of differential ranges for all five aforementioned light charged particles and all elements from hydrogen to uranium. The KFKSPEC and KFKXN libraries are based essentially on nuclear model calculations using the statistical evaporation model superimposed with the pre-equilibrium contribution as implemented in the Lawrence Livermore National Laboratory ALICE code. The KFKSPEC library includes 633 isotopes, of which 55 are in their isomeric states, and contains 63,300 spectra of the (n,x) type with almost 1.5 million data points. The KFKXN library also includes 633 isotopes and contains all (x,n) and partly (x,2n) cross sections for 4431 reactions with ∼ 106,000 data points. The KFKSTOP library is considered complete and has 11,040 data points. 42 refs., 2 figs., 4 tabs

  14. A multilevel in space and energy solver for multigroup diffusion eigenvalue problems

    Directory of Open Access Journals (Sweden)

    Ben C. Yee

    2017-09-01

    Full Text Available In this paper, we present a new multilevel in space and energy diffusion (MSED method for solving multigroup diffusion eigenvalue problems. The MSED method can be described as a PI scheme with three additional features: (1 a grey (one-group diffusion equation used to efficiently converge the fission source and eigenvalue, (2 a space-dependent Wielandt shift technique used to reduce the number of PIs required, and (3 a multigrid-in-space linear solver for the linear solves required by each PI step. In MSED, the convergence of the solution of the multigroup diffusion eigenvalue problem is accelerated by performing work on lower-order equations with only one group and/or coarser spatial grids. Results from several Fourier analyses and a one-dimensional test code are provided to verify the efficiency of the MSED method and to justify the incorporation of the grey diffusion equation and the multigrid linear solver. These results highlight the potential efficiency of the MSED method as a solver for multidimensional multigroup diffusion eigenvalue problems, and they serve as a proof of principle for future work. Our ultimate goal is to implement the MSED method as an efficient solver for the two-dimensional/three-dimensional coarse mesh finite difference diffusion system in the Michigan parallel characteristics transport code. The work in this paper represents a necessary step towards that goal.

  15. Specifications for a two-dimensional multi-group scattering code: ALCI; Specification d'un code de diffusion multigroupe a deux dimensions: ALCI

    Energy Technology Data Exchange (ETDEWEB)

    Bayard, J P; Guillou, A; Lago, B; Bureau du Colombier, M J; Guillou, G; Vasseur, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-02-01

    This report describes the specifications of the ALCI programme. This programme resolves the system of difference equations similar to the homogeneous problem of multigroup neutron scattering, with two dimensions in space, in the three geometries XY, RZ, R{theta}. It is possible with this method to calculate geometric and composition criticalities and also to calculate the accessory problem on demand. The maximum number of points dealt with is 6000. The maximum permissible number of groups is 12. The internal iterations are treated by the method of alternating directions. The external iterations are accelerated using the extrapolation method due to Tchebychev. (authors) [French] Ce rapport decrit les specifications du programme ALCI. Ce programme resout le systeme d'equations aux differences approchant le probleme homogene de la diffusion neutronique multigroupe, a deux dimensions d'espace, dans les trois geometries XY, RZ, R{theta}. Il permet des calculs de criticalite geometrique et de composition et calcule sur demande le probleme adjoint. Le nombre maximum de points traites est de 6000. Le nombre maximum de groupes permis est de 12. Les iterations interieure sont traitees par la methode des directions alternees. Les iterations exterieures sont accelerees par la methode d'extrapolation de Tchebychev. (auteurs)

  16. An analytical multigroup benchmark for (n, γ) and (n, n', γ) verification of diffusion theory algorithms

    International Nuclear Information System (INIS)

    Ganapol, B.D.

    2011-01-01

    Highlights: → Coupled neutron and gamma transport is considered in the multigroup diffusion approximation. → The model accommodates fission, up- and down-scattering and common neutron-gamma interactions. → The exact solution to the diffusion equation in a heterogeneous media of any number of regions is found. → The solution is shown to parallel the one-group case in a homogeneous medium. → The discussion concludes with a heterogeneous, 2 fuel-plate 93.2% enriched reactor fuel benchmark demonstration. - Abstract: The angular flux for the 'rod model' describing coupled neutron/gamma (n, γ) diffusion has a particularly straightforward analytical representation when viewed from the perspective of a one-group homogeneous medium. Cast in the form of matrix functions of a diagonalizable matrix, the solution to the multigroup equations in heterogeneous media is greatly simplified. We shall show exactly how the one-group homogeneous medium solution leads to the multigroup solution.

  17. On efficiently computing multigroup multi-layer neutron reflection and transmission conditions

    International Nuclear Information System (INIS)

    Abreu, Marcos P. de

    2007-01-01

    In this article, we present an algorithm for efficient computation of multigroup discrete ordinates neutron reflection and transmission conditions, which replace a multi-layered boundary region in neutron multiplication eigenvalue computations with no spatial truncation error. In contrast to the independent layer-by-layer algorithm considered thus far in our computations, the algorithm here is based on an inductive approach developed by the present author for deriving neutron reflection and transmission conditions for a nonactive boundary region with an arbitrary number of arbitrarily thick layers. With this new algorithm, we were able to increase significantly the computational efficiency of our spectral diamond-spectral Green's function method for solving multigroup neutron multiplication eigenvalue problems with multi-layered boundary regions. We provide comparative results for a two-group reactor core model to illustrate the increased efficiency of our spectral method, and we conclude this article with a number of general remarks. (author)

  18. ENEA-Bologna production and testing of Jeff-3.1 multi-group cross section libraries for nuclear fission applications

    International Nuclear Information System (INIS)

    Pescarini, M.; Orsi, R.; Sinitsa, V.

    2008-01-01

    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)

  19. Angular finite volume method for solving the multigroup transport equation with piecewise average scattering cross sections

    International Nuclear Information System (INIS)

    Calloo, A.; Vidal, J.F.; Le Tellier, R.; Rimpault, G.

    2011-01-01

    This paper deals with the solving of the multigroup integro-differential form of the transport equation for fine energy group structure. In that case, multigroup transfer cross sections display strongly peaked shape for light scatterers and the current Legendre polynomial expansion is not well-suited to represent them. Furthermore, even if considering an exact scattering cross sections representation, the scattering source in the discrete ordinates method (also known as the Sn method) being calculated by sampling the angular flux at given directions, may be wrongly computed due to lack of angular support for the angular flux. Hence, following the work of Gerts and Matthews, an angular finite volume solver has been developed for 2D Cartesian geometries. It integrates the multigroup transport equation over discrete volume elements obtained by meshing the unit sphere with a product grid over the polar and azimuthal coordinates and by considering the integrated flux per solid angle element. The convergence of this method has been compared to the S_n method for a highly anisotropic benchmark. Besides, piecewise-average scattering cross sections have been produced for non-bound Hydrogen atoms using a free gas model for thermal neutrons. LWR lattice calculations comparing Legendre representations of the Hydrogen scattering multigroup cross section at various orders and piecewise-average cross sections for this same atom are carried out (while keeping a Legendre representation for all other isotopes). (author)

  20. Semi-continuous and multigroup models in extended kinetic theory

    International Nuclear Information System (INIS)

    Koller, W.

    2000-01-01

    The aim of this thesis is to study energy discretization of the Boltzmann equation in the framework of extended kinetic theory. In case that external fields can be neglected, the semi- continuous Boltzmann equation yields a sound basis for various generalizations. Semi-continuous kinetic equations describing a three component gas mixture interacting with monochromatic photons as well as a four component gas mixture undergoing chemical reactions are established and investigated. These equations reflect all major aspects (conservation laws, equilibria, H-theorem) of the full continuous kinetic description. For the treatment of the spatial dependence, an expansion of the distribution function in terms of Legendre polynomials is carried out. An implicit finite differencing scheme is combined with the operator splitting method. The obtained numerical schemes are applied to the space homogeneous study of binary chemical reactions and to spatially one-dimensional laser-induced acoustic waves. In the presence of external fields, the developed overlapping multigroup approach (with the spline-interpolation as its extension) is well suited for numerical studies. Furthermore, two formulations of consistent multigroup approaches to the non-linear Boltzmann equation are presented. (author)

  1. Article 74 sect. 1 no. 24 (air pollution control) as a competent fundament for Renewable Energy Resources Act and thermal energy law; Art. 74 Abs. Nr. 24 GG (Luftreinhaltung) als Kompetenzgrundlage fuer EEG und EEWaermeG

    Energy Technology Data Exchange (ETDEWEB)

    Soesemann, F. [Kammergericht, Berlin (Germany); Humboldt-Univ. Berlin (Germany); Ecologic gGmbH, Berlin (Germany)

    2008-08-15

    Since the reform of federalism, no examination of requirement is necessary in accordance with article 72 sect. 2 Basic Law in order to support federal laws on article 74 sect. 1 no. 24 Basic Law. It is in the interest of the federal law giver to legislate the climate protection such as Renewable Energy Resources Act and thermal energy law alone on the basis of the authority standard in order to avoid the examination of requirement. In the sense of the article 74 sect. 1 no. 24 Basic Law, the climate protection is considered as air pollution control and serves to the reduction of greenhouse gases.

  2. Production and testing of the VITAMIN-B6 fine-group and the BUGLE-93 broad-group neutron/photon cross-section libraries derived from ENDF/B-VI nuclear data

    International Nuclear Information System (INIS)

    Ingersoll, D.T.; White, J.E.; Wright, R.Q.; Hunter, H.T.; Slater, C.O.; Greene, N.M.; MacFarlane, R.E.

    1993-01-01

    A new multigroup cross-section library based on ENDF/B-VI data has been produced and tested for light water reactor shielding and reactor pressure vessel dosimetry applications. The broad-group library is designated BUGLE-93. The processing methodology is consistent with ANSI/ANS 6.1.2, since 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 is designated VITAMIN-B6. An extensive integral data testing effort was also performed. In general, results using the new data show significant improvements relative to earlier ENDF data

  3. Social comparison and perceived breach of psychological contract: their effects on burnout in a multigroup analysis.

    Science.gov (United States)

    Cantisano, Gabriela Topa; Domínguez, J Francisco Morales; García, J Luis Caeiro

    2007-05-01

    This study focuses on the mediator role of social comparison in the relationship between perceived breach of psychological contract and burnout. A previous model showing the hypothesized effects of perceived breach on burnout, both direct and mediated, is proposed. The final model reached an optimal fit to the data and was confirmed through multigroup analysis using a sample of Spanish teachers (N = 401) belonging to preprimary, primary, and secondary schools. Multigroup analyses showed that the model fit all groups adequately.

  4. Systematics of the southern African genus Ixia (Iridaceae: Crocoideae: 4. Revision of sect. Dichone

    Directory of Open Access Journals (Sweden)

    P. Goldblatt

    2012-12-01

    Full Text Available The southern African genus Ixia L. comprises ± 90 species from the winter-rainfall zone of the subcontinent. Ixia sect. Dichone (Salisb. ex Baker Goldblatt & J.C.Manning, one of four sections in the genus and currently including 10 species and three varieties, is distinguished by the following floral characters: lower part of the perianth tube filiform and tightly clasping the style; filaments not decurrent; upper part of the perianth tube short to vestigial; style branches involute-tubular and stigmatic only at the tips; and so-called subdidymous anthers. We review the taxonomy of the section, providing complete descriptions and distribution maps, and a key to the species. I. amethystina Manning & Goldblatt is recognized to be a later synonym of I. brevituba G.J.Lewis. Most collections currently included under that name represent another species, here described as I. rigida. We recognize five additional species in the section: early summer-blooming I. altissima from the Cedarberg; I. bifolia from the Caledon District; I. flagellaris, a stoloniferous species from the Cedarberg; I. simulans from the western Langeberg; and I. tenuis from the Piketberg. We also raise to species rank I. micrandra var. confusa and var. minor, as I. confusa and I. minor respectively. Foliar and associated floral variation in the widespread I. scillaris has led us to recognize two new subspecies among its northern populations, broad leaved subsp. latifolia and the dwarfed, smaller flowered subsp. toximontana; subsp. scillaris is restricted to the immediate southwestern Cape, from Darling to Somerset West. Sect. Dichone now has 17 species and two subspecies.

  5. The problem of resonance self-shielding effect in neutron multigroup calculations

    International Nuclear Information System (INIS)

    Wang Qingming; Huang Jinghua

    1991-01-01

    It is not allowed to neglect the resonance self-shielding effect in hybrid blanket and fast reactor neutron designs. The authors discussed the importance as well as the method of considering the resonance self-shielding effect in hybrid blanket and fast reactor neutron multigroup calculations

  6. Microstructural and dielectric properties of Ba0.6Sr0.4Ti1-xZrxO3 based combinatorial thin film capacitors library

    International Nuclear Information System (INIS)

    Liu Guozhen; Wolfman, Jerome; Autret-Lambert, Cecile; Sakai, Joe; Roger, Sylvain; Gervais, Monique; Gervais, Francois

    2010-01-01

    Epitaxial growth of Ba 0.6 Sr 0.4 Ti 1-x Zr x O 3 (0≤x≤0.3) composition spread thin film library on SrRuO 3 /SrTiO 3 layer by combinatorial pulsed laser deposition (PLD) is reported. X-ray diffraction and energy dispersive x-ray spectroscopy studies showed an accurate control of the film phase and composition by combinatorial PLD. A complex evolution of the microstructure and morphology with composition of the library is described, resulting from the interplay between epitaxial stress, increased chemical pressure, and reduced elastic energy upon Zr doping. Statistical and temperature-related capacitive measurements across the library showed unexpected variations in the dielectric properties. Doping windows with enhanced permittivity and tunability are identified, and correlated to microstructural properties.

  7. Geographical isolation caused the diversification of the Mediterranean thorny cushion-like Astragalus L. sect. Tragacantha DC. (Fabaceae).

    Science.gov (United States)

    Hardion, Laurent; Dumas, Pierre-Jean; Abdel-Samad, Farah; Bou Dagher Kharrat, Magda; Surina, Bostjan; Affre, Laurence; Médail, Frédéric; Bacchetta, Gianluigi; Baumel, Alex

    2016-04-01

    Understanding the origin and evolution of Mediterranean vascular flora within the long-term context of climate change requires a continuous study of historical biogeography supported by molecular phylogenetic approaches. Here we provide new insights into the fascinating but often overlooked diversification of Mediterranean xerophytic plants. Growing in some of the most stressing Mediterranean environments, i.e. coastal and mountainous opened habitats, the circum-Mediterranean Astragalus L. sect. Tragacantha DC. (Fabaceae) gathers several thorny cushion-like taxa. These have been the subjects of recent taxonomical studies, but they have not yet been investigated within a comprehensive molecular framework. Bayesian phylogenetics applied to rDNA ITS sequences reveal that the diversification of A. sect. Tragacantha has roots dating back to the Pliocene, and the same data also indicate an eastern-western split giving rise to the five main lineages that exist today. In addition, AFLP fingerprinting supports an old east-west pattern of vicariance that completely rules out the possibility of a recent eastern origin for western taxa. The observed network of genetic relationships implies that contrary to what is widely claimed in the taxonomic literature, it is range fragmentation, as opposed to a coastal-to-mountain ecological shift, that is likely the main driver of diversification. Copyright © 2016 Elsevier Inc. All rights reserved.

  8. Exploring the power of rice (O. sativa x O. rufipogon) chromosome segment substitution line libraries

    Science.gov (United States)

    Transgressive variation was reported as an increase in grain yield for several rice (Oryza sativa x O. rufipogon) advanced backcross mapping populations. The objective of this study was to develop chromosome segment substitution line (CSSL) libraries to further dissect the reported transgressive var...

  9. COMPAR, NJOY, GROUPIE, FLANGE-2, ETOG-3, XLACS Multigroup Cross-Sections General Comparison

    International Nuclear Information System (INIS)

    Anaf, Jaime; Chalhoub, E.S.

    1990-01-01

    1 - Description of program or function: A system for comparing multigroup cross sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS. This system comprises the COMPAR program and interface (auxiliary) programs developed for each of the programs under consideration. These are REDCOMP for GROUPIE, FLACOMP for FLANGE-II, ETOCOMP for ETOG-3 and XLACOMP for XLACS. For the NJOY program there is RGENDF, a program developed apart from this system. It is a modular system in which the inclusion of new multigroup cross section generating program requires no more than the development of a new interface module. 2 - Method of solution: Refer to comments in main routine. 3 - Restrictions on the complexity of the problem: Refer to comments in main routine

  10. TRIDENT: a two-dimensional, multigroup, triangular mesh discrete ordinates, explicit neutron transport code

    International Nuclear Information System (INIS)

    Seed, T.J.; Miller, W.F. Jr.; Brinkley, F.W. Jr.

    1977-03-01

    TRIDENT solves the two-dimensional-multigroup-transport equations in rectangular (x-y) and cylindrical (r-z) geometries using a regular triangular mesh. Regular and adjoint, inhomogeneous and homogeneous (k/sub eff/ and eigenvalue searches) problems subject to vacuum, reflective, white, or source boundary conditions are solved. General anisotropic scattering is allowed and anisotropic-distributed sources are permitted. The discrete-ordinates approximation is used for the neutron directional variables. An option is included to append a fictitious source to the discrete-ordinates equations that is defined such that spherical-harmonics solutions (in x-y geometry) or spherical-harmonics-like solutions (in r-z geometry) are obtained. A spatial-finite-element method is used in which the angular flux is expressed as a linear polynomial in each triangle that is discontinous at triangle boundaries. Unusual Features of the program: Provision is made for creation of standard interface output files for S/sub N/ constants, angle-integrated (scalar) fluxes, and angular fluxes. Standard interface input files for S/sub N/ constants, inhomogeneous sources, cross sections, and the scalar flux may be read. Flexible edit options as well as a dump and restart capability are provided

  11. A spectral nodal method for eigenvalue SN transport problems in two-dimensional rectangular geometry for energy multigroup nuclear reactor global calculations

    International Nuclear Information System (INIS)

    Silva, Davi Jose M.; Alves Filho, Hermes; Barros, Ricardo C.

    2015-01-01

    A spectral nodal method is developed for multigroup x,y-geometry discrete ordinates (S N ) eigenvalue problems for nuclear reactor global calculations. This method uses the conventional multigroup SN discretized spatial balance nodal equations with two non-standard auxiliary equations: the spectral diamond (SD) auxiliary equations for the discretization nodes inside the fuel regions, and the spectral Green's function (SGF) auxiliary equations for the non-multiplying regions, such as the baffle and the reactor. This spectral nodal method is derived from the analytical general solution of the SN transverse integrated nodal equations with constant approximations for the transverse leakage terms within each discretization node. The SD and SGF auxiliary equations have parameters, which are determined to preserve the homogeneous and the particular components of these local general solutions. Therefore, we refer to the offered method as the hybrid SD-SGF-Constant Nodal (SD-SGF-CN) method. The S N discretized spatial balance equations, together with the SD and the SGF auxiliary equations form the SD-SGF-CN equations. We solve the SD-SGF-CN equations by using the one-node block inversion inner iterations (NBI), wherein the most recent estimates for the incoming group node-edge average or prescribed boundary conditions are used to evaluate the outgoing group node-edge average fluxes in the directions of the S N transport sweeps, for each estimate of the dominant eigenvalue in the conventional Power outer iterations. We show in numerical calculations that the SD-SGF-CN method is very accurate for coarse-mesh multigroup S N eigenvalue problems, even though the transverse leakage terms are approximated rather simply. (author)

  12. Angular finite volume method for solving the multigroup transport equation with piecewise average scattering cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Calloo, A.; Vidal, J.F.; Le Tellier, R.; Rimpault, G., E-mail: ansar.calloo@cea.fr, E-mail: jean-francois.vidal@cea.fr, E-mail: romain.le-tellier@cea.fr, E-mail: gerald.rimpault@cea.fr [CEA, DEN, DER/SPRC/LEPh, Saint-Paul-lez-Durance (France)

    2011-07-01

    This paper deals with the solving of the multigroup integro-differential form of the transport equation for fine energy group structure. In that case, multigroup transfer cross sections display strongly peaked shape for light scatterers and the current Legendre polynomial expansion is not well-suited to represent them. Furthermore, even if considering an exact scattering cross sections representation, the scattering source in the discrete ordinates method (also known as the Sn method) being calculated by sampling the angular flux at given directions, may be wrongly computed due to lack of angular support for the angular flux. Hence, following the work of Gerts and Matthews, an angular finite volume solver has been developed for 2D Cartesian geometries. It integrates the multigroup transport equation over discrete volume elements obtained by meshing the unit sphere with a product grid over the polar and azimuthal coordinates and by considering the integrated flux per solid angle element. The convergence of this method has been compared to the S{sub n} method for a highly anisotropic benchmark. Besides, piecewise-average scattering cross sections have been produced for non-bound Hydrogen atoms using a free gas model for thermal neutrons. LWR lattice calculations comparing Legendre representations of the Hydrogen scattering multigroup cross section at various orders and piecewise-average cross sections for this same atom are carried out (while keeping a Legendre representation for all other isotopes). (author)

  13. FENDL/MG-2.0 and FENDL/MC-2.0. The processed cross-section libraries for neutron photon transport calculations. Version 1, March 1997. Summary documentation

    International Nuclear Information System (INIS)

    Wienke, H.; Herman, M.

    1998-01-01

    Evaluated neutron reaction data and photon-atom interaction cross sections for materials contained in the general purpose Fusion Evaluated Nuclear Data Library (FENDL/E2.0) have been processed with the NJOY code system into VITAMIN-J multigroup structure, for use in discrete-ordinates transport codes, and into continuous energy ACE format, for use in the Monte Carlo transport code MCNP. This document summarizes the resulting data libraries FENDL/MG-2.0 version 1 and FENDL/MC-2.0 version 1. The data are available costfree from the IAEA Nuclear Data Section online or on magnetic tape. (author)

  14. A multigroup treatment of radiation transport

    International Nuclear Information System (INIS)

    Tahir, N.A.; Laing, E.W.; Nicholas, D.J.

    1980-12-01

    A multi-group radiation package is outlined which will accurately handle radiation transfer problems in laser-produced plasmas. Bremsstrahlung, recombination and line radiation are included as well as fast electron Bremsstrahlung radiation. The entire radiation field is divided into a large number of groups (typically 20), which diffuse radiation energy in real space as well as in energy space, the latter occurring via electron-radiation interaction. Using this model a radiation transport code will be developed to be incorporated into MEDUSA. This modified version of MEDUSA will be used to study radiative preheat effects in laser-compression experiments at the Central Laser Facility, Rutherford Laboratory. The model is also relevant to heavy ion fusion studies. (author)

  15. Verification and validation of a multi-temperature JEFF 3.1 library for MCNP(X) - JEF/DOC-1099

    Energy Technology Data Exchange (ETDEWEB)

    Haeck, W; Verboomen, B

    2005-11-15

    A set of library files can be considered to be a validated application library when the following requirements are met: the library processing has been verified; the library files have been tested on internationally accepted benchmarks (validation); every aspect in their creation has been documented The verification process ensures that no problems were encountered during processing (due to faulty data, bad formatting, . . . ) and that corrective actions are taken or a warning is issued if a problem was found. It is important to note that the verification process starts with the selection of the NJOY processing path and all relevant parameters, even before the first library fie is produced. The third requirement (document everything) is closely linked to this part of the verification process. Because the library is produced for use with MCNP(X), it must be validated for use with MCNP(X) to ensure that it gives accurate and reliable results. This can be done by testing the library against internationally accepted benchmarks. And again, this part must be well documented. Documenting every aspect in the creation of the library is important for future reference. A data preprocessing code (in our case NJOY) should never be considered as a black box. With the documentation, a user should be capable of producing exactly the same library and understand every aspect of its creation. Before even starting with the production of a library, we should examine what has been done in the past in the field of library preparation and verification. This is why we started by looking into the documentation of older libraries to see what NJOY processing path they had chosen, what values they used for important parameters such as the reconstruction tolerance, . . . Other important information is the type of verification tests that were performed and what types of benchmarks were used to validate the library. And during this process the most important question is why. We should learn from

  16. Verification and validation of a multi-temperature JEFF 3.1 library for MCNP(X) - JEF/DOC-1099

    International Nuclear Information System (INIS)

    Haeck, W.; Verboomen, B.

    2005-11-01

    A set of library files can be considered to be a validated application library when the following requirements are met: the library processing has been verified; the library files have been tested on internationally accepted benchmarks (validation); every aspect in their creation has been documented The verification process ensures that no problems were encountered during processing (due to faulty data, bad formatting, . . . ) and that corrective actions are taken or a warning is issued if a problem was found. It is important to note that the verification process starts with the selection of the NJOY processing path and all relevant parameters, even before the first library fie is produced. The third requirement (document everything) is closely linked to this part of the verification process. Because the library is produced for use with MCNP(X), it must be validated for use with MCNP(X) to ensure that it gives accurate and reliable results. This can be done by testing the library against internationally accepted benchmarks. And again, this part must be well documented. Documenting every aspect in the creation of the library is important for future reference. A data preprocessing code (in our case NJOY) should never be considered as a black box. With the documentation, a user should be capable of producing exactly the same library and understand every aspect of its creation. Before even starting with the production of a library, we should examine what has been done in the past in the field of library preparation and verification. This is why we started by looking into the documentation of older libraries to see what NJOY processing path they had chosen, what values they used for important parameters such as the reconstruction tolerance, . . . Other important information is the type of verification tests that were performed and what types of benchmarks were used to validate the library. And during this process the most important question is why. We should learn from

  17. Verification of KARMA GEOM/TRPT Module with Given Multi-group Cross Sections

    International Nuclear Information System (INIS)

    Koo, Bon Seung; Hong, Ser Gi; Song, Jae Seung

    2009-01-01

    KAERI has developed a two-dimensional multigroup transport theory code KARMA (Kernel Analyzer by Ray-tracing Method for Fuel Assembly). KARMA uses CMFD (Coarse Mesh Finite Difference) accelerated MOC (Method of Characteristics) method for burnup calculation on a single fuel pin, a fuel assembly and a core consisting of rectangular array of fuel pins. KARMA code intends to be employed as a nuclear design tool for the Korean commercial pressurizer water reactor. Prior to the application to actual assembly designs, the code has to be approved by regularity agency. Therefore, it is essential that the reliability of KARMA code should be sufficiently evaluated against well-defined benchmark problems. In this paper, verification of GEOM/TRPT modules of KARMA was performed to confirm a reliability of the KARMA transport solution via comparisons with Monte Carlo calculations by using a consistent set of multi-group macroscopic cross-sections

  18. Scalable Multi-group Key Management for Advanced Metering Infrastructure

    OpenAIRE

    Benmalek , Mourad; Challal , Yacine; Bouabdallah , Abdelmadjid

    2015-01-01

    International audience; Advanced Metering Infrastructure (AMI) is composed of systems and networks to incorporate changes for modernizing the electricity grid, reduce peak loads, and meet energy efficiency targets. AMI is a privileged target for security attacks with potentially great damage against infrastructures and privacy. For this reason, Key Management has been identified as one of the most challenging topics in AMI development. In this paper, we propose a new Scalable multi-group key ...

  19. Optimal calculational schemes for solving multigroup photon transport problem

    International Nuclear Information System (INIS)

    Dubinin, A.A.; Kurachenko, Yu.A.

    1987-01-01

    A scheme of complex algorithm for solving multigroup equation of radiation transport is suggested. The algorithm is based on using the method of successive collisions, the method of forward scattering and the spherical harmonics method, and is realized in the FORAP program (FORTRAN, BESM-6 computer). As an example the results of calculating reactor photon transport in water are presented. The considered algorithm being modified may be used for solving neutron transport problems

  20. APPLE, Plot of 1-D Multigroup Neutron Flux and Gamma Flux and Reaction Rates from ANISN

    International Nuclear Information System (INIS)

    Kawasaki, Hiromitsu; Seki, Yasushi

    1983-01-01

    A - Description of problem or function: The APPLE-2 code has the following functions: (1) It plots multi-group energy spectra of neutron and/or gamma ray fluxes calculated by ANISN, DOT-3.5, and MORSE. (2) It gives an overview plot of multi-group neutron fluxes calculated by ANISN and DOT-3.5. The scalar neutron flux phi(r,E) is plotted with the spatial parameter r linear along the Y-axis, logE along the X-axis and log phi(r,E) in the Z direction. (3) It calculates the spatial distribution and region volume integrated values of reaction rates using the scalar flux calculated with ANISN and DOT-3.5. (4) Reaction rate distribution along the R or Z direction may be plotted. (5) An overview plot of reaction rates or scalar fluxes summed over specified groups may be plotted. R(ri,zi) or phi(ri,zi) is plotted with spatial parameters r and z along the X- and Y-axes in an orthogonal coordinate system. (6) Angular flux calculated by ANISN is rearranged and a shell source at any specified spatial mesh point may be punched out in FIDO format. The shell source obtained may be employed in solving deep penetration problems with ANISN, when the entire reactor system is divided into two or more parts and the neutron fluxes in two adjoining parts are connected by using the shell source. B - Method of solution: (a) The input data specification is made as simple as possible by making use of the input data required in the radiation transport code. For example, geometry related data in ANISN and DOT are transmitted to APPLE-2 along with scalar flux data so as to reduce duplicity and errors in reproducing these data. (b) Most the input data follow the free form FIDO format developed at Oak Ridge National Laboratory and used in the ANISN code. Furthermore, the mixture specifying method used in ANISN is also employed by APPLE-2. (c) Libraries for some standard response functions required in fusion reactor design have been prepared and are made available to users of the 42-group neutron

  1. HAMMER, 1-D Multigroup Neutron Transport Infinite System Cell Calculation for Few-Group Diffusion Calculation

    International Nuclear Information System (INIS)

    Honeck, H.C.

    1984-01-01

    1 - Description of problem or function: HAMMER performs infinite lattice, one-dimensional cell multigroup calculations, followed (optionally) by one-dimensional, few-group, multi-region reactor calculations with neutron balance edits. 2 - Method of solution: Infinite lattice parameters are calculated by means of multigroup transport theory, composite reactor parameters by few-group diffusion theory. 3 - Restrictions on the complexity of the problem: - Cell calculations - maxima of: 30 thermal groups; 54 epithermal groups; 20 space points; 20 regions; 18 isotopes; 10 mixtures; 3 thermal up-scattering mixtures; 200 resonances per group; no overlap or interference; single level only. - Reactor calculations - maxima of : 40 regions; 40 mixtures; 250 space points; 4 groups

  2. The SpeX Prism Library Analysis Toolkit: Design Considerations and First Results

    Science.gov (United States)

    Burgasser, Adam J.; Aganze, Christian; Escala, Ivana; Lopez, Mike; Choban, Caleb; Jin, Yuhui; Iyer, Aishwarya; Tallis, Melisa; Suarez, Adrian; Sahi, Maitrayee

    2016-01-01

    Various observational and theoretical spectral libraries now exist for galaxies, stars, planets and other objects, which have proven useful for classification, interpretation, simulation and model development. Effective use of these libraries relies on analysis tools, which are often left to users to develop. In this poster, we describe a program to develop a combined spectral data repository and Python-based analysis toolkit for low-resolution spectra of very low mass dwarfs (late M, L and T dwarfs), which enables visualization, spectral index analysis, classification, atmosphere model comparison, and binary modeling for nearly 2000 library spectra and user-submitted data. The SpeX Prism Library Analysis Toolkit (SPLAT) is being constructed as a collaborative, student-centered, learning-through-research model with high school, undergraduate and graduate students and regional science teachers, who populate the database and build the analysis tools through quarterly challenge exercises and summer research projects. In this poster, I describe the design considerations of the toolkit, its current status and development plan, and report the first published results led by undergraduate students. The combined data and analysis tools are ideal for characterizing cool stellar and exoplanetary atmospheres (including direct exoplanetary spectra observations by Gemini/GPI, VLT/SPHERE, and JWST), and the toolkit design can be readily adapted for other spectral datasets as well.This material is based upon work supported by the National Aeronautics and Space Administration under Grant No. NNX15AI75G. SPLAT code can be found at https://github.com/aburgasser/splat.

  3. A broad-group cross-section library based on ENDF/B-VII.0 for fast neutron dosimetry Applications

    Energy Technology Data Exchange (ETDEWEB)

    Alpan, F.A. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2011-07-01

    A new ENDF/B-VII.0-based coupled 44-neutron, 20-gamma-ray-group cross-section library was developed to investigate the latest evaluated nuclear data file (ENDF) ,in comparison to ENDF/B-VI.3 used in BUGLE-96, as well as to generate an objective-specific library. The objectives selected for this work consisted of dosimetry calculations for in-vessel and ex-vessel reactor locations, iron atom displacement calculations for reactor internals and pressure vessel, and {sup 58}Ni(n,{gamma}) calculation that is important for gas generation in the baffle plate. The new library was generated based on the contribution and point-wise cross-section-driven (CPXSD) methodology and was applied to one of the most widely used benchmarks, the Oak Ridge National Laboratory Pool Critical Assembly benchmark problem. In addition to the new library, BUGLE-96 and an ENDF/B-VII.0-based coupled 47-neutron, 20-gamma-ray-group cross-section library was generated and used with both SNLRML and IRDF dosimetry cross sections to compute reaction rates. All reaction rates computed by the multigroup libraries are within {+-} 20 % of measurement data and meet the U. S. Nuclear Regulatory Commission acceptance criterion for reactor vessel neutron exposure evaluations specified in Regulatory Guide 1.190. (authors)

  4. Cytotaxonomy of North African species of Delphinium L. sect. Delphinium (Ranunculaceae

    Directory of Open Access Journals (Sweden)

    Blanché, César

    1990-05-01

    Full Text Available Both chromosome numbers and karyotype structure from 15 wild North-African populations of Delphinium L. sect. Delphinium were studied. Delphinium balansae Boiss. & Reuter, with a more symmetrical karyotype, appears as a possible perennial ancestor of' the remaining annuals with highly asymmetrical and shorter karyotype. A taxonomical reorganization of this section is proposed: two new series (ser. Macropetala ser. nova and ser. Balansae ser. nova. and two new nomenclatural combinations (D. nanum subsp. alboliliaceum and D. nanum subsp. elongatum are proposed.

