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Sample records for carlo cross-section library

  1. Graphs of the cross sections in the recommended Monte Carlo cross-section library at the Los Alamos Scientific Laboratory

    International Nuclear Information System (INIS)

    Soran, P.D.; Seamon, R.E.

    1980-05-01

    Graphs of all neutron cross sections and photon production cross sections on the Recommended Monte Carlo Cross Section (RMCCS) library have been plotted along with local neutron heating numbers. Values for anti ν, the average number of neutrons per fission, are also given

  2. Graphs of the cross sections in the Alternate Monte Carlo Cross Section library at the Los Alamos Scientific Laboratory

    International Nuclear Information System (INIS)

    Seamon, R.E.; Soran, P.D.

    1980-06-01

    Graphs of all neutron cross sections and photon production cross sections on the Alternate Monte Carlo Cross Section (AMCCS) library have been plotted along with local neutron heating numbers. The values of ν-bar, the average number of neutrons per fission, are also plotted for appropriate isotopes

  3. Comparative evaluation of photon cross section libraries for materials of interest in PET Monte Carlo simulations

    CERN Document Server

    Zaidi, H

    1999-01-01

    the many applications of Monte Carlo modelling in nuclear medicine imaging make it desirable to increase the accuracy and computational speed of Monte Carlo codes. The accuracy of Monte Carlo simulations strongly depends on the accuracy in the probability functions and thus on the cross section libraries used for photon transport calculations. A comparison between different photon cross section libraries and parametrizations implemented in Monte Carlo simulation packages developed for positron emission tomography and the most recent Evaluated Photon Data Library (EPDL97) developed by the Lawrence Livermore National Laboratory was performed for several human tissues and common detector materials for energies from 1 keV to 1 MeV. Different photon cross section libraries and parametrizations show quite large variations as compared to the EPDL97 coefficients. This latter library is more accurate and was carefully designed in the form of look-up tables providing efficient data storage, access, and management. Toge...

  4. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio.

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)

  5. Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

    Energy Technology Data Exchange (ETDEWEB)

    Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

    1996-05-01

    A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author).

  6. Development of ANJOYMC Program for Automatic Generation of Monte Carlo Cross Section Libraries

    International Nuclear Information System (INIS)

    Kim, Kang Seog; Lee, Chung Chan

    2007-03-01

    The NJOY code developed at Los Alamos National Laboratory is to generate the cross section libraries in ACE format for the Monte Carlo codes such as MCNP and McCARD by processing the evaluated nuclear data in ENDF/B format. It takes long time to prepare all the NJOY input files for hundreds of nuclides with various temperatures, and there can be some errors in the input files. In order to solve these problems, ANJOYMC program has been developed. By using a simple user input deck, this program is not only to generate all the NJOY input files automatically, but also to generate a batch file to perform all the NJOY calculations. The ANJOYMC program is written in Fortran90 and can be executed under the WINDOWS and LINUX operating systems in Personal Computer. Cross section libraries in ACE format can be generated in a short time and without an error by using a simple user input deck

  7. Generation and Verification of ENDF/B-VII.0 Cross section Libraries for Monte Carlo Calculations

    International Nuclear Information System (INIS)

    Park, Ho Jin; Kwak, Min Su; Joo, Han Gyu; Kim, Chang Hyo

    2007-01-01

    For Monte Carlo neutronics calculations, a continuous energy nuclear data library is needed. It can be generated from various evaluated nuclear data files such as ENDF/B using the ACER routine of the NJOY.code after a series of prior processing involving various other NJOY routines. Recently, a utility code, which generates the NJOY input decks in an automated mode, named ANJOYMC became available. The use of this code greatly reduces the user's effort and the possibility of input errors. In December 2006, the initial version of the ENDF/BVII nuclear data library was released. It was reported that the new data files have much better data which reduces the errors noted in the previous versions. Thus it is worthwhile to examine the performance of the new data files particularly using an independent Monte Carlo code, MCCARD and the ANJOYMC utility code. The verification of the newly generated library can be readily performed by analyzing numerous standard criticality benchmark problems

  8. Evaluated cross section libraries

    International Nuclear Information System (INIS)

    Maqurno, B.A.

    1976-01-01

    The dosimetry tape (ENDF/B-IV tape 412) was issued in a general CSEWG distribution, August 1974. The pointwise cross section data file was tested with specified reference spectra. A group averaged cross section data file (620 groups based on tape 412) was tested with the above spectra and the results are presented in this report

  9. 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)

  10. 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

  11. 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)

  12. Homogenized group cross sections by Monte Carlo

    International Nuclear Information System (INIS)

    Van Der Marck, S. C.; Kuijper, J. C.; Oppe, J.

    2006-01-01

    Homogenized group cross sections play a large role in making reactor calculations efficient. Because of this significance, many codes exist that can calculate these cross sections based on certain assumptions. However, the application to the High Flux Reactor (HFR) in Petten, the Netherlands, the limitations of such codes imply that the core calculations would become less accurate when using homogenized group cross sections (HGCS). Therefore we developed a method to calculate HGCS based on a Monte Carlo program, for which we chose MCNP. The implementation involves an addition to MCNP, and a set of small executables to perform suitable averaging after the MCNP run(s) have completed. Here we briefly describe the details of the method, and we report on two tests we performed to show the accuracy of the method and its implementation. By now, this method is routinely used in preparation of the cycle to cycle core calculations for HFR. (authors)

  13. 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

  14. 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

  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. 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)

  17. Production of neutron cross section library based on JENDL-4.0 to continuous-energy Monte Carlo code MVP and its application to criticality analysis of benchmark problems in the ICSBEP handbook

    International Nuclear Information System (INIS)

    Okumura, Keisuke; Nagaya, Yasunobu

    2011-09-01

    In May 2010, JENDL-4.0 was released from Japan Atomic Energy Agency as the updated Japanese Nuclear Data Library. It was processed by the nuclear data processing system LICEM and an arbitrary-temperature neutron cross section library MVPlib - nJ40 was produced for the neutron and photon transport calculation code MVP based on the continuous-energy Monte Carlo method. The library contains neutron cross sections for 406 nuclides on the free gas model, thermal scattering cross sections, and cross sections of pseudo fission products for burn-up calculations with MVP. Criticality benchmark calculations were carried out with MVP and MVPlib - nJ40 for about 1,000 cases of critical experiments stored in the hand book of International Criticality Safety Benchmark Evaluation Project (ICSBEP), which covers a wide variety of fuel materials, fuel forms, and neutron spectra. We report all comparison results (C/E values) of effective neutron multiplication factors between calculations and experiments to give a validation data for the prediction accuracy of JENDL-4.0 for criticalities. (author)

  18. 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)

  19. Testing of cross section libraries for TRIGA criticality benchmark

    International Nuclear Information System (INIS)

    Snoj, L.; Trkov, A.; Ravnik, M.

    2007-01-01

    Influence of various up-to-date cross section libraries on the multiplication factor of TRIGA benchmark as well as the influence of fuel composition on the multiplication factor of the system composed of various types of TRIGA fuel elements was investigated. It was observed that keff calculated by using the ENDF/B VII cross section library is systematically higher than using the ENDF/B-VI cross section library. The main contributions (∼ 2 20 pcm) are from 235 U and Zr. (author)

  20. 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

  1. Comparison of integral cross section values of several cross section libraries in the SAND-II format

    International Nuclear Information System (INIS)

    Zijp, W.L.; Nolthenius, H.J.

    1978-01-01

    A comparison of some integral cross section values for several cross section libraries in the SAND-II format is presented. The integral cross section values are calculated with aid of the spectrum functions for a Watt fission spectrum, a 1/E spectrum and a Maxwellian spectrum. The libraries which are considered here are CCC-112B, ENDF/B-IV, DETAN74, LAPENAS and CESNEF. These 5 cross section libraries used have all the SAND-II format. (author)

  2. Validation of cross sections for Monte Carlo simulation of the photoelectric effect

    CERN Document Server

    Han, Min Cheol; Pia, Maria Grazia; Basaglia, Tullio; Batic, Matej; Hoff, Gabriela; Kim, Chan Hyeong; Saracco, Paolo

    2016-01-01

    Several total and partial photoionization cross section calculations, based on both theoretical and empirical approaches, are quantitatively evaluated with statistical analyses using a large collection of experimental data retrieved from the literature to identify the state of the art for modeling the photoelectric effect in Monte Carlo particle transport. Some of the examined cross section models are available in general purpose Monte Carlo systems, while others have been implemented and subjected to validation tests for the first time to estimate whether they could improve the accuracy of particle transport codes. The validation process identifies Scofield's 1973 non-relativistic calculations, tabulated in the Evaluated Photon Data Library(EPDL), as the one best reproducing experimental measurements of total cross sections. Specialized total cross section models, some of which derive from more recent calculations, do not provide significant improvements. Scofield's non-relativistic calculations are not surp...

  3. Comparison of integral cross section values of several cross section libraries in the SAND-II format

    International Nuclear Information System (INIS)

    Zijp, W.L.; Nolthenius, H.J.

    1976-09-01

    A comparison of some integral cross-section values for several cross-section libraries in the SAND-II format is presented. The integral cross-section values are calculated with the aid of the spectrum functions for a Watt fission spectrum, a 1/E spectrum and a Maxwellian spectrum. The libraries which are considered here are CCC-112B, ENDF/B-IV, DETAN74, LAPENAS and CESNEF. These 5 cross-section libraries used have all the SAND-II format. Discrepancies between cross-sections in the different libraries are indicated but not discussed

  4. Njoy modules used at Enea, Frascati to produce an Ace format neutron cross section library from Eff-1 for the Monte Carlo Mcnp

    International Nuclear Information System (INIS)

    Petrizzi, L.

    1989-01-01

    A note is presented about the experience had in using the NJOY 87.1 module to produce an ACE format library for MCNP from the European Fusion File EFF-1. The IBM 3090 computer, MVS system at ENEA, Bologna was used. The library, called MCNP. EFF1 is at the moment available at Frascati. Few words are said about the met processing problems and the more general topics related to our activity

  5. Library of neutron cross sections of the Thermos code

    International Nuclear Information System (INIS)

    Alonso V, G.; Hernandez L, H.

    1991-10-01

    The present work is the complement of the IT.SN/DFR-017 report in which the structure and the generation of the library of the Thermos code is described. In this report the comparison among the values of the cross sections that has the current library of the Thermos code and those generated by means of the ENDF-B/NJOY it is shown. (Author)

  6. Differences between cross-section libraries for neutron dosimetry

    International Nuclear Information System (INIS)

    Tardelli, T.C.; Stecher, L.C.; Coelho, T.S.; Castro, V.A. De; Cavalieri, T.A.; Menzel, F.; Giarola, R.S.; Domingos, D.B.; Yoriyaz, H.

    2013-01-01

    Absorbed dose calculations depend on a consistent set of nuclear data used in simulations in computer codes. Nuclear data are stored in libraries, however, the information available about the differences in dose caused by different libraries are rare. The libraries are processed by a computer system to be able to be used by a radiation transport code. One of the systems capable of processing nuclear data is the NJOY system. The objective of this study is to evaluate the nuclear data libraries for neutrons available in the literature, and to quantify the differences in absorbed dose obtained using the libraries JENDL 4.0, JEFF 3.3.1 and ENDF/B.VII. The absorbed dose calculation was performed on a simple geometric model, as spheres, and in anthropomorphic model of the human body based on the ICRP-110 for neutron transport simulation using the MCNP5 code. The results were compared with literature data. The results obtained with cross sections from the libraries JEFF and ENDF/B.VII have shown to be identical in most cases, except for one case where the difference has exceeded 10%. The results obtained with JENDL library has shown to be considerably different in most cases comparing to other two libraries. Some differences were over 200%. The dose calculations showed differences between the libraries, which is justified by differences in the cross sections. It has been observed that the cross sections values of certain nuclides assume quite different values in different libraries. These differences in turn cause considerable differences in dose calculations. (author)

  7. EJ2-MCNPlib. Contents of the JEF-2.2 based neutron cross-section library for MCNP4A

    International Nuclear Information System (INIS)

    Hogenbirk, A.; Oppe, J.

    1995-05-01

    In this report a description is given of the EJ2-MCNPlib library. The EJ2-MCNPlib library is to be used for reactivity/critically calculations and general neutron/photon transport calculations with the Monte Carlo code MCNP4A. The library is based on the European JEF-2.2 nuclear data evaluation and contains data for all (i.e. 313) nuclides available on this evaluation.The cross-section data were generated using the NJOY cross-section processing code system, version 91.118. For easy reference cross-section plots are given in this report for the total, elastic and absorption cross sections for all nuclides on the EJ2-MCNPlib library. Furthermore, for verification purposes a graphical intercomparison is given of the results of standard benchmark calculations performed with JEF-2.2 cross-section data and with ENDF/B-V cross-section data (whenever available). 6 refs

  8. 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)

  9. MOX Cross-Section Libraries for ORIGEN-ARP

    International Nuclear Information System (INIS)

    Gauld, I.C.

    2003-01-01

    The use of mixed-oxide (MOX) fuel in commercial nuclear power reactors operated in Europe has expanded rapidly over the past decade. The predicted characteristics of MOX fuel such as the nuclide inventories, thermal power from decay heat, and radiation sources are required for design and safety evaluations, and can provide valuable information for non-destructive safeguards verification activities. This report describes the development of computational methods and cross-section libraries suitable for the analysis of irradiated MOX fuel with the widely-used and recognized ORIGEN-ARP isotope generation and depletion code of the SCALE (Standardized Computer Analyses for Licensing Evaluation) code system. The MOX libraries are designed to be used with the Automatic Rapid Processing (ARP) module of SCALE that interpolates appropriate values of the cross sections from a database of parameterized cross-section libraries to create a problem-dependent library for the burnup analysis. The methods in ORIGEN-ARP, originally designed for uranium-based fuels only, have been significantly upgraded to handle the larger number of interpolation parameters associated with MOX fuels. The new methods have been incorporated in a new version of the ARP code that can generate libraries for low-enriched uranium (LEU) and MOX fuel types. The MOX data libraries and interpolation algorithms in ORIGEN-ARP have been verified using a database of declared isotopic concentrations for 1042 European MOX fuel assemblies. The methods and data are validated using a numerical MOX fuel benchmark established by the Organization for Economic Cooperation and Development (OECD) Working Group on burnup credit and nuclide assay measurements for irradiated MOX fuel performed as part of the Belgonucleaire ARIANE International Program

  10. FENDL/MC. Library of continuous energy cross sections in ACE format for neutron-photon transport calculations with the Monte Carlo N-particle Transport Code system MCNP 4A. 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 for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into ACE format using the NJOY system by R.E. MacFarlane. This document summarizes the resulting continuous energy cross-section data library FENDL/MC version 1.1. The data are available cost free, upon request from the IAEA Nuclear Data Section, online or on magnetic tape. (author). 1 tab

  11. 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

  12. 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

  13. Benchmark of neutron production cross sections with Monte Carlo codes

    Science.gov (United States)

    Tsai, Pi-En; Lai, Bo-Lun; Heilbronn, Lawrence H.; Sheu, Rong-Jiun

    2018-02-01

    Aiming to provide critical information in the fields of heavy ion therapy, radiation shielding in space, and facility design for heavy-ion research accelerators, the physics models in three Monte Carlo simulation codes - PHITS, FLUKA, and MCNP6, were systematically benchmarked with comparisons to fifteen sets of experimental data for neutron production cross sections, which include various combinations of 12C, 20Ne, 40Ar, 84Kr and 132Xe projectiles and natLi, natC, natAl, natCu, and natPb target nuclides at incident energies between 135 MeV/nucleon and 600 MeV/nucleon. For neutron energies above 60% of the specific projectile energy per nucleon, the LAQGMS03.03 in MCNP6, the JQMD/JQMD-2.0 in PHITS, and the RQMD-2.4 in FLUKA all show a better agreement with data in heavy-projectile systems than with light-projectile systems, suggesting that the collective properties of projectile nuclei and nucleon interactions in the nucleus should be considered for light projectiles. For intermediate-energy neutrons whose energies are below the 60% projectile energy per nucleon and above 20 MeV, FLUKA is likely to overestimate the secondary neutron production, while MCNP6 tends towards underestimation. PHITS with JQMD shows a mild tendency for underestimation, but the JQMD-2.0 model with a modified physics description for central collisions generally improves the agreement between data and calculations. For low-energy neutrons (below 20 MeV), which are dominated by the evaporation mechanism, PHITS (which uses GEM linked with JQMD and JQMD-2.0) and FLUKA both tend to overestimate the production cross section, whereas MCNP6 tends to underestimate more systems than to overestimate. For total neutron production cross sections, the trends of the benchmark results over the entire energy range are similar to the trends seen in the dominate energy region. Also, the comparison of GEM coupled with either JQMD or JQMD-2.0 in the PHITS code indicates that the model used to describe the first

  14. New evaluated neutron cross section libraries for the GEANT4 code

    International Nuclear Information System (INIS)

    Mendoza, E.; Cano-Ott, D.; Guerrero, C.; Capote, R.

    2012-04-01

    The so-called High Precision neutron physics model implemented in the GEANT4 simulation package allows simulating the transport of neutrons with energies up to 20 MeV. It relies on the G4NDL cross section libraries, prepared by the GEANT4 collaboration from evaluated cross section files and distributed freely together with the code. Even though the performance of the G4NDL library has been improved over the time, users running complex simulations which involve the transport of neutrons do need more flexibility, in particular when assessing the uncertainties in the simulation results due to the neutron (and hence the nuclear) data library used. For this reason, a software tool has been developed for transforming any evaluated neutron cross section library in the ENDF-6 format into the G4NDL format. Furthermore, eight different releases of ENDF-B, JEFF, JENDL, CENDL and BROND national libraries have been translated into the G4NDL format and are distributed by the IAEA nuclear data service at www-nds.iaea.org/geant4. In this way, GEANT4 users have access to the complete list of standard evaluated neutron data libraries when performing Monte Carlo simulations with GEANT4. Consistency checks and a first validation of the libraries have been made following the methods described in this report. (author)

  15. Cross section library DOSCROS77 (in the SAND-II format)

    International Nuclear Information System (INIS)

    Zijp, W.L.; Nolthenius, H.J.; Borg, N.J.C.M. van der.

    1977-08-01

    The dosimetry cross section library DOSCROS77 is documented with tables, plots and cross section values averaged over a few reference spectra. This library is based on the ENDF/B-IV dosimetry file, supplemented with some other evaluations. The total number of reaction cross section sets incorporated in this library is 49 (+3 cover cross sections sets). The cross section data are available in a format which is suitable for the program SAND-II

  16. AXMIX program for cross section mixing and library arrangement

    International Nuclear Information System (INIS)

    Haynes, G.C.

    1976-03-01

    AXMIX is a FORTRAN IV computer code written to provide the user a tool for creating cross-section data sets for ANISN and DOT from cross-section sets already available on cards, nuclide-organized libraries, and group-independent data sets. Numerous options, including adjointing, P/sub n/ adjustments, and changing table length, are available to give the user broad flexibility. The number of energy groups which will fit into the core allocated is determined first. If all groups will fit, the solution is straightforward; if not, then the maximum number of groups which will fit is processed repeatedly by using direct access I/O and storage disks. Some flexibility in applying AXMIX is lost when cross sections to be processed contain upscatter. A special section on upscatter is included in the report. AXMIX is written for IBM System 360 computers with at least 150K bytes of memory. Problems of a practical nature require from 2 to 20 seconds of CPU time on a 360/91 computer. Running time is inversely proportional to the number of groups of data which will fit into core memory. I/O time is 50 to 100 times CPU time

  17. How to Use Benchmark and Cross-section Studies to Improve Data Libraries and Models

    Science.gov (United States)

    Wagner, V.; Suchopár, M.; Vrzalová, J.; Chudoba, P.; Svoboda, O.; Tichý, P.; Krása, A.; Majerle, M.; Kugler, A.; Adam, J.; Baldin, A.; Furman, W.; Kadykov, M.; Solnyshkin, A.; Tsoupko-Sitnikov, S.; Tyutyunikov, S.; Vladimirovna, N.; Závorka, L.

    2016-06-01

    Improvements of the Monte Carlo transport codes and cross-section libraries are very important steps towards usage of the accelerator-driven transmutation systems. We have conducted a lot of benchmark experiments with different set-ups consisting of lead, natural uranium and moderator irradiated by relativistic protons and deuterons within framework of the collaboration “Energy and Transmutation of Radioactive Waste”. Unfortunately, the knowledge of the total or partial cross-sections of important reactions is insufficient. Due to this reason we have started extensive studies of different reaction cross-sections. We measure cross-sections of important neutron reactions by means of the quasi-monoenergetic neutron sources based on the cyclotrons at Nuclear Physics Institute in Řež and at The Svedberg Laboratory in Uppsala. Measurements of partial cross-sections of relativistic deuteron reactions were the second direction of our studies. The new results obtained during last years will be shown. Possible use of these data for improvement of libraries, models and benchmark studies will be discussed.

  18. Neutron Cross Section Libraries for Cryogenic Aromatic Moderator Materials

    International Nuclear Information System (INIS)

    Cantargi, Florencia; Granada, J.R.; Sbaffoni, Maria Monica

    2008-01-01

    The dynamics of a set of aromatic hydrocarbons, such as benzene, toluene, mesitylene and a 3:2 mixture (by volume) of mesitylene and toluene, all of them in solid phase, was studied as potential moderator materials for cold neutron sources. Cross section libraries were generated for hydrogen bounded in those materials, at several temperatures in ACE format, and they were used in MCNP calculations to analyze their neutron production compared with traditional materials like solid methane and liquid hydrogen. In particular, cross section libraries were generated at 20 K, which is the operating temperature of the majority of the existing cold neutron sources. Although solid methane is the best moderator in terms of cold neutron production, it has very poor radiation resistance, causing spontaneous burping even at fairly low doses. Such effect is considerably reduced in the aromatic hydrocarbons. On the other hand, all of them show a similar and significant neutron production, with the exception of benzene. Even though those aromatic materials are very easy to handle, the solid phases that produce an enhanced flux of cold neutrons correspond to amorphous structures rich in low-energy excitations, and they can be created through lengthy cooling processes requiring in many cases additional annealing stages. The 3:2 mesitylene-toluene mixture, that forms in a simple and direct manner the appropriate disordered structure, constitutes an excellent cryogenic moderator material, as it is able to produce an intense flux of cold neutrons while presenting high resistance to radiation, thus conforming a new and advantageous alternative to traditional moderator materials. (authors)

  19. Propagation of cross section uncertainties in combined Monte Carlo neutronics and burnup calculations

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.; Oppe, J.; Klein Meulekamp, R.; Koning, H. [NRG - Fuels, Actinides and Isotopes group, Petten (Netherlands)

    2005-07-01

    Some years ago a methodology was developed at NRG for the calculation of 'density-to-density' and 'one-group cross section-to-density' sensitivity matrices and covariance matrices for final nuclide densities for burnup schemes consisting of multiple sets of flux/spectrum and burnup calculations. The applicability of the methodology was then demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. A recent development is the extension of this methodology to enable its application in combination with the OCTOPUS-MCNP-FISPACT/ORIGEN Monte Carlo burnup scheme. This required some extensions to the sensitivity matrix calculation tool CASEMATE. The extended methodology was applied on the 'HTR Plutonium Cell Burnup Benchmark' to calculate the uncertainties (covariances) in the final densities, as far as these uncertainties are caused by uncertainties in cross sections. Up to 600 MWd/kg these uncertainties are larger than the differences between the code systems. However, it should be kept in mind that the calculated uncertainties are based on EAF4 uncertainty data. It is not exactly clear on beforehand what a proper set of associated (MCNP) cross sections and covariances would yield in terms of final uncertainties in calculated densities. This will be investigated, by the same formalism, once these data becomes available. It should be noted that the studies performed up till the present date are mainly concerned with the influence of uncertainties in cross sections. The influence of uncertainties in the decay constants, although included in the formalism, is not considered further. Also the influence of other uncertainties (such as -geometrical- modelling approximations) has been left out of consideration for the time being. (authors)

  20. Propagation of cross section uncertainties in combined Monte Carlo neutronics and burnup calculations

    International Nuclear Information System (INIS)

    Kuijper, J.C.; Oppe, J.; Klein Meulekamp, R.; Koning, H.

    2005-01-01

    Some years ago a methodology was developed at NRG for the calculation of 'density-to-density' and 'one-group cross section-to-density' sensitivity matrices and covariance matrices for final nuclide densities for burnup schemes consisting of multiple sets of flux/spectrum and burnup calculations. The applicability of the methodology was then demonstrated by calculations of BR3 MOX pin irradiation experiments employing multi-group cross section uncertainty data from the EAF4 data library. A recent development is the extension of this methodology to enable its application in combination with the OCTOPUS-MCNP-FISPACT/ORIGEN Monte Carlo burnup scheme. This required some extensions to the sensitivity matrix calculation tool CASEMATE. The extended methodology was applied on the 'HTR Plutonium Cell Burnup Benchmark' to calculate the uncertainties (covariances) in the final densities, as far as these uncertainties are caused by uncertainties in cross sections. Up to 600 MWd/kg these uncertainties are larger than the differences between the code systems. However, it should be kept in mind that the calculated uncertainties are based on EAF4 uncertainty data. It is not exactly clear on beforehand what a proper set of associated (MCNP) cross sections and covariances would yield in terms of final uncertainties in calculated densities. This will be investigated, by the same formalism, once these data becomes available. It should be noted that the studies performed up till the present date are mainly concerned with the influence of uncertainties in cross sections. The influence of uncertainties in the decay constants, although included in the formalism, is not considered further. Also the influence of other uncertainties (such as -geometrical- modelling approximations) has been left out of consideration for the time being. (authors)

  1. 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.

  2. DOSCROS81. ECN Cross-Section Library for neutron dosimetry. Summary of contents and documentation

    International Nuclear Information System (INIS)

    Lemmel, H.D.

    1982-01-01

    This document summarizes the contents and documentation of the Cross Section Library DOSCROS81 (640 groups in an extended SAND-II format). The library is based on ENDF/B-5 dosimetry file, supplemented with some other evaluations. The total number of reaction cross section sets incorporated in this library is 70 (+3 cover cross section sets). The entire library can be obtained free of charge from the IAEA Nuclear Data Section. A revised version is called DOSCROS81A. (author)

  3. On the use of the Serpent Monte Carlo code for few-group cross section generation

    International Nuclear Information System (INIS)

    Fridman, E.; Leppaenen, J.

    2011-01-01

    Research highlights: → B1 methodology was used for generation of leakage-corrected few-group cross sections in the Serpent Monte-Carlo code. → Few-group constants generated by Serpent were compared with those calculated by Helios deterministic lattice transport code. → 3D analysis of a PWR core was performed by a nodal diffusion code DYN3D employing two-group cross section sets generated by Serpent and Helios. → An excellent agreement in the results of 3D core calculations obtained with Helios and Serpent generated cross-section libraries was observed. - Abstract: Serpent is a recently developed 3D continuous-energy Monte Carlo (MC) reactor physics burnup calculation code. Serpent is specifically designed for lattice physics applications including generation of homogenized few-group constants for full-core core simulators. Currently in Serpent, the few-group constants are obtained from the infinite-lattice calculations with zero neutron current at the outer boundary. In this study, in order to account for the non-physical infinite-lattice approximation, B1 methodology, routinely used by deterministic lattice transport codes, was considered for generation of leakage-corrected few-group cross sections in the Serpent code. A preliminary assessment of the applicability of the B1 methodology for generation of few-group constants in the Serpent code was carried out according to the following steps. Initially, the two-group constants generated by Serpent were compared with those calculated by Helios deterministic lattice transport code. Then, a 3D analysis of a Pressurized Water Reactor (PWR) core was performed by the nodal diffusion code DYN3D employing two-group cross section sets generated by Serpent and Helios. At this stage thermal-hydraulic (T-H) feedback was neglected. The DYN3D results were compared with those obtained from the 3D full core Serpent MC calculations. Finally, the full core DYN3D calculations were repeated taking into account T-H feedback and

  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. 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

  6. Creation of the equilibrium core PBMR ORIGEN-S cross section library

    International Nuclear Information System (INIS)

    Stoker, C.C.; Reitsma, F.; Karriem, Z.

    2002-01-01

    As part of the design calculations for the Pebble Bed Modular Reactor (PBMR), fuel inventories, neutron and gamma sources and decay heat needs to be determined for the fuel spheres. Using the SCALE4.4 code system, a PBMR specific cross section library was created for the ORIGEN-S depletion calculations, assuming a 10-pass refueling system for the PBMR. In this paper the rationale for the creation of the PBMR library is evaluated in terms of the spectrum dependence due to burn-up. The ORIGEN-S PBMR library was further evaluated comparing the results for different parameters calculated with the reactor analysis diffusion code VSOP and the Monte Carlo code MCNP4C. (author)

  7. Neutron cross-section library for SAND-2 and its service program

    International Nuclear Information System (INIS)

    Berzonis, M.A.; Bondars, Kh.Ya.; Lapenas, A.A.

    1978-01-01

    The logical structure of the neutron cross-section library used in the SAND-2 program complex is considered. The organization of the DSIG01 program creating and servicing the neutron cross section library is described. The DSIG 01 program is written on FORTRAN and permits to create the neutron cross section library on the ES computer magnetic discs operating under the control of the ES operating system and to perform certain manipulations therewith

  8. Monte Carlo Calculation of Thermal Neutron Inelastic Scattering Cross Section Uncertainties by Sampling Perturbed Phonon Spectra

    Science.gov (United States)

    Holmes, Jesse Curtis

    Nuclear data libraries provide fundamental reaction information required by nuclear system simulation codes. The inclusion of data covariances in these libraries allows the user to assess uncertainties in system response parameters as a function of uncertainties in the nuclear data. Formats and procedures are currently established for representing covariances for various types of reaction data in ENDF libraries. This covariance data is typically generated utilizing experimental measurements and empirical models, consistent with the method of parent data production. However, ENDF File 7 thermal neutron scattering library data is, by convention, produced theoretically through fundamental scattering physics model calculations. Currently, there is no published covariance data for ENDF File 7 thermal libraries. Furthermore, no accepted methodology exists for quantifying or representing uncertainty information associated with this thermal library data. The quality of thermal neutron inelastic scattering cross section data can be of high importance in reactor analysis and criticality safety applications. These cross sections depend on the material's structure and dynamics. The double-differential scattering law, S(alpha, beta), tabulated in ENDF File 7 libraries contains this information. For crystalline solids, S(alpha, beta) is primarily a function of the material's phonon density of states (DOS). Published ENDF File 7 libraries are commonly produced by calculation and processing codes, such as the LEAPR module of NJOY, which utilize the phonon DOS as the fundamental input for inelastic scattering calculations to directly output an S(alpha, beta) matrix. To determine covariances for the S(alpha, beta) data generated by this process, information about uncertainties in the DOS is required. The phonon DOS may be viewed as a probability density function of atomic vibrational energy states that exist in a material. Probable variation in the shape of this spectrum may be

  9. Performance assessment of new neutron cross section libraries using MCNP code and some critical benchmarks

    International Nuclear Information System (INIS)

    Bakkari, B El; Bardouni, T El.; Erradi, L.; Chakir, E.; Meroun, O.; Azahra, M.; Boukhal, H.; Khoukhi, T El.; Htet, A.

    2007-01-01

    Full text: New releases of nuclear data files made available during the few recent years. The reference MCNP5 code (1) for Monte Carlo calculations is usually distributed with only one standard nuclear data library for neutron interactions based on ENDF/B-VI. The main goal of this work is to process new neutron cross sections libraries in ACE continuous format for MCNP code based on the most recent data files recently made available for the scientific community : ENDF/B-VII.b2, ENDF/B-VI (release 8), JEFF3.0, JEFF-3.1, JENDL-3.3 and JEF2.2. In our data treatment, we used the modular NJOY system (release 99.9) (2) in conjunction with its most recent upadates. Assessment of the processed point wise cross sections libraries performances was made by means of some criticality prediction and analysis of other integral parameters for a set of reactor benchmarks. Almost all the analyzed benchmarks were taken from the international handbook of Evaluated criticality safety benchmarks experiments from OECD (3). Some revised benchmarks were taken from references (4,5). These benchmarks use Pu-239 or U-235 as the main fissionable materiel in different forms, different enrichments and cover various geometries. Monte Carlo calculations were performed in 3D with maximum details of benchmark description and the S(α,β) cross section treatment was adopted in all thermal cases. The resulting one standard deviation confidence interval for the eigenvalue is typically +/-13% to +/-20 pcm [fr

  10. Perturbation analysis for Monte Carlo continuous cross section models

    International Nuclear Information System (INIS)

    Kennedy, Chris B.; Abdel-Khalik, Hany S.

    2011-01-01

    Sensitivity analysis, including both its forward and adjoint applications, collectively referred to hereinafter as Perturbation Analysis (PA), is an essential tool to complete Uncertainty Quantification (UQ) and Data Assimilation (DA). PA-assisted UQ and DA have traditionally been carried out for reactor analysis problems using deterministic as opposed to stochastic models for radiation transport. This is because PA requires many model executions to quantify how variations in input data, primarily cross sections, affect variations in model's responses, e.g. detectors readings, flux distribution, multiplication factor, etc. Although stochastic models are often sought for their higher accuracy, their repeated execution is at best computationally expensive and in reality intractable for typical reactor analysis problems involving many input data and output responses. Deterministic methods however achieve computational efficiency needed to carry out the PA analysis by reducing problem dimensionality via various spatial and energy homogenization assumptions. This however introduces modeling error components into the PA results which propagate to the following UQ and DA analyses. The introduced errors are problem specific and therefore are expected to limit the applicability of UQ and DA analyses to reactor systems that satisfy the introduced assumptions. This manuscript introduces a new method to complete PA employing a continuous cross section stochastic model and performed in a computationally efficient manner. If successful, the modeling error components introduced by deterministic methods could be eliminated, thereby allowing for wider applicability of DA and UQ results. Two MCNP models demonstrate the application of the new method - a Critical Pu Sphere (Jezebel), a Pu Fast Metal Array (Russian BR-1). The PA is completed for reaction rate densities, reaction rate ratios, and the multiplication factor. (author)

  11. MICROX-2 cross section library based on ENDF/B-VII

    International Nuclear Information System (INIS)

    Hou, J.; Ivanov, K.; Choi, H.

    2012-01-01

    New cross section libraries of a neutron transport code MICROX-2 have been generated for advanced reactor design and fuel cycle analyses. A total of 386 nuclides were processed, including 10 thermal scattering nuclides, which are available in ENDF/B-VII release 0 nuclear data. The NJOY system and MICROR code were used to process nuclear data and convert them into MICROX-2 format. The energy group structure of the new library was optimized for both the thermal and fast neutron spectrum reactors based on Contributon and Point-wise Cross Section Driven (CPXSD) method, resulting in a total of 1173 energy groups. A series of lattice cell level benchmark calculations have been performed against both experimental measurements and Monte Carlo calculations for the effective/infinite multiplication factor and reaction rate ratios. The results of MICROX-2 calculation with the new library were consistent with those of 15 reference cases. The average errors of the infinite multiplication factor and reaction rate ratio were 0.31% δk and 1.9%, respectively. The maximum error of reaction rate ratio was 8% for 238 U-to- 235 U fission of ZEBRA lattice against the reference calculation done by MCNP5. (authors)

  12. Generation of neutron cross sections library for the Thermos code of the Fuel management System (FMS)

    International Nuclear Information System (INIS)

    Alonso V, G.; Viais J, J.

    1990-10-01

    There is developed a method to generate the library of neutron cross sections for the Thermos code by means of the database ENDF-B/IV and the NJOY code. The obtained results are compared with the version previous of the library of neutron cross sections which was processed using the version ENDF-B/III. (Author)

  13. SHAMSI, 48 group cross-section library for fusion nucleonics analysis

    International Nuclear Information System (INIS)

    Ponti, C.; Abbas, Tayyab.

    1982-01-01

    A P 3 48 group coupled neutron gamma-ray (34 N - 14 G) cross-section library is produced and validated for neutronic studies in fusion reactor blanket/shield. This report describes the library content, the procedure adopted and the results of the calculations performed for testing the cross sections

  14. C4P cross-section libraries for safety analyses with SIMMER and related studies

    International Nuclear Information System (INIS)

    Rineiski, A.; Sinitsa, V.; Gabrielli, F.; Maschek, W.

    2011-01-01

    A code and data system, C 4 P, is under development at KIT. It includes fine-group master libraries and tools for generating problem-oriented cross-section libraries, primarily for safety studies with the SIMMER code and related analyses. In the paper, the 560-group master library and problem oriented 40-group and 72-group cross-section libraries, for thermal and fast systems, respectively, are described and their performances are investigated. (author)

  15. PCS a code system for generating production cross section libraries

    International Nuclear Information System (INIS)

    Cox, L.J.

    1997-01-01

    This document outlines the use of the PCS Code System. It summarizes the execution process for generating FORMAT2000 production cross section files from FORMAT2000 reaction cross section files. It also describes the process of assembling the ASCII versions of the high energy production files made from ENDL and Mark Chadwick's calculations. Descriptions of the function of each code along with its input and output and use are given. This document is under construction. Please submit entries, suggestions, questions, and corrections to (ljc at sign llnl.gov) 3 tabs

  16. Research on the display of nuclear cross-section library

    International Nuclear Information System (INIS)

    Huang Shien; Wang Kan; Yu Ganglin

    2008-01-01

    Minutely parsed the dot cross-section format (ACE format) data of the ENDF/ B-6.8 database, which is the foundation of the program that achieved the reading and related handling of ACE format data. This program achieved the plotting, zooming and comparing display functions of nuclear cross section-energy of ENDF/B-6.8 database. It also provides the standard picture formatting file output and/or standard text formatting file output of interesting nuclear data. It accomplished some appropriate validations of this program via the comparing between program results and reference data. (authors)

  17. 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

  18. Two improved Monte Carlo photon cross section techniques

    International Nuclear Information System (INIS)

    Scudiere, M.B.

    1978-01-01

    Truncated series of Legendre coefficients and polynomials are often used in multigroup transport computer codes to describe group-to-group angular density transfer functions. Imposition of group structure on the energy continuum may create discontinuities in the first derivative of these functions. Because of the nature of these discontinuities efficient and accurate full-range polynomial expansions are not practically obtainable. Two separate and distinct methods for Monte Carlo photon transport are presented which eliminate essentially all major disadvantages of truncated expansions. In the first method, partial-range expansions are applied between the discontinuities. Here accurate low-order representations are obtained, which yield modest savings in computer charges. The second method employs unique properties of functions to replace them with a few smooth well-behaved representations. This method brings about a considerable savings in computer memory requirements. In addition, accuracy of the first method is maintained, while execution times are reduced even further

  19. Numerical computation of discrete differential scattering cross sections for Monte Carlo charged particle transport

    International Nuclear Information System (INIS)

    Walsh, Jonathan A.; Palmer, Todd S.; Urbatsch, Todd J.

    2015-01-01

    Highlights: • Generation of discrete differential scattering angle and energy loss cross sections. • Gauss–Radau quadrature utilizing numerically computed cross section moments. • Development of a charged particle transport capability in the Milagro IMC code. • Integration of cross section generation and charged particle transport capabilities. - Abstract: We investigate a method for numerically generating discrete scattering cross sections for use in charged particle transport simulations. We describe the cross section generation procedure and compare it to existing methods used to obtain discrete cross sections. The numerical approach presented here is generalized to allow greater flexibility in choosing a cross section model from which to derive discrete values. Cross section data computed with this method compare favorably with discrete data generated with an existing method. Additionally, a charged particle transport capability is demonstrated in the time-dependent Implicit Monte Carlo radiative transfer code, Milagro. We verify the implementation of charged particle transport in Milagro with analytic test problems and we compare calculated electron depth–dose profiles with another particle transport code that has a validated electron transport capability. Finally, we investigate the integration of the new discrete cross section generation method with the charged particle transport capability in Milagro.