    S'han estudiat els nombres cromosòmics i l'estructura deis cariotipus de 15 poblacions nordafricanes de Delphinium L. sect. Delphinium. Delphinium balansae Boiss. & Reuter, amb cariotipus mes simètric, apareix com a possible ancestre perenne de la resta d’espècies anuals amb cariotipus més asimètric i més curt. Es proposa una reorganització taxonòmica de la secció, d'on es descriu en dues series noves (ser. Macropetala ser. nova i ser. Balansae ser. nova. i es proposen dues combinacions noves (D. nanum subsp. alboliliaceum and D. nanum subsp. elongatum.

  5. Benchmarking of multigroup neutron cross sections libraries on neutron transmission through WWER-440 vessel

    International Nuclear Information System (INIS)

    Ilieva, K.; Belousov, S.; Apostolov, T.

    1998-01-01

    The verification of calculated neutron fluence onto the WWER-440/230 pressure vessel is very topical task in particular referring that some of this type of reactors have been operated the major part of its design lifetime. Since the induced activity from the neutron irradiation onto the elements is a simple response of neutron flux the neutron fluence verification usually is done using the measured activity of radionuclides produced during reactor operation. Calculational and experimental results of 54 Mn induced activity of scraps from inner wall of Unit 1 reactor pressure vessel after 18th cycle and detectors irradiated behind the vessel during the 18th cycle of Unit 1 at Kozloduy NPP as well as neutron flux attenuation through the WWER-440/230 pressure vessel are presented. Neutron cross sections libraries generated on the base of ENDF/B-IV and ENDF/B-VI have been used in the calculations. The comparative analysis of evaluated activities and attenuation coefficient demonstrates the better reliability of the neutron fluence calculations by the libraries based on ENDF/B-VI than by ones on ENDF/B-IV. The extreme rarity of data for the activity of scraps from the WWER-440 reactor vessel and its combination with the data for the detectors irradiated behind the vessel makes them especially attractive for verification of calculational methods of neutron fluence onto the WWER-440 vessel with dummy cassettes loading. (author)

  6. MC2-2, Calculation of Fast Neutron Spectra and Multigroup Cross-Sections from ENDF/B Data

    International Nuclear Information System (INIS)

    2001-01-01

    1 - Description of program or function: MC 2 -2 solves the neutron slowing-down equations using basic neutron data derived from ENDF/B data files to determine fundamental mode spectra for use in generating multigroup neutron cross sections. The current edition includes the ability to treat all ENDF/B-V and -VI data representations. It accommodates high-order P scattering representations and provides numerous capabilities such as isotope mixing, delayed neutron processing, free-format input, and flexibility in output data selection. This edition supersedes previous releases of the MC22 program and the earlier MC2 program. Improved physics algorithms and increased computational efficiency are incorporated. Input data files required by MC2-2 may be generated from ENDF/B data by the code ETOE-2. The hyper-fine-group integral transport theory module of MC2-2, RABANL, is an improved version of the RABBLE/RABID codes. Many of the MC2-2 modules are used in the SDX code. 2 - Methods: The extended transport P1, B1, consistent P1, and consistent B1 fundamental mode ultra-fine-group equations are solved using continuous slowing-down theory and multigroup methods. Fast and accurate resonance integral methods are used in the narrow resonance resolved and unresolved resonance treatments. A fundamental mode homogeneous unit cell calculation is performed using either a multigroup or a continuous slowing-down treatment. Multigroup neutron homogeneous cross sections are generated in an ISOTXS format for an arbitrary group structure. A hyper-fine-group integral transport slowing down calculation (RABANL) is available as an option. RABANL performs a homogeneous or heterogeneous (pin or slab) unit cell calculation over the resonance region (resolved and unresolved) and generates multigroup neutron cross sections in an ISOTXS format. Neutron cross sections are generated by RABANL for the homogeneous unit cell and for each heterogeneous region in the pin or slab unit cell calculation

  7. Development on hybrid evaluated nuclear data library HENDL1.0/MG/MC

    International Nuclear Information System (INIS)

    Xu Dezheng; Gao Chunjing; Zheng Shanliang; Liu Haibo; Zhu Xiaoxiang; Li Jingjing; Wu Yican

    2004-01-01

    A Hybrid Evaluated Nuclear Data Library (HENDL) named as HENDL1.0 has been developed by Fusion Design Study (FDS) team of Institute of Plasma Physics, Academia Sinica (ASIPP) to take into account the requirements in design and research relevant to fusion, fission and fusion-fission sub-critical hybrid reactor. HENDLI1.0 contains one basic evaluated sub-library naming HENDL1.0/E and to processed working sub-libraries naming HENDL1.0/MG and HENDL1.0/MC, respectively. Through carefully comparing, distinguishing and choosing, HENDL1.0/E integrated basic evaluated neutron data files of 213 nuclides from the several main data libraries for evaluated neutron reaction cross sections including ENDF/B-VI (USA), JEF-2.2 (OECD/NEA, Europe), JENDL-3.2 (Japan), CENDL-2 (China), BROND-2 (Russia) and FENDL-2 (IAEA/NDS, ITER program). Based on this, 175-group neutron and 42-group photon neutron-photon coupled multi-group working library HENDL1.0/MG used for discrete ordinate Sn method transport calculation (such as ANISN code) and a compact ENDF form (ACE), continuous energy structure (pointwise) neutron cross section library HENDL1.0/MC for Monte Carlo method transport simulation (as MCMP code) can be attainable with the current group constants processing system NJOY and transport cross section preparation code TRANSX referring to the Vitamin-J energy group structure. In addition, two special bases i.e. transmutation (burnup) library BURNUP. DAT and response function library RESPONSE.DAT, have been also made for fuel cycle calculation and reactivity analyses of nuclear reactor. The relevant sample testing, benchmark checking and primary confirmation are also carried out to assess the validity of multi-purpose data library HENDL1.0. (authors)

  8. Development and Characterization of Microsatellite Markers for Central American Begonia sect. Gireoudia (Begoniaceae

    Directory of Open Access Journals (Sweden)

    Alex D. Twyford

    2013-05-01

    Full Text Available Premise of the study: Transcriptome sequence data were used to design microsatellite primers for two widespread Central American Begonia species, B. heracleifolia and B. nelumbiifolia, to investigate population structure and hybridization. Methods and Results: The transcriptome from vegetative meristem tissue from the related B. plebeja was mined for microsatellite loci, and 31 primer pairs amplified in the target species. Fifteen primer pairs were combined in two multiplex PCR reactions, which amplified an average of four alleles per locus. Conclusions: The markers developed will be a valuable genetic resource for medium-throughput genotyping of Central American species of Begonia sect. Gireoudia. A subset of these markers have perfect sequence matches to Asian B. venusta, and are promising for studies in other Begonia sections.

  9. G.M.S.I. - A generalised multigroup system of calculations using the IBM 7030 (stretch) computer

    International Nuclear Information System (INIS)

    Gratton, C.P.; Smith, P.E.

    1965-02-01

    G.M.S. is a generalised system of reactor physics calculations written in the FORTRAN programming language for the IBM 7030 (STRETCH) computer. The programme will perform cell, supercell and overall reactor physics problems within the limits of one dimension using either slab or cylindrical geometry. The Winfrith DSN(1) programme has been incorporated as a subroutine and all flux distribution calculations are performed using Transport Theory. A facility for the study of fuel burn-up has been included with boundary conditions for either cell irradiations or the burn-up characteristics of the overall reactor. Particular attention has been paid to the simplicity of the input to the programme to allow a wide range of potential users and extensive edit facilities have been incorporated to allow the better understanding of the multigroup output. A graphical output option is available using the STROMBERG-CARLSON equipment and, by a series of commands, graphs of power peaking and burn-up rates are produced for the designer or, alternatively, neutron spectra and activations for comparison with experiment, The logic of the programme is written in terms of 40 neutron energy groups, the group boundaries of which may be selected by the user. Clearly the use of the system for fast reactor calculations requires that greater emphasis be given to the representation of neutron events at high energy than would be necessary for thermal reactor studies. It is envisaged therefore that eventually two library files will be made available to cover the extreme cases of fast and thermal systems and that the group boundaries of these will be selected to enable adequate treatment of intermediate spectrum reactors. At the present time the thermal reactor library file only has been prepared; the content of this library is discussed in Section 2. The group structure is comprised of twelve groups in the fast and intermediate energy region, ten groups in and around the plutonium 240 resonance at 1

  10. Multiple and asymmetrical origin of polyploid dog rose hybrids (Rosa L. sect. Caninae (DC.) Ser.) involving unreduced gametes.

    Science.gov (United States)

    Herklotz, V; Ritz, C M

    2017-08-01

    Polyploidy and hybridization are important factors for generating diversity in plants. The species-rich dog roses ( Rosa sect. Caninae ) originated by allopolyploidy and are characterized by unbalanced meiosis producing polyploid egg cells (usually 4 x ) and haploid sperm cells (1 x ). In extant natural stands species hybridize spontaneously, but the extent of natural hybridization is unknown. The aim of the study was to document the frequency of reciprocal hybridization between the subsections Rubigineae and Caninae with special reference to the contribution of unreduced egg cells (5 x ) producing 6 x offspring after fertilization with reduced (1 x ) sperm cells. We tested whether hybrids arose by independent multiple events or via a single or few incidences followed by a subsequent spread of hybrids. Population genetics of 45 mixed stands of dog roses across central and south-eastern Europe were analysed using microsatellite markers and flow cytometry. Hybrids were recognized by the presence of diagnostic alleles and multivariate statistics were used to display the relationships between parental species and hybrids. Among plants classified to subsect. Rubigineae , 32 % hybridogenic individuals were detected but only 8 % hybrids were found in plants assigned to subsect. Caninae . This bias between reciprocal crossings was accompanied by a higher ploidy level in Rubigineae hybrids, which originated more frequently by unreduced egg cells. Genetic patterns of hybrids were strongly geographically structured, supporting their independent origin. The biased crossing barriers between subsections are explained by the facilitated production of unreduced gametes in subsect. Rubigineae . Unreduced egg cells probably provide the highly homologous chromosome sets required for correct chromosome pairing in hybrids. Furthermore, the higher frequency of Rubigineae hybrids is probably influenced by abundance effects because the plants of subsect. Caninae are much more abundant

  11. Implementing an interval computation library for OCaml on x86/amd64 architectures

    OpenAIRE

    Alliot , Jean-Marc; Gotteland , Jean-Baptiste; Vanaret , Charlie; Durand , Nicolas; Gianazza , David

    2012-01-01

    International audience; In this paper, we present two implementations of interval arithmetic for OCaml on x86/amd64 architectures. The first implementation is a binding to the classical MPFI/MPFR library. It provides access to multi-precision floating-point arithmetic and multi-precision floating-point intervalarithmetic. The second implementation has been natively written in assembly language for low-level functions and in OCaml for higher-levelfunctions. It has proven as fast as classical C...

  12. Validation of the 172 group ENDFB7GX library

    International Nuclear Information System (INIS)

    Khan, Suhail Ahmad; Raj, Devesh; Karthikeyan, R.; Jagannathan, V.

    2007-01-01

    Full text: Five 172 group libraries, viz., IAEAGX, ENDFB6GX, JENDL3GX, JEFF31GX, and LWRPSGX were obtained as a part of the IAEA WIMS Library Update Project (WLUP). The first four libraries have data available for 173 nuclides up to 244 Cm. The LWRPSGX library based on JEFF3.1 point dataset is an extended library up to 252 Cf. Data for 12 more actinides and the related burnup chain were added. The five libraries were validated against known experiments in an earlier work. In general the LWRPSGX was found to be giving better results. Recently another version of 172 group library 'ENDFB7GX' has been released. In the present work we provide the results of validation of the ENDFB7GX library against the same set of experimental data and a comparison with results of other libraries. The experimental configuration data include a variety of uniform lattices with enriched UO 2 , U- metal, mixed oxide (UO 2 -PuO 2 ) fuels with H 2 O and D 2 O moderators for a wide range of enrichment, fuel diameter and ratio of moderator to fuel volume (V m /V f ). The calculations have been done using the code LATTEST which solves the single pin lattice cell problem by 1-D multi-group transport theory after cylindricalising the square or hexagonal cell boundary. The LATTEST code is an improved version of the MURLI code and is capable of providing a ready testing of any new cross section library against a set of experimental benchmark lattices collected from various sources. The calculated k eff values and certain spectral indices, where available, have been compared for all the libraries for more than hundred critical lattices. There is a general under prediction of k eff values by all libraries. The maximum under prediction is for ENDFB6GX library and the least is for JENDL3GX library. The ENDFB7GX library, in general, is found to over predict in comparison to the k eff values obtained using LWRPSGX library. While scrutinizing the basic nuclear data it was noted that the slowing down cross

  13. Finally! A valid test of configural invariance using permutation in multigroup CFA

    NARCIS (Netherlands)

    Jorgensen, T.D.; Kite, B.A.; Chen, P.-Y.; Short, S.D.; van der Ark, L.A.; Wiberg, M.; Culpepper, S.A.; Douglas, J.A.; Wang, W.-C.

    2017-01-01

    In multigroup factor analysis, configural measurement invariance is accepted as tenable when researchers either (a) fail to reject the null hypothesis of exact fit using a χ2 test or (b) conclude that a model fits approximately well enough, according to one or more alternative fit indices (AFIs).

  14. A spectral nodal method for eigenvalue S{sub N} transport problems in two-dimensional rectangular geometry for energy multigroup nuclear reactor global calculations

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Davi Jose M.; Alves Filho, Hermes; Barros, Ricardo C., E-mail: davijmsilva@yahoo.com.br, E-mail: halves@iprj.uerj.br, E-mail: rcbarros@pq.cnpq.br [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Programa de Pos-Graduacao em Modelagem Computacional

    2015-07-01

    A spectral nodal method is developed for multigroup x,y-geometry discrete ordinates (S{sub N}) eigenvalue problems for nuclear reactor global calculations. This method uses the conventional multigroup SN discretized spatial balance nodal equations with two non-standard auxiliary equations: the spectral diamond (SD) auxiliary equations for the discretization nodes inside the fuel regions, and the spectral Green's function (SGF) auxiliary equations for the non-multiplying regions, such as the baffle and the reactor. This spectral nodal method is derived from the analytical general solution of the SN transverse integrated nodal equations with constant approximations for the transverse leakage terms within each discretization node. The SD and SGF auxiliary equations have parameters, which are determined to preserve the homogeneous and the particular components of these local general solutions. Therefore, we refer to the offered method as the hybrid SD-SGF-Constant Nodal (SD-SGF-CN) method. The S{sub N} discretized spatial balance equations, together with the SD and the SGF auxiliary equations form the SD-SGF-CN equations. We solve the SD-SGF-CN equations by using the one-node block inversion inner iterations (NBI), wherein the most recent estimates for the incoming group node-edge average or prescribed boundary conditions are used to evaluate the outgoing group node-edge average fluxes in the directions of the S{sub N} transport sweeps, for each estimate of the dominant eigenvalue in the conventional Power outer iterations. We show in numerical calculations that the SD-SGF-CN method is very accurate for coarse-mesh multigroup S{sub N} eigenvalue problems, even though the transverse leakage terms are approximated rather simply. (author)

  15. Variational P1 approximations of general-geometry multigroup transport problems

    International Nuclear Information System (INIS)

    Rulko, R.P.; Tomasevic, D.; Larsen, E.W.

    1995-01-01

    A variational approximation is developed for general-geometry multigroup transport problems with arbitrary anisotropic scattering. The variational principle is based on a functional that approximates a reaction rate in a subdomain of the system. In principle, approximations that result from this functional ''optimally'' determine such reaction rates. The functional contains an arbitrary parameter α and requires the approximate solutions of a forward and an adjoint transport problem. If the basis functions for the forward and adjoint solutions are chosen to be linear functions of the angular variable Ω, the functional yields the familiar multigroup P 1 equations for all values of α. However, the boundary conditions that result from the functional depend on α. In particular, for problems with vacuum boundaries, one obtains the conventional mixed boundary condition, but with an extrapolation distance that depends continuously on α. The choice α = 0 yields a generalization of boundary conditions derived earlier by Federighi and Pomraning for a more limited class of problems. The choice α = 1 yields a generalization of boundary conditions derived previously by Davis for monoenergetic problems. Other boundary conditions are obtained by choosing different values of α. The authors discuss this indeterminancy of α in conjunction with numerical experiments

  16. Contribution to the solution of the multigroup Boltzmann equation by the determinist methods and the Monte Carlo method; Contribution a la resolution de l`equation de Bolztmann en multigroupe par les methodes deterministes et Monte-Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Li, M

    1998-08-01

    In this thesis, two methods for solving the multigroup Boltzmann equation have been studied: the interface-current method and the Monte Carlo method. A new version of interface-current (IC) method has been develop in the TDT code at SERMA, where the currents of interface are represented by piecewise constant functions in the solid angle space. The convergence of this method to the collision probability (CP) method has been tested. Since the tracking technique is used for both the IC and CP methods, it is necessary to normalize he collision probabilities obtained by this technique. Several methods for this object have been studied and implemented in our code, we have compared their performances and chosen the best one as the standard choice. The transfer matrix treatment has been a long-standing difficulty for the multigroup Monte Carlo method: when the cross-sections are converted into multigroup form, important negative parts will appear in the angular transfer laws represented by low-order Legendre polynomials. Several methods based on the preservation of the first moments, such as the discrete angles methods and the equally-probable step function method, have been studied and implemented in the TRIMARAN-II code. Since none of these codes has been satisfactory, a new method, the non equally-probably step function method, has been proposed and realized in our code. The comparisons for these methods have been done in several aspects: the preservation of the moments required, the calculation of a criticality problem and the calculation of a neutron-transfer in water problem. The results have showed that the new method is the best one in all these comparisons, and we have proposed that it should be a standard choice for the multigroup transfer matrix. (author) 76 refs.

  17. Phylogeny and biogeography of Primula sect. Armerina: implications for plant evolution under climate change and the uplift of the Qinghai-Tibet Plateau.

    Science.gov (United States)

    Ren, Guangpeng; Conti, Elena; Salamin, Nicolas

    2015-08-16

    The historical orogenesis and associated climatic changes of mountain areas have been suggested to partly account for the occurrence of high levels of biodiversity and endemism. However, their effects on dispersal, differentiation and evolution of many groups of plants are still unknown. In this study, we examined the detailed diversification history of Primula sect. Armerina, and used biogeographic analysis and macro-evolutionary modeling to investigate a series of different questions concerning the evolution of the geographical and ecological distribution of the species in this section. We sequenced five chloroplast and one nuclear genes for species of Primula sect. Armerina. Neither chloroplast nor nuclear trees support the monophyly of the section. The major incongruences between the two trees occur among closely related species and may be explained by hybridization. Our dating analyses based on the chloroplast dataset suggest that this section began to diverge from its relatives around 3.55 million years ago, largely coinciding with the last major uplift of the Qinghai-Tibet Plateau (QTP). Biogeographic analysis supports the origin of the section in the Himalayan Mountains and dispersal from the Himalayas to Northeastern QTP, Western QTP and Hengduan Mountains. Furthermore, evolutionary models of ecological niches show that the two P. fasciculata clades have significantly different climatic niche optima and rates of niche evolution, indicating niche evolution under climatic changes and further providing evidence for explaining their biogeographic patterns. Our results support the hypothesis that geologic and climatic events play important roles in driving biological diversification of organisms in the QTP area. The Pliocene uplift of the QTP and following climatic changes most likely promoted both the inter- and intraspecific divergence of Primula sect. Armerina. This study also illustrates how niche evolution under climatic changes influences biogeographic

  18. Global dynamics of multi-group SEI animal disease models with indirect transmission

    International Nuclear Information System (INIS)

    Wang, Yi; Cao, Jinde

    2014-01-01

    A challenge to multi-group epidemic models in mathematical epidemiology is the exploration of global dynamics. Here we formulate multi-group SEI animal disease models with indirect transmission via contaminated water. Under biologically motivated assumptions, the basic reproduction number R 0 is derived and established as a sharp threshold that completely determines the global dynamics of the system. In particular, we prove that if R 0 <1, the disease-free equilibrium is globally asymptotically stable, and the disease dies out; whereas if R 0 >1, then the endemic equilibrium is globally asymptotically stable and thus unique, and the disease persists in all groups. Since the weight matrix for weighted digraphs may be reducible, the afore-mentioned approach is not directly applicable to our model. For the proofs we utilize the classical method of Lyapunov, graph-theoretic results developed recently and a new combinatorial identity. Since the multiple transmission pathways may correspond to the real world, the obtained results are of biological significance and possible generalizations of the model are also discussed

  19. Interface discontinuity factors in the modal Eigenspace of the multigroup diffusion matrix

    International Nuclear Information System (INIS)

    Garcia-Herranz, N.; Herrero, J.J.; Cuervo, D.; Ahnert, C.

    2011-01-01

    Interface discontinuity factors based on the Generalized Equivalence Theory are commonly used in nodal homogenized diffusion calculations so that diffusion average values approximate heterogeneous higher order solutions. In this paper, an additional form of interface correction factors is presented in the frame of the Analytic Coarse Mesh Finite Difference Method (ACMFD), based on a correction of the modal fluxes instead of the physical fluxes. In the ACMFD formulation, implemented in COBAYA3 code, the coupled multigroup diffusion equations inside a homogenized region are reduced to a set of uncoupled modal equations through diagonalization of the multigroup diffusion matrix. Then, physical fluxes are transformed into modal fluxes in the Eigenspace of the diffusion matrix. It is possible to introduce interface flux discontinuity jumps as the difference of heterogeneous and homogeneous modal fluxes instead of introducing interface discontinuity factors as the ratio of heterogeneous and homogeneous physical fluxes. The formulation in the modal space has been implemented in COBAYA3 code and assessed by comparison with solutions using classical interface discontinuity factors in the physical space. (author)

  20. Second order time evolution of the multigroup diffusion and P1 equations for radiation transport

    International Nuclear Information System (INIS)

    Olson, Gordon L.

    2011-01-01

    Highlights: → An existing multigroup transport algorithm is extended to be second-order in time. → A new algorithm is presented that does not require a grey acceleration solution. → The two algorithms are tested with 2D, multi-material problems. → The two algorithms have comparable computational requirements. - Abstract: An existing solution method for solving the multigroup radiation equations, linear multifrequency-grey acceleration, is here extended to be second order in time. This method works for simple diffusion and for flux-limited diffusion, with or without material conduction. A new method is developed that does not require the solution of an averaged grey transport equation. It is effective solving both the diffusion and P 1 forms of the transport equation. Two dimensional, multi-material test problems are used to compare the solution methods.

  1. MVP/GMVP version 3. General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

    2017-03-01

    In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)

  2. Preparation of processed nuclear data libraries for thermal, fast and fusion research and power reactor applications. Texts of papers presented at the IAEA consultants' meeting

    International Nuclear Information System (INIS)

    Ganesan, S.

    1996-04-01

    The report contains 12 papers on nuclear data processing activities in Algeria, India, Indonesia, Italy, Japan, Republic of Korea, the Netherlands, Russia, Slovenia, United Kingdom, U.S.A., including ENDF formatted nuclear data libraries and computer code systems such as NJOY, AMPX, NSLINK, MCNP, multigroup data schemes such as WIMS, ABBN, and others. The role of the IAEA Nuclear Data Section in the establishment of nuclear data centers in developing countries is reviewed. (author). Refs, figs, tabs

  3. MC2-2: a code to calculate fast neutron spectra and multigroup cross sections

    International Nuclear Information System (INIS)

    Henryson, H. II; Toppel, B.J.; Stenberg, C.G.

    1976-06-01

    MC 2 -2 is a program to solve the neutron slowing down problem using basic neutron data derived from the ENDF/B data files. The spectrum calculated by MC 2 -2 is used to collapse the basic data to multigroup cross sections for use in standard reactor neutronics codes. Four different slowing down formulations are used by MC 2 -2: multigroup, continuous slowing down using the Goertzel-Greuling or Improved Goertzel-Greuling moderating parameters, and a hyper-fine-group integral transport calculation. Resolved and unresolved resonance cross sections are calculated accounting for self-shielding, broadening and overlap effects. This document provides a description of the MC 2 -2 program. The physics and mathematics of the neutron slowing down problem are derived and detailed information is provided to aid the MC 2 -2 user in preparing input for the program and implementation of the program on IBM 370 or CDC 7600 computers

  4. Young Adults’ Attitude Towards Advertising: a multi-group analysis by ethnicity

    OpenAIRE

    Hiram Ting; Ernest Cyril de Run; Ramayah Thurasamy

    2015-01-01

    Objective – This study aims to investigate the attitude of Malaysian young adults towards advertising. How this segment responds to advertising, and how ethnic/cultural differences moderate are assessed.Design/methodology/approach – A quantitative questionnaire is used to collect data at two universities. Purposive sampling technique is adopted to ensure the sample represents the actual population. Structural equation modelling (SEM) and multi-group analysis (MGA) are utilized in analysis.Fin...

  5. Mining the multigroup-discrete ordinates algorithm for high quality solutions

    International Nuclear Information System (INIS)

    Ganapol, B.D.; Kornreich, D.E.

    2005-01-01

    A novel approach to the numerical solution of the neutron transport equation via the discrete ordinates (SN) method is presented. The new technique is referred to as 'mining' low order (SN) numerical solutions to obtain high order accuracy. The new numerical method, called the Multigroup Converged SN (MGCSN) algorithm, is a combination of several sequence accelerators: Romberg and Wynn-epsilon. The extreme accuracy obtained by the method is demonstrated through self consistency and comparison to the independent semi-analytical benchmark BLUE. (authors)

  6. Simulation of the burnup in cell calculation using the WIMSD-5B Code considering different nuclear data libraries

    International Nuclear Information System (INIS)

    Tavares, Desirée Yael de Sena; Silva, Adilson Costa da; Lima, Zelmo Rodrigues de

    2017-01-01

    This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows to determine the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups. (author)

  7. Simulation of the burnup in cell calculation using the WIMSD-5B Code considering different nuclear data libraries

    Energy Technology Data Exchange (ETDEWEB)

    Tavares, Desirée Yael de Sena; Silva, Adilson Costa da; Lima, Zelmo Rodrigues de, E-mail: zelmolima@yahoo.com.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows to determine the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose the neutron transport theory is used in a problem with dimensional reduction, but in contrast is considered a large number of groups associated with the neutron spectrum. There are a variety of reactor physics codes that determine the nuclear parameters by solving the neutron transport equation applied to an equivalent cell representing a fuel element. The WIMSD-5B code is a deterministic code that solves the transport equation using collision probability method. The simulation of fuel burning in the cell calculation took into account different nuclear data libraries. The WIMSD-5B code supports several nuclear data libraries and in the present work the following libraries were used: IAEA, ENDFB-VII.1, JENDL3.2, JEFF3.1 and JEF2.2, all formatted for 69 energy groups. (author)

  8. Nuclear data libraries for Tripoli-3.5 code; Bibliotheques de donnees nucleaires pour le code tripoli-3.5

    Energy Technology Data Exchange (ETDEWEB)

    Vergnaud, Th

    2001-07-01

    The TRIPOLI-3 code uses multigroup nuclear data libraries generated using the NJOY-THEMIS suite of modules: for neutrons, they are produced from the ENDF/B-VI evaluations and cover the range between 20 MeV and 10{sup -5} eV, either in 315 groups and for one temperature, or in 3209 groups and for five temperatures; for gamma-rays, they are from JEF2 and are processed in groups between 14 MeV and keV. The probability tables used for the neutron transport calculations have been derived from the ENDF/B-VI evaluations using the CALENDF code. Cross sections for gamma production by neutron interaction (fission, capture or inelastic scattering) have been derived from ENDF/B-VI in 315 neutron groups and 75 gamma groups. The code also uses two response function libraries: for neutrons; based on several sources, in particular the dosimetry libraries IRDF/85 and IRDF/90; for gamma-rays it is based on the JEF2 evaluation and contains the kerma factors for all the elements and cross sections for all interactions. (author)

  9. X-231B technology demonstration for in situ treatment of contaminated soil: Contaminant characterization and three dimensional spatial modeling

    International Nuclear Information System (INIS)

    West, O.R.; Siegrist, R.L.; Mitchell, T.J.; Pickering, D.A.; Muhr, C.A.; Greene, D.W.; Jenkins, R.A.

    1993-11-01

    Fine-textured soils and sediments contaminated by trichloroethylene (TCE) and other chlorinated organics present a serious environmental restoration challenge at US Department of Energy (DOE) sites. DOE and Martin Marietta Energy Systems, Inc. initiated a research and demonstration project at Oak Ridge National Laboratory. The goal of the project was to demonstrate a process for closure and environmental restoration of the X-231B Solid Waste Management Unit at the DOE Portsmouth Gaseous Diffusion Plant. The X-231B Unit was used from 1976 to 1983 as a land disposal site for waste oils and solvents. Silt and clay deposits beneath the unit were contaminated with volatile organic compounds and low levels of radioactive substances. The shallow groundwater was also contaminated, and some contaminants were at levels well above drinking water standards. This document begins with a summary of the subsurface physical and contaminant characteristics obtained from investigative studies conducted at the X-231B Unit prior to January 1992 (Sect. 2). This is then followed by a description of the sample collection and analysis methods used during the baseline sampling conducted in January 1992 (Sect. 3). The results of this sampling event were used to develop spatial models for VOC contaminant distribution within the X-231B Unit

  10. Parallel computation of multigroup reactivity coefficient using iterative method

    Science.gov (United States)

    Susmikanti, Mike; Dewayatna, Winter

    2013-09-01

    One of the research activities to support the commercial radioisotope production program is a safety research target irradiation FPM (Fission Product Molybdenum). FPM targets form a tube made of stainless steel in which the nuclear degrees of superimposed high-enriched uranium. FPM irradiation tube is intended to obtain fission. The fission material widely used in the form of kits in the world of nuclear medicine. Irradiation FPM tube reactor core would interfere with performance. One of the disorders comes from changes in flux or reactivity. It is necessary to study a method for calculating safety terrace ongoing configuration changes during the life of the reactor, making the code faster became an absolute necessity. Neutron safety margin for the research reactor can be reused without modification to the calculation of the reactivity of the reactor, so that is an advantage of using perturbation method. The criticality and flux in multigroup diffusion model was calculate at various irradiation positions in some uranium content. This model has a complex computation. Several parallel algorithms with iterative method have been developed for the sparse and big matrix solution. The Black-Red Gauss Seidel Iteration and the power iteration parallel method can be used to solve multigroup diffusion equation system and calculated the criticality and reactivity coeficient. This research was developed code for reactivity calculation which used one of safety analysis with parallel processing. It can be done more quickly and efficiently by utilizing the parallel processing in the multicore computer. This code was applied for the safety limits calculation of irradiated targets FPM with increment Uranium.

  11. A code system to generate multigroup cross-sections using basic data

    International Nuclear Information System (INIS)

    Garg, S.B.; Kumar, Ashok

    1978-01-01

    For the neutronic studies of nuclear reactors, multigroup cross-sections derived from the basic energy point data are needed. In order to carry out the design based studies, these cross-sections should also incorporate the temperature and fuel concentration effects. To meet these requirements, a code system comprising of RESRES, UNRES, FIGERO, INSCAT, FUNMO, AVER1 and BGPONE codes has been adopted. The function of each of these codes is discussed. (author)

  12. ERRORJ, Multigroup covariance matrices generation from ENDF-6 format

    International Nuclear Information System (INIS)

    Chiba, Go

    2007-01-01

    1 - Description of program or function: ERRORJ produces multigroup covariance matrices from ENDF-6 format following mainly the methods of the ERRORR module in NJOY94.105. New version differs from previous version in the following features: Additional features in ERRORJ with respect to the NJOY94.105/ERRORR module: - expands processing for the covariance matrices of resolved and unresolved resonance parameters; - processes average cosine of scattering angle and fission spectrum; - treats cross-correlation between different materials and reactions; - accepts input of multigroup constants with various forms (user input, GENDF, etc.); - outputs files with various formats through utility NJOYCOVX (COVERX format, correlation matrix, relative error and standard deviation); - uses a 1% sensitivity method for processing of resonance parameters; - ERRORJ can process the JENDL-3.2 and 3.3 covariance matrices. Additional features of the version 2 with respect to the previous version of ERRORJ: - Since the release of version 2, ERRORJ has been modified to increase its reliability and stability, - calculation of the correlation coefficients in the resonance region, - Option for high-speed calculation is implemented, - Perturbation amount is optimised in a sensitivity calculation, - Effect of the resonance self-shielding can be considered, - a compact covariance format (LCOMP=2) proposed by N. M. Larson can be read. Additional features of the version 2.2.1 with respect to the previous version of ERRORJ: - Several routines were modified to reduce calculation time. The new one needs shorter calculation time (50-70%) than the old version without changing results. - In the U-233 and Pu-241 files of JENDL-3.3 an inconsistency between resonance parameters in MF=32 and those in MF=2 was corrected. NEA-1676/06: This version differs from the previous one (NEA-1676/05) in the following: ERRORJ2.2.1 was modified to treat the self-shielding effect accurately. NEA-1676/07: This version

  13. Study on the Control Strategy of Shifting Time Involving Multigroup Clutches

    Directory of Open Access Journals (Sweden)

    Zhen Zhu

    2016-01-01

    Full Text Available This paper focuses on the control strategy of shifting time involving multigroup clutches for a hydromechanical continuously variable transmission (HMCVT. The dynamic analyses of mathematical models are presented in this paper, and the simulation models are used to study the control strategy of HMCVT. Simulations are performed in Simulation X platform to investigate the shifting time of clutches under different operating conditions. On this basis, simulation analysis and test verification of two typical conditions, which play the decisive roles for the shifting quality, are carried out. The results show that there are differences in the shifting time of the two typical conditions. In the shifting process from the negative transmission of hydromechanical ranges to the positive transmission of hydromechanical ranges, the control strategy based on the shifting time is switching the clutches of shifting mechanism firstly and then disengaging a group of clutches of planetary gear mechanism and engaging another group of the clutches of planetary gear mechanism lastly. In the shifting process from the hydraulic range to the hydromechanical range, the control strategy based on the shifting time is switching the clutches of hydraulic shifting mechanism and planetary gear mechanism at first and then engaging the clutch of shifting mechanism.