  20. Group cross-section processing method and common nuclear group cross-section library based on JENDL-3 nuclear data file

    International Nuclear Information System (INIS)

    Hasegawa, Akira

    1991-01-01

    A common group cross-section library has been developed in JAERI. This system is called 'JSSTDL-295n-104γ (neutron:295 gamma:104) group constants library system', which is composed of a common 295n-104γ group cross-section library based on JENDL-3 nuclear data file and its utility codes. This system is applicable to fast and fusion reactors. In this paper, firstly outline of group cross-section processing adopted in Prof. GROUCH-G/B system is described in detail which is a common step for all group cross-section library generation. Next available group cross-section libraries developed in Japan based on JENDL-3 are briefly reviewed. Lastly newly developed JSSTDL library system is presented with some special attention to the JENDL-3 data. (author)

  1. 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

  2. Reconstruction of point cross-section from ENDF data file for Monte Carlo applications

    International Nuclear Information System (INIS)

    Kumawat, H.; Saxena, A.; Carminati, F.; )

    2016-12-01

    Monte Carlo neutron transport codes are one of the best tools to simulate complex systems like fission and fusion reactors, Accelerator Driven Sub-critical systems, radio-activity management of spent fuel and waste, optimization and characterization of neutron detectors, optimization of Boron Neutron Capture Therapy, imaging etc. The neutron cross-section and secondary particle emission properties are the main input parameters of such codes. The fission, capture and elastic scattering cross-sections have complex resonating structures. Evaluated Nuclear Data File (ENDF) contains these cross-sections and secondary parameters. We report the development of reconstruction procedure to generate point cross-sections and probabilities from ENDF data file. The cross-sections are compared with the values obtained from PREPRO and in some cases NJOY codes. The results are in good agreement. (author)

  3. Cross-section library and processing techniques within the SCALE system

    International Nuclear Information System (INIS)

    Westfall, R.M.

    1986-01-01

    A summary of each of the SCALE system features involved in problem-dependent cross section processing is presented. These features include criticality libraries, shielding libraries, the Standard Composition Library, the SCALE functional modules: BONAMI-S, NITAWL-S, XSDRNPM-S, ICE-S, and the Material Information Processor. The automated procedure for cross-section processing is described with examples. 15 refs

  4. Fitted temperature-corrected Compton cross sections for Monte Carlo applications and a sampling distribution

    International Nuclear Information System (INIS)

    Wienke, B.R.; Devaney, J.J.; Lathrop, B.L.

    1984-01-01

    Simple temperature-corrected cross sections, which replace the static Klein-Nishina set in a one-to-one manner, are developed for Monte Carlo applications. The reduced set is obtained from a nonlinear least-squares fit to the exact photon-Maxwellian electron cross sections by using a Klein-Nishina-like formula as the fitting equation. Two parameters are sufficient, and accurate to two decimal places, to explicitly fit the exact cross sections over a range of 0 to 100 keV in electron temperature and 0 to 1 MeV in incident photon energy. Since the fit equations are Klein-Nishina-like, existing Monte Carlo code algorithms using the Klein-Nishina formula can be trivially modified to accommodate corrections for a moving Maxwellian electron background. The simple two parameter scheme and other fits are presented and discussed and comparisons with exact predictions are exhibited. The fits are made to the total photon-Maxwellian electron cross section and the fitting parameters can be consistently used in both the energy conservation equation for photon-electron scattering and the differential cross section, as they are presently sampled in Monte Carlo photonics applications. The fit equations are motivated in a very natural manner by the asymptotic expansion of the exact photon-Maxwellian effective cross-section kernel. A probability distribution is also obtained for the corrected set of equations

  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. Cross sections of the lumped fission products for the AMZ library

    International Nuclear Information System (INIS)

    Ono, S.; Corcueca, R.P.; Nascimento, J.A.

    1985-01-01

    The preparation of the lumped fission product cross section for the AMZ library is described. For this purpose 100 nuclides were selected. The cross sections for each nuclide were generated by the NJOY code with evaluated nuclear data from ENDF/B-V, complemented with ENDF/B-IV data. A comparison is performed between the data obtained and the lumped fission product cross section of JFS-II [pt

  7. Remarks on the comparison of cross section libraries for neutron metrology

    International Nuclear Information System (INIS)

    Zijp, W.L.; Nolthenius, H.J.; Appelman, K.H.

    1977-01-01

    Cross section libraries in a 620 group structure were available from different origin: CCC-112B, DETAN-74 and ENDF/B-IV. For a few well known neutron spectra (CFRMF spectrum, ΣΣ spectrum, fission neutron spectrum, HFR neutron spectrum) a comparison was made of the available experimental reaction rates in foil detectors and the reaction rates as calculated with the different cross section libraries. This investigation is dealing with the consistency of cross section data within a library, and the consistency of activity data in actual reaction rate determinations. Some preliminary conclusions are given

  8. ACT-1000. Group activation cross-section library for WWER-1000 type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zolotarev, K I; Pashchenko, A B [National Research Centre - A.I. Leipunsky Institute for Physics and Power Engineering, Obninsk (Russian Federation)

    2001-10-01

    The ACT-1000, a problem-oriented library of group-averaged activation cross-sections for WWER-1000 type reactors, is based on evaluated microscopic cross-section data files. The ACT-1000 data library was designed for calculating induced activity for the main dose-generated nuclides contained in WWER-1000 structural materials. In preparing the ACT-1000 library, 47 group-averaged cross-section data for the 10{sup -9}-17.33 MeV energy range were used to calculate the spatial-energy neutron flux distribution. (author)

  9. Comparative analysis of the neutron cross-sections of iron from various evaluated data libraries

    International Nuclear Information System (INIS)

    Bychkov, V.M.; Vozyakov, V.V.; Manokhin, V.N.; Smoll, F.; Resner, P.; Seeliger, D.; Hermsdorf, D.

    1983-09-01

    The comparative analysis of neutron cross-sections of iron from evaluated nuclear data libraries SOKRATOR, KEDAK, ENDL is done in energy interval from 0.025 eV to 20 MeV. Some of iron cross-sections from SOKRATOR library are revised and new data, which are obtained by using new experimental data and more comprehensive theoretical methods, are recommended. As a result the new version of the iron neutron cross-section file (BNF-2012) is produced for SOKRATOR library. (author)

  10. FENDL/E-2.0. Evaluated nuclear data library of neutron-nucleus interaction cross sections and photon production cross sections and photon-atom interaction cross sections for fusion applications. Version 1, March 1997. Summary documentation

    International Nuclear Information System (INIS)

    Pashchenko, A.B.; Wienke, H.

    1998-01-01

    This document presents the description of a physical tape containing the basic evaluated nuclear data library of neutron-nucleus interaction cross sections, photon production cross sections and photon-atom interaction cross sections for fusion applications. It is part of the evaluated nuclear data library for fusion applications FENDL-2. The data are available cost-free from the Nuclear Data Section upon request. The data can also be retrieved by the user via online access through international computer networks. (author)

  11. Preliminary assessment of Geant4 HP models and cross section libraries by reactor criticality benchmark calculations

    DEFF Research Database (Denmark)

    Cai, Xiao-Xiao; Llamas-Jansa, Isabel; Mullet, Steven

    2013-01-01

    Geant4 is an open source general purpose simulation toolkit for particle transportation in matter. Since the extension of the thermal scattering model in Geant4.9.5 and the availability of the IAEA HP model cross section libraries, it is now possible to extend the application area of Geant4......, U and O in uranium dioxide, Al metal, Be metal, and Fe metal. The native HP cross section library G4NDL does not include data for elements with atomic number larger than 92. Therefore, transuranic elements, which have impacts for a realistic reactor, can not be simulated by the combination of the HP...... models and the G4NDL library. However, cross sections of those missing isotopes were made available recently through the IAEA project “new evaluated neutron cross section libraries for Geant4”....

  12. Creation of problem-dependent Doppler-broadened cross sections in the KENO Monte Carlo code

    International Nuclear Information System (INIS)

    Hart, Shane W.D.; Celik, Cihangir; Maldonado, G. Ivan; Leal, Luiz

    2016-01-01

    Highlights: • A quick method of Doppler broadening one- and two-dimensional cross sections has been added to KENO. • The method uses a finite difference method to broaden data to user defined temperatures. • Various problems and benchmarks were run to showcase results. • Results with the Doppler broadened cross sections are closer to benchmark results. - Abstract: This paper introduces a quick method for improving the accuracy of Monte Carlo simulations by generating one- and two-dimensional cross sections at a user-defined temperature before performing transport calculations. A finite difference method is used to Doppler-broaden cross sections to the desired temperature, and unit-base interpolation is done to generate the probability distributions for double differential two-dimensional thermal moderator cross sections at any arbitrarily user-defined temperature. The accuracy of these methods is tested using a variety of contrived problems. In addition, various benchmarks at elevated temperatures are modeled, and results are compared with benchmark results. The problem-dependent cross sections are observed to produce eigenvalue estimates that are closer to the benchmark results than those without the problem-dependent cross sections.

  13. Improvement of decay and cross-section data libraries for activation calculations

    International Nuclear Information System (INIS)

    Attaya, H.

    1993-01-01

    A new decay data library has been completed. The new library contains up-to-date decay information (half-lives, branching ratios, decay energies, γ's energies and intensities). Activation responses such as the air and water biological hazard potentials, the waste disposal rating, and the biological dose are also included in this library. Recently developed cross-section libraries have been acquired to be used together with the decay data library

  14. New Standard Evaluated Neutron Cross Section Libraries for the GEANT4 Code and First Verification

    CERN Document Server

    Mendoza, Emilio; Koi, Tatsumi; Guerrero, Carlos

    2014-01-01

    The Monte Carlo simulation of the interaction of neutrons with matter relies on evaluated nuclear data libraries and models. The evaluated libraries are compilations of measured physical parameters (such as cross sections) combined with predictions of nuclear model calculations which have been adjusted to reproduce the experimental data. The results obtained from the simulations depend largely on the accuracy of the underlying nuclear data used, and thus it is important to have access to the nuclear data libraries available, either of general use or compiled for specific applications, and to perform exhaustive validations which cover the wide scope of application of the simulation code. In this paper we describe the work performed in order to extend the capabilities of the GEANT4 toolkit for the simulation of the interaction of neutrons with matter at neutron energies up to 20 MeV and a first verification of the results obtained. Such a work is of relevance for applications as diverse as the simulation of a n...

  15. Quest for precision in hadronic cross sections at low energy: Monte Carlo tools vs. experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Actis, S [Paul-Scherrer-Institute Wuerenlingen and Villigen, Villigen (Switzerland); Arbuzov, A [Joint Institute for Nuclear Research, Dubna (Russian Federation). Bogoliubov Lab. of Theoretical Physics; Balossini, G [Pavia Univ. (Italy). Dipt. di Fisica Nucleare e Teorica; INFN, Pavia [IT; and others

    2009-12-15

    We present the achievements of the last years of the experimental and theoretical groups working on hadronic cross section measurements at the low energy e{sup +}e{sup -} colliders in Beijing, Frascati, Ithaca, Novosibirsk, Stanford and Tsukuba and on {tau} decays. We sketch the prospects in these fields for the years to come. We emphasise the status and the precision of the Monte Carlo generators used to analyse the hadronic cross section measurements obtained as well with energy scans as with radiative return, to determine luminosities and {tau} decays. The radiative corrections fully or approximately implemented in the various codes and the contribution of the vacuum polarisation are discussed. (orig.)

  16. Quest for precision in hadronic cross sections at low energy: Monte Carlo tools vs. experimental data

    International Nuclear Information System (INIS)

    Actis, S.; Arbuzov, A.

    2009-12-01

    We present the achievements of the last years of the experimental and theoretical groups working on hadronic cross section measurements at the low energy e + e - colliders in Beijing, Frascati, Ithaca, Novosibirsk, Stanford and Tsukuba and on τ decays. We sketch the prospects in these fields for the years to come. We emphasise the status and the precision of the Monte Carlo generators used to analyse the hadronic cross section measurements obtained as well with energy scans as with radiative return, to determine luminosities and τ decays. The radiative corrections fully or approximately implemented in the various codes and the contribution of the vacuum polarisation are discussed. (orig.)

  17. 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

  18. Neutron cross-section libraries in the AMPX master interface format for thermal and fast reactors

    International Nuclear Information System (INIS)

    Bjerke, M.A.; Webster, C.C.

    1981-12-01

    Neutron cross-section libraries in the AMPX master interface format have been created for three reactor types. Included are an 84-group library for use with light-water reactors, a 27-group library for use with heavy-water CANDU reactors and a 126-group library for use with liquid metal fast breeder reactors. In general, ENDF/B data were used in the creation of these libraries, and the nuclides included in each library should be sufficient for most neutronic analyses of reactors of that type. Each library has been used successfully in fuel depletion calculations

  19. Development of modern CANDU PHWR cross-section libraries for SCALE

    International Nuclear Information System (INIS)

    Shoman, Nathan T.; Skutnik, Steven E.

    2016-01-01

    Highlights: • New ORIGEN libraries for CANDU 28 and 37-element fuel assemblies have been created. • These new reactor data libraries are based on modern ENDF/B-VII.0 cross-section data. • The updated CANDU data libraries show good agreement with radiochemical assay data. • Eu-154 overestimated when using ENDF-VII.0 due to a lower thermal capture cross-section. - Abstract: A new set of SCALE fuel lattice models have been developed for the 28-element and 37-element CANDU fuel assembly designs using modern cross-section data from ENDF-B/VII.0 in order to produce new reactor data libraries for SCALE/ORIGEN depletion analyses. These new libraries are intended to provide users with a convenient means of evaluating depletion of CANDU fuel assemblies using ORIGEN through pre-generated cross sections based on SCALE lattice physics calculations. The performance of the new CANDU ORIGEN libraries in depletion analysis benchmarks to radiochemical assay data were compared to the previous version of the CANDU libraries provided with SCALE (based on WIMS-AECL models). Benchmark comparisons with available radiochemical assay data indicate that the new cross-section libraries perform well at matching major actinide species (U/Pu), which are generally within 1–4% of experimental values. The library also showed similar or better results over the WIMS-AECL library regarding fission product species and minor actinoids (Np, Am, and Cm). However, a notable exception was in calculated inventories of "1"5"4Eu and "1"5"5Eu, where the new library employing modern nuclear data (ENDF/B-VII.0) performed substantially poorer than the previous WIMS-AECL library (which used ENDF-B/VI.8 cross-sections for these species). The cause for this discrepancy appears to be due to differences in the "1"5"4Eu thermal capture cross-section between ENDF/B-VI.8 and ENDF/B-VII.0, an effect which is exacerbated by the highly thermalized flux of a CANDU heavy water reactor compared to that of a typical

  20. Library of neutron reaction cross-sections in the ABBN-93 constant system

    International Nuclear Information System (INIS)

    Zabrodskaya, S.V.; Korchagina, Zh.A.; Koshcheev, V.N.; Nikolaev, M.N.; Tsibulya, A.M.

    2001-01-01

    The library of neutron reaction group cross-sections in the ABBN-93 constant set is described. The format used for data representation, the content and purpose of the sub-libraries and their practical application in the SCALE criticality safety estimation system are discussed. (author)

  1. 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)

  2. 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)

  3. Up to date cross sections library for Thermos and Record codes

    International Nuclear Information System (INIS)

    Hernandez Lopez, H.

    1993-01-01

    Reactor cell analysis is the first step in determining reactor core behavior and is required in the reload licensing process. For best results, reactor cell analysis should be carried out with libraries of up to date, accurate cross sections produced with well described methods from standard evaluated nuclear data. At first step in this work were determined the library structure for RECORD and THERMOS and were prepared the cross sections libraries using the NJOY nuclear data processing system and the ENDF-B/IV evaluated nuclear data. These libraries were used by the codes and some samples were perform, the result show some differences against the results obtained using the previous libraries. By other hand the libraries contain various adjustments to correct for deficiencies in nuclear data or analytical methods. These adjustments doesn't have any documentation, although some of them were identified in this work. (Author). 25 refs, 78 figs, 55 tabs

  4. A unified Monte Carlo approach to fast neutron cross section data evaluation.

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.; Nuclear Engineering Division

    2008-03-03

    A unified Monte Carlo (UMC) approach to fast neutron cross section data evaluation that incorporates both model-calculated and experimental information is described. The method is based on applications of Bayes Theorem and the Principle of Maximum Entropy as well as on fundamental definitions from probability theory. This report describes the formalism, discusses various practical considerations, and examines a few numerical examples in some detail.

  5. Evaluated cross-section libraries and kerma factors for neutrons up to 100 MeV on 12C

    International Nuclear Information System (INIS)

    Chadwick, M.B.; Blann, M.; Cox, L.; Young, P.G.; Meigooni, A.

    1995-01-01

    A program is being carried out at Lawrence Livermore National Laboratory to develop high-energy evaluated nuclear data libraries for use in Monte Carlo simulations of cancer radiation therapy. In this report we describe evaluated cross sections and kerma factors for neutrons with incident energies up to 100 MeV on 12 C. The aim of this effort is to incorporate advanced nuclear physics modeling methods, with new experimental measurements, to generate cross section libraries needed for an accurate simulation of dose deposition in fast neutron therapy. The evaluated libraries are based mainly on nuclear model calculations, benchmarked to experimental measurements where they exist. We use the GNASH code system, which includes Hauser-Feshbach, preequilibrium, and direct reaction mechanisms. The libraries tabulate elastic and nonelastic cross sections, angle-energy correlated production spectra for light ejectiles with A≤and kinetic energies given to light ejectiles and heavy recoil fragments. The major steps involved in this effort are: (1) development and validation of nuclear models for incident energies up to 100 MeV; (2) collation of experimental measurements, including new results from Louvain-la-Nueve and Los Alamos; (3) extension of the Livermore ENDL formats for representing high-energy data; (4) calculation and evaluation of nuclear data; and (5) validation of the libraries. We describe the evaluations in detail, with particular emphasis on our new high-energy modeling developments. Our evaluations agree well with experimental measurements of integrated and differential cross sections. We compare our results with the recent ENDF/B-VI evaluation which extends up to 32 MeV

  6. 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

  7. FENDL/E. Evaluated nuclear data library of neutron nuclear interaction cross-sections and photon production cross-sections and photon-atom interaction cross sections for fusion applications. Version 1.1 of November 1994

    International Nuclear Information System (INIS)

    Pashchenko, A.B.; Wienke, H.; Ganesan, S.; McLaughlin, P.K.

    1996-01-01

    This document presents the description of a physical tape containing the basic evaluated nuclear data library of neutron nuclear interaction cross-sections and photon production cross-sections and photon-atom interaction cross-sections for fusion applications. It is part of FENDL, the evaluated nuclear data library for fusion applications. The nuclear data are available cost-free for distribution to interested scientists upon request. The data can also be retrieved by the user via online access through international computer networks. (author). 11 refs, 1 tab

  8. Basis calculation of phase cross section library in a low power fast reactor neutronic simulation

    International Nuclear Information System (INIS)

    Jachic, J.

    1993-09-01

    In order to implement the utilization of the efficient multidimensional cubic SPLINE interpolation, we determine the phase library bases for net like relevant state components. A generic cubic surface and a weighted plane pertinent alternative interpolating methods used capable to generate cross sections values for fixed coordinates from cell code calculated data points is used. It is verified that the phase library bases increases or decrease smoothly and monotonically with the spectrum asymmetry and total flux buckling. This justifies its use in cross section updating avoiding cell calculations. (author)

  9. Energy-balance check for continuous energy cross section library CENACE-1.0

    International Nuclear Information System (INIS)

    Zhao Qiujuan; Wu Haicheng; Ge Zhigang

    2014-01-01

    In order to verify the reliability of the multiple-temperature continuous energy cross section library CENACE-1.0 when used for calculating nuclear heating in reactor core, NJOY99/HEATR module and auxiliary code chkACEheat developed locally were used to perform energy-balance check for all materials in the library. The test results show that the pass rate of KERMA factors and heat production cross sections of the CENACE-1.0 library is better than that of the other ACE libraries used as comparison. However, unreasonable KERMA factors still exist in various evaluation libraries, and methods to directly revise the calculation results of KERMA factors need to be developed. (authors)

  10. ZZ DOSCROS, Neutron Cross-Section Library for Spectra Unfolding and Integral Parameter Evaluation

    International Nuclear Information System (INIS)

    Zijp, Willem L.; Nolthenius, Henk J.; Rieffe, Henk Ch.

    1987-01-01

    1 - Description of problem or function: Format: SAND-II; Number of groups: 640 fine group cross section values; Nuclides: Li, B, F, Na, Mg, Al, S, Sc, Ti, Cr, Mn, Fe, Co, Ni, Cu, Zn, As, Br, Nb, Mo, Rh, Pd, Ag, In, Sb, I, Cs, La, Eu, Sm, Dy, Lu, Ta, W, Re, Au, Th, U, Np, Pu. Origin: ENDF/B-V mainly, ENDF/B-IV, INDL/V. This library forms in combination with the DAMSIG81 library a convenient source of evaluated energy dependent cross section sets which may be used in the determination of neutron spectra by means of adjustment (or unfolding) procedures or which can be used for the determination of integral parameters (such as damage-to-activation ratio) useful in characterising the neutron spectra. The energy dependent fine group cross section data are presented in a 640 group structure of the SAND-II type. This group structure has 45 energy groups per energy decade below 1 MeV and a group width of 100 KeV above 1 MeV. The total energy span of this group structure is from 10 -10 MeV to 20 MeV. The library has the SAND-II format, which implies that a special part of the library has to contain cover cross section data sets. These cross section data sets are required in the SAND-II program for taking into account the influence of special detector surroundings which may be used during an irradiation. 2 - Method of solution: The selection of the reactions from the evaluated nuclear data libraries was determined by various properties of the reactions for neutron metrology. For this reason all the well- known reactions of the ENDF/B-V dosimetry file are included but these data are supplemented with cross section sets for less well known metrology reactions which may become of interest

  11. Description of the ENDF-NJOY system for the generation of cross sections libraries

    International Nuclear Information System (INIS)

    Alonso V, G.

    1991-01-01

    The physics of nuclear reactors requires of a great number of data to be able to evaluate the different phenomena that happen in a nuclear reactor; these data are mainly the microscopic neutron cross sections, but it is also required of data of radioactive decay and of nuclear structure for a great number of materials as well as of the cross sections of the photons and the production of these for the neutron interaction. These data group in nuclear databases, being the main ones: ENDF Nuclear Evaluated File, ENDL Dates Nuclear Evaluated Library it Dates (of the Laboratory Lawrence Livermore). JENDL Japanese Nuclear Evaluated Library Dates. Soviet SOKRATOR Nuclear Evaluated KEDAF Nuclear Karlsruhe File Dates. JEF Join Evaluated File (coordinated by NEA Data Bank). The existent codes that execute neutron and photon calculations require libraries of data that are very different some of other and of the databases. Of here that it is required of a series of processing codes that use the database like enter and its generate a secondary library of cross sections, which is read as enter for a code of spectra generation. Generally average cross sections by group are obtained; this library is that it is used in the codes that execute neutron calculations. (Author)

  12. Point 2004 A Temperature Dependent ENDF/B-VI, Release 8 Cross Section Library

    International Nuclear Information System (INIS)

    Cullen, D E

    2004-01-01

    The ENDF/B data library has recently been updated and is now freely available through the National Nuclear Data Center (NNDC), Brookhaven National Laboratory. This most recent library is identified as ENDF/B-VI, Release 8. Release 8 completely supersedes all preceding releases. Release 8 will be the last release of ENDF/B-VI; the next release of ENDF/B data will be for the new ENDF/B-VII library. As distributed the ENDF/B-VI, Release 8 data includes cross sections represented in the form of a combination of resonance parameters and/or tabulated energy dependent cross sections, nominally at 0 Kelvin temperature. For use in applications this library has been processed into the form of temperature dependent cross sections at eight neutron reactor like temperatures, between 0 and 2100 Kelvin, in steps of 300 Kelvin. It has also been processed to five astrophysics like temperatures, 1, 10, 100 eV, 1 and 10 keV. For reference purposes, 300 Kelvin is approximately 1/40 eV, so that 1 eV is approximately 12,000 Kelvin. At each temperature the cross sections are tabulated and linearly interpolable in energy. All results are in the computer independent ENDF/B-VI character format [1], which allows the data to be easily transported between computers. In its processed form this library is approximately 4.3 gigabyte in size and is distributed on a single DVD

  13. Analysis of fusion neutronics calculations and appraisal of UW cross-section library

    International Nuclear Information System (INIS)

    Xie Jianping; Li Xingzhong; Ying Chuntong

    1989-01-01

    A series of calculations for different cases (especially for the values of tritium breeding ratio T, and the fuel breeding ratio F in the blanket of a hybrid reactor) were carried out by using ANISN program and UW cross-section library. The comparison with other results in China and abroad kalso was done. It was shownwn that the installation and execution of ANISN program on ELXSI machine at Tsinghua University are successful, and the UW cross-section library is reliable. It may be used for fusion neutronics calculation in the future. The paper also points out that the difference between the calculations and by the authors are due to jthe different in cross-section data used

  14. ECNJEFI. A JEFI based 219-group neutron cross-section library: User's manual

    International Nuclear Information System (INIS)

    Stad, R.C.L. van der; Gruppelaar, H.

    1992-07-01

    This manual describes the contents of the ECNJEF1 library. The ECNJEF1 library is a JEF1.1 based 219-group AMPX-Master library for reactor calculations with the AMPX/SCALE-system, e.g. the PASC-3 system as implemented at the Netherlands Energy Research Foundation in Petten, Netherlands. The group cross-section data were generated with NJOY and NPTXS/XLACS-2 from the AMPX system. The data on the ECNJEF1 library allows resolved-resonance treatment by NITAWL and/or unresolved resonance self-shielding by BONAMI. These codes are based upon the Nordheim and Bondarenko methods, respectively. (author). 10 refs., 7 tabs

  15. 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

  16. POINT 2011: ENDF/B-VII.1 Beta2 Temperature Dependent Cross Section Library

    Energy Technology Data Exchange (ETDEWEB)

    Cullen, D E

    2011-04-07

    This report is one in the series of 'POINT' reports that over the years have presented temperature dependent cross sections for the then current version of ENDF/B. In each case I have used my personal computer at home and publicly available data and codes. I have used these in combination to produce the temperature dependent cross sections used in applications and presented in this report. I should mention that today anyone with a personal computer can produce these results. The latest ENDF/B-VII.1 beta2 data library was recently and is now freely available through the National Nuclear Data Center (NNDC), Brookhaven National Laboratory. This release completely supersedes all preceding releases of ENDF/B. As distributed the ENDF/B-VII.1 data includes cross sections represented in the form of a combination of resonance parameters and/or tabulated energy dependent cross sections, nominally at 0 Kelvin temperature. For use in our applications the ENDF/B-VII.1 library has been processed into cross sections at eight neutron reactor like temperatures, between 0 and 2100 Kelvin, in steps of 300 Kelvin (the exception being 293.6 Kelvin, for exact room temperature at 20 Celsius). It has also been processed to five astrophysics like temperatures, 1, 10, 100 eV, 1 and 10 keV. For reference purposes, 300 Kelvin is approximately 1/40 eV, so that 1 eV is approximately 12,000 Kelvin. At each temperature the cross sections are tabulated and linearly interpolable in energy. All results are in the computer independent ENDF-6 character format [R2], which allows the data to be easily transported between computers. In its processed form the POINT 2011 library is approximately 16 gigabyte in size and is distributed on one compressed DVDs (see, below for the details of the contents of each DVD).

  17. Bonderenko self-shielded cross sections and multiband parameters derived from the LLL Evaluated-Nuclear-Data Library (ENDL)

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1978-01-01

    Bonderenko self-shielded cross sections and multiband parameters from the Lawrence Livermore Laboratory Evaluated-Nuclear-Data Library (ENDL) as of July 4, 1978 are presented. These data include total, elastic, capture, and fission cross sections in the TART 175 group structure. Multiband parameters are listed. Bonderenko self-shielded cross section and the multiband parameters are presented on microfiche

  18. TOPICS-B, Neutron and Gamma Cross-Sections Library Handling in FIDO Format

    International Nuclear Information System (INIS)

    Wasastjerna, Frej

    2003-01-01

    1 - Description of program or function: The program is intended to manipulate working format neutron and/or gamma cross section libraries, carrying out such operations as mixing materials, deleting unneeded groups, inserting response cross sections or whatever the user may require. It has been designed to make it easy to include new modules to cope with new requirements. The cross section libraries involved should preferably be in ANISN format, but if they are not, this too can be handled by adding new modules as needed. This program is intended to supersede TOPICS (NEA-1406). TOPICS was intended for interactive use, but experience has shown that using it is somewhat difficult. Therefore it was redesigned for batch use (the input is written to a file and the program is then run using that file, instead of reading input directly from the keyboard). 2 - Method of solution: Each required operation is performed by a separate module (a set of subprograms). 3 - Restrictions on the complexity of the problem: Essentially none, variable dimensioning is used. However, TOPICS-B is not intended to be applied to basic nuclear data libraries (such as the ENDF/B series) or to flexible format libraries (e.g., the VITAMIN series). It is intended only for working format libraries like the BUGLE series

  19. Role of ''standard'' fine-group cross section libraries in shielding analysis

    International Nuclear Information System (INIS)

    Weisbin, C.R.; Roussin, R.W.; Oblow, E.M.; Cullen, D.E.; White, J.E.; Wright, R.Q.

    1977-01-01

    The Divisions of Magnetic Fusion Energy (DMFE) and Reactor Development and Demonstration (DRDD) of the United States Energy Research and Development Administration (ERDA) have jointly sponsored the development of a 171 neutron, 36 gamma ray group pseudo composition independent cross section library based upon ENDF/B-IV. This library (named VITAMIN-C and packaged by RSIC as DLC-41) is intended to be generally applicable to fusion blanket and LMFBR core and shield analysis. The purpose of this paper is to evaluate this library as a possible candidate for specific designation as a ''standard'' in light of American Nuclear Society standards for fine-group cross section data sets. The rationale and qualification procedure for such a standard are discussed. Finally, current limitations and anticipated extensions to this processed data file are described

  20. ORACLE: an adjusted cross-section and covariance library for fast-reactor analysis

    International Nuclear Information System (INIS)

    Yeivin, Y.; Marable, J.H.; Weisbin, C.R.; Wagschal, J.J.

    1980-01-01

    Benchmark integral-experiment values from six fast critical-reactor assemblies and two standard neutron fields are combined with corresponding calculations using group cross sections based on ENDF/B-V in a least-squares data adjustment using evaluated covariances from ENDF/B-V and supporting covariance evaluations. Purpose is to produce an adjusted cross-section and covariance library which is based on well-documented data and methods and which is suitable for fast-reactor design. By use of such a library, data- and methods-related biases of calculated performance parameters should be reduced and uncertainties of the calculated values minimized. Consistency of the extensive data base is analyzed using the chi-square test. This adjusted library ORACLE will be available shortly

  1. FENDL/A-2.0. Neutron activation cross section data library for fusion applications

    International Nuclear Information System (INIS)

    Pashchenko, A.B.; Wienke, H.; Kopecky, J.; Sublet, J.C. Sublet; Forrest, R.A.

    1997-01-01

    This document describes the contents of a comprehensive neutron cross section data library for 13,006 neutron activation reactions with 739 target nuclides from H (A=1,Z=1) to Cm (A=248,Z=96), in the incident energy range up to 20 MeV. FENDL/A-2 is a sublibrary of FENDL-2, the second revision of the evaluated nuclear data library for fusion applications. It is supplemented by a decay data library FENDL/D-2 in ENDF-6 format for 1867 nuclides. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape upon request. (author)

  2. AXMIX, ANISN Cross-Sections Mixing, Transport Corrections, Data Library Management

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Nature of physical problem solved: Mixing, changing table length, adjoining, making scattering order adjustments (PN delta function subtraction), and transport corrections of ANISN-type cross sections, and management of cross section data sets and libraries. 2 - Method of solution: The number of energy groups which will fit into the core allocated is determined first. If all groups will fit, the solution is straightforward. If not, then the maximum number of groups which will fit is processed repeatedly using direct access I/O and storage disks. 3 - Restrictions on the complexity of the problem: Some flexibility in applying AXMIX is lost when cross sections to be processed contain up-scatter. A special section on up-scatter is therefore included in the report

  3. A probability-conserving cross-section biasing mechanism for variance reduction in Monte Carlo particle transport calculations

    OpenAIRE

    Mendenhall, Marcus H.; Weller, Robert A.

    2011-01-01

    In Monte Carlo particle transport codes, it is often important to adjust reaction cross sections to reduce the variance of calculations of relatively rare events, in a technique known as non-analogous Monte Carlo. We present the theory and sample code for a Geant4 process which allows the cross section of a G4VDiscreteProcess to be scaled, while adjusting track weights so as to mitigate the effects of altered primary beam depletion induced by the cross section change. This makes it possible t...

  4. Generation of the library of neutron cross sections for the Record code of the Fuel Management System (FMS)

    International Nuclear Information System (INIS)

    Alonso V, G.; Hernandez L, H.

    1991-11-01

    On the basis of the library structure of the RECORD code a method to generate the neutron cross sections by means of the ENDF-B/IV database and the NJOY code has been developed. The obtained cross sections are compared with those of the current library which was processed using the ENDF-B/III version. (Author)

  5. Improvements on burnup chain model and group cross section library in the SRAC system

    International Nuclear Information System (INIS)

    Akie, Hiroshi; Okumura, Keisuke; Takano, Hideki; Ishiguro, Yukio; Kaneko, Kunio.

    1992-01-01

    Data and functions of the cell burnup calculation of the SRAC system were revised to improve mainly the accuracy of the burnup calculation of high conversion light water reactors (HCLWRs). New burnup chain models were developed in order to treat fission products (FPs) and actinide nuclides in detail. Group cross section library, SRACLIB-JENDL2, was generated based on JENDL-2 nuclear data file. In generating this library, emphasis was placed on FPs and actinides. Also revised were the data such as the average energy release per fission for various actinides. These improved data were verified by performing the burnup analysis of PWR spent fuels. Some new functions were added to the SRAC system for the convenience to yield macroscopic cross sections used in the core burnup process. (author)

  6. 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

  7. ARP: A PC-compatible scheme for generating ORIGEN-S cross section library

    International Nuclear Information System (INIS)

    Leal, L.C.; Hermann, O.W.; Parks, C.V.

    1995-01-01

    The SAS2H sequence of the SCALE code system has been widely used for treating problems related to the characterization of nuclear systems for disposal, storage, and shipment. The calculations, in general, consist of determining the isotope compositions of the different materials present in the problem as a function of time, which subsequently enable determination of the heat generation and radiation source terms. In the SAS2H scheme, time-dependent material concentrations are obtained using the ORIGEN-S code based on a point-depletion calculation that utilizes problem-dependent cross-section libraries generated by distinct codes of the SAS2H sequence. In this paper we will be concerned with the methodology utilized in the SAS2H control module to create cross-section libraries for point-depletion calculations with the ORIGEN-S code. A brief description of the SAS2H scheme will be given, and a new capability, the automatic rapid processing (ARP), for generating problem-dependent ORIGEN-S cross-section libraries will be presented. Use of ARP can enable execution of ORIGEN-S on a personal computer with identical accuracy to that obtained with SAS2H

  8. 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

  9. Validation of KENO V.a. and two cross-section libraries for criticality calculations of low-enriched uranium systems

    International Nuclear Information System (INIS)

    Easter, M.E.

    1985-07-01

    The SCALE code system, utilizing the Monte Carlo computer code KENO V.a, was employed to calculate 37 critical experiments. The critical assemblies had 235 U enrichments of 5% or less and cover a variety of geometries and materials. Values of k/sub eff/ were calculated using two different results using either of the cross-section libraries. The 16-energy-group Hansen-Roach and the 27-energy-group ENDF/B-IV cross-section libraries, available in SCALE, were used in this validation study, and both give good results for the experiments considered. It is concluded that the code and cross sections are adequate for low-enriched uranium systems and that reliable criticality safety calculations can be made for such systems provided the limits of validated applicability are not exceeded

  10. Cross sections needed for investigations into track phenomena and Monte-Carlo calculations

    International Nuclear Information System (INIS)

    Paretzke, H.G.

    1983-01-01

    Investigations into basic radiation action mechanisms as well as into applied radiation transport problems (e.g. electron microscopy) greatly benefit from detailed computer simulations of charged particle track structures in matter. The first and in fact most important and most difficult step in any such calculation is the derivation of reliable cross sections for the most relevant interaction processes in the material(s) under consideration. The second step in radiation transport calculations is the testing of results or intermediate results for quantitative or qualitative consistency with other experimental or theoretical information (e.g. yields, backscatter factors). This paper discusses the types of the most important collision cross sections for studies on track phenomena by detailed Monte-Carlo calculations, the necessary accuracy of such data and various means of consistency checks of calculated results. This will be done mainly with examples taken from radiation physics as applied to dosimetric and biological problems (i.e. to gaseous and condensed targets). 12 references, 8 figures

  11. 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)

  12. 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

  13. 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

  14. Organization of cross-section data in the Monte Carlo code SPARTAN

    International Nuclear Information System (INIS)

    Bending, R.C.

    1974-01-01

    The Monte Carlo code SPARTAN is a general-purpose code intended for neutron or gamma transport calculations. The code is designed to accept physics data from a number of external libraries (which may be used singly or in combination) and to use this data with as little alteration as possible. Data obtained from one or several libraries is placed in an interface file on magnetic tape or disk, using a general hierarchical structure which allows particular data items to be assessed in a straightforward way. The interface file, with or without additional data from cards, is regarded as a data source for the main Monte Carlo calculation. A summary of the functional forms, sampling distributions, and particle interaction laws which are available at present, and some of the mathematical methods used are described. 5 references. (U.S.)

  15. IAEA nuclear data for applications: Cross section standards and the reference input parameter library (RIPL)

    International Nuclear Information System (INIS)

    Capote Noy, Roberto; Nichols, Alan L.; Pronyaev, Vladimir G.

    2003-01-01

    develop a library of validated nuclear-model input parameters, referred to as the Reference Input Parameter Library (RIPL). The first stage of this work was initiated in 1994 and the second step began in 1998, both as IAEA CRPs. A consistent library of recommended nuclear theoretical input parameters is now available (RIPL-2) that includes a large amount of theoretical information suitable for nuclear reaction calculations, along with a number of computer codes for parameter retrieval and related calculations. A third further phase of this project has been recently initiated in order to extend the applicability of the RIPL library to cross sections for reactions on nuclei far from the line of stability, incident energies up to 200 MeV, and reactions induced by charged particles. (authors)

  16. A probability-conserving cross-section biasing mechanism for variance reduction in Monte Carlo particle transport calculations

    Energy Technology Data Exchange (ETDEWEB)

    Mendenhall, Marcus H., E-mail: marcus.h.mendenhall@vanderbilt.edu [Vanderbilt University, Department of Electrical Engineering, P.O. Box 351824B, Nashville, TN 37235 (United States); Weller, Robert A., E-mail: robert.a.weller@vanderbilt.edu [Vanderbilt University, Department of Electrical Engineering, P.O. Box 351824B, Nashville, TN 37235 (United States)

    2012-03-01

    In Monte Carlo particle transport codes, it is often important to adjust reaction cross-sections to reduce the variance of calculations of relatively rare events, in a technique known as non-analog Monte Carlo. We present the theory and sample code for a Geant4 process which allows the cross-section of a G4VDiscreteProcess to be scaled, while adjusting track weights so as to mitigate the effects of altered primary beam depletion induced by the cross-section change. This makes it possible to increase the cross-section of nuclear reactions by factors exceeding 10{sup 4} (in appropriate cases), without distorting the results of energy deposition calculations or coincidence rates. The procedure is also valid for bias factors less than unity, which is useful in problems that involve the computation of particle penetration deep into a target (e.g. atmospheric showers or shielding studies).

  17. A probability-conserving cross-section biasing mechanism for variance reduction in Monte Carlo particle transport calculations

    International Nuclear Information System (INIS)

    Mendenhall, Marcus H.; Weller, Robert A.