  14. Soft x-ray driven ablation and its positive use for a new efficient acceleration

    International Nuclear Information System (INIS)

    Yabe, Takashi; Kiyokawa, Shuji; Mochizuki, Takayasu; Sakabe, Shuji; Yamanaka, Chiyoe

    1983-01-01

    The ablation process driven by soft X-ray is investigated by one-dimensional hydrodynamic code coupled with LTE, average ion model and multi-group radiation package. The following two major results are obtained: (1) the ablation pressure and mass ablation rate scalings, and (2) a new acceleration scheme which positively uses the unique property of soft X-ray transport. (author)

  15. A boundary integral equation for boundary element applications in multigroup neutron diffusion theory

    International Nuclear Information System (INIS)

    Ozgener, B.

    1998-01-01

    A boundary integral equation (BIE) is developed for the application of the boundary element method to the multigroup neutron diffusion equations. The developed BIE contains no explicit scattering term; the scattering effects are taken into account by redefining the unknowns. Boundary elements of the linear and constant variety are utilised for validation of the developed boundary integral formulation

  16. ZZ FENDL-2, Evaluated Nuclear Data Library for Fusion Neutronics Applications

    International Nuclear Information System (INIS)

    2005-01-01

    Description: FENDL: Fusion Evaluated Nuclear Data Library. Materials/nuclides: H 1 , H 2 , H 3 , He 3 , He 4 , Li 6 , Li 7 , Be 9 , B 10 , B 11 , C 12 , N 14 , N 15 , O 16 , F 19 , Na 23 , Mg nat , Al 27 , Si 28 , Si 29 , Si 30 , P 31 , S nat , Cl 35 , Cl 37 , K nat , Ca nat , 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 , Zr nat , Nb 93 , Mo 92 , Mo 94 , Mo 95 , Mo 96 , Mo 97 , Mo 98 , Mo 100 , Sn nat , Ta 181 , W 182 , W 183 , W 184 , W 186 , Au 197 , Pb 206 , Pb 207 , Pb 208 , Bi 209 . Photo-atomic data. IAEA1364/02: FENDL version 2.0 consists of the following sub-libraries: - ACTIVATION (FENDL/A-2.0)- neutron activation cross sections for 13006 reactions on 739 targets ranging from 1-H up to 248-Cm at incident energies up to 20 MeV. Pointwise and processed data in different formats are included. Plots are available. - DECAY (FENDL/D-2.0) - decay properties (decay type, decay energy, and half life) for 1867 nuclides and isomers. FENDL/D-2.0 sub-library is complementary to the activation sub-library. Pointwise and processed data are included. - DOSIMETRY (FENDL/DS-2.0) - neutron cross sections to be used for reactor neutron dosimetry by foil activation, radiation damage cross-sections, and benchmark neutron spectra. This sub-library is identical to the International Reactor Dosimetry File (IRDF-90). Pointwise and processed data are included. - FUSION (FENDL/C-2.0) - charged-particle cross sections for the following fusion reactions: 1-H 2 (d,n)2-He 3 , 1-H 2 (d,p)1-H 3 , 2-He 3 (d,p)2-He 4 , 1-H-3(t,2n)2-He 4 , and 1-H 3 (d,n)2-He 4 . Pointwise and processed data are included. - TRANSPORT - validated basic nuclear data (neutron-nucleus interaction including photon production, and photon-atom interaction cross sections) for 57 nuclides relevant for fusion. In addition to the pointwise data (FENDL/E-2.0), the sub-library

  17. Comparison of dissimilarity measures for cluster analysis of X-ray diffraction data from combinatorial libraries

    Science.gov (United States)

    Iwasaki, Yuma; Kusne, A. Gilad; Takeuchi, Ichiro

    2017-12-01

    Machine learning techniques have proven invaluable to manage the ever growing volume of materials research data produced as developments continue in high-throughput materials simulation, fabrication, and characterization. In particular, machine learning techniques have been demonstrated for their utility in rapidly and automatically identifying potential composition-phase maps from structural data characterization of composition spread libraries, enabling rapid materials fabrication-structure-property analysis and functional materials discovery. A key issue in development of an automated phase-diagram determination method is the choice of dissimilarity measure, or kernel function. The desired measure reduces the impact of confounding structural data issues on analysis performance. The issues include peak height changes and peak shifting due to lattice constant change as a function of composition. In this work, we investigate the choice of dissimilarity measure in X-ray diffraction-based structure analysis and the choice of measure's performance impact on automatic composition-phase map determination. Nine dissimilarity measures are investigated for their impact in analyzing X-ray diffraction patterns for a Fe-Co-Ni ternary alloy composition spread. The cosine, Pearson correlation coefficient, and Jensen-Shannon divergence measures are shown to provide the best performance in the presence of peak height change and peak shifting (due to lattice constant change) when the magnitude of peak shifting is unknown. With prior knowledge of the maximum peak shifting, dynamic time warping in a normalized constrained mode provides the best performance. This work also serves to demonstrate a strategy for rapid analysis of a large number of X-ray diffraction patterns in general beyond data from combinatorial libraries.

  18. COMPAR: system to compare multigroup cross sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 AND XLACS

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1987-11-01

    A system, composed by the computer programs COMPAR and its interfaces, developed for comparing multigroup cross sections calculated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS, is presented. (author)

  19. REX1-87, Multigroup Neutron Cross-Sections from ENDF/B

    International Nuclear Information System (INIS)

    Gopalakrishnan, V.; Ganesan, S.

    1988-01-01

    1 - Description of program or function: The program calculates self- shielding factors for reactor applications from a pre-processed (linearized) evaluated nuclear data file in the ENDF/B format. 2 - Method of solution: Bondarenko definition of multigroup self- shielding factors invoking narrow resonance treatment is used. 3 - Restrictions on the complexity of the problem: a) Maximum no. of energy group is 620. b) Only the built-in forms of the weighting functions can be chosen. c) The program is strictly limited to resolved resonance region from physical considerations

  20. AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

    Energy Technology Data Exchange (ETDEWEB)

    Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.

    1992-10-01

    AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available.

  1. AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

    International Nuclear Information System (INIS)

    Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.

    1992-10-01

    AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available

  2. MVP/GMVP Version 3. General purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods (Translated document)

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu; Okumura, Keisuke; Sakurai, Takeshi; Mori, Takamasa

    2017-03-01

    In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two Monte Carlo codes MVP (continuous-energy method) and GMVP (multigroup method) have been developed at Japan Atomic Energy Agency. The codes have adopted a vectorized algorithm and have been developed for vector-type supercomputers. They also support parallel processing with a standard parallelization library MPI and thus a speed-up of Monte Carlo calculations can be achieved on general computing platforms. The first and second versions of the codes were released in 1994 and 2005, respectively. They have been extensively improved and new capabilities have been implemented. The major improvements and new capabilities are as follows: (1) perturbation calculation for effective multiplication factor, (2) exact resonant elastic scattering model, (3) calculation of reactor kinetics parameters, (4) photo-nuclear model, (5) simulation of delayed neutrons, (6) generation of group constants. This report describes the physical model, geometry description method used in the codes, new capabilities and input instructions. (author)

  3. Survey of computer codes which produce multigroup data from ENDF/B-IV

    International Nuclear Information System (INIS)

    Greene, N.M.

    1975-01-01

    The features of three code systems that produce multigroup neutron data are contrasted. This includes the ETOE-2/MC 2 -2/SDX, MINX/SPHINX and AMPX code packages. These systems all contain a fairly extensive set of processing capabilities with the current evaluated nuclear data files--ENDF/B. They were designed with different goals and applications in mind. This paper discusses some of their differences and the implications for particular situations

  4. AFCI-2.0 Neutron Cross Section Covariance Library

    Energy Technology Data Exchange (ETDEWEB)

    Herman, M.; Herman, M; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S.; Mughabghab, S.F.; Sonzogni, A.; Talou, P.; Chadwick, M.B.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Yount, P.G.

    2011-03-01

    The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R&D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78 structural

  5. AFCI-2.0 Neutron Cross Section Covariance Library

    International Nuclear Information System (INIS)

    Herman, M.; Oblozinsky, P.; Mattoon, C.M.; Pigni, M.; Hoblit, S.; Mughabghab, S.F.; Sonzogni, A.; Talou, P.; Chadwick, M.B.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Yount, P.G.

    2011-01-01

    The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R and D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78

  6. From Fourier Transforms to Singular Eigenfunctions for Multigroup Transport

    International Nuclear Information System (INIS)

    Ganapol, B.D.

    2001-01-01

    A new Fourier transform approach to the solution of the multigroup transport equation with anisotropic scattering and isotropic source is presented. Through routine analytical continuation, the inversion contour is shifted from the real line to produce contributions from the poles and cuts in the complex plane. The integrand along the branch cut is then recast in terms of matrix continuum singular eigenfunctions, demonstrating equivalence of Fourier transform inversion and the singular eigenfunction expansion. The significance of this paper is that it represents the initial step in revealing the intimate connection between the Fourier transform and singular eigenfunction approaches as well as serves as a basis for a numerical algorithm

  7. Simulate-HEX - The multi-group diffusion equation in hexagonal-z geometry

    International Nuclear Information System (INIS)

    Lindahl, S. O.

    2013-01-01

    The multigroup diffusion equation is solved for the hexagonal-z geometry by dividing each hexagon into 6 triangles. In each triangle, the Fourier solution of the wave equation is approximated by 8 plane waves to describe the intra-nodal flux accurately. In the end an efficient Finite Difference like equation is obtained. The coefficients of this equation depend on the flux solution itself and they are updated once per power/void iteration. A numerical example demonstrates the high accuracy of the method. (authors)

  8. DiSCaMB: a software library for aspherical atom model X-ray scattering factor calculations with CPUs and GPUs.

    Science.gov (United States)

    Chodkiewicz, Michał L; Migacz, Szymon; Rudnicki, Witold; Makal, Anna; Kalinowski, Jarosław A; Moriarty, Nigel W; Grosse-Kunstleve, Ralf W; Afonine, Pavel V; Adams, Paul D; Dominiak, Paulina Maria

    2018-02-01

    It has been recently established that the accuracy of structural parameters from X-ray refinement of crystal structures can be improved by using a bank of aspherical pseudoatoms instead of the classical spherical model of atomic form factors. This comes, however, at the cost of increased complexity of the underlying calculations. In order to facilitate the adoption of this more advanced electron density model by the broader community of crystallographers, a new software implementation called DiSCaMB , 'densities in structural chemistry and molecular biology', has been developed. It addresses the challenge of providing for high performance on modern computing architectures. With parallelization options for both multi-core processors and graphics processing units (using CUDA), the library features calculation of X-ray scattering factors and their derivatives with respect to structural parameters, gives access to intermediate steps of the scattering factor calculations (thus allowing for experimentation with modifications of the underlying electron density model), and provides tools for basic structural crystallographic operations. Permissively (MIT) licensed, DiSCaMB is an open-source C++ library that can be embedded in both academic and commercial tools for X-ray structure refinement.

  9. The Relationship Between Antecedent and Consequence Factors with Safety Behaviour in PT.X

    OpenAIRE

    Fitriani, Apris; Nawawiwetu, Erwin Dyah

    2017-01-01

    Background : Safety behaviour is an act worker to minimize the possibility of accidents in workplace. Based on the Antecedents-Behaviour-Consequence (ABC) theory, safety behaviour of worker related with the antecedent and consequence factors. Purpose : The purpose of this research was to study the association between antecedent and consequence factors with safety behaviour of workers in Ring Frame Unit Spinning II PT. X. Methods : This was an observational descriptive research with cross sect...

  10. Specifications for a two-dimensional multi-group scattering code: ALCI

    International Nuclear Information System (INIS)

    Bayard, J.P.; Guillou, A.; Lago, B.; Bureau du Colombier, M.J.; Guillou, G.; Vasseur, Ch.

    1965-02-01

    This report describes the specifications of the ALCI programme. This programme resolves the system of difference equations similar to the homogeneous problem of multigroup neutron scattering, with two dimensions in space, in the three geometries XY, RZ, RΘ. It is possible with this method to calculate geometric and composition criticalities and also to calculate the accessory problem on demand. The maximum number of points dealt with is 6000. The maximum permissible number of groups is 12. The internal iterations are treated by the method of alternating directions. The external iterations are accelerated using the extrapolation method due to Tchebychev. (authors) [fr

  11. Collapsing of multigroup cross sections in optimization problems solved by means of the maximum principle of Pontryagin

    International Nuclear Information System (INIS)

    Anton, V.

    1979-05-01

    A new formulation of multigroup cross section collapsing based on the conservation of point or zone value of hamiltonian is presented. This attempt is proper to optimization problems solved by means of maximum principle of Pontryagin. (author)

  12. Algorithm development and verification of UASCM for multi-dimension and multi-group neutron kinetics model

    International Nuclear Information System (INIS)

    Si, S.

    2012-01-01

    The Universal Algorithm of Stiffness Confinement Method (UASCM) for neutron kinetics model of multi-dimensional and multi-group transport equations or diffusion equations has been developed. The numerical experiments based on transport theory code MGSNM and diffusion theory code MGNEM have demonstrated that the algorithm has sufficient accuracy and stability. (authors)

  13. Continuous energy Monte Carlo method based homogenization multi-group constants calculation

    International Nuclear Information System (INIS)

    Li Mancang; Wang Kan; Yao Dong

    2012-01-01

    The efficiency of the standard two-step reactor physics calculation relies on the accuracy of multi-group constants from the assembly-level homogenization process. In contrast to the traditional deterministic methods, generating the homogenization cross sections via Monte Carlo method overcomes the difficulties in geometry and treats energy in continuum, thus provides more accuracy parameters. Besides, the same code and data bank can be used for a wide range of applications, resulting in the versatility using Monte Carlo codes for homogenization. As the first stage to realize Monte Carlo based lattice homogenization, the track length scheme is used as the foundation of cross section generation, which is straight forward. The scattering matrix and Legendre components, however, require special techniques. The Scattering Event method was proposed to solve the problem. There are no continuous energy counterparts in the Monte Carlo calculation for neutron diffusion coefficients. P 1 cross sections were used to calculate the diffusion coefficients for diffusion reactor simulator codes. B N theory is applied to take the leakage effect into account when the infinite lattice of identical symmetric motives is assumed. The MCMC code was developed and the code was applied in four assembly configurations to assess the accuracy and the applicability. At core-level, A PWR prototype core is examined. The results show that the Monte Carlo based multi-group constants behave well in average. The method could be applied to complicated configuration nuclear reactor core to gain higher accuracy. (authors)

  14. Hydrogen transport in a toroidal plasma using multigroup discrete-ordinates methodology

    International Nuclear Information System (INIS)

    Wienke, B.R.; Miller, W.F. Jr.; Seed, T.J.

    1979-01-01

    Neutral hydrogen transport in a fully ionized two-dimensional tokamak plasma was examined using discrete ordinates and contrasted with earlier analyses. In particular, curvature effects induced by toroidal geometries and ray effects caused by possible source localization were investigated. From an overview of the multigroup discrete-ordinates approximation, methodology in two-dimensional cylindrical geometry is detailed, mesh and plasma zoning procedures are sketched, and the piecewise polynomial solution algorithm on a triangular domain is obtained. Toroidal effects and comparisons as related to reaction rates and perticle spectra are examined for various model and source configurations

  15. Multigroup constants for charged particle elastic nuclear (plus interference) scattering of light isotopes

    International Nuclear Information System (INIS)

    Cullen, D.E.; Perkins, S.T.

    1977-01-01

    Multi-group averaged reaction rates and transfer matrices were calculated for charged particle induced elastic nuclear (plus interference) scattering. Results are presented using a ten group structure for all twenty-five permutations of projectile and target for the following charged particles: p, d, t, 3 He and alpha. Transfer matrices are presented in a simplified form for both incident projectile and the knock-ons; these matrices explicitly conserve energy

  16. ENDF/B-III scattering law library

    International Nuclear Information System (INIS)

    Goulo, V.; Lemmel, H.D.

    1989-12-01

    This library contains scattering law data S(α,β) for 10 materials used for thermal reactor and shielding calculations: H 2 O, D 2 O, Be, BeO, C, CH 2 , C 6 H 6 , UO 2 , ZrH x (for H), ZrH x (for Zr) for temperatures from 296 to 1000 or 1200 deg. K. Data are in ENDF/B-3 format. The library is available from the IAEA Nuclear Data Section on magnetic tape, costfree upon request. (author). Figs and tabs

  17. COMPAR: A system for comparing multigroup cross-sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1988-02-01

    A system consisting of the COMPAR computer program and its interfaces which was developed for comparing multigroup cross-sections generated by NJOY, GROUPIE, FLANGE-II, ETOG-3 and XLACS is presented. (author). 13 refs

  18. Some reciprocity-like relations in multi-group neutron diffusion and transport theory over bare homogeneous regions

    International Nuclear Information System (INIS)

    Modak, R.S.; Sahni, D.C.

    1996-01-01

    Some simple reciprocity-like relations that exist in multi-group neutron diffusion and transport theory over bare homogeneous regions are presented. These relations do not involve the adjoint solutions and are directly related to numerical schemes based on an explicit evaluation of the fission matrix. (author)

  19. Comparing Safari Tech Books Online and Books24x7 E-Book Collections: A Case Study from the University of British Columbia Library

    Science.gov (United States)

    Barsky, Eugene; Schattman, Lisa; Greenwood, Aleteia

    2009-01-01

    Most academic libraries are seeking to provide electronic access to the very dynamic and changing field of technology related material. "Safari Tech Books Online" and "Books24x7" are the major e-book collections in this area. We compared the "Safari Tech Books Online" and "Books24x7" e-book packages as to…

  20. Determination of Intermediate Resonance Parameter with RMET21 for nTRACER

    International Nuclear Information System (INIS)

    Sohail, Muhammad; Kim, Myung Hyun

    2012-01-01

    Ray Tracing based code nTRACER is being developed in Seoul National University that has the capability of 3-dimensional whole core neutron transport calculation. As a part of development of multi-group neutron cross section library for nTRACER, the current work is intended to accurately determine intermediate resonance parameters. Beside the systematic calculation of subgroup parameters for resonance self shielding calculation, intermediate resonance parameters itself can be as important as the multi-group neutron cross section in the library and its overall accuracy. In this paper lambda factors were computed using RMET21 from ENDF/B-VII.1 for nTRACER to investigate its dependence on temperature and background cross section and replaced with lambda factors from HELIOS multi-group library. The procedure used for determining the intermediate resonance parameter for the isotope under study is introduced in the next section. Oxygen being one of the primary nuclide in PWR fuel has been selected for intermediate resonance parameters calculation

  1. Improved Fiber Bragg Grating Array OFFH-CDMA System Using a Novel Frequency-Overlapping Multigroup Method

    Science.gov (United States)

    Peng, Wei-Ren; Lin, Wen-Piao; Chi, Sien

    2006-03-01

    The authors propose a novel frequency-overlapping multigroup scheme for a passive all-optical fast-frequency hopped code-division multiple-access (OFFH-CDMA) system based on fiber Bragg grating array (FBGA). In the conventional scheme, the users are assigned those codes constructed on the nonoverlapping frequency slots, and therefore the bandgaps between the adjacent gratings are wasted. To make a more efficient use of the optical spectrum, the proposed scheme divided the users into several groups, and assigned the codes, which interleaved to each other to the different groups. In addition to the higher utilization of the spectrum, the interleaved nature of the frequency allocations of different groups will make the groups less correlated and, hence, lower the multiple-access interference (MAI). The corresponding codeset and its constraints for this new scheme are also developed and analyzed. The performance of the system in terms of the correlation functions and bit error rate (BER) are given in both the conventional and the proposed schemes. The numerical results show that, with the multigroup scheme, performance is much improved compared to the conventional scheme.

  2. Adjustement of multigroup cross sections using fast reactor integral data

    International Nuclear Information System (INIS)

    Renke, C.A.C.

    1982-01-01

    A methodology for the adjustment of multigroup cross section is presented, structured with aiming to compatibility the limitated number of measured values of integral parameters known and disponible, and the great number of cross sections to be adjusted the group of cross section used is that obtained from the Carnaval II calculation system, understanding as formular the sets of calculation methods and data bases. The adjustment is realized, using the INCOAJ computer code, developed in function of one statistical formulation, structural from the bayer considerations, taking in account the measurement processes of cross section and integral parameters defined on statistical bases. (E.G.) [pt

  3. Extension and Verification of the Cross-Section Library for the VVER-1000 Surveillance Specimen Region

    International Nuclear Information System (INIS)

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

    2011-01-01

    The objective of this work is a generation of new version of the BGL multigroup cross-section to extend the region of its applicability. The existing library version is problem oriented for VVER-1000 type of reactors and was generated by collapsing of the VITAMIN-B6 problem independent cross-section fine-group library applying the VVER-1000 reactor middle plane spectrum in cylindrical geometry. The new version BGLex additionally contains cross-sections averaged on the corresponding spectra of the surveillance specimen's (SS) region for VVER-1000 type of reactors. Comparative analysis of the neutron spectra for different one-dimensional geometry models that could be applied for the cross-section collapsing using the software package SCALE, showed a high sensitivity of the results to the geometry model. That is why a neutron importance assessment was done for the SS region using the adjoint solution calculated by the two-dimensional code DORT and problem-independent library VITAMIN-B6. The one-dimensional geometry model applied to the cross-section collapsing were determined by the material limits above the reactor core in axial direction z as for every material a homogenization in radial direction was done. The material homogenization in radial direction was done by material weighing taking into account the adjoint solution as well as the neutron source. The one-dimensional geometry model comprising the homogenized weighed materials was applied for the cross-section generation of the fine-group library VITAMIN-B6 to the broad-group structure of BGL library. The new version BGLex was extended with cross-sections for the SS region. Verification and validation of the new version BGLex is forthcoming. It includes comparison between the calculated results with the new version BGLex and the libraries BGL and VITAMIN-B6 and comparison with experimental results. (author)

  4. Extension and Verification of the Cross-Section Library for the VVER- 1000 Surveillance Specimen Region

    International Nuclear Information System (INIS)

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

    2011-01-01

    The objective of this work is a generation of new version of the BGL multigroup cross-section to extend the region of its applicability. The existing library version is problem oriented for VVER-1000 type of reactors and was generated by collapsing of the VITAMIN-B6 problem independent cross-section fine-group library applying the VVER-1000 reactor middle plane spectrum in cylindrical geometry. The new version BGLex additionally contains cross-sections averaged on the corresponding spectra of the surveillance specimen's (SS) region for VVER-1000 type of reactors. Comparative analysis of the neutron spectra for different one-dimensional geometry models that could be applied for the cross-section collapsing using the software package SCALE, showed a high sensitivity of the results to the geometry model. That is why a neutron importance assessment was done for the SS region using the adjoint solution calculated by the two-dimensional code DORT and problem-independent library VITAMIN-B6. The one-dimensional geometry model applied to the cross-section collapsing were determined by the material limits above the reactor core in axial direction z as for every material a homogenization in radial direction was done. The material homogenization in radial direction was done by material weighing taking into account the adjoint solution as well as the neutron source. The one-dimensional geometry model comprising the homogenized weighed materials was applied for the cross-section generation of the fine-group library VITAMIN-B6 to the broad-group structure of BGL library. The new version BGLex was extended with cross-sections for the SS region. Verification and validation of the new version BGLex is forthcoming. It includes comparison between the calculated results with the new version BGLex and the libraries BGL and VITAMIN-B6 and comparison with experimental results. (author)

  5. Group-decoupled multi-group pin power reconstruction utilizing nodal solution 1D flux profiles

    International Nuclear Information System (INIS)

    Yu, Lulin; Lu, Dong; Zhang, Shaohong; Wang, Dezhong

    2014-01-01

    Highlights: • A direct fitting multi-group pin power reconstruction method is developed. • The 1D nodal solution flux profiles are used as the condition. • The least square fit problem is analytically solved. • A slowing down source improvement method is applied. • The method shows good accuracy for even challenging problems. - Abstract: A group-decoupled direct fitting method is developed for multi-group pin power reconstruction, which avoids both the complication of obtaining 2D analytic multi-group flux solution and any group-coupled iteration. A unique feature of the method is that in addition to nodal volume and surface average fluxes and corner fluxes, transversely-integrated 1D nodal solution flux profiles are also used as the condition to determine the 2D intra-nodal flux distribution. For each energy group, a two-dimensional expansion with a nine-term polynomial and eight hyperbolic functions is used to perform a constrained least square fit to the 1D intra-nodal flux solution profiles. The constraints are on the conservation of nodal volume and surface average fluxes and corner fluxes. Instead of solving the constrained least square fit problem numerically, we solve it analytically by fully utilizing the symmetry property of the expansion functions. Each of the 17 unknown expansion coefficients is expressed in terms of nodal volume and surface average fluxes, corner fluxes and transversely-integrated flux values. To determine the unknown corner fluxes, a set of linear algebraic equations involving corner fluxes is established via using the current conservation condition on all corners. Moreover, an optional slowing down source improvement method is also developed to further enhance the accuracy of the reconstructed flux distribution if needed. Two test examples are shown with very good results. One is a four-group BWR mini-core problem with all control blades inserted and the other is the seven-group OECD NEA MOX benchmark, C5G7

  6. Testing a new multigroup inference approach to reconstructing past environmental conditions

    Directory of Open Access Journals (Sweden)

    Maria RIERADEVALL

    2008-08-01

    Full Text Available A new, quantitative, inference model for environmental reconstruction (transfer function, based for the first time on the simultaneous analysis of multigroup species, has been developed. Quantitative reconstructions based on palaeoecological transfer functions provide a powerful tool for addressing questions of environmental change in a wide range of environments, from oceans to mountain lakes, and over a range of timescales, from decades to millions of years. Much progress has been made in the development of inferences based on multiple proxies but usually these have been considered separately, and the different numeric reconstructions compared and reconciled post-hoc. This paper presents a new method to combine information from multiple biological groups at the reconstruction stage. The aim of the multigroup work was to test the potential of the new approach to making improved inferences of past environmental change by improving upon current reconstruction methodologies. The taxonomic groups analysed include diatoms, chironomids and chrysophyte cysts. We test the new methodology using two cold-environment training-sets, namely mountain lakes from the Pyrenees and the Alps. The use of multiple groups, as opposed to single groupings, was only found to increase the reconstruction skill slightly, as measured by the root mean square error of prediction (leave-one-out cross-validation, in the case of alkalinity, dissolved inorganic carbon and altitude (a surrogate for air-temperature, but not for pH or dissolved CO2. Reasons why the improvement was less than might have been anticipated are discussed. These can include the different life-forms, environmental responses and reaction times of the groups under study.

  7. Using Multi-Group Confirmatory Factor Analysis to Evaluate Cross-Cultural Research: Identifying and Understanding Non-Invariance

    Science.gov (United States)

    Brown, Gavin T. L.; Harris, Lois R.; O'Quin, Chrissie; Lane, Kenneth E.

    2017-01-01

    Multi-group confirmatory factor analysis (MGCFA) allows researchers to determine whether a research inventory elicits similar response patterns across samples. If statistical equivalence in responding is found, then scale score comparisons become possible and samples can be said to be from the same population. This paper illustrates the use of…

  8. NUMERICAL MULTIGROUP TRANSIENT ANALYSIS OF SLAB NUCLEAR REACTOR WITH THERMAL FEEDBACK

    Directory of Open Access Journals (Sweden)

    Filip Osuský

    2016-12-01

    Full Text Available The paper describes a new numerical code for multigroup transient analyses with thermal feedback. The code is developed at Institute of Nuclear and Physical Engineering. It is necessary to carefully investigate transient states of fast neutron reactors, due to recriticality issues after accident scenarios. The code solves numerical diffusion equation for 1D problem with possible neutron source incorporation. Crank-Nicholson numerical method is used for the transient states. The investigated cases are describing behavior of PWR fuel assembly inside of spent fuel pool and with the incorporated neutron source for better illustration of thermal feedback.

  9. TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticality studies

    International Nuclear Information System (INIS)

    Ermumcu, G.; Gonnord, J.; Nimal, J.C.

    1980-01-01

    TRIMARAN is developed for safety analysis of nuclear components containing fissionable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method, in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies

  10. TRIMARAN: a three dimensional multigroup P1 Monte Carlo code for criticallity studies

    International Nuclear Information System (INIS)

    Ermuncu, G.; Gonnord, J.; Nimal, J.C.

    1980-04-01

    TRIMARAN is developed for safety analysis of nuclar components containing fissionnable materials: shipping casks, storage and cooling pools, manufacture and reprocessing plants. It solves the transport equation by Monte Carlo method in general three dimensional geometry with multigroup P1 approximation. A special representation of cross sections and numbers has been developed in order to reduce considerably the computing cost and allow this three dimensional code to compete with standard numerical program used in parametric studies

  11. Multi-group dynamic quantum secret sharing with single photons

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Hongwei [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China); Ma, Haiqiang, E-mail: hqma@bupt.edu.cn [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China); Wei, Kejin [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China); Yang, Xiuqing [School of Science, Beijing Jiaotong University, Beijing 100044 (China); Qu, Wenxiu; Dou, Tianqi; Chen, Yitian; Li, Ruixue; Zhu, Wu [School of Science and State Key Laboratory of Information Photonics and Optical Communications, Beijing University of Posts and Telecommunications, Beijing 100876 (China)

    2016-07-15

    In this letter, we propose a novel scheme for the realization of single-photon dynamic quantum secret sharing between a boss and three dynamic agent groups. In our system, the boss can not only choose one of these three groups to share the secret with, but also can share two sets of independent keys with two groups without redistribution. Furthermore, the security of communication is enhanced by using a control mode. Compared with previous schemes, our scheme is more flexible and will contribute to a practical application. - Highlights: • A multi-group dynamic quantum secret sharing with single photons scheme is proposed. • Any one of the groups can be chosen to share secret through controlling the polarization of photons. • Two sets of keys can be shared simultaneously without redistribution.

  12. Impact of different libraries on the performance calculation of a modul-type pebble bed HTR

    International Nuclear Information System (INIS)

    Ohlig, U.; Brockmann, H.; Haas, K.A.; Teuchert, E.

    1991-01-01

    A new multigroup library for the GAM-THERMOS spectrum codes has been compiled from the sources ENDF/B-V and JEF-1. The progress in comparison to the 20 years old standard library has been studied for one specific reactor design of the Modular High Temperature Reactor. The study covers various aspects of the performance of the reactor both for the initial core and for the equilibrium cycle. For the multiplication factor k eff the different amounts to Δ k eff = 0.0164 in the startup reactor, which is mainly due to changes in the cross sections of 235 U. At the turn to the equilibrium cycle the difference reduces to Δ k eff = 0.0017 as due to various opposite tendencies in the data of the many involved nuclides. The change in the mass balance of the fissile materials is about 5%. The impact on the temperature coefficients is in the order of 4%, and the influence on other safety related properties of the reactor is lower than about 1 or 2 percent, confirming the confidence in formerly received results. (author). 10 refs, 3 figs, 6 tabs

  13. Skyshine analysis using various nuclear data files

    International Nuclear Information System (INIS)

    Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N.; Nomura, Y.; Tsubosaka, A.

    2000-01-01

    The calculations of the spacial distributions of dose rate for neutron and secondary photons, thermal neutron fluxes and space-energy distributions of neutron and photons near the air-ground interface were performed by MCNP and DORT codes. Different nuclear data files were used (ENDF/B-IV, ENDF/B-VI, FENDL-2, JENDL-3.2). Either the standard pointwise libraries (MCNP) or special libraries prepared by NJOY code from ENDF/B and others' files were used. Prepared multigroup coupled neutron and photon cross sections libraries for DORT code had CASK-40 group energy structures. The libraries contain pointwise or multigroup cross sections data for all elements included in the atmosphere and ground composition. The validation of the calculated results was performed with using the experimental data obtained for the series of measurements at RA reactor. (author)

  14. Readiness of librarians in public libraries towards integration of ...

    African Journals Online (AJOL)

    The study is designed to x-ray the level of preparedness of librarians in Nigerian Public libraries towards integrating social media to the provision of library and information services (LIS). The survey research method was adopted using public libraries in south-east geo-political zone of Nigeria. The population of study ...

  15. Investigations of safety-related parameters applying a new multi-group diffusion code for HTR transients

    International Nuclear Information System (INIS)

    Kasselmann, S.; Druska, C.; Lauer, A.

    2010-01-01

    The energy spectra of fast and thermal neutrons from fission reactions in the FZJ code TINTE are modelled by two broad energy groups. Present demands for increased numerical accuracy led to the question of how precise the 2-group approximation is compared to a multi-group model. Therefore a new simulation program called MGT (Multi Group TINTE) has recently been developed which is able to handle up to 43 energy groups. Furthermore, an internal spectrum calculation for the determination of cross-sections can be performed for each time step and location within the reactor. In this study the multi-group energy models are compared to former calculations with only two energy groups. Different scenarios (normal operation and design-basis accidents) have been defined for a high temperature pebble bed reactor design with annular core. The effect of an increasing number of energy groups on safety-related parameters like the fuel and coolant temperature, the nuclear heat source or the xenon concentration is studied. It has been found that for the studied scenarios the use of up to 8 energy groups is a good trade-off between precision and a tolerable amount of computing time. (orig.)

  16. Sub-cell balanced nodal expansion methods using S4 eigenfunctions for multi-group SN transport problems in slab geometry

    International Nuclear Information System (INIS)

    Hong, Ser Gi; Lee, Deokjung

    2015-01-01

    A highly accurate S 4 eigenfunction-based nodal method has been developed to solve multi-group discrete ordinate neutral particle transport problems with a linearly anisotropic scattering in slab geometry. The new method solves the even-parity form of discrete ordinates transport equation with an arbitrary S N order angular quadrature using two sub-cell balance equations and the S 4 eigenfunctions of within-group transport equation. The four eigenfunctions from S 4 approximation have been chosen as basis functions for the spatial expansion of the angular flux in each mesh. The constant and cubic polynomial approximations are adopted for the scattering source terms from other energy groups and fission source. A nodal method using the conventional polynomial expansion and the sub-cell balances was also developed to be used for demonstrating the high accuracy of the new methods. Using the new methods, a multi-group eigenvalue problem has been solved as well as fixed source problems. The numerical test results of one-group problem show that the new method has third-order accuracy as mesh size is finely refined and it has much higher accuracies for large meshes than the diamond differencing method and the nodal method using sub-cell balances and polynomial expansion of angular flux. For multi-group problems including eigenvalue problem, it was demonstrated that the new method using the cubic polynomial approximation of the sources could produce very accurate solutions even with large mesh sizes. (author)

  17. Multigroup transport calculations of critical and fuel assemblies with taking into account the scattering anisotropy

    International Nuclear Information System (INIS)

    Rubin, I.E.; Dneprovskaya, N.M.

    2005-01-01

    A technique for calculation of reactor lattices by means of the transmission probabilities with taking into account the scattering anisotropy is generalized for the multigroup case. The errors of the calculated multiplication coefficients and energy release distributions do noe exceed practically the errors, of these values, obtained by the Monte Carlo method. The proposed method is most effective when determining the small difference effects [ru

  18. Validation of multigroup neutron cross sections for the Advanced Neutron Source against the FOEHN critical experimental measurements

    International Nuclear Information System (INIS)

    Smith, L.A.; Gehin, J.C.; Worley, B.A.; Renier, J.P.

    1994-01-01

    The FOEHN critical experiments were analyzed to validate the use of multigroup cross sections in the design of the Advanced Neutron Source. Eleven critical configurations were evaluated using the KENO, DORT, and VENTURE neutronics codes. Eigenvalue and power density profiles were computed and show very good agreement with measured values

  19. Analysis of Environmental Friendly Library Based on the Satisfaction and Service Quality: study at Library “X”

    Science.gov (United States)

    Herdiansyah, Herdis; Satriya Utama, Andre; Safruddin; Hidayat, Heri; Gema Zuliana Irawan, Angga; Immanuel Tjandra Muliawan, R.; Mutia Pratiwi, Diana

    2017-10-01

    One of the factor that influenced the development of science is the existence of the library, which in this case is the college libraries. Library, which is located in the college environment, aims to supply collections of literatures to support research activities as well as educational for students of the college. Conceptually, every library now starts to practice environmental principles. For example, “X” library as a central library claims to be an environmental friendly library for practicing environmental friendly management, but the X library has not inserted the satisfaction and service aspect to the users, including whether it is true that environmental friendly process is perceived by library users. Satisfaction can be seen from the comparison between expectations and reality of library users. This paper analyzes the level of library user satisfaction with library services in the campus area and the gap between expectations and reality felt by the library users. The result of the research shows that there is a disparity between the hope of library management, which is sustainable and environmentally friendly with the reality in the management of the library, so that it has not given satisfaction to the users yet. The gap value of satisfaction that has the biggest difference is in the library collection with the value of 1.57; while for the smallest gap value is in the same service to all students with a value of 0.67.

  20. RADHEAT-V3, a code system for generating coupled neutron and gamma-ray group constants and analyzing radiation transport

    International Nuclear Information System (INIS)

    Koyama, Kinji; Taji, Yukichi; Miyasaka, Shun-ichi; Minami, Kazuyoshi.

    1977-07-01

    The modular code system RADHEAT is for producing coupled multigroup neutron and gamma-ray cross section sets, analyzing the neutron and gamma-ray transport, and calculating the energy deposition and atomic displacements due to these radiations in a nuclear reactor or shield. The basic neutron cross sections and secondary gamma-ray production data are taken from ENDF/B and POPOP4 libraries respectively. The system (1) generates multigroup neutron cross sections, energy deposition coefficients and atomic displacement factors due to neutron reactions, (2) generates multigroup gamma-ray cross sections and energy transfer coefficients, (3) generates secondary gamma-ray production cross sections, (4) combines these cross sections into the coupled set, (5) outputs and updates the multigroup cross section libraries in convenient formats for other transport codes, (6) analyzes the neutron and gamma-ray transport and calculates the energy deposition and the number density of atomic displacements in a medium, (7) collapses the cross sections to a broad-group structure, by option, using the weighting functions obtained by one-dimensional transport calculation, and (8) plots, by option, multigroup cross sections, and neutron and gamma-ray distributions. Definitions of the input data required in various options of the code system are also given. (auth.)

  1. Reactor physics of light water power reactors of the WWER-1000 type in the SKODA Trust. Part 2

    International Nuclear Information System (INIS)

    Zbytovsky, A.; Vyskocil, V.; Vacek, J.; Krysl, V.

    1979-01-01

    The system is described of multigroup transport programs and the respective data libraries used at the SKODA WORKS for calculating few-group diffusion constants. The importance is stressed of precisioning nuclear data. The programs for multigroup calculations are mostly written in FORTRAN-1900. The structure of microcalculations is described and the possible cost savings are indicated. (J.P.)

  2. Contribution to the validation of the Apollo code library for thermal neutron reactors

    International Nuclear Information System (INIS)

    Tellier, H.; Van der Gucht, C.; Vanuxeem, J.

    1988-03-01

    The neutron nuclear data which are needed by reactor physicists to perform core calculation are brought together in the evaluated files. The files are processed to provide multigroup cross sections. The accuracy of the core calculations depends on the initial data which are sometimes not accurate enough. Therefore the reactor physicists carry out integral experiments. We show in this paper, how the use of these integral experiments and the application of the tendency research method can improve the accuracy of the neutron data. This technique was applied to the validation of the Apollo code library. For this purpose 60 buckling measurements (34 for uranium fuel multiplying media and 26 for plutonium fuel multiplying media) and 42 spent fuel analysis were used. Small modifications of the initial data are proposed. The final values are compared which recent recommended values of microscopic data and the agreement is good [fr

  3. Processing and benchmarking of evaluated nuclear data file/b-viii.0β4 cross-section library by analysis of a series of critical experimental benchmark using the monte carlo code MCNP(X and NJOY2016

    Directory of Open Access Journals (Sweden)

    Kabach Ouadie

    2017-12-01

    Full Text Available To validate the new Evaluated Nuclear Data File (ENDF/B-VIII.0β4 library, 31 different critical cores were selected and used for a benchmark test of the important parameter keff. The four utilized libraries are processed using Nuclear Data Processing Code (NJOY2016. The results obtained with the ENDF/B-VIII.0β4 library were compared against those calculated with ENDF/B-VI.8, ENDF/B-VII.0, and ENDF/B-VII.1 libraries using the Monte Carlo N-Particle (MCNP(X code. All the MCNP(X calculations of keff values with these four libraries were compared with the experimentally measured results, which are available in the International Critically Safety Benchmark Evaluation Project. The obtained results are discussed and analyzed in this paper.

  4. The X-shooter Spectral Library and Carbon stars

    NARCIS (Netherlands)

    Gonneau, A.; Lançon, A.; Trager, S. C.; Chen, Y.; Peletier, R.; Aringer, B.; Nowotny, W.; Cambrésy, L.; Martins, F.; Nuss, E.; Palacios, A.

    2013-01-01

    Until recently, most empirical stellar spectral libraries were limited to a certain wavelength range or combined data from different stars, taken by different instruments of which some have low spectral resolution, limiting for instance our ability to analyze galaxies jointly in the ultraviolet, the

  5. Effective marketing of library and information services: problems and ...

    African Journals Online (AJOL)

    Effective marketing of library services is an important aspect of Librarianship that helps ... Therefore, this work x-rays the problems and prospects of effective marketing of library and ... EMAIL FREE FULL TEXT EMAIL FREE FULL TEXT

  6. The isotope density inverse problem in multigroup neutron transport

    International Nuclear Information System (INIS)

    Zazula, J.M.

    1981-01-01

    The inverse problem for stationary multigroup anisotropic neutron transport is discussed in order to search for isotope densities in multielement medium. The spatial- and angular-integrated form of neutron transport equation, in terms of the flux in a group - density of an element spatial correlation, leads to a set of integral functionals for the densities weighted by the group fluxes. Some methods of approximation to make the problem uniquently solvable are proposed. Particularly P 0 angular flux information and the spherically-symetrical geometry of an infinite medium are considered. The numerical calculation using this method related to sooner evaluated direct problem data gives promising agreement with primary densities. This approach would be the basis for further application in an elemental analysis of a medium, using an isotopic neutron source and a moving, energy-dependent neutron detector. (author)

  7. Analysis of sensitive questions across cultures : An application of multigroup item randomized response theory to sexual attitudes and behavior

    NARCIS (Netherlands)

    de Jong, M.G.; Pieters, R.; Stremersch, S.

    2012-01-01

    Answers to sensitive questions are prone to social desirability bias. If not properly addressed, the validity of the research can be suspect. This article presents multigroup item randomized response theory (MIRRT) to measure self-reported sensitive topics across cultures. The method was

  8. Skyshine analysis using various nuclear data files

    Energy Technology Data Exchange (ETDEWEB)

    Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Nomura, Y.; Tsubosaka, A. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-03-01

    The calculations of the spacial distributions of dose rate for neutron and secondary photons, thermal neutron fluxes and space-energy distributions of neutron and photons near the air-ground interface were performed by MCNP and DORT codes. Different nuclear data files were used (ENDF/B-IV, ENDF/B-VI, FENDL-2, JENDL-3.2). Either the standard pointwise libraries (MCNP) or special libraries prepared by NJOY code from ENDF/B and others' files were used. Prepared multigroup coupled neutron and photon cross sections libraries for DORT code had CASK-40 group energy structures. The libraries contain pointwise or multigroup cross sections data for all elements included in the atmosphere and ground composition. The validation of the calculated results was performed with using the experimental data obtained for the series of measurements at RA reactor. (author)

  9. Linear triangle finite element formulation for multigroup neutron transport analysis with anisotropic scattering

    Energy Technology Data Exchange (ETDEWEB)

    Lillie, R.A.; Robinson, J.C.

    1976-05-01

    The discrete ordinates method is the most powerful and generally used deterministic method to obtain approximate solutions of the Boltzmann transport equation. A finite element formulation, utilizing a canonical form of the transport equation, is here developed to obtain both integral and pointwise solutions to neutron transport problems. The formulation is based on the use of linear triangles. A general treatment of anisotropic scattering is included by employing discrete ordinates-like approximations. In addition, multigroup source outer iteration techniques are employed to perform group-dependent calculations. The ability of the formulation to reduce substantially ray effects and its ability to perform streaming calculations are demonstrated by analyzing a series of test problems. The anisotropic scattering and multigroup treatments used in the development of the formulation are verified by a number of one-dimensional comparisons. These comparisons also demonstrate the relative accuracy of the formulation in predicting integral parameters. The applicability of the formulation to nonorthogonal planar geometries is demonstrated by analyzing a hexagonal-type lattice. A small, high-leakage reactor model is analyzed to investigate the effects of varying both the spatial mesh and order of angular quadrature. This analysis reveals that these effects are more pronounced in the present formulation than in other conventional formulations. However, the insignificance of these effects is demonstrated by analyzing a realistic reactor configuration. In addition, this final analysis illustrates the importance of incorporating anisotropic scattering into the finite element formulation. 8 tables, 29 figures.

  10. Linear triangle finite element formulation for multigroup neutron transport analysis with anisotropic scattering

    International Nuclear Information System (INIS)

    Lillie, R.A.; Robinson, J.C.

    1976-05-01

    The discrete ordinates method is the most powerful and generally used deterministic method to obtain approximate solutions of the Boltzmann transport equation. A finite element formulation, utilizing a canonical form of the transport equation, is here developed to obtain both integral and pointwise solutions to neutron transport problems. The formulation is based on the use of linear triangles. A general treatment of anisotropic scattering is included by employing discrete ordinates-like approximations. In addition, multigroup source outer iteration techniques are employed to perform group-dependent calculations. The ability of the formulation to reduce substantially ray effects and its ability to perform streaming calculations are demonstrated by analyzing a series of test problems. The anisotropic scattering and multigroup treatments used in the development of the formulation are verified by a number of one-dimensional comparisons. These comparisons also demonstrate the relative accuracy of the formulation in predicting integral parameters. The applicability of the formulation to nonorthogonal planar geometries is demonstrated by analyzing a hexagonal-type lattice. A small, high-leakage reactor model is analyzed to investigate the effects of varying both the spatial mesh and order of angular quadrature. This analysis reveals that these effects are more pronounced in the present formulation than in other conventional formulations. However, the insignificance of these effects is demonstrated by analyzing a realistic reactor configuration. In addition, this final analysis illustrates the importance of incorporating anisotropic scattering into the finite element formulation. 8 tables, 29 figures

  11. AUS module MIRANDA - a data preparation code based on multiregion resonance theory

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1977-07-01

    MIRANDA is a data preparation module of the AUS reactor neutronics scheme and is used to prepare multigroup cross-section data which are pertinent to a particular reactor system from a general purpose multigroup library of cross sections. The cross-section library has been prepared from ENDF/B and includes temperature dependent data and resonance cross sections represented by subgroup parameters. The MIRANDA module includes a multiregion resonance calculation in slab, cylinder or cluster geometry, a homogeneous Bsub(L) flux solution, and a group condensation facility. Interaction with other AUS modules, particularly for burnup calculations, is provided. (Author)

  12. MULTI - A multigroup or multipoint P{sub 3} programme for calculating thermal neutron spectra in a reactor cell

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M V [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1968-06-15

    Programme MULTI calculates the space energy distribution of thermal neutrons in a multizone, cylindrical, infinitely long reactor lattice by using the multigroup or multipoint P{sub 3} approximation. This report presents a short description of the algorithm and the programme and gives the instructions for its exploitation. (author)

  13. FPDCYS and FPSPEC: computer programs for calculating fission-product beta and gamma multigroup spectra from ENDF/B-IV data

    International Nuclear Information System (INIS)

    Stamatelatos, M.G.; England, T.R.

    1977-05-01

    FPDCYS and FPSPEC are two FORTRAN computer programs used at the Los Alamos Scientific Laboratory (LASL), in conjunction with the CINDER-10 program, for calculating cumulative fission-product beta and/or gamma multigroup spectra in arbitrary energy structures, and for arbitrary neutron irradiation periods and cooling times. FPDCYS processes ENDF/B-IV fission-product decay energy data to generate multigroup beta and gamma spectra from individual ENDF/B-IV fission-product nuclides. FPSPEC further uses these spectra and the corresponding nuclide activities calculated by the CINDER-10 code to produce cumulative beta and gamma spectra in the same energy grids in which FPDCYS generates individual isotope decay spectra. The code system consisting of CINDER-10, FPDCYS, and FPSPEC has been used for comparisons with experimental spectra and continues to be used at LASL for generating spectra in special user-oriented group structures. 3 figures

  14. MIRANDA - a module based on multiregion resonance theory for generating cross sections within the AUS neutronics code system

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1985-12-01

    MIRANDA is the cross-section generation module of the AUS neutronics code system used to prepare multigroup cross-section data which are pertinent to a particular study from a general purpose multigroup library of cross sections. Libraries have been prepared from ENDF/B which are suitable for thermal and fast fission reactors and for fusion blanket studies. The libraries include temperature dependent data, resonance cross sections represented by subgroup parameters and may contain photon as well as neutron data. The MIRANDA module includes a multiregion resonance calculation in slab, cylinder or cluster geometry, a homogeneous B L flux solution, and a group condensation facility. This report documents the modifications to an earlier version of MIRANDA and provides a complete user's manual

  15. Influences of W Content on the Phase Transformation Properties and the Associated Stress Change in Thin Film Substrate Combinations Studied by Fabrication and Characterization of Thin Film V1- xW xO2 Materials Libraries.

    Science.gov (United States)

    Wang, Xiao; Rogalla, Detlef; Ludwig, Alfred

    2018-04-09

    The mechanical stress change of VO 2 film substrate combinations during their reversible phase transformation makes them promising for applications in micro/nanoactuators. V 1- x W x O 2 thin film libraries were fabricated by reactive combinatorial cosputtering to investigate the effects of the addition of W on mechanical and other transformation properties. High-throughput characterization methods were used to systematically determine the composition spread, crystalline structure, surface topography, as well as the temperature-dependent phase transformation properties, that is, the hysteresis curves of the resistance and stress change. The study indicates that as x in V 1- x W x O 2 increases from 0.007 to 0.044 the crystalline structure gradually shifts from the VO 2 (M) phase to the VO 2 (R) phase. The transformation temperature decreases by 15 K/at. % and the resistance change is reduced to 1 order of magnitude, accompanied by a wider transition range and a narrower hysteresis with a minimal value of 1.8 K. A V 1- x W x O 2 library deposited on a Si 3 N 4 /SiO 2 -coated Si cantilever array wafer was used to study simultaneously the temperature-dependent stress change σ( T) of films with different W content through the phase transformation. Compared with σ( T) of ∼700 MPa of a VO 2 film, σ( T) in V 1- x W x O 2 films decreases to ∼250 MPa. Meanwhile, σ( T) becomes less abrupt and occurs over a wider temperature range with decreased transformation temperatures.

  16. Improvement of the efficiency of two-dimensional multigroup transport calculations assuming isotropic reflection with multilevel spatial discretisation

    International Nuclear Information System (INIS)

    Stankovski, Z.; Zmijarevic, I.

    1987-06-01

    This paper presents two approximations used in multigroup two-dimensional transport calculations in large, very homogeneous media: isotropic reflection together with recently proposed group-dependent spatial representations. These approximations are implemented as standard options in APOLLO 2 assembly transport code. Presented example calculations show that significant savings in computational costs are obtained while preserving the overall accuracy

  17. Production and Testing of the VITAMIN-B7 Fine-Group and BUGLE-B7 Broad-Group Coupled Neutron/Gamma Cross-Section Libraries Derived from ENDF/B-VII.0 Nuclear Data

    Energy Technology Data Exchange (ETDEWEB)

    Risner, J. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wiarda, D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Dunn, M. E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, T. M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Peplow, D. E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Patton, B. W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2011-09-30

    New coupled neutron-gamma cross-section libraries have been developed for use in light water reactor (LWR) shielding applications, including pressure vessel dosimetry calculations. The libraries, which were generated using Evaluated Nuclear Data File/B Version VII Release 0 (ENDF/B-VII.0), use the same fine-group and broad-group energy structures as the VITAMIN-B6 and BUGLE-96 libraries. The processing methodology used to generate both libraries is based on the methods used to develop VITAMIN-B6 and BUGLE-96 and is consistent with ANSI/ANS 6.1.2. The ENDF data were first processed into the fine-group pseudo-problem-independent VITAMIN-B7 library and then collapsed into the broad-group BUGLE-B7 library. The VITAMIN-B7 library contains data for 391 nuclides. This represents a significant increase compared to the VITAMIN-B6 library, which contained data for 120 nuclides. The BUGLE-B7 library contains data for the same nuclides as BUGLE-96, and maintains the same numeric IDs for those nuclides. The broad-group data includes nuclides which are infinitely dilute and group collapsed using a concrete weighting spectrum, as well as nuclides which are self-shielded and group collapsed using weighting spectra representative of important regions of LWRs. The verification and validation of the new libraries includes a set of critical benchmark experiments, a set of regression tests that are used to evaluate multigroup crosssection libraries in the SCALE code system, and three pressure vessel dosimetry benchmarks. Results of these tests confirm that the new libraries are appropriate for use in LWR shielding analyses and meet the requirements of Regulatory Guide 1.190.

  18. The Nodal Polynomial Expansion method to solve the multigroup diffusion equations

    International Nuclear Information System (INIS)

    Ribeiro, R.D.M.

    1983-03-01

    The methodology of the solutions of the multigroup diffusion equations and uses the Nodal Polynomial Expansion Method is covered. The EPON code was developed based upon the above mentioned method for stationary state, rectangular geometry, one-dimensional or two-dimensional and for one or two energy groups. Then, one can study some effects such as the influence of the baffle on the thermal flux by calculating the flux and power distribution in nuclear reactors. Furthermore, a comparative study with other programs which use Finite Difference (CITATION and PDQ5) and Finite Element (CHD and FEMB) Methods was undertaken. As a result, the coherence, feasibility, speed and accuracy of the methodology used were demonstrated. (Author) [pt

  19. The solution of the multigroup neutron transport equation using spherical harmonics

    International Nuclear Information System (INIS)

    Fletcher, K.

    1981-01-01

    A solution of the multi-group neutron transport equation in up to three space dimensions is presented. The flux is expanded in a series of unnormalised spherical harmonics. Using the various recurrence formulae a linked set of first order differential equations is obtained for the moments psisup(g)sub(lm)(r), γsup(g)sub(lm)(r). Terms with odd l are eliminated resulting in a second order system which is solved by two methods. The first is a finite difference formulation using an iterative procedure, secondly, in XYZ and XY geometry a finite element solution is given. Results for a test problem using both methods are exhibited and compared. (orig./RW) [de

  20. A massively parallel algorithm for the collision probability calculations in the Apollo-II code using the PVM library

    International Nuclear Information System (INIS)

    Stankovski, Z.

    1995-01-01

    The collision probability method in neutron transport, as applied to 2D geometries, consume a great amount of computer time, for a typical 2D assembly calculation evaluations. Consequently RZ or 3D calculations became prohibitive. In this paper we present a simple but efficient parallel algorithm based on the message passing host/node programing model. Parallelization was applied to the energy group treatment. Such approach permits parallelization of the existing code, requiring only limited modifications. Sequential/parallel computer portability is preserved, witch is a necessary condition for a industrial code. Sequential performances are also preserved. The algorithm is implemented on a CRAY 90 coupled to a 128 processor T3D computer, a 16 processor IBM SP1 and a network of workstations, using the Public Domain PVM library. The tests were executed for a 2D geometry with the standard 99-group library. All results were very satisfactory, the best ones with IBM SP1. Because of heterogeneity of the workstation network, we did ask high performances for this architecture. The same source code was used for all computers. A more impressive advantage of this algorithm will appear in the calculations of the SAPHYR project (with the future fine multigroup library of about 8000 groups) with a massively parallel computer, using several hundreds of processors. (author). 5 refs., 6 figs., 2 tabs

  1. A massively parallel algorithm for the collision probability calculations in the Apollo-II code using the PVM library

    International Nuclear Information System (INIS)

    Stankovski, Z.

    1995-01-01

    The collision probability method in neutron transport, as applied to 2D geometries, consume a great amount of computer time, for a typical 2D assembly calculation about 90% of the computing time is consumed in the collision probability evaluations. Consequently RZ or 3D calculations became prohibitive. In this paper the author presents a simple but efficient parallel algorithm based on the message passing host/node programmation model. Parallelization was applied to the energy group treatment. Such approach permits parallelization of the existing code, requiring only limited modifications. Sequential/parallel computer portability is preserved, which is a necessary condition for a industrial code. Sequential performances are also preserved. The algorithm is implemented on a CRAY 90 coupled to a 128 processor T3D computer, a 16 processor IBM SPI and a network of workstations, using the Public Domain PVM library. The tests were executed for a 2D geometry with the standard 99-group library. All results were very satisfactory, the best ones with IBM SPI. Because of heterogeneity of the workstation network, the author did not ask high performances for this architecture. The same source code was used for all computers. A more impressive advantage of this algorithm will appear in the calculations of the SAPHYR project (with the future fine multigroup library of about 8000 groups) with a massively parallel computer, using several hundreds of processors

  2. Approximate analytical solution of two-dimensional multigroup P-3 equations

    International Nuclear Information System (INIS)

    Matausek, M.V.; Milosevic, M.

    1981-01-01

    Iterative solution of multigroup spherical harmonics equations reduces, in the P-3 approximation and in two-dimensional geometry, to a problem of solving an inhomogeneous system of eight ordinary first order differential equations. With appropriate boundary conditions, these equations have to be solved for each energy group and in each iteration step. The general solution of the corresponding homogeneous system of equations is known in analytical form. The present paper shows how the right-hand side of the system can be approximated in order to derive a particular solution and thus an approximate analytical expression for the general solution of the inhomogeneous system. This combined analytical-numerical approach was shown to have certain advantages compared to the finite-difference method or the Lie-series expansion method, which have been used to solve similar problems. (author)

  3. Resolution of the multigroup scattering equation in a one-dimensional geometry and subsidiary calculations: the MUDE code; Resolution de l'equation multigroupe de la diffusion dans une geometrie a une dimension et calculs annexes: code MUDE

    Energy Technology Data Exchange (ETDEWEB)

    Bore, C; Dandeu, Y; Saint-Amand, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-07-01

    MUDE is a nuclear code written in FORTRAN II for IBM 7090-7094. It resolves a system of difference equations approximating to the one-dimensional multigroup neutron scattering problem. More precisely, this code makes it possible to: 1. Calculate the critical condition of a reactor (k{sub eff}, critical radius, critical composition) and the corresponding fluxes; 2. Calculate the associated fluxes and various subsidiary results; 3. Carry out perturbation calculations; 4. Study the propagation of fluxes at a distance; 5. Estimate the relative contributions of the cross sections (macroscopic or microscopic); 6. Study the changes with time of the composition of the reactor. (authors) [French] MUDE est un code nucleaire ecrit en FORTRAN II pour IBM 7090-7094. Il resout un systeme d'equations aux differences approchant le probleme de diffusion neutronique multigroupe a une dimension. Plus precisement ce code permet de: 1. Calculer la condition critique d'un reacteur (k{sub eff}, rayon critique, composition critique) et les flux correspondants; 2. Calculer les flux adjoints et divers resultats connexes; 3. Effectuer des calculs de perturbation; 4. Etudier la propagation des flux a longue distance; 5. Ponderer des sections efficaces (macroscopiques ou microscopiques); 6. Etudier l'evolution de la composition du reacteur au cours du temps. (auteurs)

  4. 2009 A Survey of Libraries in Gombe State By Maimuna Izah

    African Journals Online (AJOL)

    parastatals, organizations, schools companies, etc in Gombe State to determine which ones have libraries. The ... Libraries play pivotal role in the .... National Orientation Agency X. 11 .... Schools Enrollment by Gender and percentage.

  5. System THEMIS. Cross sections processing system from ENDF/B

    Energy Technology Data Exchange (ETDEWEB)

    Gonnord, J.

    1983-09-01

    The THEMIS system allowed to prepare a self punctual and multigroup library for codes solving the TRIPOLI-PROMETHEE transport equation, allowing comparisons with different methods and approximations. The contents of the THEMIS data base was fixed from its use by the PROMETHEE system (punctual Monte Carlo calculations, multigroup calculations, uncertainties analysis and sensitivity studies). The main characteristics of the THEMIS cross section processing system are briefly presented.

  6. Global dynamics of a novel multi-group model for computer worms

    International Nuclear Information System (INIS)

    Gong Yong-Wang; Song Yu-Rong; Jiang Guo-Ping

    2013-01-01

    In this paper, we study worm dynamics in computer networks composed of many autonomous systems. A novel multi-group SIQR (susceptible-infected-quarantined-removed) model is proposed for computer worms by explicitly considering anti-virus measures and the network infrastructure. Then, the basic reproduction number of worm R 0 is derived and the global dynamics of the model are established. It is shown that if R 0 is less than or equal to 1, the disease-free equilibrium is globally asymptotically stable and the worm dies out eventually, whereas, if R 0 is greater than 1, one unique endemic equilibrium exists and it is globally asymptotically stable, thus the worm persists in the network. Finally, numerical simulations are given to illustrate the theoretical results. (general)

  7. The Power to Detect Sex Differences in IQ Test Scores Using Multi-Group Covariance and Means Structure Analyses

    Science.gov (United States)

    Molenaar, Dylan; Dolan, Conor V.; Wicherts, Jelle M.

    2009-01-01

    Research into sex differences in general intelligence, g, has resulted in two opposite views. In the first view, a g-difference is nonexistent, while in the second view, g is associated with a male advantage. Past research using Multi-Group Covariance and Mean Structure Analysis (MG-CMSA) found no sex difference in g. This failure raised the…

  8. Approximate analytical solution of two-dimensional multigroup P-3 equations

    International Nuclear Information System (INIS)

    Matausek, M.V.; Milosevic, M.

    1981-01-01

    Iterative solution of multigroup spherical harmonics equations reduces, in the P-3 approximation and in two-dimensional geometry, to a problem of solving an inhomogeneous system of eight ordinary first order differential equations. With appropriate boundary conditions, these equations have to be solved for each energy group and in each iteration step. The general solution of the corresponding homogeneous system of equations is known in analytical form. The present paper shows how the right-hand side of the system can be approximated in order to derive a particular solution and thus an approximate analytical expression for the general solution of the inhomogeneous system. This combined analytical-numerical approach was shown to have certain advantages compared to the finite-difference method or the Lie-series expansion method, which have been used to solve similar problems. (orig./RW) [de

  9. SIRIUS - A one-dimensional multigroup analytic nodal diffusion theory code

    Energy Technology Data Exchange (ETDEWEB)

    Forslund, P. [Westinghouse Atom AB, Vaesteraas (Sweden)

    2000-09-01

    In order to evaluate relative merits of some proposed intranodal cross sections models, a computer code called Sirius has been developed. Sirius is a one-dimensional, multigroup analytic nodal diffusion theory code with microscopic depletion capability. Sirius provides the possibility of performing a spatial homogenization and energy collapsing of cross sections. In addition a so called pin power reconstruction method is available for the purpose of reconstructing 'heterogeneous' pin qualities. consequently, Sirius has the capability of performing all the calculations (incl. depletion calculations) which are an integral part of the nodal calculation procedure. In this way, an unambiguous numerical analysis of intranodal cross section models is made possible. In this report, the theory of the nodal models implemented in sirius as well as the verification of the most important features of these models are addressed.

  10. Gray and multigroup radiation transport models for two-dimensional binary stochastic media using effective opacities

    International Nuclear Information System (INIS)

    Olson, Gordon L.

    2016-01-01

    One-dimensional models for the transport of radiation through binary stochastic media do not work in multi-dimensions. Authors have attempted to modify or extend the 1D models to work in multidimensions without success. Analytic one-dimensional models are successful in 1D only when assuming greatly simplified physics. State of the art theories for stochastic media radiation transport do not address multi-dimensions and temperature-dependent physics coefficients. Here, the concept of effective opacities and effective heat capacities is found to well represent the ensemble averaged transport solutions in cases with gray or multigroup temperature-dependent opacities and constant or temperature-dependent heat capacities. In every case analyzed here, effective physics coefficients fit the transport solutions over a useful range of parameter space. The transport equation is solved with the spherical harmonics method with angle orders of n=1 and 5. Although the details depend on what order of solution is used, the general results are similar, independent of angular order. - Highlights: • Gray and multigroup radiation transport is done through 2D stochastic media. • Approximate models for the mean radiation field are found for all test problems. • Effective opacities are adjusted to fit the means of stochastic media transport. • Test problems include temperature dependent opacities and heat capacities • Transport solutions are done with angle orders n=1 and 5.

  11. Development of a polynomial nodal model to the multigroup transport equation in one dimension

    International Nuclear Information System (INIS)

    Feiz, M.

    1986-01-01

    A polynomial nodal model that uses Legendre polynomial expansions was developed for the multigroup transport equation in one dimension. The development depends upon the least-squares minimization of the residuals using the approximate functions over the node. Analytical expressions were developed for the polynomial coefficients. The odd moments of the angular neutron flux over the half ranges were used at the internal interfaces, and the Marshak boundary condition was used at the external boundaries. Sample problems with fine-mesh finite-difference solutions of the diffusion and transport equations were used for comparison with the model

  12. Crispv programme

    International Nuclear Information System (INIS)

    Marinkovicj, N.

    CRISPV (Criticality and Spectrum code) is a multigroup neutron spectrum code for homogeneous reactor cores and is actually a somewhat modified version of the original CRISP programme. It is a combination of DATAPREP-II and BIGG-II programmes. It is assumed that the reactor cell is a cylindrical fuel rod in the light or heavy water moderator. DATEPREP-II CODE forms the multigroup data for homogeneous reactor and prepares the input parameters for the BIGG-II code. It has its own nuclear data library on a separate tape in binary mode. BIGG-II code is a multigroup neutron spectrum and criticality code for a homogenized medium. It has as well its own separate data library. In the CRISPV programme the overlay structure enables automatic handling of data calculated in the DATAPREP-II programme and needed in the BIGG-II core. Both programmes are written in FORTRAN for CDC 3600. Using the programme is very efficient and simple

  13. Solution of the multigroup diffusion equation for two-dimensional triangular regions by finite Fourier transformation

    International Nuclear Information System (INIS)

    Takeshi, Y.; Keisuke, K.