    2012-01-01

    In Monte Carlo particle transport codes, it is often important to adjust reaction cross-sections to reduce the variance of calculations of relatively rare events, in a technique known as non-analog Monte Carlo. We present the theory and sample code for a Geant4 process which allows the cross-section of a G4VDiscreteProcess to be scaled, while adjusting track weights so as to mitigate the effects of altered primary beam depletion induced by the cross-section change. This makes it possible to increase the cross-section of nuclear reactions by factors exceeding 10 4 (in appropriate cases), without distorting the results of energy deposition calculations or coincidence rates. The procedure is also valid for bias factors less than unity, which is useful in problems that involve the computation of particle penetration deep into a target (e.g. atmospheric showers or shielding studies).

  18. Transport methods: general. 2. Monte Carlo Particle Transport in Media with Exponentially Varying Time-Dependent Cross Sections

    International Nuclear Information System (INIS)

    Brown, Forrest B.; Martin, William R.

    2001-01-01

    We have investigated Monte Carlo schemes for analyzing particle transport through media with exponentially varying time-dependent cross sections. For such media, the cross sections are represented in the form Σ(t) = Σ 0 e -at (1) or equivalently as Σ(x) = Σ 0 e -bx (2) where b = av and v is the particle speed. For the following discussion, the parameters a and b may be either positive, for exponentially decreasing cross sections, or negative, for exponentially increasing cross sections. For most time-dependent Monte Carlo applications, the time and spatial variations of the cross-section data are handled by means of a stepwise procedure, holding the cross sections constant for each region over a small time interval Δt, performing the Monte Carlo random walk over the interval Δt, updating the cross sections, and then repeating for a series of time intervals. Continuously varying spatial- or time-dependent cross sections can be treated in a rigorous Monte Carlo fashion using delta-tracking, but inefficiencies may arise if the range of cross-section variation is large. In this paper, we present a new method for sampling collision distances directly for cross sections that vary exponentially in space or time. The method is exact and efficient and has direct application to Monte Carlo radiation transport methods. To verify that the probability density function (PDF) is correct and that the random-sampling procedure yields correct results, numerical experiments were performed using a one-dimensional Monte Carlo code. The physical problem consisted of a beam source impinging on a purely absorbing infinite slab, with a slab thickness of 1 cm and Σ 0 = 1 cm -1 . Monte Carlo calculations with 10 000 particles were run for a range of the exponential parameter b from -5 to +20 cm -1 . Two separate Monte Carlo calculations were run for each choice of b, a continuously varying case using the random-sampling procedures described earlier, and a 'conventional' case where the

  19. Implementing of AMPX-II system for a univac computer neutron cross-section libraries

    International Nuclear Information System (INIS)

    Sancho, J.; Verdu, G.; Serradell, V.

    1984-01-01

    The AMPX-II system, developed at ORNL, is constituted by a modular set of computer programs, for generation and handling of several nuclear data libraries. The processing starts from ENDF/B library. Along this paper, we refer mainly to the modules related with neutron cross section libraries: master, working and weighted. These modules have been implemented recently for a UNIVAC 1100/60 computer in the Universidad Politecnica de Valencia (Spain). In order to run the programs in that machine it has been necessary to introduce a number of modifications into their programing structure. The main difficulties found in this work and the need of verification for the new versions are also pointed out. We also refer to the results obtained from the execution of a set of little sample problems. (author)

  20. Monte Carlo calculation of the cross-section of single event upset induced by 14MeV neutrons

    International Nuclear Information System (INIS)

    Li, H.; Deng, J.Y.; Chang, D.M.

    2005-01-01

    High-density static random access memory may experience single event upsets (SEU) in neutron environments. We present a new method to calculate the SEU cross-section. Our method is based on explicit generation and transport of the secondary reaction products and detailed accounting for energy loss by ionization. Instead of simulating the behavior of the circuit, we use the Monte Carlo method to simulate the process of energy deposition in sensitive volumes. Thus, we do not need to know details about the circuit. We only need a reasonable guess for the size of the sensitive volumes. In the Monte Carlo simulation, the cross-section of SEU induced by 14MeV neutrons is calculated. We can see that the Monte Carlo simulation not only can provide a new method to calculate SEU cross-section, but also can give a detailed description about random process of the SEU

  1. Specifications for adjusted cross section and covariance libraries based upon CSEWG fast reactor and dosimetry benchmarks

    International Nuclear Information System (INIS)

    Weisbin, C.R.; Marable, J.H.; Collins, P.J.; Cowan, C.L.; Peelle, R.W.; Salvatores, M.

    1979-06-01

    The present work proposes a specific plan of cross section library adjustment for fast reactor core physics analysis using information from fast reactor and dosimetry integral experiments and from differential data evaluations. This detailed exposition of the proposed approach is intended mainly to elicit review and criticism from scientists and engineers in the research, development, and design fields. This major attempt to develop useful adjusted libraries is based on the established benchmark integral data, accurate and well documented analysis techniques, sensitivities, and quantified uncertainties for nuclear data, integral experiment measurements, and calculational methodology. The adjustments to be obtained using these specifications are intended to produce an overall improvement in the least-squares sense in the quality of the data libraries, so that calculations of other similar systems using the adjusted data base with any credible method will produce results without much data-related bias. The adjustments obtained should provide specific recommendations to the data evaluation program to be weighed in the light of newer measurements, and also a vehicle for observing how the evaluation process is converging. This report specifies the calculational methodology to be used, the integral experiments to be employed initially, and the methods and integral experiment biases and uncertainties to be used. The sources of sensitivity coefficients, as well as the cross sections to be adjusted, are detailed. The formulae for sensitivity coefficients for fission spectral parameters are developed. A mathematical formulation of the least-square adjustment problem is given including biases and uncertainties in methods

  2. Testing of the IRDF-90 cross-section library in benchmark neutron spectra

    International Nuclear Information System (INIS)

    Nolthenius, H.J.; Zsolnay, E.M.; Szondi, E.J.

    1993-09-01

    The new version of the International Reactor Dosimetry File IRDF-90 (called ''Version April 1993'') has been tested by calculation of average cross-sections and their uncertainties in a coarse three energy group structure and by neutron spectrum adjustments in reference neutron spectra. This paper presents the results obtained and compares them with the corresponding ones of the old IRDF-85 and with the data of the Nuclear Data Guide for Reactor Neutron Metrology. The applicability of the new library in the field of neutron metrology is discussed. (orig.)

  3. 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

  4. VITAMIN E: a multipurpose ENDF/B-V coupled neutron-gamma cross section library

    International Nuclear Information System (INIS)

    Barhen, J.; Cacuci, D.G.; Ford, W.E. III; Roussin, R.W.; Wagschal, J.J.; Weisbin, C.R.; White, J.E.; Wright, R.Q.

    1979-01-01

    The US Department of Energy Office of Fusion Energy and the Division of Reactor Research and Technology jointly sponsored the development of a coupled fine-group cross section library (VITAMIN-C). The experience gained in the generation, validation, and utilization of the VITAMIN-C library along with its broad range of applicability has led to the request for updating this data set using ENDF/B-V. Additional support in this regard has been provided by the Defense Nuclear Agency (DNA) and by EPRI in support of weapons analyses and light water reactor shielding and dosimetry problems, respectively. The rationale for developing the multipurpose ENDF/B-V-based VITAMIN-E library is presented, with special emphasis on new models used in the data generation algorithms. The library specifications and testing procedures are also discussed in detail. The distribution of the VITAMIN-E library is currently subject to the same restrictions as the distribution of the ENDF/B-V data. 2 tables

  5. PACER: a Monte Carlo time-dependent spectrum program for generating few-group diffusion-theory cross sections

    International Nuclear Information System (INIS)

    Candelore, N.R.; Kerrick, W.E.; Johnson, E.G.; Gast, R.C.; Dei, D.E.; Fields, D.L.

    1982-09-01

    The PACER Monte Carlo program for the CDC-7600 performs fixed source or eigenvalue calculations of spatially dependent neutron spectra in rod-lattice geometries. The neutron flux solution is used to produce few group, flux-weighted cross sections spatially averaged over edit regions. In general, PACER provides environmentally dependent flux-weighted few group microscopic cross sections which can be made time (depletion) dependent. These cross sections can be written in a standard POX output file format. To minimize computer storage requirements, PACER allows separate spectrum and edit options. PACER also calculates an explicit (n, 2n) cross section. The PACER geometry allows multiple rod arrays with axial detail. This report provides details of the neutron kinematics and the input required

  6. ROSFOND based heating-damage cross sections sub-library: Preliminary uncertainty assessment

    International Nuclear Information System (INIS)

    Sinitsa, V.V.

    2016-01-01

    The accuracy of radiation damage calculations for the most important LWR component, the reactor pressure vessel (RPV), directly linked with the RPV End-of-Life (EoL) prediction which is in its turn connected with fundamental nuclear safety aspects and relevant economic impacts. In this connection, for nearly ten years the ENEA-Bologna Nuclear Data Group conducts the nuclear data processing and validation activities addressed to update the specialized broad-group coupled neutron/photon working cross section libraries for shielding and radiation damage calculations through NJOY and Bologna revised version of SCAMPI data processing systems. A number of working group-wise data libraries has been prepared and transferred to the ENEA Data Bank for dissemination. Several years ago the NRC ”Kurchatov Institute” has reset the GRUCON project, originally designed to provide group constants for fast nuclear reactor calculations [12], with aim to expand its application area and to use in the WWER safety tasks, in particular, in the RPV radiation damage analyses. By means of updated GRUCON and NJOY-99 processing codes, and calculation procedure, developed in the NDG of ENEA Bologna, a sample of kerma&damage energy point-wise data sub-libraries from different evaluated data libraries has been generated. On the base of this sample, the quantitative assessment of kerma/dpa data precision in the RPV calculations is obtained

  7. 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

  8. ZZ ENDL82, Evaluated Charged Particle, Neutron, Photon Cross-Section Library

    International Nuclear Information System (INIS)

    2001-01-01

    Description of program or function: - Format: Described in the manual; - Number of groups: (energies between 100 eV and 100 MeV); - Nuclides: 94 (Z 1 to 99); - Origin: LLNL Evaluated Nuclear Data Library. ENDL82 is a collection of evaluated data for neutron-induced reactions, photon interactions with matter, and charged-particle-induced reactions. It is maintained in a computer-oriented system. All interpolable quantities for neutron-induced reactions are presented so that linear interpolation between successive entries yields values that are consistent with stated experimental errors, where experiments exist, or that adhere to an assumed law, such as 1/v energy dependence, within a small fraction (typically 1%). In the case of an assumed energy-dependence law for cross sections, this is accomplished by creating a large number of (energy, cross section) pairs by computer and subsequently thinning the points to a specified accuracy, using the subroutine THINER. All angular distributions are differential probabilities normalized to an integral of unity over the cosine of the scattering angle. All energy distributions of secondary particles are presented as normalized Legendre polynomial representations. The linear interpolation will construct an acceptable angular distribution at an intermediate energy

  9. 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.

  10. 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.

  11. 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.

  12. BARC 75 - A 75 group neutron-photon coupled cross-section library with P5- anisotropic scattering matrices

    International Nuclear Information System (INIS)

    Garg, S.B.

    1990-01-01

    A 75 group neutron-photon coupled cross-section library has been developed for 42 reactor nuclides utilizing the basic cross-section files - ENDF/B-IV for neutrons and DLC-7F for photons. 50 neutron energy groups and gamma energy groups are included in this library which should be well suited to carry out safety, shielding and core physics studies of nuclear reactors based on fission or fusion processes. This library is also adequate for oil logging and mineral exploration investigations. (author). 11 refs., 3 tabs

  13. Comparison of CASMO and NESSEL few group cross section libraries and their usage in DYN3D

    International Nuclear Information System (INIS)

    Kuchin, A.; Ovdiyenko, Y.; Loetsch, T.

    2007-01-01

    This work presents comparative analysis of two group diffusion cross section libraries which were generated by NESSEL-4 and CASMO-4 lattice codes. Diffusion parameters were calculated for VVER-1000 fuel assemblies with stainless steel spacing grids and guiding tubes. These cross section sets were introduced into reactor core code DYN3D and tested on the base of real reactor core states. In this case operation data of the first three fuel cycles of 6-th unit of Zaporizhzhya NPP were used

  14. A CUMULATIVE MIGRATION METHOD FOR COMPUTING RIGOROUS TRANSPORT CROSS SECTIONS AND DIFFUSION COEFFICIENTS FOR LWR LATTICES WITH MONTE CARLO

    Energy Technology Data Exchange (ETDEWEB)

    Zhaoyuan Liu; Kord Smith; Benoit Forget; Javier Ortensi

    2016-05-01

    A new method for computing homogenized assembly neutron transport cross sections and dif- fusion coefficients that is both rigorous and computationally efficient is proposed in this paper. In the limit of a homogeneous hydrogen slab, the new method is equivalent to the long-used, and only-recently-published CASMO transport method. The rigorous method is used to demonstrate the sources of inaccuracy in the commonly applied “out-scatter” transport correction. It is also demonstrated that the newly developed method is directly applicable to lattice calculations per- formed by Monte Carlo and is capable of computing rigorous homogenized transport cross sections for arbitrarily heterogeneous lattices. Comparisons of several common transport cross section ap- proximations are presented for a simple problem of infinite medium hydrogen. The new method has also been applied in computing 2-group diffusion data for an actual PWR lattice from BEAVRS benchmark.

  15. 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)

  16. ZZ CANDULIB-AECL, Burnup-Dependent ORIGEN-S Cross-Section Libraries for Candu Reactor Fuels

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Historical background and information: - 28-element fuel cross-section library: Format: Designed for use with the ORIGEN-S isotope generation and depletion code. Materials: Co, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Ta, W, Re, Au, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDSF, ENDF/B-IV, -V and -VI Weighting spectrum: determined using WIMS-AECL transport code. - 37-element fuel cross-section library: Format: Designed for use with the ORIGEN-S isotope generation and depletion code. Materials: Co, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Ta, W, Re, Au, Th, Pa, U, Np, Pu, Am, Cm. Origin: ENDSF, ENDF/B-IV, -V and -VI Weighting spectrum: determined using WIMS-AECL transport code. In 1995, updated ORIGEN-S cross-section libraries were created as part of a program to upgrade and standardize the computer codes and nuclear data employed for used fuel characterization. This effort was funded through collaboration between Atomic Energy of Canada Limited and the Canadian Nuclear Power Utilities, under the Candu Owners Group (COG). The updated cross sections were generated using the WIMS-AECL lattice code and ENDF/B-V and -VI based data to provide cross section consistency with reactor physics codes. 2 - Application of the data: The libraries in this data collection are designed for characterising used fuel from Candu pressurized heavy water reactors. Two libraries are provided: one for the standard 28-element fuel bundle design, the other for the 37-element fuel bundle design. The libraries were generated for typical reactor operating conditions. The libraries are designed for use with the ORIGEN-S isotope generation and depletion code. 3 - Source and scope of data: The Candu libraries are updated with cross sections from a variety of different sources. Capture

  17. 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)

  18. Experience in developing and using the VITAMIN-C 171-neutron, 36-gamma-ray group cross-section library

    International Nuclear Information System (INIS)

    Roussin, R.W.; Weisbin, C.R.; White, J.E.; Wright, R.Q.; Greene, N.M.; Ford, W.E. III; Wright, J.B.; Diggs, B.R.

    1978-01-01

    The Department of Energy (DOE) Division of Magnetic Fusion Energy (DMFE) and Reactor Research and Technology (DRRT) jointly sponsored the development of a coupled, fine-group cross-section library. The 171-neutron, 36-gamma-ray group library is intended to be applicable to fusion reactor neutronics and LMFBR core and shield analysis. Versions of the library are available from the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory in both AMPX and CCCC formats. Computer codes for energy group collapsing, interpolation on Bondarenko factors for resonance self-shielding and temperature corrections, and various other useful data manipulations are available. The experience gained in the utilization of this library is discussed. Indications are that this venture, which is designed to allow users to derive problem-dependent cross sections from a fine-group master library, has been a success

  19. A Monte Carlo calculation of the pionium break-up probability with different sets of pionium target cross sections

    International Nuclear Information System (INIS)

    Santamarina, C; Schumann, M; Afanasyev, L G; Heim, T

    2003-01-01

    Chiral perturbation theory predicts the lifetime of pionium, a hydrogen-like π + π - atom, to better than 3% precision. The goal of the DIRAC experiment at CERN is to obtain and check this value experimentally by measuring the break-up probability of pionium in a target. In order to accurately measure the lifetime one needs to know the relationship between the break-up probability and the lifetime to 1% accuracy. We have obtained this dependence by modelling the evolution of pionic atoms in the target using Monte Carlo methods. The model relies on the computation of the pionium-target-atom interaction cross sections. Three different sets of pionium-target cross sections with varying degrees of complexity were used: from the simplest first-order Born approximation involving only the electrostatic interaction to a more advanced approach, taking into account multiphoton exchanges and relativistic effects. We conclude that, in order to obtain the pionium lifetime to 1% accuracy from the break-up probability, the pionium-target cross sections must be known with the same accuracy for the low excited bound states of the pionic atom. This result has been achieved, for low Z targets, with the two most precise cross section sets. For large Z targets only the set accounting for multiphoton exchange satisfies the condition

  20. IAEA consultants' meeting on selection of evaluations for the FENDL/A-2 activation cross section library. Summary report

    International Nuclear Information System (INIS)

    Pashchenko, A.B.

    1996-02-01

    FENDL/A is a nuclear data library of neutron activation cross-sections prepared for use in nuclear fusion reactor development. The present report contains recommendations for the creation of a second improved version of FENDL/A, including a list of 400 neutron reactions to be considered with priority. (author)

  1. Library of neutron cross sections of the Thermos code; Biblioteca de secciones eficaces de neutrones del codigo Thermos

    Energy Technology Data Exchange (ETDEWEB)

    Alonso V, G; Hernandez L, H [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1991-10-15

    The present work is the complement of the IT.SN/DFR-017 report in which the structure and the generation of the library of the Thermos code is described. In this report the comparison among the values of the cross sections that has the current library of the Thermos code and those generated by means of the ENDF-B/NJOY it is shown. (Author)

  2. Neutron secondary-particle production cross sections and their incorporation into Monte-Carlo transport codes

    International Nuclear Information System (INIS)

    Brenner, D.J.; Prael, R.E.; Little, R.C.

    1987-01-01

    Realistic simulations of the passage of fast neutrons through tissue require a large quantity of cross-sectional data. What are needed are differential (in particle type, energy and angle) cross sections. A computer code is described which produces such spectra for neutrons above ∼14 MeV incident on light nuclei such as carbon and oxygen. Comparisons have been made with experimental measurements of double-differential secondary charged-particle production on carbon and oxygen at energies from 27 to 60 MeV; they indicate that the model is adequate in this energy range. In order to utilize fully the results of these calculations, they should be incorporated into a neutron transport code. This requires defining a generalized format for describing charged-particle production, putting the calculated results in this format, interfacing the neutron transport code with these data, and charged-particle transport. The design and development of such a program is described. 13 refs., 3 figs

  3. Differences between cross-section libraries for neutron dosimetry; Diferencas entre bibliotecas de secoes de choque para dosimetria de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Tardelli, T.C.; Stecher, L.C.; Coelho, T.S.; Castro, V.A. De; Cavalieri, T.A.; Menzel, F.; Giarola, R.S.; Domingos, D.B.; Yoriyaz, H., E-mail: tiago.tardelli@gmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Engenharia Nuclear

    2013-08-15

    Absorbed dose calculations depend on a consistent set of nuclear data used in simulations in computer codes. Nuclear data are stored in libraries, however, the information available about the differences in dose caused by different libraries are rare. The libraries are processed by a computer system to be able to be used by a radiation transport code. One of the systems capable of processing nuclear data is the NJOY system. The objective of this study is to evaluate the nuclear data libraries for neutrons available in the literature, and to quantify the differences in absorbed dose obtained using the libraries JENDL 4.0, JEFF 3.3.1 and ENDF/B.VII. The absorbed dose calculation was performed on a simple geometric model, as spheres, and in anthropomorphic model of the human body based on the ICRP-110 for neutron transport simulation using the MCNP5 code. The results were compared with literature data. The results obtained with cross sections from the libraries JEFF and ENDF/B.VII have shown to be identical in most cases, except for one case where the difference has exceeded 10%. The results obtained with JENDL library has shown to be considerably different in most cases comparing to other two libraries. Some differences were over 200%. The dose calculations showed differences between the libraries, which is justified by differences in the cross sections. It has been observed that the cross sections values of certain nuclides assume quite different values in different libraries. These differences in turn cause considerable differences in dose calculations. (author)

  4. Preparation of lumped fission product (FP) cross sections for a multigroup library

    International Nuclear Information System (INIS)

    Ono, S.; Corcuera, R.P.