    1983-01-01

    The multigroup neutron diffusion equation for two-dimensional triangular geometry is solved by the finite Fourier transformation method. Using the zero-th-order equation of the integral equation derived by this method, simple algebraic expressions for the flux are derived and solved by the alternating direction implicit method. In sample calculations for a benchmark problem of a fast breeder reactor, it is shown that the present method gives good results with fewer mesh points than the usual finite difference method

  14. A Multigroup diffusion solver using pseudo transient continuation for a radiation-hydrodynamic code with patch-based AMR

    Energy Technology Data Exchange (ETDEWEB)

    Shestakov, A I; Offner, S R

    2006-09-21

    We present a scheme to solve the nonlinear multigroup radiation diffusion (MGD) equations. The method is incorporated into a massively parallel, multidimensional, Eulerian radiation-hydrodynamic code with adaptive mesh refinement (AMR). The patch-based AMR algorithm refines in both space and time creating a hierarchy of levels, coarsest to finest. The physics modules are time-advanced using operator splitting. On each level, separate 'level-solve' packages advance the modules. Our multigroup level-solve adapts an implicit procedure which leads to a two-step iterative scheme that alternates between elliptic solves for each group with intra-cell group coupling. For robustness, we introduce pseudo transient continuation ({Psi}tc). We analyze the magnitude of the {Psi}tc parameter to ensure positivity of the resulting linear system, diagonal dominance and convergence of the two-step scheme. For AMR, a level defines a subdomain for refinement. For diffusive processes such as MGD, the refined level uses Dirichet boundary data at the coarse-fine interface and the data is derived from the coarse level solution. After advancing on the fine level, an additional procedure, the sync-solve (SS), is required in order to enforce conservation. The MGD SS reduces to an elliptic solve on a combined grid for a system of G equations, where G is the number of groups. We adapt the 'partial temperature' scheme for the SS; hence, we reuse the infrastructure developed for scalar equations. Results are presented. We consider a multigroup test problem with a known analytic solution. We demonstrate utility of {Psi}tc by running with increasingly larger timesteps. Lastly, we simulate the sudden release of energy Y inside an Al sphere (r = 15 cm) suspended in air at STP. For Y = 11 kT, we find that gray radiation diffusion and MGD produce similar results. However, if Y = 1 MT, the two packages yield different results. Our large Y simulation contradicts a long-standing theory

  15. A Multigroup diffusion Solver Using Pseudo Transient Continuation for a Radiaiton-Hydrodynamic Code with Patch-Based AMR

    Energy Technology Data Exchange (ETDEWEB)

    Shestakov, A I; Offner, S R

    2007-03-02

    We present a scheme to solve the nonlinear multigroup radiation diffusion (MGD) equations. The method is incorporated into a massively parallel, multidimensional, Eulerian radiation-hydrodynamic code with adaptive mesh refinement (AMR). The patch-based AMR algorithm refines in both space and time creating a hierarchy of levels, coarsest to finest. The physics modules are time-advanced using operator splitting. On each level, separate 'level-solve' packages advance the modules. Our multigroup level-solve adapts an implicit procedure which leads to a two-step iterative scheme that alternates between elliptic solves for each group with intra-cell group coupling. For robustness, we introduce pseudo transient continuation ({Psi}tc). We analyze the magnitude of the {Psi}tc parameter to ensure positivity of the resulting linear system, diagonal dominance and convergence of the two-step scheme. For AMR, a level defines a subdomain for refinement. For diffusive processes such as MGD, the refined level uses Dirichet boundary data at the coarse-fine interface and the data is derived from the coarse level solution. After advancing on the fine level, an additional procedure, the sync-solve (SS), is required in order to enforce conservation. The MGD SS reduces to an elliptic solve on a combined grid for a system of G equations, where G is the number of groups. We adapt the 'partial temperature' scheme for the SS; hence, we reuse the infrastructure developed for scalar equations. Results are presented. We consider a multigroup test problem with a known analytic solution. We demonstrate utility of {Psi}tc by running with increasingly larger timesteps. Lastly, we simulate the sudden release of energy Y inside an Al sphere (r = 15 cm) suspended in air at STP. For Y = 11 kT, we find that gray radiation diffusion and MGD produce similar results. However, if Y = 1 MT, the two packages yield different results. Our large Y simulation contradicts a long-standing theory

  16. MVP/GMVP 2: general purpose Monte Carlo codes for neutron and photon transport calculations based on continuous energy and multigroup methods

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu; Okumura, Keisuke; Mori, Takamasa; Nakagawa, Masayuki

    2005-06-01

    In order to realize fast and accurate Monte Carlo simulation of neutron and photon transport problems, two vectorized Monte Carlo codes MVP and GMVP have been developed at JAERI. MVP is based on the continuous energy model and GMVP is on the multigroup model. Compared with conventional scalar codes, these codes achieve higher computation speed by a factor of 10 or more on vector super-computers. Both codes have sufficient functions for production use by adopting accurate physics model, geometry description capability and variance reduction techniques. The first version of the codes was released in 1994. They have been extensively improved and new functions have been implemented. The major improvements and new functions are (1) capability to treat the scattering model expressed with File 6 of the ENDF-6 format, (2) time-dependent tallies, (3) reaction rate calculation with the pointwise response function, (4) flexible source specification, (5) continuous-energy calculation at arbitrary temperatures, (6) estimation of real variances in eigenvalue problems, (7) point detector and surface crossing estimators, (8) statistical geometry model, (9) function of reactor noise analysis (simulation of the Feynman-α experiment), (10) arbitrary shaped lattice boundary, (11) periodic boundary condition, (12) parallelization with standard libraries (MPI, PVM), (13) supporting many platforms, etc. This report describes the physical model, geometry description method used in the codes, new functions and how to use them. (author)

  17. LABAN-PEL: a two-dimensional, multigroup diffusion, high-order response matrix code

    International Nuclear Information System (INIS)

    Mueller, E.Z.

    1991-06-01

    The capabilities of LABAN-PEL is described. LABAN-PEL is a modified version of the two-dimensional, high-order response matrix code, LABAN, written by Lindahl. The new version extends the capabilities of the original code with regard to the treatment of neutron migration by including an option to utilize full group-to-group diffusion coefficient matrices. In addition, the code has been converted from single to double precision and the necessary routines added to activate its multigroup capability. The coding has also been converted to standard FORTRAN-77 to enhance the portability of the code. Details regarding the input data requirements and calculational options of LABAN-PEL are provided. 13 refs

  18. New data libraries and physics data management tools

    International Nuclear Information System (INIS)

    Han, M; Kim, C H; Seo, H; Pia, M G; Saracco, P; Augelli, M; Hauf, S; Kuster, M; Moneta, L; Quintieri, L

    2011-01-01

    A number of physics data libraries for Monte Carlo simulation are reviewed. The development of a package for the management of physics data is described: its design, implementation and computational benchmarks. This package improves the data management tools originally developed for Geant4 electromagnetic physics models based on data libraries. The implementation exploits recent evolutions of the C++ libraries appearing in the C++0x draft, which are intended for inclusion in the next C++ ISO Standard. The new tools improve the computational performance of physics data management.

  19. An Experiment of Robust Parallel Algorithm for the Eigenvalue problem of a Multigroup Neutron Diffusion based on modified FETI-DP : Part 2

    International Nuclear Information System (INIS)

    Chang, Jonghwa

    2014-01-01

    Today, we can use a computer cluster consist of a few hundreds CPUs with reasonable budget. Such computer system enables us to do detailed modeling of reactor core. The detailed modeling will improve the safety and the economics of a nuclear reactor by eliminating un-necessary conservatism or missing consideration. To take advantage of such a cluster computer, efficient parallel algorithms must be developed. Mechanical structure analysis community has studied the domain decomposition method to solve the stress-strain equation using the finite element methods. One of the most successful domain decomposition method in terms of robustness is FETI-DP. We have modified the original FETI-DP to solve the eigenvalue problem for the multi-group diffusion problem in previous study. In this study, we report the result of recent modification to handle the three-dimensional subdomain partitioning, and the sub-domain multi-group problem. Modified FETI-DP algorithm has been successfully applied for the eigenvalue problem of multi-group neutron diffusion equation. The overall CPU time is decreasing as number of sub-domains (partitions) is increasing. However, there may be a limit in decrement due to increment of the number of primal points will increase the CPU time spent by the solution of the global equation. Even distribution of computational load (criterion a) is important to achieve fast computation. The subdomain partition can be effectively performed using suitable graph theory partition package such as MeTIS

  20. An Experiment of Robust Parallel Algorithm for the Eigenvalue problem of a Multigroup Neutron Diffusion based on modified FETI-DP : Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jonghwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Today, we can use a computer cluster consist of a few hundreds CPUs with reasonable budget. Such computer system enables us to do detailed modeling of reactor core. The detailed modeling will improve the safety and the economics of a nuclear reactor by eliminating un-necessary conservatism or missing consideration. To take advantage of such a cluster computer, efficient parallel algorithms must be developed. Mechanical structure analysis community has studied the domain decomposition method to solve the stress-strain equation using the finite element methods. One of the most successful domain decomposition method in terms of robustness is FETI-DP. We have modified the original FETI-DP to solve the eigenvalue problem for the multi-group diffusion problem in previous study. In this study, we report the result of recent modification to handle the three-dimensional subdomain partitioning, and the sub-domain multi-group problem. Modified FETI-DP algorithm has been successfully applied for the eigenvalue problem of multi-group neutron diffusion equation. The overall CPU time is decreasing as number of sub-domains (partitions) is increasing. However, there may be a limit in decrement due to increment of the number of primal points will increase the CPU time spent by the solution of the global equation. Even distribution of computational load (criterion a) is important to achieve fast computation. The subdomain partition can be effectively performed using suitable graph theory partition package such as MeTIS.

  1. Block copolymer libraries: modular versatility of the macromolecular Lego system.

    Science.gov (United States)

    Lohmeijer, Bas G G; Wouters, Daan; Yin, Zhihui; Schubert, Ulrich S

    2004-12-21

    The synthesis and characterization of a new 4 x 4 library of block copolymers based on polystyrene and poly(ethylene oxide) connected by an asymmetrical octahedral bis(terpyridine) ruthenium complex at the block junction are described, while initial studies on the thin film morphology of the components of the library are presented by the use of Atomic Force Microscopy, demonstrating the impact of a library approach to derive structure-property relationships.

  2. System THEMIS. Cross sections processing system from ENDF/B

    International Nuclear Information System (INIS)

    Gonnord, J.

    1983-09-01

    The THEMIS system allowed to prepare a self punctual and multigroup library for codes solving the TRIPOLI-PROMETHEE transport equation, allowing comparisons with different methods and approximations. The contents of the THEMIS data base was fixed from its use by the PROMETHEE system (punctual Monte Carlo calculations, multigroup calculations, uncertainties analysis and sensitivity studies). The main characteristics of the THEMIS cross section processing system are briefly presented [fr

  3. A multi-region boundary element method for multigroup neutron diffusion calculations

    International Nuclear Information System (INIS)

    Ozgener, H.A.; Ozgener, B.

    2001-01-01

    For the analysis of a two-dimensional nuclear system consisting of a number of homogeneous regions (termed cells), first the cell matrices which depend solely on the material composition and geometrical dimension of the cell (hence on the cell type) are constructed using a boundary element formulation based on the multigroup boundary integral equation. For a particular nuclear system, the cell matrices are utilized in the assembly of the global system matrix in block-banded form using the newly introduced concept of virtual side. For criticality calculations, the classical fission source iteration is employed and linear system solutions are by the block Gaussian-elimination algorithm. The numerical applications show the validity of the proposed formulation both through comparison with analytical solutions and assessment of benchmark problem results against alternative methods

  4. AMPX: a modular system for multigroup cross-section generation and manipulation

    International Nuclear Information System (INIS)

    Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Diggs, B.R.; Webster, C.C.; Lucius, J.L.; White, J.E.; Wright, R.Q.; Westfall, R.M.

    1978-01-01

    The AMPX system, developed at the Oak Ridge National Laboratory over the past seven years, is a collection of computer programs in a modular arrangement. Starting with ENDF-formatted nuclear data files, the system includes a full range of features needed to produce and use multigroup neutron, gamma-ray production, and gamma-ray interaction cross-section data. The balance between production and analysis is roughly even; thus, the system serves a wide variety of needs. The modularity is particularly attractive, since it allows the user to choose an arbitrary execution sequence from the approximately 40 to 50 modules available in the system. The modularity also allows selection from different treatments; e.g., the Nordheim method, a full-blown integral transport calculation, the Bondarenko method, or other alternative can be selected for resonance shielding. 2 figures

  5. Status of MAPA (Modular Accelerator Physics Analysis) and the Tech-X Object-Oriented Accelerator Library

    Science.gov (United States)

    Cary, J. R.; Shasharina, S.; Bruhwiler, D. L.

    1998-04-01

    The MAPA code is a fully interactive accelerator modeling and design tool consisting of a GUI and two object-oriented C++ libraries: a general library suitable for treatment of any dynamical system, and an accelerator library including many element types plus an accelerator class. The accelerator library inherits directly from the system library, which uses hash tables to store any relevant parameters or strings. The GUI can access these hash tables in a general way, allowing the user to invoke a window displaying all relevant parameters for a particular element type or for the accelerator class, with the option to change those parameters. The system library can advance an arbitrary number of dynamical variables through an arbitrary mapping. The accelerator class inherits this capability and overloads the relevant functions to advance the phase space variables of a charged particle through a string of elements. Among other things, the GUI makes phase space plots and finds fixed points of the map. We discuss the object hierarchy of the two libraries and use of the code.

  6. The SpeX Prism Library for Ultracool Dwarfs: A Resource for Stellar, Exoplanet and Galactic Science and Student-Led Research

    Science.gov (United States)

    Burgasser, Adam

    The NASA Infrared Telescope Facility's (IRTF) SpeX spectrograph has been an essential tool in the discovery and characterization of ultracool dwarf (UCD) stars, brown dwarfs and exoplanets. Over ten years of SpeX data have been collected on these sources, and a repository of low-resolution (R 100) SpeX prism spectra has been maintained by the PI at the SpeX Prism Spectral Libraries website since 2008. As the largest existing collection of NIR UCD spectra, this repository has facilitated a broad range of investigations in UCD, exoplanet, Galactic and extragalactic science, contributing to over 100 publications in the past 6 years. However, this repository remains highly incomplete, has not been uniformly calibrated, lacks sufficient contextual data for observations and sources, and most importantly provides no data visualization or analysis tools for the user. To fully realize the scientific potential of these data for community research, we propose a two-year program to (1) calibrate and expand existing repository and archival data, and make it virtual-observatory compliant; (2) serve the data through a searchable web archive with basic visualization tools; and (3) develop and distribute an open-source, Python-based analysis toolkit for users to analyze the data. These resources will be generated through an innovative, student-centered research model, with undergraduate and graduate students building and validating the analysis tools through carefully designed coding challenges and research validation activities. The resulting data archive, the SpeX Prism Library, will be a legacy resource for IRTF and SpeX, and will facilitate numerous investigations using current and future NASA capabilities. These include deep/wide surveys of UCDs to measure Galactic structure and chemical evolution, and probe UCD populations in satellite galaxies (e.g., JWST, WFIRST); characterization of directly imaged exoplanet spectra (e.g., FINESSE), and development of low

  7. CINESP - computational program of spatial kinetics for nuclear reactors in the one-two dimension multigroup diffusion theory

    International Nuclear Information System (INIS)

    Santos, R.S. dos

    1993-01-01

    This paper presents a computational program to solve numerically the reactor kinetics equations in the multigroup diffusion theory. One or two-dimensional problems in cylindrical or Cartesian geometries, with any number of energy and delayed-neutron precursors groups are dealt with. The main input and output of the program are briefly discussed. Various results demonstrate the accuracy and versatility of the program, when compared with other kinetics programs. (author)

  8. Using the probability method for multigroup calculations of reactor cells in a thermal energy range

    International Nuclear Information System (INIS)

    Rubin, I.E.; Pustoshilova, V.S.

    1984-01-01

    The possibility of using the transmission probability method with performance inerpolation for determining spatial-energy neutron flux distribution in cells of thermal heterogeneous reactors is considered. The results of multigroup calculations of several uranium-water plane and cylindrical cells with different fuel enrichment in a thermal energy range are given. A high accuracy of results is obtained with low computer time consumption. The use of the transmission probability method is particularly reasonable in algorithms of the programmes compiled computer with significant reserve of internal memory

  9. A Laplace transform method for energy multigroup hybrid discrete ordinates

    International Nuclear Information System (INIS)

    Segatto, C.F.; Vilhena, M.T.; Barros, R.C.

    2010-01-01

    In typical lattice cells where a highly absorbing, small fuel element is embedded in the moderator, a large weakly absorbing medium, high-order transport methods become unnecessary. In this work we describe a hybrid discrete ordinates (S N) method for energy multigroup slab lattice calculations. This hybrid S N method combines the convenience of a low-order S N method in the moderator with a high-order S N method in the fuel. The idea is based on the fact that in weakly absorbing media whose physical size is several neutron mean free paths in extent, even the S 2 method (P 1 approximation), leads to an accurate result. We use special fuel-moderator interface conditions and the Laplace transform (LTS N ) analytical numerical method to calculate the two-energy group neutron flux distributions and the thermal disadvantage factor. We present numerical results for a range of typical model problems.

  10. Programs Lucky and LuckyC - 3D parallel transport codes for the multi-group transport equation solution for XYZ geometry by Pm Sn method

    International Nuclear Information System (INIS)

    Moriakov, A.; Vasyukhno, V.; Netecha, M.; Khacheresov, G.

    2003-01-01

    Powerful supercomputers are available today. MBC-1000M is one of Russian supercomputers that may be used by distant way access. Programs LUCKY and LUCKY C were created to work for multi-processors systems. These programs have algorithms created especially for these computers and used MPI (message passing interface) service for exchanges between processors. LUCKY may resolved shielding tasks by multigroup discreet ordinate method. LUCKY C may resolve critical tasks by same method. Only XYZ orthogonal geometry is available. Under little space steps to approximate discreet operator this geometry may be used as universal one to describe complex geometrical structures. Cross section libraries are used up to P8 approximation by Legendre polynomials for nuclear data in GIT format. Programming language is Fortran-90. 'Vector' processors may be used that lets get a time profit up to 30 times. But unfortunately MBC-1000M has not these processors. Nevertheless sufficient value for efficiency of parallel calculations was obtained under 'space' (LUCKY) and 'space and energy' (LUCKY C ) paralleling. AUTOCAD program is used to control geometry after a treatment of input data. Programs have powerful geometry module, it is a beautiful tool to achieve any geometry. Output results may be processed by graphic programs on personal computer. (authors)

  11. Development of a reference spent fuel library of 17x17 PWR fuel assemblies

    International Nuclear Information System (INIS)

    Rossa, Riccardo; Borella, Alessandro; Van der Meer, Klaas

    2013-01-01

    One of the most common ways to investigate new Non-Destructive Assays (NDA) for the spent fuel assemblies are Monte Carlo simulations. In order to build realistic models the user must define in an accurate way the material compositions and the source terms in the system. This information can be obtained using burnup codes such as ORIGEN-ARP and ALEPH2.2, developed at SCK-CEN. These software applications allow the user to select the irradiation history of the fuel assembly and to calculate the corresponding isotopic composition and neutron/gamma emissions as a function of time. In the framework of the development of an innovative NDA for spent fuel verifications, SCK•CEN built an extensive fuel library for 17x17 PWR assemblies, using both ORIGEN-ARP and ALEPH2.2. The parameters considered in the calculations were initial enrichment, discharge burnup, and cooling time. The combination of these variables allows to obtain more than 1500 test cases. Considering the broad range of the parameters, the fuel library can be used for other purposes apart from spent fuel verifications, for instance for the direct disposal in geological repositories. In addition to the isotopic composition of the spent fuel, the neutron and photon emissions were also calculated and compared between the two codes. The comparison of the isotopic composition showed a good agreement between the codes for most of the relevant isotopes in the spent fuel. However, specific isotopes as well as neutron and gamma spectra still need to be investigated in detail.

  12. Longitudinal multigroup invariance analysis of the satisfaction with food-related life scale in university students.

    Science.gov (United States)

    Schnettler, Berta; Miranda, Horacio; Miranda-Zapata, Edgardo; Salinas-Oñate, Natalia; Grunert, Klaus G; Lobos, Germán; Sepúlveda, José; Orellana, Ligia; Hueche, Clementina; Bonilla, Héctor

    2017-06-01

    This study examined longitudinal measurement invariance in the Satisfaction with Food-related Life (SWFL) scale using follow-up data from university students. We examined this measure of the SWFL in different groups of students, separated by various characteristics. Through non-probabilistic longitudinal sampling, 114 university students (65.8% female, mean age: 22.5) completed the SWFL questionnaire three times, over intervals of approximately one year. Confirmatory factor analysis was used to examine longitudinal measurement invariance. Two types of analysis were conducted: first, a longitudinal invariance by time, and second, a multigroup longitudinal invariance by sex, age, socio-economic status and place of residence during the study period. Results showed that the 3-item version of the SWFL exhibited strong longitudinal invariance (equal factor loadings and equal indicator intercepts). Longitudinal multigroup invariance analysis also showed that the 3-item version of the SWFL displays strong invariance by socio-economic status and place of residence during the study period over time. Nevertheless, it was only possible to demonstrate equivalence of the longitudinal factor structure among students of both sexes, and among those older and younger than 22 years. Generally, these findings suggest that the SWFL scale has satisfactory psychometric properties for longitudinal measurement invariance in university students with similar characteristics as the students that participated in this research. It is also possible to suggest that satisfaction with food-related life is associated with sex and age. Copyright © 2017 Elsevier Ltd. All rights reserved.

  13. LibCPIXE: A PIXE simulation open-source library for multilayered samples

    International Nuclear Information System (INIS)

    Pascual-Izarra, C.; Barradas, N.P.; Reis, M.A.

    2006-01-01

    Most particle induced X-ray emission (PIXE) data analysis codes are not focused on handling multilayered samples. We have developed an open-source library called 'LibCPIXE', for PIXE data analysis. It is written in standard C and implements functions for simulating X-ray yields of PIXE spectra taken from arbitrary samples, including multilayered targets. The library is designed to be fast, portable, modular and scalable, as well as to facilitate its incorporation into any existing program. In order to demonstrate the capabilities of the library, a program called CPIXE was developed and used to analyze various real samples involving both bulk and layered samples. Just as the library, the CPIXE source code is freely available under the General Public License. We demonstrate that it runs both under GNU/Linux systems as well as under MS Windows. There is in principle no limitation to port it to other platforms

  14. Multiregion, multigroup collision probability method with white boundary condition for light water reactor thermalization calculations

    International Nuclear Information System (INIS)

    Ozgener, B.; Ozgener, H.A.

    2005-01-01

    A multiregion, multigroup collision probability method with white boundary condition is developed for thermalization calculations of light water moderated reactors. Hydrogen scatterings are treated by Nelkin's kernel while scatterings from other nuclei are assumed to obey the free-gas scattering kernel. The isotropic return (white) boundary condition is applied directly by using the appropriate collision probabilities. Comparisons with alternate numerical methods show the validity of the present formulation. Comparisons with some experimental results indicate that the present formulation is capable of calculating disadvantage factors which are closer to the experimental results than alternative methods

  15. Multigroup perturbation model for kinetic analysis of nuclear reactors

    International Nuclear Information System (INIS)

    Souza, G.M.

    1989-01-01

    The scope of this work is the development of a multigroup perturbation theory for the purpose of Kinetic and dynamic analysis of nuclear reactors. The equations that describe the reactor behavior were presented in all generality and written in the shorthand notation of matrices and vectors. In the derivation of those equations indetermined operators and discretizing factors were introduced and then determined by comparision with conventional equations. Fick's Law was developed in higher orders for neutron and importance current density. The solution of the direct and adjoint fields were represented by combination of the eigenfunctions of the B and B* operators and the eigenvalue modulus equality was established mathematically. In the derivation of the reactivity expression the B operator perturbation was split in two non coupled to the flux form and level. The prompt neutrons effective mean life was derived from reactor equations and importance conservation. The establishment of the Nordheim's equation, although modified, was based on Gandini. Finally, a mathematical interpretation of the flux-trap region was avented. (author)

  16. About the Library - Betty Petersen Memorial Library

    Science.gov (United States)

    branch library of the NOAA Central Library. The library serves the NOAA Science Center in Camp Springs , Maryland. History and Mission: Betty Petersen Memorial Library began as a reading room in the NOAA Science Science Center staff and advises the library on all aspects of the library program. Library Newsletters

  17. Library based x-ray scatter correction for dedicated cone beam breast CT

    International Nuclear Information System (INIS)

    Shi, Linxi; Zhu, Lei; Vedantham, Srinivasan; Karellas, Andrew

    2016-01-01

    Purpose: The image quality of dedicated cone beam breast CT (CBBCT) is limited by substantial scatter contamination, resulting in cupping artifacts and contrast-loss in reconstructed images. Such effects obscure the visibility of soft-tissue lesions and calcifications, which hinders breast cancer detection and diagnosis. In this work, we propose a library-based software approach to suppress scatter on CBBCT images with high efficiency, accuracy, and reliability. Methods: The authors precompute a scatter library on simplified breast models with different sizes using the GEANT4-based Monte Carlo (MC) toolkit. The breast is approximated as a semiellipsoid with homogeneous glandular/adipose tissue mixture. For scatter correction on real clinical data, the authors estimate the breast size from a first-pass breast CT reconstruction and then select the corresponding scatter distribution from the library. The selected scatter distribution from simplified breast models is spatially translated to match the projection data from the clinical scan and is subtracted from the measured projection for effective scatter correction. The method performance was evaluated using 15 sets of patient data, with a wide range of breast sizes representing about 95% of general population. Spatial nonuniformity (SNU) and contrast to signal deviation ratio (CDR) were used as metrics for evaluation. Results: Since the time-consuming MC simulation for library generation is precomputed, the authors’ method efficiently corrects for scatter with minimal processing time. Furthermore, the authors find that a scatter library on a simple breast model with only one input parameter, i.e., the breast diameter, sufficiently guarantees improvements in SNU and CDR. For the 15 clinical datasets, the authors’ method reduces the average SNU from 7.14% to 2.47% in coronal views and from 10.14% to 3.02% in sagittal views. On average, the CDR is improved by a factor of 1.49 in coronal views and 2.12 in sagittal

  18. Library based x-ray scatter correction for dedicated cone beam breast CT

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Linxi; Zhu, Lei, E-mail: leizhu@gatech.edu [Nuclear and Radiological Engineering and Medical Physics Programs, The George W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, Georgia 30332 (United States); Vedantham, Srinivasan; Karellas, Andrew [Department of Radiology, University of Massachusetts Medical School, Worcester, Massachusetts 01655 (United States)

    2016-08-15

    Purpose: The image quality of dedicated cone beam breast CT (CBBCT) is limited by substantial scatter contamination, resulting in cupping artifacts and contrast-loss in reconstructed images. Such effects obscure the visibility of soft-tissue lesions and calcifications, which hinders breast cancer detection and diagnosis. In this work, we propose a library-based software approach to suppress scatter on CBBCT images with high efficiency, accuracy, and reliability. Methods: The authors precompute a scatter library on simplified breast models with different sizes using the GEANT4-based Monte Carlo (MC) toolkit. The breast is approximated as a semiellipsoid with homogeneous glandular/adipose tissue mixture. For scatter correction on real clinical data, the authors estimate the breast size from a first-pass breast CT reconstruction and then select the corresponding scatter distribution from the library. The selected scatter distribution from simplified breast models is spatially translated to match the projection data from the clinical scan and is subtracted from the measured projection for effective scatter correction. The method performance was evaluated using 15 sets of patient data, with a wide range of breast sizes representing about 95% of general population. Spatial nonuniformity (SNU) and contrast to signal deviation ratio (CDR) were used as metrics for evaluation. Results: Since the time-consuming MC simulation for library generation is precomputed, the authors’ method efficiently corrects for scatter with minimal processing time. Furthermore, the authors find that a scatter library on a simple breast model with only one input parameter, i.e., the breast diameter, sufficiently guarantees improvements in SNU and CDR. For the 15 clinical datasets, the authors’ method reduces the average SNU from 7.14% to 2.47% in coronal views and from 10.14% to 3.02% in sagittal views. On average, the CDR is improved by a factor of 1.49 in coronal views and 2.12 in sagittal

  19. MORET: Version 4.B. A multigroup Monte Carlo criticality code

    International Nuclear Information System (INIS)

    Jacquet, Olivier; Miss, Joachim; Courtois, Gerard

    2003-01-01

    MORET 4 is a three dimensional multigroup Monte Carlo code which calculates the effective multiplication factor (keff) of any configurations more or less complex as well as reaction rates in the different volumes of the geometry and the leakage out of the system. MORET 4 is the Monte Carlo code of the APOLLO2-MORET 4 standard route of CRISTAL, the French criticality package. It is the most commonly used Monte Carlo code for French criticality calculations. During the last four years, the MORET 4 team has developed or improved the following major points: modernization of the geometry, implementation of perturbation algorithms, source distribution convergence, statistical detection of stationarity, unbiased variance estimation and creation of pre-processing and post-processing tools. The purpose of this paper is not only to present the new features of MORET but also to detail clearly the physical models and the mathematical methods used in the code. (author)

  20. Analytic solutions of the multigroup space-time reactor kinetics equations

    International Nuclear Information System (INIS)

    Lee, C.E.; Rottler, S.

    1986-01-01

    The development of analytical and numerical solutions to the reactor kinetics equations is reviewed. Analytic solutions of the multigroup space-time reactor kinetics equations are developed for bare and reflected slabs and spherical reactors for zero flux, zero current and extrapolated endpoint boundary conditions. The material properties of the reactors are assumed constant in space and time, but spatially-dependent source terms and initial conditions are investigated. The system of partial differential equations is reduced to a set of linear ordinary differential equations by the Laplace transform method. These equations are solved by matrix Green's functions yielding a general matrix solution for the neutron flux and precursor concentration in the Laplace transform space. The detailed pole structure of the Laplace transform matrix solutions is investigated. The temporally- and spatially-dependent solutions are determined from the inverse Laplace transform using the Cauchy residue theorem, the theorem of Frobenius, a knowledge of the detailed pole structure and matrix operators. (author)

  1. An Experiment of Robust Parallel Algorithm for the Eigenvalue problem of a Multigroup Neutron Diffusion based on modified FETI-DP

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jonghwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Parallelization of Monte Carlo simulation is widely adpoted. There are also several parallel algorithms developed for the SN transport theory using the parallel wave sweeping algorithm and for the CPM using parallel ray tracing. For practical purpose of reactor physics application, the thermal feedback and burnup effects on the multigroup cross section should be considered. In this respect, the domain decomposition method(DDM) is suitable for distributing the expensive cross section calculation work. Parallel transport code and diffusion code based on the Raviart-Thomas mixed finite element method was developed. However most of the developed methods rely on the heuristic convergence of flux and current at the domain interfaces. Convergence was not attained in some cases. Mechanical stress computation community has also work on the DDM to solve the stress-strain equation using the finite element methods. The most successful domain decomposition method in terms of robustness is FETI-DP. We have modified the original FETI-DP to solve the eigenvalue problem for the multigroup diffusion problem in this study.

  2. America's Star Libraries, 2010: Top-Rated Libraries

    Science.gov (United States)

    Lyons, Ray; Lance, Keith Curry

    2010-01-01

    The "LJ" Index of Public Library Service 2010, "Library Journal"'s national rating of public libraries, identifies 258 "star" libraries. Created by Ray Lyons and Keith Curry Lance, and based on 2008 data from the IMLS, it rates 7,407 public libraries. The top libraries in each group get five, four, or three stars. All included libraries, stars or…

  3. Epitope selection from an uncensored peptide library displayed on avian leukosis virus

    International Nuclear Information System (INIS)

    Khare, Pranay D.; Rosales, Ana G.; Bailey, Kent R.; Russell, Stephen J.; Federspiel, Mark J.

    2003-01-01

    Phage display libraries have provided an extraordinarily versatile technology to facilitate the isolation of peptides, growth factors, single chain antibodies, and enzymes with desired binding specificities or enzymatic activities. The overall diversity of peptides in phage display libraries can be significantly limited by Escherichia coli protein folding and processing machinery, which result in sequence censorship. To achieve an optimal diversity of displayed eukaryotic peptides, the library should be produced in the endoplasmic reticulum of eukaryotic cells using a eukaryotic display platform. In the accompanying article, we presented experiments that demonstrate that polypeptides of various sizes could be efficiently displayed on the envelope glycoproteins of a eukaryotic virus, avian leukosis virus (ALV), and the displayed polypeptides could efficiently attach to cognate receptors without interfering with viral attachment and entry into susceptible cells. In this study, methods were developed to construct a model library of randomized eight amino acid peptides using the ALV eukaryotic display platform and screen the library for specific epitopes using immobilized antibodies. A virus library with approximately 2 x 10 6 different members was generated from a plasmid library of approximately 5 x 10 6 diversity. The sequences of the randomized 24 nucleotide/eight amino acid regions of representatives of the plasmid and virus libraries were analyzed. No significant sequence censorship was observed in producing the virus display library from the plasmid library. Different populations of peptide epitopes were selected from the virus library when different monoclonal antibodies were used as the target. The results of these two studies clearly demonstrate the potential of ALV as a eukaryotic platform for the display and selection of eukaryotic polypeptides libraries

  4. Morfologia dos tricomas em Tibouchina sect. Pleroma (D. Don cogn. (melastomataceae

    Directory of Open Access Journals (Sweden)

    Paulo José Fernandes Guimarães

    1999-01-01

    Full Text Available Melastomataceae é a família que apresenta a maior diversidade de tricomas. Dr. Wurdack em seu "Atlas of hairs" reconheceu 46 tipos de tricomas para as Melastomataceae neotropicais com o uso de MEV, e referiu 17 destes para o gênero Tibouchina. Em Melastomataceae a morfologia dos tricomas tem sido extensivamente usada na delimitação dos táxons, assim como em estudos que antecederam ao trabalho de Cogniaux, a monografia das Melastomataceae brasileiras publicada em 1885. Em Tibouchina sect. Pleroma a morfologia dos tricomas foi substancial para a circunscrição de várias espécies. O uso da MEV permitiu uma caracterização dos 15 tipos diferentes de tricomas observados nas 41 espécies examinadas. Os diferentes tipos são descritos com base no aspecto morfológico tanto em MO como em MEV e ilustrados através de eletromicrografias de varredura e fotomicrografias.The Melastomataceae display greater trichome diversity. Dr. Wurdack, in his Atlas of hairs, recognized 46 types of trichomes for the Neotropical Melastomataceae through the Scanning Electron Microscope (SEM, and referred 17 of them to the genus Tibouchina. The trichome morphology has been extensively used in the delimitation of the taxa in Melastomataceae, as well as in previous studies by Cogniaux, the monograph of Brazilian Melastomataceae published in 1885. The morphology of the trichomes was essential for the delimitation of several species in Tibouchina section Pleroma. The use of SEM provided the best characterization of the 15 different types of trichomes recognized among the 41 species examined. These were described based on the morphological appearance, both optical microscopy and scanning electron microscopy, and illustrated by scanning electron micrographs and photomicrographs.

  5. Featured Library: Parrish Library

    OpenAIRE

    Kirkwood, Hal P, Jr

    2015-01-01

    The Roland G. Parrish Library of Management & Economics is located within the Krannert School of Management at Purdue University. Between 2005 - 2007 work was completed on a white paper that focused on a student-centered vision for the Management & Economics Library. The next step was a massive collection reduction and a re-envisioning of both the services and space of the library. Thus began a 3 phase renovation from a 2 floor standard, collection-focused library into a single floor, 18,000s...

  6. Nuclear data libraries for gamma analysis programs, including special purpose libraries

    International Nuclear Information System (INIS)

    Westmeier, W.; Reus, U.; Siemon, K.; Westmeier, H.

    2002-01-01

    A complete set of gamma ray and other data relating to the decay of all known nuclides has been compiled from the literature and electronic sources. The data were critically reviewed, evaluated and the consistency was checked. The total catalogue is comprised of data-sets for 3841 nuclides and isomers with 106,355 gamma ray and 19,347 X ray entries. A subset of 1,627 nuclides with 32,456 gamma ray entries is extracted into a database, which provides the basis for the generation of nuclide libraries for the application in spectrometry programmes. The WINDOWS based program NUC M AN was developed which serves for the updating and editing of the database as well as the generation of task-specific user libraries for special purposes. The literature cut-off date for this catalogue is January 1, 1998. (author)

  7. High-Throughput Synthesis and Characterization of Eu Doped Ba xSr2- xSiO4 Thin Film Phosphors.

    Science.gov (United States)

    Frost, Sara; Guérin, Samuel; Hayden, Brian E; Soulié, Jean-Philippe; Vian, Chris

    2018-06-20

    High-throughput techniques have been employed for the synthesis and characterization of thin film phosphors of Eu-doped Ba x Sr 2- x SiO 4 . Direct synthesis from evaporation of the constituent elements under a flux of atomic oxygen on a sapphire substrate at 850 °C was used to directly produce thin film libraries (415 nm thickness) of the crystalline orthosilicate phase with the desired compositional variation (0.24 > x > 1.86). The orthosilicate phase could be synthesized as a pure, or predominantly pure, phase. Annealing the as synthesized library in a reducing atmosphere resulted in the reduction of the Eu while retaining the orthosilicate phase, and resulted in a materials thin film library where fluorescence excited by blue light (450 nm) was observable by the naked eye. Parallel screening of the fluorescence from the combinatorial libraries of Eu doped Ba x Sr 2- x SiO 4 has been implemented by imaging the fluorescent radiation over the library using a monochrome digital camera using a series of color filters. Informatics tools have been developed to allow the 1931 CIE color coordinates and the relative quantum efficiencies of the materials library to be rapidly assessed and mapped against composition, crystal structure and phase purity. The range of compositions gave values of CIE x between 0.17 and 0.52 and CIE y between 0.48 and 0.69 with relative efficiencies in the range 2.0 × 10 -4 -7.6 × 10 -4 . Good agreement was obtained between the thin film phosphors and the fluorescence characteristics of a number of corresponding bulk phosphor powders. The thermal quenching of fluorescence in the thin film libraries was also measured in the temperature range 25-130 °C: The phase purity of the thin film was found to significantly influence both the relative quantum efficiency and the thermal quenching of the fluorescence.

  8. Taking into account the impact of attrition on the assessment of response shift and true change: a multigroup structural equation modeling approach

    NARCIS (Netherlands)

    Verdam, M.G.E.; Oort, F.J.; van der Linden, Y.M.; Sprangers, M.A.G.

    2015-01-01

    Purpose: Missing data due to attrition present a challenge for the assessment and interpretation of change and response shift in HRQL outcomes. The objective was to handle such missingness and to assess response shift and ‘true change’ with the use of an attrition-based multigroup structural

  9. Taking into account the impact of attrition on the assessment of response shift and true change: a multigroup structural equation modeling approach

    NARCIS (Netherlands)

    Verdam, Mathilde G. E.; Oort, Frans J.; van der Linden, Yvette M.; Sprangers, Mirjam A. G.

    2015-01-01

    Missing data due to attrition present a challenge for the assessment and interpretation of change and response shift in HRQL outcomes. The objective was to handle such missingness and to assess response shift and 'true change' with the use of an attrition-based multigroup structural equation

  10. Clarifying the butter Boletes: a new genus, Butyriboletus, is established to accommodate Boletus sect. Appendiculati, and six new species are described.

    Science.gov (United States)

    Arora, David; Frank, Jonathan L

    2014-01-01

    The butter boletes (Boletus s.l. sect. Appendiculati) are an economically important group of ectomycorrhizal fungi whose basidiocarps have a yellow tube layer that often bruises blue, yellow reticulate stipe, mild flavor and firm yellow-tinged flesh that may or may not turn blue when exposed. Morphological characters and molecular data (ITS and LSU) place this group in a separate phylogenetic clade from Boletus sensu stricto. Here we establish a new genus, Butyriboletus, to accommodate 14 species of butter boletes that range from Asia to Europe, north Africa and North America. We recombine eight previously described butter bolete species and we describe six new species: four from western USA (Bu. persolidus, Bu. primiregius, Bu. autumnigius, Bu. querciregius) and two from Yunnan, China (Bu. yicibus, Bu. sanicibus). © 2014 by The Mycological Society of America.

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  12. A self-consistent nodal method in response matrix formalism for the multigroup diffusion equations

    International Nuclear Information System (INIS)

    Malambu, E.M.; Mund, E.H.

    1996-01-01

    We develop a nodal method for the multigroup diffusion equations, based on the transverse integration procedure (TIP). The efficiency of the method rests upon the convergence properties of a high-order multidimensional nodal expansion and upon numerical implementation aspects. The discrete 1D equations are cast in response matrix formalism. The derivation of the transverse leakage moments is self-consistent i.e. does not require additional assumptions. An outstanding feature of the method lies in the linear spatial shape of the local transverse leakage for the first-order scheme. The method is described in the two-dimensional case. The method is validated on some classical benchmark problems. (author)

  13. An accurate solution of point reactor neutron kinetics equations of multi-group of delayed neutrons

    International Nuclear Information System (INIS)

    Yamoah, S.; Akaho, E.H.K.; Nyarko, B.J.B.

    2013-01-01

    Highlights: ► Analytical solution is proposed to solve the point reactor kinetics equations (PRKE). ► The method is based on formulating a coefficient matrix of the PRKE. ► The method was applied to solve the PRKE for six groups of delayed neutrons. ► Results shows good agreement with other traditional methods in literature. ► The method is accurate and efficient for solving the point reactor kinetics equations. - Abstract: The understanding of the time-dependent behaviour of the neutron population in a nuclear reactor in response to either a planned or unplanned change in the reactor conditions is of great importance to the safe and reliable operation of the reactor. In this study, an accurate analytical solution of point reactor kinetics equations with multi-group of delayed neutrons for specified reactivity changes is proposed to calculate the change in neutron density. The method is based on formulating a coefficient matrix of the homogenous differential equations of the point reactor kinetics equations and calculating the eigenvalues and the corresponding eigenvectors of the coefficient matrix. A small time interval is chosen within which reactivity relatively stays constant. The analytical method was applied to solve the point reactor kinetics equations with six-groups delayed neutrons for a representative thermal reactor. The problems of step, ramp and temperature feedback reactivities are computed and the results compared with other traditional methods. The comparison shows that the method presented in this study is accurate and efficient for solving the point reactor kinetics equations of multi-group of delayed neutrons

  14. Multi-Population Invariance with Dichotomous Measures: Combining Multi-Group and MIMIC Methodologies in Evaluating the General Aptitude Test in the Arabic Language

    Science.gov (United States)

    Sideridis, Georgios D.; Tsaousis, Ioannis; Al-harbi, Khaleel A.

    2015-01-01

    The purpose of the present study was to extend the model of measurement invariance by simultaneously estimating invariance across multiple populations in the dichotomous instrument case using multi-group confirmatory factor analytic and multiple indicator multiple causes (MIMIC) methodologies. Using the Arabic version of the General Aptitude Test…

  15. Robust Sub-nanomolar Library Preparation for High Throughput Next Generation Sequencing.

    Science.gov (United States)

    Wu, Wells W; Phue, Je-Nie; Lee, Chun-Ting; Lin, Changyi; Xu, Lai; Wang, Rong; Zhang, Yaqin; Shen, Rong-Fong

    2018-05-04

    Current library preparation protocols for Illumina HiSeq and MiSeq DNA sequencers require ≥2 nM initial library for subsequent loading of denatured cDNA onto flow cells. Such amounts are not always attainable from samples having a relatively low DNA or RNA input; or those for which a limited number of PCR amplification cycles is preferred (less PCR bias and/or more even coverage). A well-tested sub-nanomolar library preparation protocol for Illumina sequencers has however not been reported. The aim of this study is to provide a much needed working protocol for sub-nanomolar libraries to achieve outcomes as informative as those obtained with the higher library input (≥ 2 nM) recommended by Illumina's protocols. Extensive studies were conducted to validate a robust sub-nanomolar (initial library of 100 pM) protocol using PhiX DNA (as a control), genomic DNA (Bordetella bronchiseptica and microbial mock community B for 16S rRNA gene sequencing), messenger RNA, microRNA, and other small noncoding RNA samples. The utility of our protocol was further explored for PhiX library concentrations as low as 25 pM, which generated only slightly fewer than 50% of the reads achieved under the standard Illumina protocol starting with > 2 nM. A sub-nanomolar library preparation protocol (100 pM) could generate next generation sequencing (NGS) results as robust as the standard Illumina protocol. Following the sub-nanomolar protocol, libraries with initial concentrations as low as 25 pM could also be sequenced to yield satisfactory and reproducible sequencing results.

  16. Library 3.0 intelligent libraries and apomediation

    CERN Document Server

    Kwanya, Tom; Underwood, Peter

    2015-01-01

    The emerging generation of research and academic library users expect the delivery of user-centered information services. 'Apomediation' refers to the supporting role librarians can give users by stepping in when users need help. Library 3.0 explores the ongoing debates on the "point oh” phenomenon and its impact on service delivery in libraries. This title analyses Library 3.0 and its potential in creating intelligent libraries capable of meeting contemporary needs, and the growing role of librarians as apomediators. Library 3.0 is divided into four chapters. The first chapter introduces and places the topic in context. The second chapter considers "point oh” libraries. The third chapter covers library 3.0 librarianship, while the final chapter explores ways libraries can move towards '3.0'.

  17. Spectrum of the multigroup neutron transport operator for bounded spatial domains

    International Nuclear Information System (INIS)

    Larsen, E.W.

    1979-01-01

    The spectrum of the multigroup neutron transport operator A is studied for bounded spatial regions D which consist of a finite number of material subregions. Our main results provide simple conditions on the material cross sections which guarantee that (1) A possesses eigenvalues in the finite plane; (2) A possesses a ''leading'' eigenvalue lambda 0 which is real, not less than the real part of any other eigenvalue, and to which there corresponds at least one nonnegative eigenfunction psi/sub lambda/0; and (3) A possesses a ''dominant'' eigenvalue lambda 0 which is real, simple, greater than the real part of any other eigenvalue, and whose eigenfunction psi/sub lambda/0 satisfies psi/sub lambda/0> or =0 and ∫psi/sub lambda/0d 2 Ω>0. We give examples to illustrate the results and to show that a leading eigenvalue need not be simple, nor its eigenfunction(s) positive

  18. Taking into account the impact of attrition on the assessment of response shift and true change: a multigroup structural equation modeling approach.

    Science.gov (United States)

    Verdam, Mathilde G E; Oort, Frans J; van der Linden, Yvette M; Sprangers, Mirjam A G

    2015-03-01

    Missing data due to attrition present a challenge for the assessment and interpretation of change and response shift in HRQL outcomes. The objective was to handle such missingness and to assess response shift and 'true change' with the use of an attrition-based multigroup structural equation modeling (SEM) approach. Functional limitations and health impairments were measured in 1,157 cancer patients, who were treated with palliative radiotherapy for painful bone metastases, before [time (T) 0], every week after treatment (T1 through T12), and then monthly for up to 2 years (T13 through T24). To handle missing data due to attrition, the SEM procedure was extended to a multigroup approach, in which we distinguished three groups: short survival (3-5 measurements), medium survival (6-12 measurements), and long survival (>12 measurements). Attrition after third, sixth, and 13th measurement occasions was 11, 24, and 41 %, respectively. Results show that patterns of change in functional limitations and health impairments differ between patients with short, medium, or long survival. Moreover, three response-shift effects were detected: recalibration of 'pain' and 'sickness' and reprioritization of 'physical functioning.' If response-shift effects would not have been taken into account, functional limitations and health impairments would generally be underestimated across measurements. The multigroup SEM approach enables the analysis of data from patients with different patterns of missing data due to attrition. This approach does not only allow for detection of response shift and assessment of true change across measurements, but also allow for detection of differences in response shift and true change across groups of patients with different attrition rates.

  19. Croatian library leaders’ views on (their library quality

    Directory of Open Access Journals (Sweden)

    Kornelija Petr Balog

    2014-04-01

    Full Text Available The purpose of this paper is to determine and describe the library culture in Croatian public libraries. Semi-structured interviews with 14 library directors (ten public and four academic were conducted. The tentative discussion topics were: definition of quality, responsibility for quality, satisfaction with library services, familiarization with user perspective of library and librarians, monitoring of user expectations and opinions. These interviews incorporate some of the findings of the project Evaluation of library and information services: public and academic libraries. The project investigates library culture in Croatian public and academic libraries and their preparedness for activities of performance measurement. The interviews reveal that library culture has changed positively in the past few years and that library leaders have positive attitude towards quality and evaluation activities. Library culture in Croatian libraries is a relatively new concept and as such was not actively developed and/or created. This article looks into the library culture of Croatian libraries, but at the same time investigates whether there is any trace of culture of assessment in them. Also, this article brings the latest update on views, opinions and atmosphere in Croatian public and academic libraries.

  20. Preparation of human Melanocortin-4 receptor agonist libraries: linear peptides X-Y-DPhe7-Arg8-Trp(or 2-Nal)9-Z-NH2.

    Science.gov (United States)

    Cheung, Adrian Wai-Hing; Qi, Lida; Gore, Vijay; Chu, Xin-Jie; Bartkovitz, David; Kurylko, Grazyna; Swistok, Joseph; Danho, Waleed; Chen, Li; Yagaloff, Keith

    2005-12-15

    Two libraries of hMC4R agonists, X-Y-DPhe(7)-Arg(8)-2-Nal(9)-Z-NH(2) and X-Y-DPhe(7)-Arg(8)-Trp(9)-Z-NH(2), totaling 185 peptides were prepared using Irori radiofrequency tagging technology and Argonaut Quest 210 Synthesizer, where X stands for N-caps, Y for His(6) surrogates and Z for Gly(10) surrogates. As a result of this study, His-modified pentapeptides with Trp were found to be more hMC4R potent than the corresponding 2-Nal analogs, novel N-caps and Gly surrogates were identified and 19 new peptides which are potent hMC4R agonists (EC(50) 1-15nM) and selective against hMC1R were discovered.

  1. Travelling Wave Solutions in Multigroup Age-Structured Epidemic Models

    Science.gov (United States)

    Ducrot, Arnaut; Magal, Pierre; Ruan, Shigui

    2010-01-01

    Age-structured epidemic models have been used to describe either the age of individuals or the age of infection of certain diseases and to determine how these characteristics affect the outcomes and consequences of epidemiological processes. Most results on age-structured epidemic models focus on the existence, uniqueness, and convergence to disease equilibria of solutions. In this paper we investigate the existence of travelling wave solutions in a deterministic age-structured model describing the circulation of a disease within a population of multigroups. Individuals of each group are able to move with a random walk which is modelled by the classical Fickian diffusion and are classified into two subclasses, susceptible and infective. A susceptible individual in a given group can be crisscross infected by direct contact with infective individuals of possibly any group. This process of transmission can depend upon the age of the disease of infected individuals. The goal of this paper is to provide sufficient conditions that ensure the existence of travelling wave solutions for the age-structured epidemic model. The case of two population groups is numerically investigated which applies to the crisscross transmission of feline immunodeficiency virus (FIV) and some sexual transmission diseases.

  2. cDNA library construction of two human Demodexspecies.

    Science.gov (United States)

    Niu, DongLing; Wang, RuiLing; Zhao, YaE; Yang, Rui; Hu, Li; Lei, YuYang; Dan, WeiChao

    2017-06-01

    The research of Demodex, a type of pathogen causing various dermatoses in animals and human beings, is lacking at RNA level. This study aims at extracting RNA and constructing cDNA library for Demodex. First, P. cuniculiand D. farinaewere mixed to establish homogenization method for RNA extraction. Second, D. folliculorumand D. breviswere collected and preserved in Trizol, which were mixed with D. farinaerespectively to extract RNA. Finally, cDNA library was constructed and its quality was assessed. The results indicated that for D. folliculorum& D. farinae, the recombination rate of cDNA library was 90.67% and the library titer was 7.50 × 104 pfu/ml. 17 of the 59 positive clones were predicted to be of D. folliculorum; For D. brevis& D. farinae, the recombination rate was 90.96% and the library titer was 7.85 x104 pfu/ml. 40 of the 59 positive clones were predicted to be of D. brevis. Further detection by specific primers demonstrated that mtDNA cox1, cox3and ATP6 detected from cDNA libraries had 96.52%-99.73% identities with the corresponding sequences in GenBank. In conclusion, the cDNA libraries constructed for Demodexmixed with D. farinaewere successful and could satisfy the requirements for functional genes detection.

  3. Summary description of the Babcock and Wilcox integrated nuclear design system

    International Nuclear Information System (INIS)

    Wittkopf, W.A.

    1976-03-01

    The Babcock and Wilcox integrated nuclear design system is divided into three broad areas: basic nuclear data processing, applications data processing, and nuclear design calculations. In basic nuclear data processing, basic nuclear data are collected, evaluated, and processed into a specified fine-energy mesh multigroup data file called a Master Library. In applications data processing, data for selected materials are retrieved from the Master Library and processed into an optimally structured, multigroup Production Library. Using these data and input descriptions of cells or regions, neutron spectra are generated and few-group constants are computed and fitted as a function of fuel burnup, initial enrichment, temperature, etc. In nuclear design calculations, few-group cross-section fits and descriptions of each core region and core geometry are input to a diffusion-depletion program or a nodal program that computes core reactivity, core power distribution, control rod worth, fuel cycle studies, core operating limitations, etc

  4. Solution of the Multigroup-Diffusion equation by the response matrix method

    International Nuclear Information System (INIS)

    Oliveira, C.R.E.

    1980-10-01

    A preliminary analysis of the response matrix method is made, considering its application to the solution of the multigroup diffusion equations. The one-dimensional formulation is presented and used to test some flux expansions, seeking the application of the method to the two-dimensional problem. This formulation also solves the equations that arise from the integro-differential synthesis algorithm. The slow convergence of the power method, used to solve the eigenvalue problem, and its acceleration by means of the Chebyshev polynomial method, are also studied. An algorithm for the estimation of the dominance ratio is presented, based on the residues of two successive iteration vectors. This ratio, which is not known a priori, is fundamental for the efficiency of the method. Some numerical problems are solved, testing the 1D formulation of the response matrix method, its application to the synthesis algorithm and also, at the same time, the algorithm to accelerate the source problem. (Author) [pt

  5. Libraries for users services in academic libraries

    CERN Document Server

    Alvite, Luisa

    2010-01-01

    This book reviews the quality and evolution of academic library services. It revises service trends offered by academic libraries and the challenge of enhancing traditional ones such as: catalogues, repositories and digital collections, learning resources centres, virtual reference services, information literacy and 2.0 tools.studies the role of the university library in the new educational environment of higher educationrethinks libraries in academic contextredefines roles for academic libraries

  6. Frequencies of X-ray and fast neutron induced chromosome translocations in human peripheral blood lymphocytes as detected by in situ hybridization using chromosome specific DNA libraries

    International Nuclear Information System (INIS)

    Natarajan, A.T.; Darroudi, F.; Vermeulen, S.; Wiegant, J.

    1992-01-01

    DNA libraries of six human chromosomes were used to detect translocations in human lymphocytes induced by different doses of X-rays and fast neutrons. Results show that with X-rays, one can detect about 1.5 to 2.0 fold more translocations in comparison to dicentrics, whereas following fast neutron irradiation, the difference between these two classes of aberrations are significantly different at high doses. In addition, triple fluorescent in situ hybridization technique was used to study the frequencies of radiation-induced translocations involving a specific chromosome. Chromosome number 1 was found to be involved in translocations more frequently than chromosomes number 2, 3, 4, 8 and X. (author). 10 refs., 1 fig., 2 tabs

  7. An improved yeast transformation method for the generation of very large human antibody libraries.

    Science.gov (United States)

    Benatuil, Lorenzo; Perez, Jennifer M; Belk, Jonathan; Hsieh, Chung-Ming

    2010-04-01

    Antibody library selection by yeast display technology is an efficient and highly sensitive method to identify binders to target antigens. This powerful selection tool, however, is often hampered by the typically modest size of yeast libraries (approximately 10(7)) due to the limited yeast transformation efficiency, and the full potential of the yeast display technology for antibody discovery and engineering can only be realized if it can be coupled with a mean to generate very large yeast libraries. We describe here a yeast transformation method by electroporation that allows for the efficient generation of large antibody libraries up to 10(10) in size. Multiple components and conditions including CaCl(2), MgCl(2), sucrose, sorbitol, lithium acetate, dithiothreitol, electroporation voltage, DNA input and cell volume have been tested to identify the best combination. By applying this developed protocol, we have constructed a 1.4 x 10(10) human spleen antibody library essentially in 1 day with a transformation efficiency of 1-1.5 x 10(8) transformants/microg vector DNA. Taken together, we have developed a highly efficient yeast transformation method that enables the generation of very large and productive human antibody libraries for antibody discovery, and we are now routinely making 10(9) libraries in a day for antibody engineering purposes.

  8. Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

    International Nuclear Information System (INIS)

    Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

    1986-01-01

    For a variety of applications, e.g., accelerator shielding design, neutrons in radiotherapy, radiation damage studies, etc., it is necessary to carry out transport calculations involving medium-energy (greater than or equal to20 MeV) neutrons. A previous paper described neutron-photon multigroup cross sections in the ANISN format for neutrons from thermal to 400 MeV. In the present paper the cross-section data presented previously have been revised to make them agree with available experimental data. 7 refs., 1 fig

  9. SCHOOL COMMUNITY PERCEPTION OF LIBRARY APPS AGAINTS LIBRARY EMPOWERMENT

    Directory of Open Access Journals (Sweden)

    Achmad Riyadi Alberto

    2017-07-01

    Full Text Available Abstract. This research is motivated by the development of information and communication technology (ICT in the library world so rapidly that allows libraries in the present to develop its services into digital-based services. This study aims to find out the school community’s perception of library apps developed by Riche Cynthia Johan, Hana Silvana, and Holin Sulistyo and its influence on library empowerment at the library of SD Laboratorium Percontohan UPI Bandung. Library apps in this research belong to the context of m-libraries, which is a library that meets the needs of its users by using mobile platforms such as smartphones,computers, and other mobile devices. Empowerment of library is the utilization of all aspects of the implementation of libraries to the best in order to achieve the expected goals. An analysis of the schoolcommunity’s perception of library apps using the Technology Acceptance Model (TAM includes: ease of use, usefulness, usability, usage trends, and real-use conditions. While the empowerment of the library includes aspects: information empowerment, empowerment of learning resources, empowerment of human resources, empowerment of library facilities, and library promotion. The research method used in this research is descriptive method with quantitative approach. Population and sample in this research is school community at SD Laboratorium Percontohan UPI Bandung. Determination of sample criteria by using disproportionate stratified random sampling with the number of samples of 83 respondents. Data analysis using simple linear regression to measure the influence of school community perception about library apps to library empowerment. The result of data analysis shows that there is influence between school community perception about library apps to library empowerment at library of SD Laboratorium Percontohan UPI Bandung which is proved by library acceptance level and library empowerment improvement.

  10. Open Source Opens Doors: Repurposing Library Software to Facilitate Faculty Research and Collaboration

    Directory of Open Access Journals (Sweden)

    Sandra L. Stump

    2013-10-01

    Full Text Available Normal 0 false false false EN-US X-NONE X-NONE Asked to convert a faculty-created Microsoft Word document of biblical references found within popular films into a searchable database for scholars, the Albright College library staff helped create a multi-access database called Bible in the Reel World. The database relied on student workers for inputting data, used MARC standard formatting for future portability, and encouraged interactive feedback, enabling scholars to submit comments and suggest additional films and references. Using the open source integrated library system Koha, MarcEdit software, and free record exporting from IMDb, library staff created a fully-searchable database for researchers and scholars to examine the use of scripture in popular film.

  11. USGS Digital Spectral Library splib06a

    Science.gov (United States)

    Clark, Roger N.; Swayze, Gregg A.; Wise, Richard A.; Livo, K. Eric; Hoefen, Todd M.; Kokaly, Raymond F.; Sutley, Stephen J.

    2007-01-01

    ), one simply needs a diagnostic absorption band. The mapping system uses continuum-removed reference spectral features fitted to features in observed spectra. Spectral features for such algorithms can be obtained from a spectrum of a sample containing large amounts of contaminants, including those that add other spectral features, as long as the shape of the diagnostic feature of interest is not modified. If, however, the data are needed for radiative transfer models to derive mineral abundances from reflectance spectra, then completely uncontaminated spectra are required. This library contains spectra that span a range of quality, with purity indicators to flag spectra for (or against) particular uses. Acquiring spectral measurements and performing sample characterizations for this library has taken about 15 person-years of effort. Software to manage the library and provide scientific analysis capability is provided (Clark, 1980, 1993). A personal computer (PC) reader for the library is also available (Livo and others, 1993). The program reads specpr binary files (Clark, 1980, 1993) and plots spectra. Another program that reads the specpr format is written in IDL (Kokaly, 2005). In our view, an ideal spectral library consists of samples covering a very wide range of materials, has large wavelength range with very high precision, and has enough sample analyses and documentation to establish the quality of the spectra. Time and available resources limit what can be achieved. Ideally, for each mineral, the sample analysis would include X-ray diffraction (XRD), electron microprobe (EM) or X-ray fluorescence (XRF), and petrographic microscopic analyses. For some minerals, such as iron oxides, additional analyses such as Mossbauer would be helpful. We have found that to make the basic spectral measurements, provide XRD, EM or XRF analyses, and microscopic analyses, document the results, and complete an entry of one spectral library sample, all takes about

  12. Use of metal/uranium mixtures to explore data uncertainties

    International Nuclear Information System (INIS)

    Parks, C.V.; Jordan, W.C.; Petrie, L.M.; Wright, R.Q.

    1995-01-01

    A table of k ∞ values for three homogenized metal/ 235 U systems calculated using both MCNP and the SCALE code system was presented in Ref. 3. The homogenized metal/ 235 U ratios were selected such that the MCNP analyses for each mixture provided k ∞ ≅ 1.0. The metals considered were Al, Zr, and Fe. These simplified systems were created in an effort to ease an investigation of discrepant results obtained using MCNP and SCALE to analyze large, dry systems of metal-clad, highly enriched fuel assemblies. Reference 3 has received considerable attention at ORNL and elsewhere because the reported k ∞ values varied by as much as 38% between the MCNP results and those of SCALE. The ORNL approach was to analyze the systems using a broad range of codes and data and to seek an understanding of the discrepancies by studying differences in the basic data and processing methods. The continuous-energy codes and data applied in the ORNL study were (1) MCNP, using ENDF/B-V, ENDF/B-VI, and LANL data evaluations, (2) VIM, using ENDF/B-V data, and (3) MONK, using a 8,200-point library based on UKNDL evaluations and a preliminary JEF library. The VIM code provides treatment of unresolved resonances; MCNP does not. The MONK analyses provided a result using both an independent code and independent data evaluations. Although accessing continuous-energy data typically requires the use of Monte Carlo codes, 1-D deterministic codes can be used to accurately calculate K ∞ values using a variety of multigroup data libraries and processing methods. The multigroup codes used in the study were MC and the CSAS1X sequence of the SCALE system. Both systems provide problem-dependent resonance processing of cross-section data and available fine-group libraries were used for the analyses. Broad-group libraries were not studied in any depth because there were non-readily available for intermediate-energy systems

  13. The Weapons Laboratory Technical Library: Automating with ’Stilas’

    Science.gov (United States)

    1990-03-01

    version of the system to LC in October 1988. -X- A small business specializing in library automation, SIRSI was founded in 1979 by library and...computer specialists, and has a strong reputation based upon the success of their UNIX-based Unicorn Collection Management System. SIRSI offers a complete...system based on the Unicorn and BRS/ Search systems. The contracted STILAS package includes UNISYS hardware, software written in the C language

  14. Aspherical-atom modeling of coordination compounds by single-crystal X-ray diffraction allows the correct metal atom to be identified.

    Science.gov (United States)

    Dittrich, Birger; Wandtke, Claudia M; Meents, Alke; Pröpper, Kevin; Mondal, Kartik Chandra; Samuel, Prinson P; Amin Sk, Nurul; Singh, Amit Pratap; Roesky, Herbert W; Sidhu, Navdeep

    2015-02-02

    Single-crystal X-ray diffraction (XRD) is often considered the gold standard in analytical chemistry, as it allows element identification as well as determination of atom connectivity and the solid-state structure of completely unknown samples. Element assignment is based on the number of electrons of an atom, so that a distinction of neighboring heavier elements in the periodic table by XRD is often difficult. A computationally efficient procedure for aspherical-atom least-squares refinement of conventional diffraction data of organometallic compounds is proposed. The iterative procedure is conceptually similar to Hirshfeld-atom refinement (Acta Crystallogr. Sect. A- 2008, 64, 383-393; IUCrJ. 2014, 1,61-79), but it relies on tabulated invariom scattering factors (Acta Crystallogr. Sect. B- 2013, 69, 91-104) and the Hansen/Coppens multipole model; disordered structures can be handled as well. Five linear-coordinate 3d metal complexes, for which the wrong element is found if standard independent-atom model scattering factors are relied upon, are studied, and it is shown that only aspherical-atom scattering factors allow a reliable assignment. The influence of anomalous dispersion in identifying the correct element is investigated and discussed. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  15. Multigroup Path Analysis of the Influence of Healthcare Quality, by Different Health Insurance Types.

    Science.gov (United States)

    Hong, Yong-Rock; Holcomb, Derek; Ballard, Michael; Schwartz, Laurel

    Winds of change have been blowing in the U.S. healthcare system since passage of the Affordable Care Act. Examining differences between individuals covered by different types of insurance is essential if healthcare executives are to develop new strategies in response to the emerging health insurance market. In this study, we used multigroup path analysis models to examine the moderating effects of health insurance on direct and indirect associations with general health status, satisfaction with received care, financial burden, and perceived value of the healthcare system. Data were obtained from the 2012 Medical Expenditure Panel Survey and analyzed according to the types of insurance: private, public, and military. With the satisfactory fit of the model (χ = 2,532.644, df = 96, p spending.

  16. Welcome to the National Wetlands Research Center Library: Not Just Another Library-A Special Library

    Science.gov (United States)

    Broussard, Linda

    2007-01-01

    Libraries are grouped into four major types: public, school, academic, and special. The U.S. Geological Survey's (USGS) National Wetlands Research Center (NWRC) library is classified as a special library because it is sponsored by the Federal government, and the collections focus on a specific subject. The NWRC library is the only USGS library dedicated to wetland science. Library personnel offer expert research services to meet the informational needs of NWRC scientists, managers, and support personnel. The NWRC library participates in international cataloging and resource sharing, which allows libraries from throughout the world to borrow from its collections. This sharing facilitates the research of other governmental agencies, universities, and those interested in the study of wetlands.

  17. Assessing Library Automation and Virtual Library Development in Four Academic Libraries in Oyo, Oyo State, Nigeria

    Science.gov (United States)

    Gbadamosi, Belau Olatunde

    2011-01-01

    The paper examines the level of library automation and virtual library development in four academic libraries. A validated questionnaire was used to capture the responses from academic librarians of the libraries under study. The paper discovers that none of the four academic libraries is fully automated. The libraries make use of librarians with…

  18. [cDNA library construction from panicle meristem of finger millet].