    1984-01-01

    A method for the calculation of lumped Fission Product (FP) cross sections has been developed. The group constants fo each nuclide are generated by NJOY code, based on ENDF/B-V data. In this first version, cross section of 28 nuclides are lumped for typical characteristics of Binary Breeder Reactor (BBR). One energy group calculations are made for a 1000 MWe fast reactor to verify the influence of burnup, number of FP and fuel composition on the lumped fission product cross sections. (Author) [pt

  5. 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

  6. Generation of ENDF/B-IV based 35 group neutron cross-section library and its application in criticality studies

    International Nuclear Information System (INIS)

    Garg, S.B.; Sinha, A.

    1985-01-01

    A 35 group cross-section library with P/sub 3/-anisotropic scattering matrices and resonance self-shielding factors has been generated from the basic ENDF/B-IV cross-section files for 57 elements. This library covers the neutron energy range from 0.005 ev to 15 MeV and is well suited for the neutronics and safety analysis of fission, fusion and hybrid systems. The library is contained in two well known files, namely, ISOTXS and BRKOXS. In order to test the efficacy of this library and to bring out the importance of resonance self-shielding, a few selected fast critical assemblies representing large dilute oxide and carbide fueled uranium and plutonium based systems have been analysed. These assemblies include ZPPR/sub 2/, ZPR-3-48, ZPR-3-53, ZPR-6-6A, ZPR-6-7, ZPR-9-31 and ZEBRA-2 and are amongst those recommended by the US Nuclear Data Evaluation Working Group for testing the accuracy of cross-sections. The evaluated multiplication constants of these assemblies compare favourably with those calculated by others

  7. The effective differential cross section for elastic scattering of electrons by atoms and its use for Monte Carlo simulation of electron passage through matter

    International Nuclear Information System (INIS)

    Sheikin, E G

    2010-01-01

    The effective differential cross section (DCS) for elastic scattering of electrons by atoms is proposed that reproduces known energy dependences for the first and second transport cross sections but provides a total elastic cross section that is significantly small compared with the known energy dependences. The number of elastic collisions of electrons in matter when using the effective DCS in Monte Carlo simulations is significantly lower than that when using the real DCS. The results of our Monte Carlo simulation of electron propagation in aluminium using the proposed DCS are in good agreement with experimental data.

  8. 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)

  9. Development and benchmark of high energy continuous-energy neutron cross Section library HENDL-ADS/MC

    International Nuclear Information System (INIS)

    Chen Chong; Wang Minghuang; Zou Jun; Xu Dezheng; Zeng Qin

    2012-01-01

    The ADS (accelerator driven sub-critical system) has great energy spans, complex energy spectrum structures and strong physical effects. Hence, the existing nuclear data libraries can't fully meet the needs of nuclear analysis in ADS. In order to do nuclear analysis for ADS system, a point-wise data library HENDL-ADS/MC (hybrid evaluated nuclear data library) was produced by FDS team. Meanwhile, to test the availability and reliability of the HENDL-ADS/MC data library, a series of shielding and critical safety benchmarks were performed. To validate and qualify the reliability of the high-energy cross section for HENDL-ADS/MC library further, a series of high neutronics integral experiments have been performed. The testing results confirm the accuracy and reliability of HENDL-ADS/MC. (authors)

  10. Integral test for Np237 and Am241 cross sections in JENDL, ENDF and JEF libraries

    International Nuclear Information System (INIS)

    Iwasaki, Tomohiko; Unesaki, Hironobu; Kitada, Takanori

    2002-01-01

    Experiments using Kyoto University critical assembly (KUCA) were performed for measuring the capture and fission reaction rates of 237 Np and 241 Am. A back-to-back fission chamber was employed for the measurement of the fission rate of 237 Np and 241 Am relative to 235 U. The capture rate of 237 Np relative to 197 Au was measured by using activation method. Eleven cores, of which the spectra were changed systematically, were mocked up for the present measurements. Five cores among the eleven were utilized for the fission reaction rate measurement. The experiment was analyzed using the Monte Carlo code MVP, the transport code TWOTRAN and the diffusion code CITATION using the libraries based on JENDL3.2, ENDF/B-VI and JEF2.2. As the results, for 237 Np, JENDL3.2 showed good agreement for both capture and fission. However, for the fission rate of 241 Am, JENDL3.2 underestimates 15-20%. On the other hand, ENDF/B-VI and JEF2.2 show different C/Es for 237 Np and 241 Am. (author)

  11. The needs for program and cross-section library improvement in calculation of neutron-induced activity inventories

    International Nuclear Information System (INIS)

    Yavshitz, S.G.; Rubchenya, V.A.; Rimski-Korsakov, A.A.

    1993-01-01

    The authors demonstrate the possibility of an approach to evaluate the radioactive inventory - induced activity of structural materials and surface contamination of reactor components, that will fit well into ORIGEN code structure and could be used on a modest PC directly on the decommissioning site. This approach would also require only one well tested set of pre-calculated and adjusted by experiment cross-section libraries (averaged by typical neutron spectra outside the reactor core). 15 refs, 1 fig

  12. Generation of one energy group cross section library with MC2 computer code

    International Nuclear Information System (INIS)

    Cunha Menezes Filho, A. da; Souza, A.L. de.

    1982-01-01

    One group temperature dependent cross sections are generated via MC 2 for Pu-242, Ni-58, Fe-56, U-235, U-238, Pu-239, Pu-240, Pu-241, Be-9 e Th-232. The influence of the buckling and the weighting functions is studied throught calculations of an important integral parameter: the critical radius. (author) [pt

  13. 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

  14. Advanced Monte Carlo procedure for the IFMIF d-Li neutron source term based on evaluated cross section data

    International Nuclear Information System (INIS)

    Simakov, S.P.; Fischer, U.; Moellendorff, U. von; Schmuck, I.; Konobeev, A.Yu.; Korovin, Yu.A.; Pereslavtsev, P.

    2002-01-01

    A newly developed computational procedure is presented for the generation of d-Li source neutrons in Monte Carlo transport calculations based on the use of evaluated double-differential d+ 6,7 Li cross section data. A new code M c DeLicious was developed as an extension to MCNP4C to enable neutronics design calculations for the d-Li based IFMIF neutron source making use of the evaluated deuteron data files. The M c DeLicious code was checked against available experimental data and calculation results of M c DeLi and MCNPX, both of which use built-in analytical models for the Li(d, xn) reaction. It is shown that M c DeLicious along with newly evaluated d+ 6,7 Li data is superior in predicting the characteristics of the d-Li neutron source. As this approach makes use of tabulated Li(d, xn) cross sections, the accuracy of the IFMIF d-Li neutron source term can be steadily improved with more advanced and validated data

  15. Advanced Monte Carlo procedure for the IFMIF d-Li neutron source term based on evaluated cross section data

    CERN Document Server

    Simakov, S P; Moellendorff, U V; Schmuck, I; Konobeev, A Y; Korovin, Y A; Pereslavtsev, P

    2002-01-01

    A newly developed computational procedure is presented for the generation of d-Li source neutrons in Monte Carlo transport calculations based on the use of evaluated double-differential d+ sup 6 sup , sup 7 Li cross section data. A new code M sup c DeLicious was developed as an extension to MCNP4C to enable neutronics design calculations for the d-Li based IFMIF neutron source making use of the evaluated deuteron data files. The M sup c DeLicious code was checked against available experimental data and calculation results of M sup c DeLi and MCNPX, both of which use built-in analytical models for the Li(d, xn) reaction. It is shown that M sup c DeLicious along with newly evaluated d+ sup 6 sup , sup 7 Li data is superior in predicting the characteristics of the d-Li neutron source. As this approach makes use of tabulated Li(d, xn) cross sections, the accuracy of the IFMIF d-Li neutron source term can be steadily improved with more advanced and validated data.

  16. Nuclear data, cross section libraries and their application in nuclear technology

    International Nuclear Information System (INIS)

    1985-01-01

    These proceedings contain the articles presented at the named seminar. The articles deal with evaluated nuclear data libraries, computer codes for neutron transport and reactor calculations using nuclear data libraries, and the application of nuclear data libraries for the calculation of the interaction of neutron beams with materials. (HSI)

  17. Modernization of Cross Section Library for VVER-1000 Type Reactors Internals and Pressure Vessel Dosimetry

    Directory of Open Access Journals (Sweden)

    Voloschenko Andrey

    2016-01-01

    Full Text Available The broad-group library BGL1000_B7 for neutron and gamma transport calculations in VVER-1000 internals, RPV and shielding was carried out on a base of fine-group library v7-200n47g from SCALE-6 system. The comparison of the library BGL1000_B7 with the library v7-200n47g and the library BGL1000 (the latter is using for VVER-1000 calculations is demonstrated on several calculation and experimental tests.

  18. ZZ FCXSEC, Coupled Cross-Section Library for Shielding from VITAMIN-C in AMPX, ANISN Format

    International Nuclear Information System (INIS)

    1985-01-01

    1 - Description of problem or function: Format: (a) and (b) AMPX, (c) and (d) ANISN; Number of groups: (a) Fine-group 171 neutron and 36 gamma-ray; (b) Broad-group 22 neutron and 21 gamma-ray; (c) Broad-group microscopic (22n-21 gamma); (d) Broad-group macroscopic; Nuclides: Mixtures: H 2 O, Borated water, Concrete, D 2 O, Lithium hydride, Boral, Dry air, Nitric acid, Uranium dioxide, S 3 0 4 , UF 6 TBP in dodecane, Sm 2 O 3 , Eu 2 O 3 , Gd 2 O 3 , Gd(NO 3 ) 3 in water, WB2, Spen fuel oxide, Thorium oxide, Uranium metal, Silver zeolite. Individual materials: C, Na, Al, Fe, Zircaloy, Cd Nb, Mo, Pb, Be, Ti, V, Mn, Co, Cu, Sn, Ta. Origin: VITAMIN-C; Weighting spectrum: From 1.1109+5 eV to 1.7333+7 eV → 239 Pu thermal fission; From 4.1399-1 eV to 1.1109+5 eV → 1/E; From 1.0000-5 eV to 4.1399-1 eV → Maxwellian. FSXSEC is a collection of cross section libraries to be used for nuclear fuel cycle shielding calculations, generated from the pseudo-composition-independent VITAMIN-C cross section library: (a) A composition-dependent self-shielded fine-group library with 171 neutron groups and 36 gamma groups, and a broad-group library with 22 neutron and 21 gamma groups for AMPX. (b) A broad-group microscopic and a broad-group macroscopic library in ANISN format. 2 - Method of solution: To generate library (a), AMPX modules BONAMI, CHOX, and MALOCS were used. To generate library (b), AMPX modules NITAWL and AXMIX were used

  19. Comparison of CASMO and NESSEL few group cross section libraries and their usage in DYN3D

    International Nuclear Information System (INIS)

    Kuchin, A.; Ovdiyenko, Y.; Loetsch, T.

    2007-01-01

    This work presents comparative analysis of two group diffusion cross sections libraries which were generated by NESSEL-4 and CASMO-4 lattice codes. Diffusion parameters were calculated for WWER-1000 fuel assemblies with stainless steel spacing grids and guiding tubes. These cross section sets were introduced into reactor core code DYN3D and tested on the base of real reactor core states. In this case operation data of the first three fuel cycles of sixth unit of Zaporizhzhya NPP were used. The work was performed in the framework of the order BMU SR 2511 - 862 500/09, UA-2575. The report describes the opinion and view of the contractor - TUV ENERGIE CONSULT - and does not necessarily represent the opinion of the ordering party - Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (Authors)

  20. Two-level MOC calculation scheme in APOLLO2 for cross-section library generation for LWR hexagonal assemblies

    International Nuclear Information System (INIS)

    Petrov, Nikolay; Todorova, Galina; Kolev, Nikola; Damian, Frederic

    2011-01-01

    The accurate and efficient MOC calculation scheme in APOLLO2, developed by CEA for generating multi-parameterized cross-section libraries for PWR assemblies, has been adapted to hexagonal assemblies. The neutronic part of this scheme is based on a two-level calculation methodology. At the first level, a multi-cell method is used in 281 energy groups for cross-section definition and self-shielding. At the second level, precise MOC calculations are performed in a collapsed energy mesh (30-40 groups). In this paper, the application and validation of the two-level scheme for hexagonal assemblies is described. Solutions for a VVER assembly are compared with TRIPOLI4® calculations and direct 281g MOC solutions. The results show that the accuracy is close to that of the 281g MOC calculation while the CPU time is substantially reduced. Compared to the multi-cell method, the accuracy is markedly improved. (author)

  1. Design and producing of fine-group cross section library HENDL3.0/FG for subcritical system

    International Nuclear Information System (INIS)

    Zou, J.; Zeng, Q.; Xu, D.; Hu, L.; Long, P.

    2012-01-01

    To improve the accuracy of the neutron analyses for subcritical system with thermal fission blanket, a coupled neutron and photon (315 n + 42γ) fine-group cross section library HENDL3.0/FG based on ENDF/B-VII, JEFF3.1 and JENDL3.3 was produced by FDS team. In order to test the availability and reliability of the HENDL3.0/FG data library, shielding and critical safety benchmarks were performed with VisualBUS code. The testing results indicated that the discrepancy between calculation and experimental values of nuclear parameters fell in a reasonable range. It showed that the nuclear data library had accuracy and availability. (authors)

  2. 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

  3. The activation cross section library UKACT1 and the inventory code FISPACT

    International Nuclear Information System (INIS)

    Forrest, R.A.

    1989-01-01

    The UK activation library for fusion applications, UKACT1, supersedes the existing UKCTRIIIA library. It contains neutron induced reaction data for 8719 reactions on 625 target nuclides. The library is used by the inventory code FISPACT which is a modified version of the existing code FISPIN. A library of decay information for all the 1314 nuclides involved is also required for calculations and this is also briefly described. UKACT1 will be used for irradiation calculations and as the starting point for a new version which will contain improved data for the most important reactions. These will be identified using the sensitivity subroutine in FISPACT. 16 refs, 1 fig., 2 tabs

  4. POINT 2012: ENDF/B-VII.1 Final Temperature Dependent Cross Section Library

    International Nuclear Information System (INIS)

    Cullen, D.E.

    2012-01-01

    This report is one in the series of 'POINT' reports that over the years have presented temperature dependent cross sections for the then current version of ENDF/B [R1]. In each case I have used my personal computer at home and publicly available data and codes: (1) publicly available nuclear data (the current ENDF/B data, available on-line at the National Nuclear Data Center, Brookhaven National Laboratory, http://www.nndc.bnl.gov/) and, (2) publicly available computer codes (the current PREPRO codes, available on-line at the Nuclear Data Section, IAEA, Vienna, Austria, http://www-nds.iaea.or.at/ndspub/endf/prepro/) and, (3) My own personal computer located in my home. I have used these in combination to produce the temperature dependent cross sections used in applications and described in this report. I should mention that today anyone with a personal computer can produce these results: by its very nature I consider this data to be born in the public domain.

  5. POINT 2012: ENDF/B-VII.1 Final Temperature Dependent Cross Section Library

    Energy Technology Data Exchange (ETDEWEB)

    Cullen, D E

    2012-02-26

    This report is one in the series of 'POINT' reports that over the years have presented temperature dependent cross sections for the then current version of ENDF/B [R1]. In each case I have used my personal computer at home and publicly available data and codes: (1) publicly available nuclear data (the current ENDF/B data, available on-line at the National Nuclear Data Center, Brookhaven National Laboratory, http://www.nndc.bnl.gov/) and, (2) publicly available computer codes (the current PREPRO codes, available on-line at the Nuclear Data Section, IAEA, Vienna, Austria, http://www-nds.iaea.or.at/ndspub/endf/prepro/) and, (3) My own personal computer located in my home. I have used these in combination to produce the temperature dependent cross sections used in applications and described in this report. I should mention that today anyone with a personal computer can produce these results: by its very nature I consider this data to be born in the public domain.

  6. Calculation of atom displacement cross section for structure material

    International Nuclear Information System (INIS)

    Liu Ping; Xu Yiping

    2015-01-01

    The neutron radiation damage in material is an important consideration of the reactor design. The radiation damage of materials mainly comes from atom displacements of crystal structure materials. The reaction cross sections of charged particles, cross sections of displacements per atom (DPA) and KERMA are the basis of radiation damage calculation. In order to study the differences of DPA cross sections with different codes and different evaluated nuclear data libraries, the DPA cross sections for structure materials were calculated with UNF and NJOY codes, and the comparisons of results were given. The DPA cross sections from different evaluated nuclear data libraries were compared. And the comparison of DPA cross sections between NJOY and Monte Carlo codes was also done. The results show that the differences among these evaluated nuclear data libraries exist. (authors)

  7. Development of the adjusted nuclear cross-section library based on JENDL-3.2 for large FBR

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Ishikawa, Makoto; Numata, Kazuyuki

    1999-04-01

    JNC (and PNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were reflected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In 1991, the adjusted library based on JENDL-2, JFS-3-J2 (ADJ91R), was developed, and it has been used on the design research for FBR. As an evaluated nuclear library, however, JENDL-3.2 is recently used. Therefore, the authors developed an adjusted library based on JENDL-3.2 which is called JFS-3-J3.2(ADJ98). It is known that the adjusted library based on JENDL-2 overestimated the sodium void reactivity worth by 10-20%. It is expected that the adjusted library based on JENDL-3.2 solve the problem. The adjusted library JFS-3-J3.2(ADJ98) was produced with the same method as the adjusted library JFS-3-J2(ADJ91R) and used more integral parameters of JUPITER experiments than the adjusted library JFS-3-J2(ADJ91R). This report also describes the design accuracy estimation on a 600 MWe class FBR with the adjusted library JFS-3-J3.2(ADJ98). Its main nuclear design parameters (multiplication factor, burn-up reactivity loss, breeding ratio, etc.) except the sodium void reactivity worth which are calculated with the adjusted library JFS-3-J3.2(ADJ98) are almost the same as those predicted with JFS-3-J2(ADJ91R). As for the sodium void reactivity, the adjusted library JFS-3-J3.2(ADJ98) estimates about 4% smaller than the JFS-3-J2(ADJ91R) because of the change of the basic nuclear library from JENDL-2 to JENDL-3.2. (author)

  8. Continuous energy cross section library for MCNP/MCNPX based on JENDL high energy file 2007. FXJH7

    International Nuclear Information System (INIS)

    Sasa, Toshinobu; Sugawara, Takanori; Fukahori, Tokio; Kosako, Kazuaki

    2008-11-01

    The latest JENDL High Energy File (JENDL/HE) was released in 2007 to respond the requirements of reaction data in high energy range up to several GeV to design accelerator facilities such as accelerator-driven systems and research complex like J-PARC. To apply the JENDL/HE-2007 file to the design study, the cross section library of FXJH7 series was constructed from the JENDL/HE file for the calculation using MCNP and MCNPX codes which are widely used in the field of nuclear reactors, fusion reactors, accelerator facilities, medical applications, and so on. In this report, the outline of the JENDL/HE-2007 file, modification of nuclear data processing code NJOY99, construction of FXJH7 library and test calculations for shielding and eigenvalue analyses are summarized. (author)

  9. Comparison of Standard Light Water Reactor Cross-Section Libraries using the United States Nuclear Regulatory Commission Boiling Water Reactor Benchmark Problem

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a comparison of contemporary and historical light water reactor shielding and pressure vessel dosimetry cross-section libraries for a boiling water reactor calculational benchmark problem. The calculational benchmark problem was developed at Brookhaven National Laboratory by the request of the U. S. Nuclear Regulatory Commission. The benchmark problem was originally evaluated by Brookhaven National Laboratory using the Oak Ridge National Laboratory discrete ordinates code DORT and the BUGLE-93 cross-section library. In this paper, the Westinghouse RAPTOR-M3G three-dimensional discrete ordinates code was used. A variety of cross-section libraries were used with RAPTOR-M3G including the BUGLE93, BUGLE-96, and BUGLE-B7 cross-section libraries developed at Oak Ridge National Laboratory and ALPAN-VII.0 developed at Westinghouse. In comparing the calculated fast reaction rates using the four aforementioned cross-section libraries in the pressure vessel capsule, for six dosimetry reaction rates, a maximum relative difference of 8% was observed. As such, it is concluded that the results calculated by RAPTOR-M3G are consistent with the benchmark and further that the different vintage BUGLE cross-section libraries investigated are largely self-consistent.