    Science.gov (United States)

    Radchuk, V; Pirko, Ia V; Isaenkov, S V; Emets, A I; Blium, Ia B

    2014-01-01

    The protocol for production of full-size cDNA using SuperScript Full-Length cDNA Library Construction Kit II (Invitrogen) was tested and high quality cDNA library from meristematic tissue of finger millet panicle (Eleusine coracana (L.) Gaertn) was created. The titer of obtained cDNA library comprised 3.01 x 10(5) CFU/ml in avarage. In average the length of cDNA insertion consisted about 1070 base pairs, the effectivity of cDNA fragment insertions--99.5%. The selective sequencing of cDNA clones from created library was performed. The sequences of cDNA clones were identified with usage of BLAST-search. The results of cDNA library analysis and selective sequencing represents prove good functionality and full length character of inserted cDNA clones. Obtained cDNA library from meristematic tissue of finger millet panicle represents good and valuable source for isolation and identification of key genes regulating metabolism and meristematic development and for mining of new molecular markers to conduct out high quality genetic investigations and molecular breeding as well.

  19. A lumped parameter method of characteristics approach and multigroup kernels applied to the subgroup self-shielding calculation in MPACT

    International Nuclear Information System (INIS)

    Stimpson, Shane G.; Liu, Yuxuan; Collins, Benjamin S.; Clarno, Kevin T.

    2017-01-01

    An essential component of the neutron transport solver is the resonance self-shielding calculation used to determine equivalence cross sections. The neutron transport code, MPACT, is currently using the subgroup self-shielding method, in which the method of characteristics (MOC) is used to solve purely absorbing fixed-source problems. Recent efforts incorporating multigroup kernels to the MOC solvers in MPACT have reduced runtime by roughly 2×. Applying the same concepts for self-shielding and developing a novel lumped parameter approach to MOC, substantial improvements have also been made to the self-shielding computational efficiency without sacrificing any accuracy. These new multigroup and lumped parameter capabilities have been demonstrated on two test cases: (1) a single lattice with quarter symmetry known as VERA (Virtual Environment for Reactor Applications) Progression Problem 2a and (2) a two-dimensional quarter-core slice known as Problem 5a-2D. From these cases, self-shielding computational time was reduced by roughly 3–4×, with a corresponding 15–20% increase in overall memory burden. An azimuthal angle sensitivity study also shows that only half as many angles are needed, yielding an additional speedup of 2×. In total, the improvements yield roughly a 7–8× speedup. Furthermore given these performance benefits, these approaches have been adopted as the default in MPACT.

  20. A lumped parameter method of characteristics approach and multigroup kernels applied to the subgroup self-shielding calculation in MPACT

    Directory of Open Access Journals (Sweden)

    Shane Stimpson

    2017-09-01

    Full Text Available An essential component of the neutron transport solver is the resonance self-shielding calculation used to determine equivalence cross sections. The neutron transport code, MPACT, is currently using the subgroup self-shielding method, in which the method of characteristics (MOC is used to solve purely absorbing fixed-source problems. Recent efforts incorporating multigroup kernels to the MOC solvers in MPACT have reduced runtime by roughly 2×. Applying the same concepts for self-shielding and developing a novel lumped parameter approach to MOC, substantial improvements have also been made to the self-shielding computational efficiency without sacrificing any accuracy. These new multigroup and lumped parameter capabilities have been demonstrated on two test cases: (1 a single lattice with quarter symmetry known as VERA (Virtual Environment for Reactor Applications Progression Problem 2a and (2 a two-dimensional quarter-core slice known as Problem 5a-2D. From these cases, self-shielding computational time was reduced by roughly 3–4×, with a corresponding 15–20% increase in overall memory burden. An azimuthal angle sensitivity study also shows that only half as many angles are needed, yielding an additional speedup of 2×. In total, the improvements yield roughly a 7–8× speedup. Given these performance benefits, these approaches have been adopted as the default in MPACT.

  1. Design, synthesis, and evaluation of an alpha-helix mimetic library targeting protein-protein interactions.

    Science.gov (United States)

    Shaginian, Alex; Whitby, Landon R; Hong, Sukwon; Hwang, Inkyu; Farooqi, Bilal; Searcey, Mark; Chen, Jiandong; Vogt, Peter K; Boger, Dale L

    2009-04-22

    The design and solution-phase synthesis of an alpha-helix mimetic library as an integral component of a small-molecule library targeting protein-protein interactions are described. The iterative design, synthesis, and evaluation of the candidate alpha-helix mimetic was initiated from a precedented triaryl template and refined by screening the designs for inhibition of MDM2/p53 binding. Upon identifying a chemically and biologically satisfactory design and consistent with the screening capabilities of academic collaborators, the corresponding complete library was assembled as 400 mixtures of 20 compounds (20 x 20 x 20-mix), where the added subunits are designed to mimic all possible permutations of the naturally occurring i, i + 4, i + 7 amino acid side chains of an alpha-helix. The library (8000 compounds) was prepared using a solution-phase synthetic protocol enlisting acid/base liquid-liquid extractions for purification on a scale that insures its long-term availability for screening campaigns. Screening of the library for inhibition of MDM2/p53 binding not only identified the lead alpha-helix mimetic upon which the library was based, but also suggests that a digestion of the initial screening results that accompany the use of such a comprehensive library can provide insights into the nature of the interaction (e.g., an alpha-helix mediated protein-protein interaction) and define the key residues and their characteristics responsible for recognition.

  2. MARKETING LIBRARY SERVICES IN ACADEMIC LIBRARIES: A ...

    African Journals Online (AJOL)

    MARKETING LIBRARY SERVICES IN ACADEMIC LIBRARIES: A TOOL FOR SURVIVAL IN THE ... This article discusses the concept of marketing library and information services as an ... EMAIL FREE FULL TEXT EMAIL FREE FULL TEXT

  3. Libraries and Accessibility: Istanbul Public Libraries Case

    Directory of Open Access Journals (Sweden)

    Gül Yücel

    2016-12-01

    Full Text Available In the study; the assessment of accessibility has been conducted in Istanbul public libraries within the scope of public area. Public libraries commonly serve with its user of more than 20 million in total, spread to the general of Turkey, having more than one thousand branches in the centrums and having more than one million registered members. The building principles and standards covering the subjects such as the selection of place, historical and architectural specification of the region, distance to the centre of population and design in a way that the disabled people could benefit from the library services fully have been determined with regulations in the construction of new libraries. There are works for the existent libraries such as access for the disabled, fire safety precautions etc. within the scope of the related standards. Easy access by everyone is prioritized in the public libraries having a significant role in life-long learning. The purpose of the study is to develop solution suggestions for the accessibility problems in the public libraries. The study based on the eye inspection and assessments carried out within the scope of accessibility in the public libraries subsidiary to Istanbul Culture and Tourism Provincial Directorate Library and Publications Department within the provincial borders of Istanbul. The arrangements such as reading halls, study areas, book shelves etc. have been examined within the frame of accessible building standards. Building entrances, ramps and staircases, horizontal and vertical circulation of building etc. have been taken into consideration within the scope of accessible building standards. The subjects such as the reading and studying areas and book shelf arrangements for the library have been assessed within the scope of specific buildings. There are a total of 34 public libraries subsidiary to Istanbul Culture and Tourism Provincial Directorate on condition that 20 ea. of them are in the

  4. Program of evaluations for nuclear data centers

    International Nuclear Information System (INIS)

    Cunha Menezes Filho, A. da.

    1980-01-01

    Standard problems for evaluating basic data libraries of multigroup constant generation, and codes for calculating integral parameters existing or will exist in the near future in the advanced study division (EAV), are defined. (M.C.K.) [pt

  5. Public libraries in the library regions in the year 2009

    Directory of Open Access Journals (Sweden)

    Milena Bon

    2011-01-01

    Full Text Available Purpose: Regional public libraries were initiated in 2003 to connect professional activities of libraries within regional networks and to ensure coordinated library development in a region in cooperation with the Library System Development Centre at the National and University Library performing a coordinating role. The article analyses the performance of public libraries and their integration in regional library networks in order to find out the level of development of conditions of performance of public libraries.Methodology/approach: Statistical data for the year 2009 were the basis for the overview of library activities of ten library regions with regard to applicable legislation and library standards. The level of regional library activities is compared to the socio-economic situation of statistical regions thus representing a new approach to the presentation of Slovenian’s public libraries’ development.Results: Absolute values indicate better development of nine libraries in the central Slovenia region while relative values offer a totally different picture. Four libraries in the region of Nova Gorica prove the highest level of development.Research limitation: Research is limited to the year 2009 and basic statistical analysis.Originality/practical implications: Findings of the analysis are useful for public libraries to plan their development strategy within a region and for financial bodies to provide for adequate financing for library activities in a specific region. The basic condition for successful public library performance is the even and harmonized development of conditions of performance as recommended by library standards.

  6. FENDL2/A-MCNP, FENDL2/A-VITJE and FENDL2/A-VITJFLAT. The processed FENDL-2 neutron activation cross-section data files. Summary documentation

    International Nuclear Information System (INIS)

    Pashchenko, A.B.; Wienke, H.

    1997-01-01

    This document summarizes the libraries of neutron activation cross-section data processed into the following three formats: continuous energy format as used by the Monte Carlo neutron/photon transport code MCNP4A; VITAMIN-J 175 multigroup format weighted with the VITAMIN-E weighting spectrum as used by the transmutation codes REAC*2/3 and FOUR ACES; VITAMIN-J 175 multigroup ENDF-6 format, with a flat weighting spectrum. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape. (author)

  7. Two-dimensional semi-analytic nodal method for multigroup pin power reconstruction

    International Nuclear Information System (INIS)

    Seung Gyou, Baek; Han Gyu, Joo; Un Chul, Lee

    2007-01-01

    A pin power reconstruction method applicable to multigroup problems involving square fuel assemblies is presented. The method is based on a two-dimensional semi-analytic nodal solution which consists of eight exponential terms and 13 polynomial terms. The 13 polynomial terms represent the particular solution obtained under the condition of a 2-dimensional 13 term source expansion. In order to achieve better approximation of the source distribution, the least square fitting method is employed. The 8 exponential terms represent a part of the analytically obtained homogeneous solution and the 8 coefficients are determined by imposing constraints on the 4 surface average currents and 4 corner point fluxes. The surface average currents determined from a transverse-integrated nodal solution are used directly whereas the corner point fluxes are determined during the course of the reconstruction by employing an iterative scheme that would realize the corner point balance condition. The outgoing current based corner point flux determination scheme is newly introduced. The accuracy of the proposed method is demonstrated with the L336C5 benchmark problem. (authors)

  8. General solution of the multigroup spherical harmonics equations in R-Z geometry

    International Nuclear Information System (INIS)

    Matausek, M.

    1983-01-01

    In the present paper the generalization is performed of the procedure to solve multigroup spherical harmonics equations, which has originally been proposed and developed foe one-dimensional systems in cylindrical or spherical geometry, and later extended for special case of a two-dimensional system in r-z geometry. The expressions are derived for the axial and the radial dependence of the group values of the neutron flux moments, in the P-3 approximation of the spherical harmonics method, in a cylindrically symmetrical system with an arbitrary number of material regions in both r and z directions. In the special case of an axially homogeneous system, these expressions reduce to the relations derived previously. The analysis is performed of the possibilities to satisfy the boundary conditions in the case when the system considered represents an elementary reactor lattice cell and in the case when the system represents a reactor as a whole. The computational effort is estimated for system of a given configuration. (author)

  9. ANISN-FONTENAY, 1-D Planar, Spherical, Cylindrical Neutron Transport and Gamma Transport with Deep Penetration

    International Nuclear Information System (INIS)

    Devillers, C.

    1973-01-01

    1 - Nature of physical problem solved: The ANISN system treats neutron and gamma transport in one-dimensional plane, spherical and cylinder geometry. The multigroup cross sections prepared by the programs LIANE and SUPERTOG are processed by the program RETTOG, which produces a binary library with Legendre expansions. The binary library can be updated and edited with the program LGR/B. The photon multigroup cross sections are created with the program GAMLEG/A. If the bulk of the data is too large, the program TAPEMA produces a special group-by-group library. The volume sources are calculated from a reduced set of input data and punched in a format suitable for input to ANISN, using the program PRESOU. The program ANISN calculates fluxes by groups, space intervals, angle and any number of reaction rates. The energy and space dependent fluxes are stored on tape and can be reprocessed, edited and plotted with the program ANISEX, which also permits to calculate supplementary reaction rates. The program ANISN can condense cross sections into a reduced number of groups. The ANISN system is used as a reference system for the evaluation of approximation methods (space-diffusion or point- kernel) or for the preparation of multigroup libraries for 2- dimensional transport codes (DOT). In particular it is used for shielding problems with high attenuation in water reactors and fast reactors. 2 - Method of solution: Method of discrete ordinates. The program has been designed to treat deep penetration with detailed calculation of spectrum as function of angle. Tests for pointwise convergence have also been introduced. 3 - Restrictions on the complexity of the problem: The complexity of the problem is limited by the storage size

  10. Libraries Today, Libraries Tomorrow: Contemporary Library Practices and the Role of Library Space in the L

    Directory of Open Access Journals (Sweden)

    Ana Vogrinčič Čepič

    2013-09-01

    Full Text Available ABSTRACTPurpose: The article uses sociological concepts in order to rethink the changes in library practices. Contemporary trends are discussed with regard to the changing nature of working habits, referring mostly to the new technology, and the (emergence of the third space phenomenon. The author does not regard libraries only as concrete public service institutions, but rather as complex cultural forms, taking in consideration wider social context with a stress on users’ practices in relation to space.Methodology/approach: The article is based on the (self- observation of the public library use, and on the (discourse analysis of internal library documents (i.e. annual reports and plans and secondary sociological literature. As such, the cultural form approach represents a classic method of sociology of culture.Results: The study of relevant material in combination with direct personal experiences reveals socio-structural causes for the change of users’ needs and habits, and points at the difficulty of spatial redefinition of libraries as well as at the power of the discourse.Research limitations: The article is limited to an observation of users’ practices in some of the public libraries in Ljubljana and examines only a small number of annual reports – the discoveries are then further debated from the sociological perspective.Originality/practical implications: The article offers sociological insight in the current issues of the library science and tries to suggest a wider explanation that could answer some of the challenges of the contemporary librarianship.

  11. The NJOY nuclear data processing system Version 91

    International Nuclear Information System (INIS)

    MacFarlane, R.E.; Muir, D.W.

    1994-10-01

    The NJOY nuclear data processing system is a comprehensive computer code package for producing pointwise and multigroup cross sections and related quantities from elevated nuclear data in the ENDF format, including the latest US library, ENDF/B-VI. The NJOY code can work with neutrons, photons, and charged particles, and it can produce libraries for a wide variety of particle transport and reactor analysis codes

  12. Nuclear data newsletter. No. 2

    International Nuclear Information System (INIS)

    1980-10-01

    This issue contains bibliographic data describing: neutron nuclear data (CINDA); charged particle reaction data; experimental nuclear reaction data; evaluated nuclear data library (ENDF/B); libraries in formats other than ENDF/B, such as BOSPOR-78 (evaluated data for threshold neutron reactions), SOKRATOR (new USSR evaluations), INDL/A (IAEA nuclear data library for actinides). It includes the following descriptions of the recently received libraries: DAMSIG-77, DOCROS-77, JUELGAM-79, GSI gamma-ray library, KEDAK-3, PNESD, UKNDL-80, etc. A summary description and documentation of the multigroup cross section data libraries which exist at the IAEA is cited and new documents of interest are listed

  13. Nuclear data newsletter. No. 2

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-10-01

    This issue contains bibliographic data describing: neutron nuclear data (CINDA); charged particle reaction data; experimental nuclear reaction data; evaluated nuclear data library (ENDF/B); libraries in formats other than ENDF/B, such as BOSPOR-78 (evaluated data for threshold neutron reactions), SOKRATOR (new USSR evaluations), INDL/A (IAEA nuclear data library for actinides). It includes the following descriptions of the recently received libraries: DAMSIG-77, DOCROS-77, JUELGAM-79, GSI gamma-ray library, KEDAK-3, PNESD, UKNDL-80, etc. A summary description and documentation of the multigroup cross section data libraries which exist at the IAEA is cited and new documents of interest are listed.

  14. National Libraries Section. General Research Libraries Division. Papers.

    Science.gov (United States)

    International Federation of Library Associations, The Hague (Netherlands).

    Papers on national library services and activities, which were presented at the 1983 International Federation of Library Associations (IFLA) conference, include: (1) "The National Library of China in its Gradual Application of Modern Technology," a discussion by Zhu Nan and Zhu Yan (China) of microform usage and library automation; (2)…

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  16. 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.

  17. LIBRARY SKILL INSTRUCTION IN NIGERIAN ACADEMIC LIBRARIES

    African Journals Online (AJOL)

    DJFLEX

    www.globaljournalseries.com; Info@globaljournalseries.com. LIBRARY SKILL INSTRUCTION IN NIGERIAN ACADEMIC. LIBRARIES. P. C. AZIAGBA AND E. H. UZOEZI. (Received 10, September 2009; Revision Accepted 8, February 2010). ABSTRACT. This survey was undertaken to portray the level of library involvement ...

  18. Application of the Laplace transform method for the albedo boundary conditions in multigroup neutron diffusion eigenvalue problems in slab geometry

    International Nuclear Information System (INIS)

    Petersen, Claudio Zen; Vilhena, Marco T.; Barros, Ricardo C.

    2009-01-01

    In this paper the application of the Laplace transform method is described in order to determine the energy-dependent albedo matrix that is used in the boundary conditions multigroup neutron diffusion eigenvalue problems in slab geometry for nuclear reactor global calculations. In slab geometry, the diffusion albedo substitutes without approximation the baffle-reflector system around the active domain. Numerical results to typical test problems are shown to illustrate the accuracy and the efficiency of the Chebysheff acceleration scheme. (orig.)

  19. E-library Implementation in Library University of Riau

    Science.gov (United States)

    Yuhelmi; Rismayeti

    2017-12-01

    This research aims to see how the e-book implementation in Library University of Riau and the obstacle in its implementation. In the Globalization era, digital libraries should be developed or else it will decrease the readers’ interest, with the recent advanced technology, digital libraries are one of the learning tools that can be used to finding an information through the internet access, hence digital libraries or commonly known as E-Library is really helping the students and academic community in finding information. The methods that used in this research is Observation, Interview, and Literature Study. The respondents in this research are the staff who involved in the process of digitization in Library University of Riau. The result of this research shows that implementation of e-library in Library University of Riau is already filled the user needs for now, although there is obstacle faced just like technical problems for example the internet connection speed and the technical problem to convert the format from Microsoft Word .doc to Adobe.pdf

  20. LabVIEW Library to EPICS Channel Access

    CERN Document Server

    Liyu, Andrei; Thompson, Dave H

    2005-01-01

    The Spallation Neutron Source (SNS) accelerator systems will deliver a 1.0 GeV, 1.4 MW proton beam to a liquid mercury target for neutron scattering research. The accelerator complex consists of a 1 GeV linear accelerator, an accumulator ring and associated transport lines. The SNS diagnostics platform is PC-based and will run Windows for its OS and LabVIEW as its programming language. Data acquisition hardware will be based on PCI cards. There will be about 300 rack-mounted computers. The Channel Access (CA) protocol of the Experimental Physics and Industrial Control System (EPICS) is the SNS control system communication standard. This paper describes the approaches, implementation, and features of LabVIEW library to CA for Windows, Linux, and Mac OS X. We also discuss how the library implements the asynchronous CA monitor routine using LabVIEW's occurrence mechanism instead of a callback function (which is not available in LabVIEW). The library is used to acquire accelerator data and applications have been ...

  1. Marketing library and information services in academic libraries in ...

    African Journals Online (AJOL)

    Marketing library and information services in academic libraries in Niger State, Nigeria. ... This study was designed to investigate the marketing of library services in academic libraries in Niger state, ... EMAIL FULL TEXT EMAIL FULL TEXT

  2. Boekbesprekings : Book Reviews

    African Journals Online (AJOL)

    1974-05-18

    May 18, 1974 ... It is consequently a we come addItl(;lTI to the orthopaedic library. The ~rst sectIOn of.the book gives a very concise and lucid. synopsIS of.the pertInent mechanical facts. and the second part deals WIth the rele,:,ant biological aspects of the musculo- skeletal system..The t?ird and final section provides a brief.

  3. Macroscopic multigroup constants for accelerator driven system core calculation

    International Nuclear Information System (INIS)

    Heimlich, Adino; Santos, Rubens Souza dos

    2011-01-01

    The high-level wastes stored in facilities above ground or shallow repositories, in close connection with its nuclear power plant, can take almost 106 years before the radiotoxicity became of the order of the background. While the disposal issue is not urgent from a technical viewpoint, it is recognized that extended storage in the facilities is not acceptable since these ones cannot provide sufficient isolation in the long term and neither is it ethical to leave the waste problem to future generations. A technique to diminish this time is to transmute these long-lived elements into short-lived elements. The approach is to use an Accelerator Driven System (ADS), a sub-critical arrangement which uses a Spallation Neutron Source (SNS), after separation the minor actinides and the long-lived fission products (LLFP), to convert them to short-lived isotopes. As an advanced reactor fuel, still today, there is a few data around these type of core systems. In this paper we generate macroscopic multigroup constants for use in calculations of a typical ADS fuel, take into consideration, the ENDF/BVI data file. Four energy groups are chosen to collapse the data from ENDF/B-VI data file by PREPRO code. A typical MOX fuel cell is used to validate the methodology. The results are used to calculate one typical subcritical ADS core. (author)

  4. Personal Virtual Libraries

    Science.gov (United States)

    Pappas, Marjorie L.

    2004-01-01

    Virtual libraries are becoming more and more common. Most states have a virtual library. A growing number of public libraries have a virtual presence on the Web. Virtual libraries are a growing addition to school library media collections. The next logical step would be personal virtual libraries. A personal virtual library (PVL) is a collection…

  5. America's Star Libraries

    Science.gov (United States)

    Lyons, Ray; Lance, Keith Curry

    2009-01-01

    "Library Journal"'s new national rating of public libraries, the "LJ" Index of Public Library Service, identifies 256 "star" libraries. It rates 7,115 public libraries. The top libraries in each group get five, four, or three Michelin guide-like stars. All included libraries, stars or not, can use their scores to learn from their peers and improve…

  6. A multi-group and preemptable scheduling of cloud resource based on HTCondor

    Science.gov (United States)

    Jiang, Xiaowei; Zou, Jiaheng; Cheng, Yaodong; Shi, Jingyan

    2017-10-01

    and LHAASO. The result indicates that multi-group and preemptable resource scheduling is efficient to support multi-group and soft preemption. Additionally, the permission controlling component has been used in the local computing cluster, supporting for experiment JUNO, CMS and LHAASO, and the scale will be expanded to more experiments at the first half year, including DYW, BES and so on. Its evidence that the permission controlling is efficient.

  7. Obtaining incremental multigroup cross sections for CANDU super cells with reactivity devices

    International Nuclear Information System (INIS)

    Balaceanu, V.; Constantin, M.

    2001-01-01

    In the last 20 years a multigroup methodology WIMS - PIJXYZ (WP) was developed and validated at INR Pitesti for obtaining incremental cross sections for reactivity devices in CANDU reactors. This is an alternate methodology to the CANDU classic methodology (experimentally adjusted) based on the POWDERPUFS and MULTICELL computer codes. The 2D supercell calculation performed with the WIMS code, that is a NEA Data Bank transport code, and which produces multigroup cross sections (on 18 energy groups) for CANDU supercell material (standard and perturbed, with and without reactivity devices). To obtain an as correct as possible 3D modelling for the CANDU supercells containing reactivity devices, the WIMS cross sections are used as input data for the PIJXYZ code, thus obtaining homogenized cross sections for CANDU supercells. PIJXYZ is an integral transport code based on the formalism of the first collision probabilities. It is analogue to the SHETAN code and it was created for neutron analyzes at cell level for CANDU type reactors were the reactivity devices are perpendicular to the fuel channels. The coordinate system used in PIJXYZ is a mixed one, namely a rectangular-cylindrical system. The geometric model used in PIJXYZ is presented. The fuel beam is represented by a horizontal cylinder and the reactivity device by a vertical one both cylinders being immersed in the moderator. Two supercell types were considered: a perturbed supercell (containing a reactivity device) and the standard supercell were the place of reactivity device is occupied by the moderator. The incremental cross sections for reactivity device are obtained as differences between the homogenized over supercell cross sections (with reactivity device) and homogenized over standards supercell (without device) cross sections. The PIJXYZ computation may be done on an energy cutting with 2 up to 18 groups. The validation of VIMS - PIJXYZ was done on the basis of several benchmark and by comparison with

  8. Solution for the multigroup neutron space kinetics equations by the modified Picard algorithm

    International Nuclear Information System (INIS)

    Tavares, Matheus G.; Petersen, Claudio Z.; Schramm, Marcelo; Zanette, Rodrigo

    2017-01-01

    In this work, we used a modified Picards method to solve the Multigroup Neutron Space Kinetics Equations (MNSKE) in Cartesian geometry. The method consists in assuming an initial guess for the neutron flux and using it to calculate a fictitious source term in the MNSKE. A new source term is calculated applying its solution, and so on, iteratively, until a stop criterion is satisfied. For the solution of the fast and thermal neutron fluxes equations, the Laplace Transform technique is used in time variable resulting in a rst order linear differential matrix equation, which are solved by classical methods in the literature. After each iteration, the scalar neutron flux and the delayed neutron precursors are reconstructed by polynomial interpolation. We obtain the fluxes and precursors through Numerical Inverse Laplace Transform using the Stehfest method. We present numerical simulations and comparisons with available results in literature. (author)

  9. Solution for the multigroup neutron space kinetics equations by the modified Picard algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Tavares, Matheus G.; Petersen, Claudio Z., E-mail: matheus.gulartetavares@gmail.com [Universidade Federal de Pelotas (UFPEL), Capao do Leao, RS (Brazil). Departamento de Matematica e Estatistica; Schramm, Marcelo, E-mail: schrammmarcelo@gmail.com [Universidade Federal de Pelotas (UFPEL), RS (Brazil). Centro de Engenharias; Zanette, Rodrigo, E-mail: rodrigozanette@hotmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Instituto de Matematica e Estatistica

    2017-07-01

    In this work, we used a modified Picards method to solve the Multigroup Neutron Space Kinetics Equations (MNSKE) in Cartesian geometry. The method consists in assuming an initial guess for the neutron flux and using it to calculate a fictitious source term in the MNSKE. A new source term is calculated applying its solution, and so on, iteratively, until a stop criterion is satisfied. For the solution of the fast and thermal neutron fluxes equations, the Laplace Transform technique is used in time variable resulting in a rst order linear differential matrix equation, which are solved by classical methods in the literature. After each iteration, the scalar neutron flux and the delayed neutron precursors are reconstructed by polynomial interpolation. We obtain the fluxes and precursors through Numerical Inverse Laplace Transform using the Stehfest method. We present numerical simulations and comparisons with available results in literature. (author)

  10. HEXAGA-III-120, -30. Three dimensional multi-group neutron diffusion programmes for a uniform triangular mesh with arbitrary group scattering

    International Nuclear Information System (INIS)

    Woznicki, Z.I.

    1983-07-01

    This report presents the HEXAGA-III-programme solving multi-group time-independent real and/or adjoint neutron diffusion equations for three-dimensional-triangular-z-geometry. The method of solution is based on the AGA two-sweep iterative method belonging to the family of factorization techniques. An arbitrary neutron scattering model is permitted. The report written for users provides the description of the programme input and output and the use of HEXAGA-III is illustrated by a sample reactor problem. (orig.) [de

  11. Combined analytical-numerical procedure to solve multigroup spherical harmonics equations in two-dimensional r-z geometry

    International Nuclear Information System (INIS)

    Matausek, M.V.; Milosevic, M.

    1986-01-01

    In the present paper a generalization is performed of a procedure to solve multigroup spherical harmonics equations, which has originally been proposed and developed for one-dimensional systems in cylindrical or spherical geometry, and later extended for a special case of a two-dimensional system in r-z geometry. The expressions are derived for the axial and the radial dependence of the group values of the neutron flux moments, in the P-3 approximation of the spherical harmonics method, in a cylindrically symmetrical system with an arbitrary number of material regions in both r- and z-directions. In the special case of an axially homogeneous system, these expressions reduce to the relations derived previously. (author)

  12. More library mashups exploring new ways to deliver library data

    CERN Document Server

    2015-01-01

    Nicole Engard follows up her ground-breaking 2009 book Library Mashups with a fresh collection of mashup projects that virtually any library can emulate, customize, and build upon. In More Library Mashups, Engard and 24 creative library professionals describe how they are mashing up free and inexpensive digital tools and techniques to improve library services and meet everyday (and unexpected) challenges. Examples from libraries of all types are designed to help even non-programmers share and add value to digital content, update and enhance library websites and collections, mashup catalog data, connect to the library's automation system, and use emerging tools like Serendip-o-matic, Umlaut, and Libki to engage users, staff, and the community.

  13. The Virtual Library in Action

    International Nuclear Information System (INIS)

    Addis, Louise

    1999-01-01

    The SLAC Library has for many years provided SPIRES-HEP, a 300,000 record bibliographic database, to the world particle physics community via the Internet as well as through clone sites in Europe and Japan. The 1991 introduction of the e-print archives at LANL coupled with the World-Wide-Web (WWW) from CERN suddenly made it possible to provide easy linkage between bibliographic database records and the actual full-text of papers. The SLAC Library has turned this possibility into reality by converting hundreds of TeX source documents each month into viewable postscript complete with figures. These (now more than 20,000) postscript files are linked to the HEP database, and the full-text is rendered universally visible via WWW. We discuss the project, the collaboration of physicists and librarians, what is easy, what is hard, and our vision for the future

  14. The Virtual Library in Action

    Energy Technology Data Exchange (ETDEWEB)

    Addis, Louise

    1999-07-02

    The SLAC Library has for many years provided SPIRES-HEP, a 300,000 record bibliographic database, to the world particle physics community via the Internet as well as through clone sites in Europe and Japan. The 1991 introduction of the e-print archives at LANL coupled with the World-Wide-Web (WWW) from CERN suddenly made it possible to provide easy linkage between bibliographic database records and the actual full-text of papers. The SLAC Library has turned this possibility into reality by converting hundreds of TeX source documents each month into viewable postscript complete with figures. These (now more than 20,000) postscript files are linked to the HEP database, and the full-text is rendered universally visible via WWW. We discuss the project, the collaboration of physicists and librarians, what is easy, what is hard, and our vision for the future.

  15. Library Science Education: A New Role for Academic Libraries

    Science.gov (United States)

    Wesley, Threasa L.

    2018-01-01

    Many individuals working in library and information organizations do not hold a master of library science (MLS) degree or other specialized library science credential. Recognizing that this professional gap could be addressed by diversified educational opportunities, the W. Frank Steely Library at Northern Kentucky University in Highland Heights…

  16. VENTURE: a code block for solving multigroup neutronics problems applying the finite-difference diffusion-theory approximation to neutron transport

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.

    1975-10-01

    The computer code block VENTURE, designed to solve multigroup neutronics problems with application of the finite-difference diffusion-theory approximation to neutron transport (or alternatively simple P 1 ) in up to three-dimensional geometry is described. A variety of types of problems may be solved: the usual eigenvalue problem, a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations, or an indirect criticality search on nuclide concentrations, or on dimensions. First-order perturbation analysis capability is available at the macroscopic cross section level

  17. Analysis of benchmark lattices with endf/b-vi, jef-2.2 and jendl-3 data

    International Nuclear Information System (INIS)

    Saglam, M.

    1995-01-01

    The NJOY Nuclear Data Processing System has been used to process the ENDF/B-VI , JEF-2.2 and JENDL-3 Nuclear Cross Section Data Bases into multigroup form. A brief description of the data bases is given and the assumptions made in processing the data from evaluated nuclear data file format to multigroup format are presented. The differences and similarities of the Evaluated Nuclear Data Files have been investigated by producing four group cross sections by using the GROUPIE code and calculating thermal, fission spectrum averaged and 2200 m/s cross sections and resonance integrals using the INTER cale. It has been shown that the evaluated data for U238 in JEF and ENDF/B-VI are principally the same while in case of U235 the same is true for JENDL and ENDF/B-VI. The evaluations for U233 and Th232 are different for all three ENDF files. Several utility codes have been written to convert the multigroup library into a WIMS-D4 compatible binary library. The performance and suitability of the generated libraries have been tested with the use of metal tueled TRX lattices, uranium oxide fueled BAPL lattices and Th232-U233 fueled BNL lattices. The use ot a new thermal scattering matrix for Hydrogen from ENDF/B-VI increased keff for 0.5 o/ while the use of ENDF/B-VI U238 decreased it for 2.5 %. Although the original WIMS library performed well for Ihe effective multiplication factor of the lattices there is an improvement for the epithermal to thermal capture rate of U238 while using new data in the TRX and BAPL lattices. The effect of the fission spectrum is investigated for the BNL lattices and it is shown that using U233 fission spectrum instead of the original U235 spectrum gives a keff which agrees better with the experimental value. The results obtained by using new multigroup data are generally acceptable and in the experimental error range. They especially improve the prediction of the reaction rate dependent benchmark parameters

  18. General Maister's Library in the University of Maribor Library: Gaining and Preservation of the General Maister's Private Library in the University of Maribor Library

    Directory of Open Access Journals (Sweden)

    Vlasta Stavbar

    2015-12-01

    Full Text Available Rudolf Maister’s book legacy, which is kept in the University of Maribor Library, is the complete legacy of the Maister’s private book collection– the “maistirana”. The Maister’s Library is in the University of Maribor Library possession since 1998, when General Maister’s heirs and the Library signed the deed of donation. 5.945 units of library materials present a special collection, which is divided from the regular library holdings and is kept in a specially designed room. General Maister’s Library is an exception in the University of Maribor Library, since the general practice of libraries is not an exclusive maintenance of the provenience principle and of the original organisation. Because the Maister’s Library consists of materials that are important when researching the Slovene literary and cultural past, it only seems reasonably to strive for the authentic organisation and preserving the library as one integral whole.In order to understand the meaning and the importance of this special library collection for the Slovene cultural heritage, we have to highlight Rudolf Maister as a passionate book lover and as the owner of one of the greatest and the most beautiful private libraries in Slovenia. General Maister started collecting books and organising his collection back in 1912 in Ljubljana; the collection was rearranged thoroughly during the time he was in Maribor; it survived the moves during World War II and after it and it finally found its place in the General Maister’s Library under the patronage of the University of Maribor Library. The rearranging of the shelving system and sorting of the materials patterned after the authentic shelving system in the General’s Maribor apartment took place in 2013. Until then the books were not arranged according to the authentic shelving system. The estimated shelving system reconstruction has reconstructed Maister’s shelving system as authentically as possible, for it is not

  19. Library usage patterns in the electronic information environment. Electronic journals, Use studies, Libraries, Medical libraries

    Directory of Open Access Journals (Sweden)

    B. Franklin

    2004-01-01

    Full Text Available This paper examines the methodology and results from Web-based surveys of more than 15,000 networked electronic services users in the United States between July 1998 and June 2003 at four academic health sciences libraries and two large main campus libraries serving a variety of disciplines. A statistically valid methodology for administering simultaneous Web-based and print-based surveys using the random moments sampling technique is discussed and implemented. Results from the Web-based surveys showed that at the four academic health sciences libraries, there were approximately four remote networked electronic services users for each in-house user. This ratio was even higher for faculty, staff, and research fellows at the academic health sciences libraries, where more than five remote users for each in-house user were recorded. At the two main libraries, there were approximately 1.3 remote users for each in-house user of electronic information. Sponsored research (grant funded research accounted for approximately 32% of the networked electronic services activity at the health sciences libraries and 16% at the main campus libraries. Sponsored researchers at the health sciences libraries appeared to use networked electronic services most intensively from on-campus, but not from in the library. The purpose of use for networked electronic resources by patrons within the library is different from the purpose of use of those resources by patrons using the resources remotely. The implications of these results on how librarians reach decisions about networked electronic resources and services are discussed.

  20. Phylogeny of holoparasitic Orobanche (Orobanchaceae) inferred from nuclear ITS sequences.

    Science.gov (United States)

    Schneeweiss, Gerald M; Colwell, Alison; Park, Jeong-Mi; Jang, Chang-Gee; Stuessy, Tod F

    2004-02-01

    Orobanche is the largest genus among the holoparasitic members of Orobanchaceae. We present the first molecular phylogenetic analysis (using nuclear ITS sequences) that includes members of all sections of Orobanche, Gymnocaulis, Myzorrhiza, Trionychon, and Orobanche. Orobanche is not monophyletic, but falls into two lineages: (1) the Orobanche group comprises Orobanche sect. Orobanche and the small Near Asian genus Diphelypaea and is characterized by a chromosome base number of x=19 and (2) the Phelipanche group contains Orobanche sects. Gymnocaulis, Myzorrhiza, and Trionychon and possesses a chromosome base number of x=12. The relationships between these two groups and to other genera such as Boschniakia or Cistanche remain unresolved. Within the Orobanche group, Orobanche macrolepis and Orobanche anatolica (including Orobanche colorata) constitute two phylogenetically distinct lineages. Intrasectional structurings proposed by some authors for O. sect. Orobanche are not confirmed by the molecular data. In most cases, intraspecific sequence divergence between accessions, if present, is negligible and not correlated with morphological or ecological traits. In a few cases, however, there is evidence for the presence of cryptic taxa.

  1. Phylogeny of holoparasitic Orobanche (Orobanchaceae) inferred from nuclear ITS sequences

    Science.gov (United States)

    Schneeweiss, G.M.; Colwell, A.; Park, J.-M.; Jang, C.-G.; Stuessy, Tod F.

    2004-01-01

    Orobanche is the largest genus among the holoparasitic members of Orobanchaceae. We present the first molecular phylogenetic analysis (using nuclear ITS sequences) that includes members of all sections of Orobanche, Gymnocaulis, Myzorrhiza, Trionychon, and Orobanche. Orobanche is not monophyletic, but falls into two lineages: (1) the Orobanche group comprises Orobanche sect. Orobanche and the small Near Asian genus Diphelypaea and is characterized by a chromosome base number of x = 19 and (2) the Phelipanche group contains Orobanche sects. Gymnocaulis, Myzorrhiza, and Trionychon and possesses a chromosome base number of x = 12. The relationships between these two groups and to other genera such as Boschniakia or Cistanche remain unresolved. Within the Orobanche group, Orobanche macrolepis and Orobanche anatolica (including Orobanche colorata) constitute two phylogenetically distinct lineages. Intrasectional structurings proposed by some authors for O. sect. Orobanche are not confirmed by the molecular data. In most cases, intraspecific sequence divergence between accessions, if present, is negligible and not correlated with morphological or ecological traits. In a few cases, however, there is evidence for the presence of cryptic taxa. ?? 2003 Elsevier Science (USA). All rights reserved.

  2. Library Subject Guides: A Case Study of Evidence-Informed Library Development

    Science.gov (United States)

    Wakeham, Maurice; Roberts, Angharad; Shelley, Jane; Wells, Paul

    2012-01-01

    This paper describes the process whereby a university library investigated the value of its subject guides to its users. A literature review and surveys of library staff, library users and other libraries were carried out. Existing library subject guides and those of other higher education libraries were evaluated. The project team reported…

  3. Library subject guides: A case study of evidence-informed library development

    OpenAIRE

    Wakeham, Maurice; Roberts, Angharad; Shelley, Jane; Wells, Paul

    2012-01-01

    This paper describes the process whereby a university library investigated the value of its subject guides to its users. A literature review and surveys of library staff, library users and other libraries were carried out. Existing library subject guides and those of other higher education libraries were evaluated. The project team reported regularly to the wider library team throughout the year long investigation and made recommendations. The project served to develop the professional skills...

  4. ERRORJ. Covariance processing code system for JENDL. Version 2

    International Nuclear Information System (INIS)

    Chiba, Gou

    2003-09-01

    ERRORJ is the covariance processing code system for Japanese Evaluated Nuclear Data Library (JENDL) that can produce group-averaged covariance data to apply it to the uncertainty analysis of nuclear characteristics. ERRORJ can treat the covariance data for cross sections including resonance parameters as well as angular distributions and energy distributions of secondary neutrons which could not be dealt with by former covariance processing codes. In addition, ERRORJ can treat various forms of multi-group cross section and produce multi-group covariance file with various formats. This document describes an outline of ERRORJ and how to use it. (author)

  5. ZZ MATXSLIBJ33, JENDL-3.3 based, 175 N-42 photon groups (VITAMIN-J) MATXS library for discrete ordinates multi-group

    International Nuclear Information System (INIS)

    Kosako, K.; Yamano, N.; Fukahori, T.; Shibata, K.; Hasegawa, A.

    2006-01-01

    1 - Description of program or function: JENDL-3.3 based, 175 neutron-42 photon groups (VITAMIN-J) MATXS library for discrete ordinates multi-group transport codes. Format: MATXS. Number of groups: 175 neutron, 42 gamma-ray. Nuclides: 337 nuclides contained in JENDL-3.3: 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, S-32, S-33, S-34, S-36, Cl-35, Cl-37, Ar-40, 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, Ge-70, Ge-72, Ge-73, Ge-74, Ge-76, As-75, Se-74, Se-76, Se-77, Se-78, Se-79, 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-86, Sr-87, Sr-88, Sr-89, Sr-90, Y-89, Y-91, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Nb-93, Nb-94, Nb-95, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Tc-99, Ru-96, Ru-98, Ru-99, Ru-100, Ru-101, Ru-102, Ru-103, Ru-104, 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-110m, Cd-106, Cd-108, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-116, In-113, In-115, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-126, Sb-121, Sb-123, Sb-124, Sb-125, Te-120, Te-122, Te-123, Te-124, Te-125, Te-126, Te-127m, Te-128, Te-129m, Te-130, I-127, I-129, 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-136, Cs-137, Ba-130, Ba-132, Ba-134, Ba-135, Ba-136, Ba-137, Ba-138, Ba-140, La-138, La-139, Ce-140, Ce-141, Ce-142, Ce-144, Pr-141, Pr-143, Nd-142, Nd-143, Nd-144, Nd-145, Nd-146, Nd-147, Nd-148, Nd-150, Pm-147, Pm-148, Pm-148m, Pm-149, Sm-144, Sm-147, Sm-148, Sm-149, Sm-150, Sm-151, Sm-152, Sm-153, Sm-154, Eu-151, Eu-152, Eu-153, Eu-154, Eu-155, Eu

  6. A consistent multigroup model for radiative transfer and its underlying mean opacities

    International Nuclear Information System (INIS)

    Turpault, Rodolphe

    2005-01-01

    In some regimes, such as in plasma physics or in super orbital atmospheric entry of space objects, the effects of radiation are crucial and can tremendously modify the hydrodynamics of the gas. In such cases, it is therefore important to have a good prediction of the radiative variables. However, full transport solutions of these multi-dimensional, time-dependent problems are too expensive to get to be involved in a coupled configuration. It is hence necessary to develop other models for radiation that are cheap, yet accurate enough to give good predictions of the radiative effects. We will herein introduce the multigroup-M1 model and look at its characteristics and in particular try to separate the angular error from the frequential one since these two approximation play very different roles. The angular behaviour of the model will be tested on a case proposed by Su and Olson and used by Olson et al. to compare various moments and (flux-limited) diffusion models. For the frequency behaviour, we use a simplified flame test-case and show the importance of taking good mean opacities

  7. The history and library statistics of JAEA library activities

    Energy Technology Data Exchange (ETDEWEB)

    Itabashi, Keizo [Japan Atomic Energy Agency, Intellectual Resources Dept., Tokai, Ibaraki (Japan)

    2012-03-15

    The history and library statistics of the Japan Atomic Energy Agency library activity were summarized. Former Japan Atomic Energy Research Institute and the former Japan Nuclear Cycle Development Institute merged in October, 2005, and Japan Atomic Energy Agency is established. Properly speaking, the library statistics of old two corporations should have been summarized, but statistics of the Japan Nuclear Cycle Development Institute is not yet obtained. Then, although it is stated as the Japan Atomic Energy Agency library, it limits to the description about the old Japan Atomic Energy Research Institute library before 2004. (author)

  8. The history and library statistics of JAEA library activities

    International Nuclear Information System (INIS)

    Itabashi, Keizo

    2012-03-01

    The history and library statistics of the Japan Atomic Energy Agency library activity were summarized. Former Japan Atomic Energy Research Institute and the former Japan Nuclear Cycle Development Institute merged in October, 2005, and Japan Atomic Energy Agency is established. Properly speaking, the library statistics of old two corporations should have been summarized, but statistics of the Japan Nuclear Cycle Development Institute is not yet obtained. Then, although it is stated as the Japan Atomic Energy Agency library, it limits to the description about the old Japan Atomic Energy Research Institute library before 2004. (author)

  9. The Infrared Telescope Facility (IRTF) Spectral Library: Cool Stars

    Science.gov (United States)

    Rayner, John T.; Cushing, Michael C.; Vacca, William D.

    2009-12-01

    We present a 0.8-5 μm spectral library of 210 cool stars observed at a resolving power of R ≡ λ/Δλ ~ 2000 with the medium-resolution infrared spectrograph, SpeX, at the 3.0 m NASA Infrared Telescope Facility (IRTF) on Mauna Kea, Hawaii. The stars have well-established MK spectral classifications and are mostly restricted to near-solar metallicities. The sample not only contains the F, G, K, and M spectral types with luminosity classes between I and V, but also includes some AGB, carbon, and S stars. In contrast to some other spectral libraries, the continuum shape of the spectra is measured and preserved in the data reduction process. The spectra are absolutely flux calibrated using the Two Micron All Sky Survey photometry. Potential uses of the library include studying the physics of cool stars, classifying and studying embedded young clusters and optically obscured regions of the Galaxy, evolutionary population synthesis to study unresolved stellar populations in optically obscured regions of galaxies and synthetic photometry. The library is available in digital form from the IRTF Web site.

  10. Continuous-energy version of KENO V.a for criticality safety applications

    International Nuclear Information System (INIS)

    Dunn, Michael E.; Greene, N. Maurice; Petrie, Lester M.

    2003-01-01

    KENO V.a is a multigroup Monte Carlo code that solves the Boltzmann transport equation and is used extensively in the criticality safety community to calculate the effective multiplication factor of systems with fissionable material. In this work, a continuous-energy or pointwise version of KENO V.a has been developed by first designing a new continuous-energy cross-section format and then by developing the appropriate Monte Carlo transport procedures to sample the new cross-section format. In order to generate pointwise cross sections for a test library, a series of cross-section processing modules were developed and used to process 50 ENDF/B-6 Release 7 nuclides for the test library. Once the cross-section processing procedures were in place, a continuous-energy version of KENO V.a was developed and tested by calculating 21 critical benchmark experiments. The point KENO-calculated results for the 21 benchmarks are in agreement with calculated results obtained with the multigroup version of KENO V.a using the 238-group ENDF/B-5 and 199-group ENDF/B-6 Release 3 libraries. Based on the calculated results with the prototypic cross-section library, a continuous-energy version of the KENO V.a code has been successfully developed and demonstrated for modeling systems with fissionable material. (author)

  11. Wikis in Libraries

    Directory of Open Access Journals (Sweden)

    Matthew Bejune

    2007-09-01

    Full Text Available Wikis have recently been adopted to support a variety of collaborative activities within libraries. This article and its companion wiki, LibraryWikis (http://librarywikis.pbwiki.com/, seek to document the phenomenon of wikis in libraries. This subject is considered within the framework of computer-supported cooperative work (CSCW. The author identified thirty-three library wikis and developed a classification schema with four categories: (1 collaboration among libraries (45.7 percent; (2 collaboration among library staff (31.4 percent; (3 collaboration among library staff and patrons (14.3 percent; and (4 collaboration among patrons (8.6 percent. Examples of library wikis are presented within the article, as is a discussion for why wikis are primarily utilized within categories I and II and not within categories III and IV. It is clear that wikis have great utility within libraries, and the author urges further application of wikis in libraries.

  12. Generation of SCALE 6 Input Data File for Cross Section Library of PWR Spent Fuel

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Cho, Dong Keun

    2010-11-01

    In order to obtain the cross section libraries of the Korean Pressurized water reactor (PWR) spent fuel (SF), SCALE 6 code input files have been generated. The PWR fuel data were obtained from the nuclear design report (NDR) of the current operating PWRs. The input file were prepared for 16 fuel types such as 4 types of Westinghouse 14x14, 3 types of OPR-1000 16x16, 4 types of Westinghouse 16x16, and 6 types of Westinghouse 17x17. For each fuel type, 5 kinds of fuel enrichments have been considered such as 1.5, 2.0 ,3.0, 4.0 and 5.0 wt%. In the SCALE 6 calculation, a ENDF-V 44 group was used. The 25 burnup step until 72000 MWD/T was used. A 1/4 symmetry model was used for 16x16 and 17x17 fuel assembly, and 1/2 symmetry model was used for 14x14 fuel assembly The generated cross section libraries will be used for the source-term analysis of the PWR SF

  13. State Virtual Libraries

    Science.gov (United States)

    Pappas, Marjorie L.

    2003-01-01

    Virtual library? Electronic library? Digital library? Online information network? These all apply to the growing number of Web-based resource collections managed by consortiums of state library entities. Some, like "INFOhio" and "KYVL" ("Kentucky Virtual Library"), have been available for a few years, but others are just starting. Searching for…

  14. Research Library

    Science.gov (United States)

    Los Alamos National Laboratory Research Library Search Site submit Contact Us | Remote Access Standards Theses/Dissertations Research Help Subject Guides Library Training Video Tutorials Alerts Research Library: delivering essential knowledge services for national security sciences since 1947 Los

  15. Library Systems: FY 2013 Public Libraries Survey (Administrative Entity)

    Data.gov (United States)

    Institute of Museum and Library Services — Find key information on library systems around the United States.These data include imputed values for libraries that did not submit information in the FY 2013 data...

  16. Library Systems: FY 2012 Public Libraries Survey (Administrative Entity)

    Data.gov (United States)

    Institute of Museum and Library Services — Find key information on library systems around the United States.These data include imputed values for libraries that did not submit information in the FY 2012 data...

  17. Multigroup analysis of nuclear elastic scattering effects in Cat-D and DD3He fusion plasmas

    International Nuclear Information System (INIS)

    Nakano, Yasuyuki; Hanada, Takahiro; Hori, Hidetoshi; Kudo, Kazuhiko; Ohta, Masao

    1987-01-01

    Effects of nuclear elastic scattering (NES) on the slowing down of charged fusion products in a typical deuterium plasma and the burn dynamics of ignited Cat-D and DD 3 He plasmas are investigated. A time-dependent multigroup method is used to take into account the effect of finite (non-zero) slowing-down time as well as the discrete nature of NES. It is shown that adequate treatment of the slowing-down process, especially consideration of NES and slowing-down time delay, is essential for an accurate prediction of the dynamic behavior and thermal instability of the plasmas. NES accelerates the temporal plasma behavior and enhances the thermal instability, leading to 20∼30 keV increase in the critical temperature. (author)

  18. Library Computing

    Science.gov (United States)

    Library Computing, 1985

    1985-01-01

    Special supplement to "Library Journal" and "School Library Journal" covers topics of interest to school, public, academic, and special libraries planning for automation: microcomputer use, readings in automation, online searching, databases of microcomputer software, public access to microcomputers, circulation, creating a…

  19. Dosimetric comparison of stopping power calibration with dual-energy CT and single-energy CT in proton therapy treatment planning

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, Jiahua [Department of Physics, University of Adelaide, Adelaide, SA 5005 (Australia); Penfold, Scott N., E-mail: scott.penfold@adelaide.edu.au [Department of Physics, University of Adelaide, Adelaide, SA 5005, Australia and Department of Medical Physics, Royal Adelaide Hospital, Adelaide, SA 5000 (Australia)

    2016-06-15

    Purpose: The accuracy of proton dose calculation is dependent on the ability to correctly characterize patient tissues with medical imaging. The most common method is to correlate computed tomography (CT) numbers obtained via single-energy CT (SECT) with proton stopping power ratio (SPR). CT numbers, however, cannot discriminate between a change in mass density and change in chemical composition of patient tissues. This limitation can have consequences on SPR calibration accuracy. Dual-energy CT (DECT) is receiving increasing interest as an alternative imaging modality for proton therapy treatment planning due to its ability to discriminate between changes in patient density and chemical composition. In the current work we use a phantom of known composition to demonstrate the dosimetric advantages of proton therapy treatment planning with DECT over SECT. Methods: A phantom of known composition was scanned with a clinical SECT radiotherapy CT-simulator. The phantom was rescanned at a lower X-ray tube potential to generate a complimentary DECT image set. A set of reference materials similar in composition to the phantom was used to perform a stoichiometric calibration of SECT CT number to proton SPRs. The same set of reference materials was used to perform a DECT stoichiometric calibration based on effective atomic number. The known composition of the phantom was used to assess the accuracy of SPR calibration with SECT and DECT. Intensity modulated proton therapy (IMPT) treatment plans were generated with the SECT and DECT image sets to assess the dosimetric effect of the imaging modality. Isodose difference maps and root mean square (RMS) error calculations were used to assess dose calculation accuracy. Results: SPR calculation accuracy was found to be superior, on average, with DECT relative to SECT. Maximum errors of 12.8% and 2.2% were found for SECT and DECT, respectively. Qualitative examination of dose difference maps clearly showed the dosimetric advantages

  20. A punched-card library of neutron cross-sections and its use in the mechanized preparation of group cross-sections for use in Monte Carlo, Carlson S{sub n} and other multi-group neutronics calculations on high-speed computers

    Energy Technology Data Exchange (ETDEWEB)

    Parker, K [Atomic Weapons Research Establishment, Aldermaston (United Kingdom)

    1962-03-15

    The AWRE punched-card library of neutron cross-sections is described together with associated IBM-7090 programmes which process this data to give group-averaged cross-sections for use in Monte Carlo, Carlson S{sub n} and other multi-group neutronics calculations. The methods developed to deal with both isotropic and anisotropic elastic scattering are described. These include the multi-group transport approximation and the full treatment of anisotropic scattering using the Legendre polynomial moments of the scattering transfer matrix. The principles of group-constant formation are considered and illustrated by describing systems of group constants suitable for fast-reactor calculations. Practical problems such as the empirical adjustment of group constants to reproduce integral results and the collapsing of a many-group set of constants to give a few-group set are discussed. (author) [French] L'auteur decrit le fichier de cartes perforees sur lesquelles on enregistre a l'Atomic Weapons Research Establishment (AWRE) les sections efficaces neutroniques ainsi que les programmes IBM-7090 associes qui sont employes pour le traitement de ces informations, en vue d'obtenir des sections efficaces moyennes par groupe pouvant servir aux calculs de neutroniques a plusieurs groupes, effectues a l'aide des methodes de Monte-Carlo, S{sub n} de Carlson et autres methodes. L'auteur expose ensuite les methodes mises au point roda etudier la diffusion elastique, tant isotrope qu'anisotrope. Elles comprennent l'approximation de transport a plusieurs groupes, ainsi que le traitement complet de la diffusion anisotrope par les moments polynomiaux de Legendre de la matrice de transfert de la diffusion. L'auteur examine les principes de la formation des constantes de groupes; a titre d'illustration, il decrit les systemes de constantes de groupes qui se pretent aux calculs de reacteurs a neutrons rapides. Il expose quelques problemes pratiques, tels que l'ajustement empirique des

  1. Libraries on the MOVE.

    Science.gov (United States)

    Edgar, Jim; And Others

    1986-01-01

    Presents papers from Illinois State Library and Shawnee Library System's "Libraries on the MOVE" conference focusing on how libraries can impact economic/cultural climate of an area. Topics addressed included information services of rural libraries; marketing; rural library development; library law; information access; interagency…

  2. Comparison of Iranian National Medical Library with digital libraries of selected countries.

    Science.gov (United States)

    Zare-Farashbandi, Firoozeh; Najafi, Nayere Sadat Soleimanzade; Atashpour, Bahare

    2014-01-01

    The important role of information and communication technologies and their influence on methods of storing, retrieving information in digital libraries, has not only changed the meanings behind classic library activates but has also created great changes in their services. However, it seems that not all digital libraries provide their users with similar services and only some of them are successful in fulfilling their role in digital environment. The Iranian National Medical library is among those that appear to come short compared to other digital libraries around the world. By knowing the different services provided by digital libraries worldwide, one can evaluate the services provided by Iranian National Medical library. The goal of this study is a comparison between Iranian National Medical library and digital libraries of selected countries. This is an applied study and uses descriptive - survey method. The statistical population is the digital libraries around the world which were actively providing library services between October and December 2011 and were selected by using the key word "Digital Library" in Google search engine. The data-gathering tool was direct access to the websites of these digital libraries. The statistical study is descriptive and Excel software was used for data analysis and plotting of the charts. The findings showed that among the 33 digital libraries investigated worldwide, most of them provided Browse (87.87%), Search (84.84%), and Electronic information retrieval (57.57%) services. The "Help" in public services (48/48%) and "Interlibrary Loan" in traditional services (27/27%) had the highest frequency. The Iranian National Medical library provides more digital services compared to other libraries but has less classic and public services and has less than half of possible public services. Other than Iranian National Medical library, among the 33 libraries investigated, the leaders in providing different services are Library of

  3. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    Energy Technology Data Exchange (ETDEWEB)

    Díez, C.J., E-mail: cj.diez@upm.es [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Cabellos, O. [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Instituto de Fusión Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Martínez, J.S. [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain)

    2015-01-15

    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.

  4. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    International Nuclear Information System (INIS)

    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

  5. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    Science.gov (United States)

    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.

  6. DIFFUSION - WRS system module number 7539 for solving a set of multigroup diffusion equations in one dimension

    International Nuclear Information System (INIS)

    Grimstone, M.J.

    1978-06-01

    The WRS Modular Programming System has been developed as a means by which programmes may be more efficiently constructed, maintained and modified. In this system a module is a self-contained unit typically composed of one or more Fortran routines, and a programme is constructed from a number of such modules. This report describes one WRS module, the function of which is to solve a set of multigroup diffusion equations for a system represented in one-dimensional plane, cylindrical or spherical geometry. The information given in this manual is of use both to the programmer wishing to incorporate the module in a programme, and to the user of such a programme. (author)

  7. Adaptive solution of the multigroup diffusion equation on irregular structured grids using a conforming finite element method formulation

    International Nuclear Information System (INIS)

    Ragusa, J. C.

    2004-01-01

    In this paper, a method for performing spatially adaptive computations in the framework of multigroup diffusion on 2-D and 3-D Cartesian grids is investigated. The numerical error, intrinsic to any computer simulation of physical phenomena, is monitored through an a posteriori error estimator. In a posteriori analysis, the computed solution itself is used to assess the accuracy. By efficiently estimating the spatial error, the entire computational process is controlled through successively adapted grids. Our analysis is based on a finite element solution of the diffusion equation. Bilinear test functions are used. The derived a posteriori error estimator is therefore based on the Hessian of the numerical solution. (authors)

  8. Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

    International Nuclear Information System (INIS)

    Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

    1986-02-01

    Multigroup cross sections (66 neutron groups and 22 photon groups) are described for neutron energies from thermal to 400 MeV. The elements considered are hydrogen, 10 B, 11 B, carbon, nitrogen, oxygen, sodium, magnesium, aluminum, silicon, sulfur, potassium, calcium, chromium, iron, nickel, tungsten, and lead. The cross section data presented are a revision of similar data presented previously. In the case of iron, transport calculations using the earlier and the revised cross sections are presented and compared, and significant differences are found. The revised cross sections are available from the Radiation Shielding information Center of the Oak Ridge National Laboratory. 32 refs., 5 figs., 3 tabs

  9. The NIST Step Class Library (Step Into the Future)

    Science.gov (United States)

    1990-09-01

    Figure 6. Excerpt from a STEP exclange file based on the Geometry model 1be NIST STEP Class Libary Page 13 An issue of concern in this...Scheifler, R., Gettys, J., and Newman, P., X Window System: C Library and Protocol Reference. Digital Press, Bedford, Mass, 1988. [Schenck90] Schenck, D

  10. The USF Libraries Virtual Library Project: A Blueprint for Development.

    Science.gov (United States)

    Metz-Wiseman, Monica; Silver, Susan; Hanson, Ardis; Johnston, Judy; Grohs, Kim; Neville, Tina; Sanchez, Ed; Gray, Carolyn

    This report of the Virtual Library Planning Committee (VLPC) is intending to serve as a blueprint for the University of South Florida (USF) Libraries as it shifts from print to digital formats in its evolution into a "Virtual Library". A comprehensive planning process is essential for the USF Libraries to make optimum use of technology,…

  11. HEXAGA-II-120, -60, -30 two-dimensional multi-group neutron diffusion programmes for a uniform triangular mesh with arbitrary group scattering

    International Nuclear Information System (INIS)

    Woznicki, Z.

    1979-06-01

    This report presents the AGA two-sweep iterative methods belonging to the family of factorization techniques in their practical application in the HEXAGA-II two-dimensional programme to obtain the numerical solution to the multi-group, time-independent, (real and/or adjoint) neutron diffusion equations for a fine uniform triangular mesh. An arbitrary group scattering model is permitted. The report written for the users provides the description of input and output. The use of HEXAGA-II is illustrated by two sample reactor problems. (orig.) [de

  12. Our Stories Transforming Our Libraries: The York County Library System

    Directory of Open Access Journals (Sweden)

    Mina Edmondson

    2016-11-01

    Full Text Available These narratives chronicle the authors’ journeys to collaborate and discover the transformative impact that stories have on library culture and library staff. This study describes a research collaboration between York County Libraries and Penn State York. In Phase I, we collected stories from library staff as the library system was being challenged to reimage public libraries for the future. The major themes and types of organizational stories identified in the initial narrative project were presented during a county-wide all-staff in-service training. The library District Consultant (first author and the Penn State professor (second author then facilitated a workshop designed to lead staff in their exploration of these topics and generate a written record of their storytelling/discussions. This data became the basis for Phase II of the project and allowed the system to strategically assess its evolving culture and identity.

  13. Burnup simulations of an inert matrix fuel using a two region, multigroup reactor physics model

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, E. [Dept. of Mechanical Engineering, Univ. of Texas at Austin, 1 Univ. Place C2200, Austin, TX 78712 (United States); Deinert, M.; Bingham Cady, K. [Dept. of Theoretical and Applied Mechanics, Cornell Univ., Ithaca, NY 14853 (United States)

    2006-07-01

    Determining the time dependent concentration of isotopes in a nuclear reactor core is of fundamental importance to analysis of nuclear fuel cycles and the impact of spent fuels on long term storage facilities. We present a fast, conceptually simple tool for performing burnup calculations applicable to obtaining isotopic balances as a function of fuel burnup. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to determine the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. The model has been tested against benchmarked results for LWRs burning UOX and MOX, as well as MONTEBURNS simulations of zirconium oxide based IMF, all with strong fidelity. As an illustrative example, VBUDS burnup calculation results for an IMF fuel are presented in this paper. (authors)

  14. Digital library initiative in an Indian research library: an experience report

    International Nuclear Information System (INIS)

    Venkadesan, S.; Narayanan, A.

    1999-01-01

    Libraries all over the world are transiting from the traditional print only era to the electronic era. The migration to an electronic library involves the understanding of several issues and requirements at the planning stage itself. This paper describes the various aspects of building a digital library system. Experience of implementing certain strategies in the IGCAR library is discussed in detail. The components of the digital library and the various access levels are discussed with examples. Finally, samples of the IGCAR library web site and the Newsletter web site are shown. (author)

  15. News from the Library: The 'long tail' Library

    CERN Multimedia

    CERN Library

    2012-01-01

    "The term 'long tail' has gained popularity in recent times as describing the retailing strategy of selling a large number of unique items with relatively small quantities sold of each usually in addition to selling fewer popular items in large quantities. The long tail was popularized by Chris Anderson, who mentioned Amazon.com, Apple and Yahoo! as examples of businesses applying this strategy." *   If we leave the business environment and move to the world of libraries, we still see this "long tail". Usually, only a small portion of a library's book collection accounts for the majority of its loans. On the other hand, there are a variety of "niche information needs" that might not be met, as libraries cannot afford to build up huge collections of documents available just-in-case. However, the networked environment of today's libraries can offer a solution. Online networks of libraries ca...

  16. Students lead the library the importance of student contributions to the academic library

    CERN Document Server

    Arnold-Garza, Sara

    2017-01-01

    In six parts-Students as Employees, Students as Curators, Students as Ambassadors, the Library as Client, Student Groups as Library Leaders, and Students as Library Designers-Students Lead the Library provides case studies of programs and initiatives that seek student input, assistance, and leadership in the academic library. Through the library, students can develop leadership skills, cultivate high levels of engagement, and offer peer learning opportunities. Through the students, libraries can create participatory design processes, enhancement and transformation of the library's core functions, and expressed library value for stakeholders.

  17. Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Clarno, Kevin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gentry, Cole [Univ. of Tennessee, Knoxville, TN (United States); Wiarda, Dorothea [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Williams, Mark L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kochunas, Brendan [Univ. of Michigan, Ann Arbor, MI (United States); Liu, Yuxuan [Univ. of Michigan, Ann Arbor, MI (United States); Palmtag, Scott [Core Physics, Inc., Wilmington, NC (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-03-01

    The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include (1) a subgroup method for resonance selfshielding and (2) a whole-core transport solver with a 2-D/1-D synthesis method. The MPACT code requires a cross section library to support all the MPACT core simulation capabilities which would be the most influencing component for simulation accuracy.

  18. THE INFRARED TELESCOPE FACILITY (IRTF) SPECTRAL LIBRARY: COOL STARS

    International Nuclear Information System (INIS)

    Rayner, John T.; Cushing, Michael C.; Vacca, William D.

    2009-01-01

    We present a 0.8-5 μm spectral library of 210 cool stars observed at a resolving power of R ≡ λ/Δλ ∼ 2000 with the medium-resolution infrared spectrograph, SpeX, at the 3.0 m NASA Infrared Telescope Facility (IRTF) on Mauna Kea, Hawaii. The stars have well-established MK spectral classifications and are mostly restricted to near-solar metallicities. The sample not only contains the F, G, K, and M spectral types with luminosity classes between I and V, but also includes some AGB, carbon, and S stars. In contrast to some other spectral libraries, the continuum shape of the spectra is measured and preserved in the data reduction process. The spectra are absolutely flux calibrated using the Two Micron All Sky Survey photometry. Potential uses of the library include studying the physics of cool stars, classifying and studying embedded young clusters and optically obscured regions of the Galaxy, evolutionary population synthesis to study unresolved stellar populations in optically obscured regions of galaxies and synthetic photometry. The library is available in digital form from the IRTF Web site.

  19. STATLIB, Interactive Statistics Program Library of Tutorial System

    International Nuclear Information System (INIS)

    Anderson, H.E.

    1986-01-01

    1 - Description of program or function: STATLIB is a conversational statistical program library developed in conjunction with a Sandia National Laboratories applied statistics course intended for practicing engineers and scientists. STATLIB is a group of 15 interactive, argument-free, statistical routines. Included are analysis of sensitivity tests; sample statistics for the normal, exponential, hypergeometric, Weibull, and extreme value distributions; three models of multiple regression analysis; x-y data plots; exact probabilities for RxC tables; n sets of m permuted integers in the range 1 to m; simple linear regression and correlation; K different random integers in the range m to n; and Fisher's exact test of independence for a 2 by 2 contingency table. Forty-five other subroutines in the library support the basic 15

  20. Libraries and Learning

    Science.gov (United States)

    Rainie, Lee

    2016-01-01

    The majority of Americans think local libraries serve the educational needs of their communities and families pretty well and library users often outpace others in learning activities. But many do not know about key education services libraries provide. This report provides statistics on library usage and presents key education services provided…