  10. 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

  11. Highly accurate nuclear and electronic stopping cross sections derived using Monte Carlo simulations to reproduce measured range data

    Science.gov (United States)

    Wittmaack, Klaus; Mutzke, Andreas

    2017-03-01

    We have examined and confirmed the previously unexplored concept of using Monte Carlo calculations in combination with measured projected ranges of ions implanted in solids to derive a quantitative description of nuclear interaction and electronic stopping. The study involved 98 ranges of 11B in Si between 1 keV and 8 MeV, contained in 12 sets of 10 different groups. Systematic errors by up to ±8% were removed to establish a refined data base with 93 ranges featuring only statistical uncertainties (±1.8%). The Monte Carlo calculations could be set up to reproduce the refined ranges with a mean ratio 1.002 ± 1.7%. The input parameters required for this very high level of agreement are as follows. Nuclear interaction is best described by the Kr-C potential, but in obligatory combination with the Lindhard-Scharff (LS) screening length. Up to 300 keV, the electronic stopping cross section is proportional to the projectile velocity, Se = kSe,LS, with k = 1.46 ± 0.01. At higher energies, Se falls progressively short of kSe,LS. Around the Bragg peak, i.e., between 0.8 and 10 MeV, Se is modeled by an adjustable function serving to tailor the peak shape properly. Calculated and measured isotope effects for ranges of 10B and 11B in Si agree within the experimental uncertainty (±0.25%). The range-based Se,R(E) reported here predicts the scarce experimental data derived from the energy loss in projectile transmission through thin Si foils to within 2% or better. By contrast, Se(E) data of available stopping power tables exhibit deviations from Se,R(E) between -40% and +14%.

  12. 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)

  13. Production and testing of HENDL-2.1/CG coarse-group cross-section library based on ENDF/B-VII.0

    International Nuclear Information System (INIS)

    Xu Dezheng; He Zhaozhong; Zou Jun; Zeng Qin

    2010-01-01

    A coarse-group coupled neutron and photon (27n + 21γ) cross-section library HENDL-2.1/CG, based on ENDF/B-VII.0 evaluate data source, has been produced by FDS Team in Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). HENDL-2.1/CG containing 350 nuclide cross-section files (from 1 H to 252 Cf) was generated in MATXS format with the NJOY processing system and then by compiling coarse-group problem-dependent format using the TRANSX code. In order to verify the availability and reliability of the HENDL-2.1/CG data library, requisite benchmark calculations were performed and compared with HENDL-2.0/MG fine-group coupled neutron and photon (175n + 42γ) cross-section library. In general, results using the coarse-group library showed similarly believable as fine-group library.

  14. Benchmarking of the FENDL-3 Neutron Cross-Section Data Library for Fusion Applications

    International Nuclear Information System (INIS)

    Fischer, U.; Kondo, K.; Angelone, M.; Batistoni, P.; Villari, R.; Bohm, T.; Sawan, M.; Walker, B.; Konno, C.

    2014-03-01

    This report summarizes the benchmark analyses performed in a joint effort of ENEA (Italy), JAEA (Japan), KIT (Germany), and the University of Wisconsin (USA) with the objective to test and qualify the neutron induced general purpose FENDL-3.0 data library for fusion applications. The benchmark approach consisted of two major steps including the analysis of a simple ITER-like computational benchmark, and a series of analyses of benchmark experiments conducted previously at the 14 MeV neutron generator facilities at ENEA Frascati, Italy (FNG) and JAEA, Tokai-mura, Japan (FNS). The computational benchmark revealed a modest increase of the neutron flux levels in the deep penetration regions and a substantial increase of the gas production in steel components. The comparison to experimental results showed good agreement with no substantial differences between FENDL-3.0 and FENDL-2.1 for most of the responses analysed. There is a slight trend, however, for an increase of the fast neutron flux in the shielding experiment and a decrease in the breeder mock-up experiments. The photon flux spectra measured in the bulk shield and the tungsten experiments are significantly better reproduced with FENDL-3.0 data. In general, FENDL-3, as compared to FENDL-2.1, shows an improved performance for fusion neutronics applications. It is thus recommended to ITER to replace FENDL-2.1 as reference data library for neutronics calculation by FENDL-3.0. (author)

  15. Quantitative and quality test of cross section library ENDF/B-b2

    International Nuclear Information System (INIS)

    Zajac, R.; Necas, V.

    2006-01-01

    This article includes a test or in other words data verification of neutron ENDF/B-VIIb2 sub library. The first part consists from the process of preparation ACE files by NJOY 99.90. The starting point of data verification describes needed patches in NJOY 99.90, which are necessary to do for correctly production of ACE files. After the obtaining ACE files follow the test of all ACE files through GODIVA - input file for MCNP. GODIVA is high enrichment sphere of U-235, where every material is added as impurity. The aim of GODIVA test is to obtain a certainty if produced ACE files are able to run through MCNP. The second part of this article begins with choose of benchmarks from 'International Handbook of Evaluated Criticality Safety Benchmark Experiments, 2005'. From this source of criticality experiments were separated some benchmarks for quality verification of ACE files by MCNP (Authors)

  16. CSRL-V ENDF/B-V 227-group neutron cross-section library and its application to thermal-reactor and criticality safety benchmarks

    International Nuclear Information System (INIS)

    Ford, W.E. III; Diggs, B.R.; Knight, J.R.; Greene, N.M.; Petrie, L.M.; Webster, C.C.; Westfall, R.M.; Wright, R.Q.; Williams, M.L.

    1982-01-01

    Characteristics and contents of the CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data) 227-neutron-group AMPX master and pointwise cross-section libraries are described. Results obtained in using CSRL-V to calculate performance parameters of selected thermal reactor and criticality safety benchmarks are discussed

  17. 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).

  18. Establishment of the BOSPOR-80 machine library of evaluated threshold reaction cross-sections and its testing by means of integral experiments

    International Nuclear Information System (INIS)

    Bychkov, V.M.; Zolotarev, K.I.; Pashchenko, A.B.; Plyaskin, V.I.

    1982-08-01

    A paper was published in 1979 containing a compilation of experimental data on the cross-sections of (n,p), (n,α) and (n,2n) threshold reactions and recommended excitation functions. A further paper considered the development of evaluation methods based on the use of theoretical model calculations, an increase in the number of recommended excitation functions, correction of the recommended cross-sections on the basis of integral experiments and allowance for recent experimental data. To satisfy the wide circle of users, BOSPOR-80 - a machine library of evaluated threshold reaction cross-sections - was set up

  19. Preparation of next generation set of group cross sections. 3

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    2002-03-01

    This fiscal year, based on the examination result about the evaluation energy range of heavy element unresolved resonance cross sections, the upper energy limit of the energy range, where ultra-fine group cross sections are produced, was raised to 50 keV, and an improvement of the group cross section processing system was promoted. At the same time, reflecting the result of studies carried out till now, a function producing delayed neutron data was added to the general-purpose group cross section processing system , thus the preparation of general purpose group cross section processing system has been completed. On the other hand, the energy structure, data constitution and data contents of next generation group cross section set were determined, and the specification of a 151 groups next generation group cross section set was defined. Based on the above specification, a concrete library format of the next generation cross section set has been determined. After having carried out the above-described work, using the general-purpose group cross section processing system , which was complete in this study, with use of the JENDL-3. 2 evaluated nuclear data, the 151 groups next generation group cross section of 92 nuclides and the ultra fine group resonance cross section library for 29 nuclides have been prepared. Utilizing the 151 groups next generation group cross section set and the ultra-fine group resonance cross-section library, a bench mark test calculation of fast reactors has been performed by using an advanced lattice calculation code. It was confirmed, by comparing the calculation result with a calculation result of continuous energy Monte Carlo code, that the 151 groups next generation cross section set has sufficient accuracy. (author)

  20. Comparative study of Monte Carlo particle transport code PHITS and nuclear data processing code NJOY for recoil cross section spectra under neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Iwamoto, Yosuke, E-mail: iwamoto.yosuke@jaea.go.jp; Ogawa, Tatsuhiko

    2017-04-01

    Because primary knock-on atoms (PKAs) create point defects and clusters in materials that are irradiated with neutrons, it is important to validate the calculations of recoil cross section spectra that are used to estimate radiation damage in materials. Here, the recoil cross section spectra of fission- and fusion-relevant materials were calculated using the Event Generator Mode (EGM) of the Particle and Heavy Ion Transport code System (PHITS) and also using the data processing code NJOY2012 with the nuclear data libraries TENDL2015, ENDF/BVII.1, and JEFF3.2. The heating number, which is the integral of the recoil cross section spectra, was also calculated using PHITS-EGM and compared with data extracted from the ACE files of TENDL2015, ENDF/BVII.1, and JENDL4.0. In general, only a small difference was found between the PKA spectra of PHITS + TENDL2015 and NJOY + TENDL2015. From analyzing the recoil cross section spectra extracted from the nuclear data libraries using NJOY2012, we found that the recoil cross section spectra were incorrect for {sup 72}Ge, {sup 75}As, {sup 89}Y, and {sup 109}Ag in the ENDF/B-VII.1 library, and for {sup 90}Zr and {sup 55}Mn in the JEFF3.2 library. From analyzing the heating number, we found that the data extracted from the ACE file of TENDL2015 for all nuclides were problematic in the neutron capture region because of incorrect data regarding the emitted gamma energy. However, PHITS + TENDL2015 can calculate PKA spectra and heating numbers correctly.

  1. Neutron and proton transmutation-activation cross section libraries to 150 MeV for application in accelerator-driven systems and radioactive ion beam target-design studies

    International Nuclear Information System (INIS)

    Koning, A.J.; Chadwick, M.B.; MacFarlane, R.E.; Mashnik, S.; Wilson, W.B.

    1998-05-01

    New transmutation-activation nuclear data libraries for neutrons and protons up to 150 MeV have been created. These data are important for simulation calculations of radioactivity, and transmutation, in accelerator-driven systems such as the production of tritium (APT) and the transmutation of waste (ATW). They can also be used to obtain cross section predictions for the production of proton-rich isotopes in (p,xn) reactions, for radioactive ion beam (RIB) target-design studies. The nuclear data in these libraries stem from two sources: for neutrons below 20 MeV, we use data from the European activation and transmutation file, EAF97; For neutrons above 20 MeV and for protons at all energies we have isotope production cross sections with the nuclear model code HMS-ALICE. This code applies the Monte Carlo Hybrid Simulation theory, and the Weisskopf-Ewing theory, to calculate cross sections. In a few cases, the HMS-ALICE results were replaced by those calculated using the GNASH code for the Los Alamos LA150 transport library. The resulting two libraries, AF150.N and AF150.P, consist of 766 nuclides each and are represented in the ENDF6-format. An outline is given of the new representation of the data. The libraries have been checked with ENDF6 preprocessing tools and have been processed with NJOY into libraries for the Los Alamos transmutation/radioactivity code CINDER. Numerous benchmark figures are presented for proton-induced excitation functions of various isotopes compared with measurements. Such comparisons are useful for validation purposes, and for assessing the accuracy of the evaluated data. These evaluated libraries are available on the WWW at: http://t2.lanl.gov/. 21 refs

  2. Testing of a JEF-1 based WIMS-D cross section library for migration area and k-infinity predictions for LWHCR lattices

    International Nuclear Information System (INIS)

    Pelloni, S.; Stepanek, J.

    1987-01-01

    The cell code WIMSD4 is used for the analysis of PROTEUS-LWHCR experiments. A library for this code which is based on the European evaluation JEF-1 was produced at EIR using the Los Alamos NJOY system with its module WIMSR and the Canadian management code WILMA. In general, this library delivered more accurate eigenvalues and reaction rates than the WIMS-Standard and WIMS81 libraries did in comparison to experimental values from PROTEUS-LWHCR Cores 1-3. However, large discrepancies (up to about 10%) occured between calculated migration areas (M 2 ). Additional investigations have been undertaken to clarify this problem, since theoretical M 2 -values are needed for deducing k-infinity in the experiments. This has been done in the context of calculations for a reference LWHCR test lattice. The following major reasons for these deviations were found. First, the self-scattering term in non-moderators (P 0 matrix) in the JEF-1 library was not transport corrected. Second, Standard and JEF-1 libraries use infinite dilute cross sections for 238 U, whereas the WIMS81 library uses fully shielded cross sections. Third, the standard library uses the 'row' formula for the transport correction, whereas the 'inflow' formula is applied in the case of JEF-1 and WIMS81 libraries. Lastly, oxygen and 238 U scattering cross sections in the fast energy range are smaller in the case of the WIMS81 library. Differences in calculated k-infinity values between the currently used library and WIMS81 (up to 3%) come (in order of importance for the reference LWHCR lattice) mainly from resonance cross sections for 240 Pu capture, 238 U capture and 239 Pu fission. Recommendations have been made for generating a new JEF-1 library using updated versions of WIMSR and WILMA. (author)

  3. 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)

  4. Assessing availability of scientific journals, databases, and health library services in Canadian health ministries: a cross-sectional study.

    Science.gov (United States)

    Léon, Grégory; Ouimet, Mathieu; Lavis, John N; Grimshaw, Jeremy; Gagnon, Marie-Pierre

    2013-03-21

    Evidence-informed health policymaking logically depends on timely access to research evidence. To our knowledge, despite the substantial political and societal pressure to enhance the use of the best available research evidence in public health policy and program decision making, there is no study addressing availability of peer-reviewed research in Canadian health ministries. To assess availability of (1) a purposive sample of high-ranking scientific journals, (2) bibliographic databases, and (3) health library services in the fourteen Canadian health ministries. From May to October 2011, we conducted a cross-sectional survey among librarians employed by Canadian health ministries to collect information relative to availability of scientific journals, bibliographic databases, and health library services. Availability of scientific journals in each ministry was determined using a sample of 48 journals selected from the 2009 Journal Citation Reports (Sciences and Social Sciences Editions). Selection criteria were: relevance for health policy based on scope note information about subject categories and journal popularity based on impact factors. We found that the majority of Canadian health ministries did not have subscription access to key journals and relied heavily on interlibrary loans. Overall, based on a sample of high-ranking scientific journals, availability of journals through interlibrary loans, online and print-only subscriptions was estimated at 63%, 28% and 3%, respectively. Health Canada had a 2.3-fold higher number of journal subscriptions than that of the provincial ministries' average. Most of the organisations provided access to numerous discipline-specific and multidisciplinary databases. Many organisations provided access to the library resources described through library partnerships or consortia. No professionally led health library environment was found in four out of fourteen Canadian health ministries (i.e. Manitoba Health, Northwest

  5. Concrete reflected cylinders of highly enriched solutions of uranyl nitrate ICSBEP Benchmark: A re-evaluation by means of MCNPX using ENDF/B-VI cross section library

    International Nuclear Information System (INIS)

    Cruzate, J.A.; Carelli, J.L.

    2011-01-01

    This work presents a theoretical re-evaluation of a set of original experiments included in the 2009 issue of the International Handbook of Evaluated Criticality Safety Benchmark Experiments, as “Concrete Reflected Cylinders of Highly Enriched Solutions of Uranyl Nitrate” (identification number: HEU-SOL-THERM- 002) [4]. The present evaluation has been made according to benchmark specifications [4], and added data taken out of the original published report [3], but applying a different approach, resulting in a more realistic calculation model. In addition, calculations have been made using the latest version of MCNPX Monte Carlo code, combined with an updated set of cross section data, the continuous-energy ENDF/B-VI library. This has resulted in a comprehensive model for the given experimental situation. Uncertainties analysis has been made based on the evaluation of experimental data presented in the HEU-SOLTHERM-002 report. Resulting calculations with the present improved physical model have been able to reproduce the criticality of configurations within 0.5%, in good agreement with experimental data. Results obtained in the analysis of uncertainties are in general agreement with those at HEU-SOL-THERM-002 benchmark document. Qualitative results from analyses made in the present work can be extended to similar fissile systems: well moderated units of 235 U solutions, reflected with concrete from all directions. Results have confirmed that neutron absorbers, even as impurities, must be taken into account in calculations if at least approximate proportions were known. (authors)

  6. Criticality benchmark comparisons leading to cross-section upgrades

    International Nuclear Information System (INIS)

    Alesso, H.P.; Annese, C.E.; Heinrichs, D.P.; Lloyd, W.R.; Lent, E.M.

    1993-01-01

    For several years criticality benchmark calculations with COG. COG is a point-wise Monte Carlo code developed at Lawrence Livermore National Laboratory (LLNL). It solves the Boltzmann equation for the transport of neutrons and photons. The principle consideration in developing COG was that the resulting calculation would be as accurate as the point-wise cross-sectional data, since no physics computational approximations were used. The objective of this paper is to report on COG results for criticality benchmark experiments in concert with MCNP comparisons which are resulting in corrections an upgrades to the point-wise ENDL cross-section data libraries. Benchmarking discrepancies reported here indicated difficulties in the Evaluated Nuclear Data Livermore (ENDL) cross-sections for U-238 at thermal neutron energy levels. This led to a re-evaluation and selection of the appropriate cross-section values from several cross-section sets available (ENDL, ENDF/B-V). Further cross-section upgrades anticipated

  7. Evaluated cross section libraries and kerma factors for neutrons up to 100 MeV on {sup 16}O and {sup 14}N

    Energy Technology Data Exchange (ETDEWEB)

    Chadwick, M.B.; Young, P.G.

    1995-07-01

    We present evaluations of the interaction of 20 to 100 MeV neutrons with oxygen and nitrogen nuclei, which follows on from our previous work on carbon. Our aim is to accurately represent integrated cross sections, inclusive emission spectra, and kerma factors, in a data library which can be used in radiation transport calculations. We apply the FKK-GNASH nuclear model code, which includes Hauser-Feshbach, preequilibrium, and direct reaction mechanisms, and use experimental measurements to optimize the calculations. We determine total, elastic, and nonelastic cross sections, angle-energy correlated emission spectra, for light ejectiles with A{<=}4 and gamma-rays, and average energy depositions. Our results for charged-particle emission spectra agree well with the measurements of Subramanian et al.. We compare kerma factors derived from our evaluated cross sections with experimental data, providing an integral benchmarking of our work. The evaluated data libraries are available as electronic files.

  8. Utilizing Monte-Carlo radiation transport and spallation cross sections to estimate nuclide dependent scaling with altitude

    Science.gov (United States)

    Argento, D.; Reedy, R. C.; Stone, J.

    2010-12-01

    Cosmogenic Nuclides (CNs) are a critical new tool for geomorphology, allowing researchers to date Earth surface events and measure process rates [1]. Prior to CNs, many of these events and processes had no absolute method for measurement and relied entirely on relative methods [2]. Continued improvements in CN methods are necessary for expanding analytic capability in geomorphology. In the last two decades, significant progress has been made in refining these methods and reducing analytic uncertainties [1,3]. Calibration data and scaling methods are being developed to provide a self consistent platform for use in interpreting nuclide concentration values into geologic data [4]. However, nuclide dependent scaling has been difficult to address due to analytic uncertainty and sparseness in altitude transects. Artificial target experiments are underway, but these experiments take considerable time for nuclide buildup in lower altitudes. In this study, a Monte Carlo method radiation transport code, MCNPX, is used to model the galactic cosmic-ray radiation impinging on the upper atmosphere and track the resulting secondary particles through a model of the Earth’s atmosphere and lithosphere. To address the issue of nuclide dependent scaling, the neutron flux values determined by the MCNPX simulation are folded in with estimated cross-section values [5,6]. Preliminary calculations indicate that scaling of nuclide production potential in free air seems to be a function of both altitude and nuclide production pathway. At 0 g/cm2 (sea-level) all neutron spallation pathways have attenuation lengths within 1% of 130 g/cm2. However, the differences in attenuation length are exacerbated with increasing altitude. At 530 g/cm2 atmospheric height (~5,500 m), the apparent attenuation lengths for aggregate SiO2(n,x)10Be, aggregate SiO2(n,x)14C and K(n,x)36Cl become 149.5 g/cm2, 151 g/cm2 and 148 g/cm2 respectively. At 700 g/cm2 atmospheric height (~8,400m - close to the highest

  9. Generation of seven group cross section library for TRIGA LEU fuel in CITATION format and benchmarking some experimental and operational data

    International Nuclear Information System (INIS)

    Sarker, M.M.; Bhuiyan, S.I.; Akramuzzaman, M.

    2007-01-01

    The principal objective of this study is to validate the seven group cross section library in CITATION format for TRIGA LEU Fuel. This presentation deals with the 'generation of a cross section library for the CITATION and its validation. We used WIMSD-5B version for the generation of all group constants. The overall strategy is: (1) use WIMS package to generate few group neutron macroscopic cross section (cell constants) for all of the materials in the core and its immediate neighborhood (2) use 3-D code CITATION to perform the global analysis of the core to study: multiplication factor, neutron flux distribution and power peaking factors. Various options available in WIMS program were studied in depth to finalize the models to generate the most appropriate group constants. For the global analysis the code CITATION and a post processing program FCAP were chosen. Thus a seven group cross section library for the calculations of TRIGA Research Reactor was generated. To investigate the validity of the generated library a critical experiment of the TRIGA research reactor was benchmarked. (author)

  10. 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)

  11. The PSIMECX medium-energy neutron activation cross-section library. Part III: Calculational methods for heavy nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Atchison, F.

    1998-09-01

    The PSIMECX library contains calculated nuclide production cross-sections from neutron-induced reactions in the energy range about 2 to 800 MeV in the following 72 stable isotopes of 24 elements: {sup 12}C, {sup 13}C, {sup 16}O, {sup 17}O, {sup 18}O, {sup 23}Na, {sup 24}Mg, {sup 25}Mg, {sup 26}Mg, {sup 27}Al, {sup 28}Si, {sup 29}Si, {sup 30}Si, {sup 31}P, {sup 32}S, {sup 33}S, {sup 34}S, {sup 36}S, {sup 35}Cl, {sup 37}Cl, {sup 39}K, {sup 40}K, {sup 41}K, {sup 40}Ca, {sup 42}Ca, {sup 43}Ca, {sup 44}Ca, {sup 46}Ca, {sup 48}Ca, {sup 46}Ti, {sup 47}Ti, {sup 48}Ti, {sup 49}Ti, {sup 50}Ti, {sup 50}V, {sup 51}V, {sup 50}Cr, {sup 52}Cr, {sup 53}Cr, {sup 54}Cr, {sup 55}Mn, {sup 54}Fe, {sup 56}Fe, {sup 57}Fe, {sup 58}Fe, {sup 58}Ni, {sup 60}Ni, {sup 61}Ni, {sup 62}Ni, {sup 64}Ni, {sup 63}Cu, {sup 65}Cu, {sup 64}Zn, {sup 66}Zn, {sup 67}Zn, {sup 68}Zn, {sup 70}Zn, {sup 92}Mo, {sup 94}Mo, {sup 95}Mo, {sup 96}Mo, {sup 97}Mo, {sup 98}Mo, {sup 100}Mo, {sup 121}Sb, {sup 123}Sb, {sup 204}Pb, {sup 206}Pb, {sup 207}Pb, {sup 208}Pb, {sup 232}Th and {sup 238}U. The energy range covers essentially all transmutation channels other than capture. The majority of the selected elements are main constituents of normal materials of construction used in and around accelerator facilities and the library is, first and foremost, designed to be a tool for the estimation of their activation in wide-band neutron fields. This third report describes and discusses the calculational methods used for the heavy nuclei. The library itself has been described in the first report of this series and the treatment for the medium and light mass nuclei is given in the second. (author)

  12. Development of automatic cross section compilation system for MCNP

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Sakurai, Kiyoshi

    1999-01-01

    A development of a code system to automatically convert cross-sections for MCNP is in progress. The NJOY code is, in general, used to convert the data compiled in the ENDF format (Evaluated Nuclear Data Files by BNL) into the cross-section libraries required by various reactor physics codes. While the cross-section library: FSXLIB-J3R2 was already converted from the JENDL-3.2 version of Japanese Evaluated Nuclear Data Library for a continuous energy Monte Carlo code MCNP, the library keeps only the cross-sections at room temperature (300 K). According to the users requirements which want to have cross-sections at higher temperature, say 600 K or 900 K, a code system named 'autonj' is under development to provide a set of cross-section library of arbitrary temperature for the MCNP code. This system can accept any of data formats adopted JENDL that may not be treated by NJOY code. The input preparation that is repeatedly required at every nuclide on NJOY execution is greatly reduced by permitting the conversion process of as many nuclides as the user wants in one execution. A few MCNP runs were achieved for verification purpose by using two libraries FSXLIB-J3R2 and the output of autonj'. The almost identical MCNP results within the statistical errors show the 'autonj' output library is correct. In FY 1998, the system will be completed, and in FY 1999, the user's manual will be published. (K. Tsuchihashi)

  13. CSRL-V: processed ENDF/B-V 227-neutron-group and pointwise cross-section libraries for criticality safety, reactor, and shielding studies

    International Nuclear Information System (INIS)

    Ford, W.E. III; Diggs, B.R.; Petrie, L.M.; Webster, C.C.; Westfall, R.M.

    1982-01-01

    A P 3 227-neutron-group cross-section library has been processed for the subsequent generation of problem-dependent fine- or broad-group cross sections for a broad range of applications, including shipping cask calculations, general criticality safety analyses, and reactor core and shielding analyses. The energy group structure covers the range 10 -5 eV - 20 MeV, including 79 thermal groups below 3 eV. The 129-material library includes processed data for all materials in the ENDF/B-V General Purpose File, several data sets prepared from LENDL data, hydrogen with water- and polyethyelene-bound thermal kernels, deuterium with C 2 O-bound thermal kernels, carbon with a graphite thermal kernel, a special 1/V data set, and a dose factor data set. The library, which is in AMPX master format, is designated CSRL-V (Criticality Safety Reference Library based on ENDF/B-V data). Also included in CSRL-V is a pointwise total, fission, elastic scattering, and (n,γ) cross-section library containing data sets for all ENDF/B-V resonance materials. Data in the pointwise library were processed with the infinite dilute approximation at a temperature of 296 0 K

  14. Physical models, cross sections, and numerical approximations used in MCNP and GEANT4 Monte Carlo codes for photon and electron absorbed fraction calculation.

    Science.gov (United States)

    Yoriyaz, Hélio; Moralles, Maurício; Siqueira, Paulo de Tarso Dalledone; Guimarães, Carla da Costa; Cintra, Felipe Belonsi; dos Santos, Adimir

    2009-11-01

    Radiopharmaceutical applications in nuclear medicine require a detailed dosimetry estimate of the radiation energy delivered to the human tissues. Over the past years, several publications addressed the problem of internal dose estimate in volumes of several sizes considering photon and electron sources. Most of them used Monte Carlo radiation transport codes. Despite the widespread use of these codes due to the variety of resources and potentials they offered to carry out dose calculations, several aspects like physical models, cross sections, and numerical approximations used in the simulations still remain an object of study. Accurate dose estimate depends on the correct selection of a set of simulation options that should be carefully chosen. This article presents an analysis of several simulation options provided by two of the most used codes worldwide: MCNP and GEANT4. For this purpose, comparisons of absorbed fraction estimates obtained with different physical models, cross sections, and numerical approximations are presented for spheres of several sizes and composed as five different biological tissues. Considerable discrepancies have been found in some cases not only between the different codes but also between different cross sections and algorithms in the same code. Maximum differences found between the two codes are 5.0% and 10%, respectively, for photons and electrons. Even for simple problems as spheres and uniform radiation sources, the set of parameters chosen by any Monte Carlo code significantly affects the final results of a simulation, demonstrating the importance of the correct choice of parameters in the simulation.

  15. 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)

  16. ZZ HPICE/F, Gamma Interaction Cross-Section Library in ENDF/B Format for Transport Calculation

    International Nuclear Information System (INIS)

    1984-01-01

    Nature of physical problem solved: Format: ENDF/B file 23; Number of groups: Point Cross Sections, energies 1 keV to 100 MeV. Nuclides: Z = 1-83, 86, 90, 92 an 94. Origin: Lawrence Livermore Laboratory; Weighting spectrum: none. The data are for use in general purpose gamma-ray transport codes. The Lawrence Livermore Laboratory has a continuing program to evaluate photon cross section. The data are given in units of (barns/atom) for energies 1 keV to 100 MeV and for elements Z = 1-83, 86, 90, 92 and 94. The MAT numbers are equal to the atomic numbers (Z). The following cross sections are tabulated: MT cross section type: 501 total; 502 coherent scattering; 504 incoherent scattering; 516 pair production (includes triplet); 603 photoelectric

  17. 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

  18. 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

  19. Monte Carlo cross section testing for thermal and intermediate 235U/238U critical assemblies, ENDF/B-V vs ENDF/B-VI

    International Nuclear Information System (INIS)

    Weinman, J.P.

    1997-06-01

    The purpose of this study is to investigate the eigenvalue sensitivity to changes in ENDF/B-V and ENDF/B-VI cross section data sets by comparing RACER vectorized Monte Carlo calculations for several thermal and intermediate spectrum critical experiments. Nineteen Oak Ridge and Rocky Flats thermal solution benchmark critical assemblies that span a range of hydrogen-to- 235 U (H/U) concentrations (2052 to 27.1) and above-thermal neutron leakage fractions (0.555 to 0.011) were analyzed. In addition, three intermediate spectrum critical assemblies (UH3-UR, UH3-NI, and HISS-HUG) were studied

  20. Validation of the Monte Carlo criticality program KENO IV and the Hansen-Roach sixteen-energy-group-cross sections for high-assay uranium systems

    International Nuclear Information System (INIS)

    Handley, G.R.; Masters, L.C.; Stachowiak, R.V.

    1981-01-01

    Validation of the Monte Carlo criticality code, KENO IV, and the Hansen-Roach sixteen-energy-group cross sections was accomplished by calculating the effective neutron multiplication constant, k/sub eff/, of 29 experimentally critical assemblies which had uranium enrichments of 92.6% or higher in the uranium-235 isotope. The experiments were chosen so that a large variety of geometries and of neutron energy spectra were covered. Problems, calculating the k/sub eff/ of systems with high-uranium-concentration uranyl nitrate solution that were minimally reflected or unreflected, resulted in the separate examination of five cases

  1. Monte Carlo Simulation of the Time-Of-Flight Technique for the Measurement of Neutron Cross-section in the Pohang Neutron Facility

    Energy Technology Data Exchange (ETDEWEB)

    An, So Hyun; Lee, Young Ouk; Lee, Cheol Woo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Young Seok [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2007-10-15

    It is essential that neutron cross sections are measured precisely for many areas of research and technique. In Korea, these experiments have been performed in the Pohang Neutron Facility (PNF) with the pulsed neutron facility based on the 100 MeV electron linear accelerator. In PNF, the neutron energy spectra have been measured for different water levels inside the moderator and compared with the results of the MCNPX calculation. The optimum size of the water moderator has been determined on the base of these results. In this study, Monte Carlo simulations for the TOF technique were performed and neutron spectra of neutrons were calculated to predict the measurements.

  2. Criticality and safety parameter studies for upgrading 3 MW TRIGA MARK II research reactor and validation of generated cross section library and computational method

    International Nuclear Information System (INIS)

    Bhuiyan, S.I.; Mondal, M.A.W.; Sarker, M.M.; Rahman, M.; Shahdatullah, M.S.; Huda, M.Q.; Chakrroborty, T.K.; Khan, M.J.H.

    2000-01-01

    This study deals with the neutronic and thermal hydraulic analysis of the 3MW TRIGA MARK II research reactor to upgrade it to a higher flux. The upgrading will need a major reshuffling and reconfiguration of the current core. To reshuffle the current core configuration, the chain of NJOY94.10 - WIMSD-5A - CITATION - PARET - MCNP4B2 codes has been used for the overall analysis. The computational methods, tools and techniques, customisation of cross section libraries, various models for cells and super cells, and a lot of associated utilities have been standardised and established/validated for the overall core analysis. Analyses using the 4-group and 7-group libraries of macroscopic cross sections generated from the 69-group WIMSD-5 library showed that a 7-group structure is more suitable for TRIGA calculations considering its LEU fuel composition. The MCNP calculations established that the CITATION calculations and the generated cross section library are reasonably good for neutronic analysis of TRIGA reactors. Results obtained from PARET demonstrated that the flux upgrade will not cause the temperature limit on the fuel to be exceeded. Also, the maximum power density remains, by a substantial margin below the level at which the departure from nucleate boiling could occur. A possible core with two additional irradiation channels around the CT is projected where almost identical thermal fluxes as in the CT are obtained. The reconfigured core also shows 7.25% thermal flux increase in the Lazy Susan. (author)

  3. Generation of the library of neutron cross sections for the Record code of the Fuel Management System (FMS); Generacion de la biblioteca de secciones eficaces de neutrones para el codigo Record del Sistema de Administracion de Combustible (FMS)

    Energy Technology Data Exchange (ETDEWEB)

    Alonso V, G; Hernandez L, H [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    1991-11-15

    On the basis of the library structure of the RECORD code a method to generate the neutron cross sections by means of the ENDF-B/IV database and the NJOY code has been developed. The obtained cross sections are compared with those of the current library which was processed using the ENDF-B/III version. (Author)

  4. 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

  5. VITAMIN-J/COVA/EFF-3 cross-section covariance matrix library and its use to analyse benchmark experiments in sinbad database

    International Nuclear Information System (INIS)

    Kodeli, Ivan-Alexander

    2005-01-01

    The new cross-section covariance matrix library ZZ-VITAMIN-J/COVA/EFF3 intended to simplify and encourage sensitivity and uncertainty analysis was prepared and is available from the NEA Data Bank. The library is organised in a ready-to-use form including both the covariance matrix data as well as processing tools:-Cross-section covariance matrices from the EFF-3 evaluation for five materials: 9 Be, 28 Si, 56 Fe, 58 Ni and 60 Ni. Other data will be included when available. -FORTRAN program ANGELO-2 to extrapolate/interpolate the covariance matrices to a users' defined energy group structure. -FORTRAN program LAMBDA to verify the mathematical properties of the covariance matrices, like symmetry, positive definiteness, etc. The preparation, testing and use of the covariance matrix library are presented. The uncertainties based on the cross-section covariance data were compared with those based on other evaluations, like ENDF/B-VI. The collapsing procedure used in the ANGELO-2 code was compared and validated with the one used in the NJOY system

  6. Development of fine-group (315n/42γ) cross section library ENDL3.0/FG for fusion-fission hybrid systems

    International Nuclear Information System (INIS)

    Zeng Qin; Zou Jun; Xu Dezhen; Jiang Jieqiong; Wang Minghuang; Wu Yican; Qiu Yuefeng; Chen Zhong; Chen Yan

    2011-01-01

    To improve the accuracy of the neutron analyses for subcritical systems with thermal fission blanket, a coupled neutron and photon (315 n + 42γ) fine-group cross section library HENDL3.0/FG based on ENDF/B-Ⅶ. 0 has been produced by FDS team. In order to test the availability and reliability of the HENDL3.0/FG data library, shielding and critical safety benchmarks were performed with VisualBUS code. The testing results indicated that the discrepancy between calculation and experimental values of nuclear parameters fell in a reasonable range. (authors)

  7. Measurements of D-T neutron induced radioactivity in plasma-facing materials and their role in qualification of activation cross-section libraries and codes

    International Nuclear Information System (INIS)

    Kumar, A.; Abdou, M.A.; Kosako, K.; Oyama, Y.; Nakamura, T.; Maekawa, H.

    1995-01-01

    The D-T neutron-induced radioactivity constitutes one of the foremost issues in fusion reactor design. The validation of activation cross-sections and decay data libraries is one of the important requirements for validating ITER design from safety and waste disposal viewpoints. An elaborate, experimental program was initiated in 1988, under USDOE-JAERI collaborative program, to validate the radioactivity codes/libraries. The measurements of decay-γ spectra from irradiated, high purity samples of Al, Si, Ti, V, Cr, Mn-Cu alloy, Fe, Co, Ni, Cu, stainless steel 316 (AISI 316), Zn, Zr, Nb, Mo, In, Sn, Ta, W, and Pb, among others, were conducted under D-T neutron fluences varying from 1.6 x 10 10 ncm -2 to 6.1 x 10 13 ncm -2 . As many as 14 neutron energy spectra were covered for a number of materials. The analysis of isotopic activities of the irradiated materials using activation cross-section libraries of four leading radioactivity codes, i.e. ACT4/THIDA-2, REAC-3, DKR-ICF, and RACC, has shown large discrepancies among the calculations, on the one hand, and between the calculations and the measurements, on the other. A discussion is also presented on definition and obtention of safety cum quality factors for various activation libraries. (orig.)

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

    Science.gov (United States)

    Frisoni, Manuela

    2017-09-01

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

  9. Assessment of Degree of Applicability of Benchmarks for Gadolinium Using KENO V.a and the 238-Group SCALE Cross-Section Library

    Energy Technology Data Exchange (ETDEWEB)

    Goluoglu, S.

    2003-12-01

    A review of the degree of applicability of benchmarks containing gadolinium using the computer code KENO V.a and the gadolinium cross sections from the 238-group SCALE cross-section library has been performed for a system that contains {sup 239}Pu, H{sub 2}O, and Gd{sub 2}O{sub 3}. The system (practical problem) is a water-reflected spherical mixture that represents a dry-out condition on the bottom of a sludge receipt and adjustment tank around steam coils. Due to variability of the mixture volume and the H/{sup 239}Pu ratio, approximations to the practical problem, referred to as applications, have been made to envelop possible ranges of mixture volumes and H/{sup 239}Pu ratios. A newly developed methodology has been applied to determine the degree of applicability of benchmarks as well as the penalty that should be added to the safety margin due to insufficient benchmarks.

  10. Single electron ionization and electron capture cross sections for (C{sup 6+}, H{sub 2}O) interaction within the Classical Trajectory Monte Carlo (CTMC) approach

    Energy Technology Data Exchange (ETDEWEB)

    Tran, H.N., E-mail: tranngochoang@tdt.edu.vn [Division of Nuclear Physics, Ton Duc Thang University, Tan Phong Ward, District 7, Ho Chi Minh City (Viet Nam); Faculty of Applied Sciences, Ton Duc Thang University, Tan Phong Ward, District 7, Ho Chi Minh City (Viet Nam); Dao, D.D. [Division of Nuclear Physics, Ton Duc Thang University, Tan Phong Ward, District 7, Ho Chi Minh City (Viet Nam); Faculty of Applied Sciences, Ton Duc Thang University, Tan Phong Ward, District 7, Ho Chi Minh City (Viet Nam); Incerti, S. [Division of Nuclear Physics, Ton Duc Thang University, Tan Phong Ward, District 7, Ho Chi Minh City (Viet Nam); Faculty of Applied Sciences, Ton Duc Thang University, Tan Phong Ward, District 7, Ho Chi Minh City (Viet Nam); CNRS, IN2P3, CENBG, UMR 5797, F-33170 Gradignan (France); Université de Bordeaux, CENBG, UMR 5797, F-33170 Gradignan (France); Bernal, M.A. [Instituto de Física Gleb Wataghin, Universidade Estadual de Campinas, SP (Brazil); Karamitros, M. [CNRS, IN2P3, CENBG, UMR 5797, F-33170 Gradignan (France); Université de Bordeaux, CENBG, UMR 5797, F-33170 Gradignan (France); Nhan Hao, T.V. [Center of Research and Development, Duy Tan University, K7/25 Quang Trung, Danang (Viet Nam); Center for Theoretical and Computational Physics, College of Education, Hue University, 34 Le Loi Street, Hue City (Viet Nam); Dang, T.M. [VNUHCM-University of Science (Viet Nam); Francis, Z. [Saint Joseph University, Beyrouth (Lebanon)

    2016-01-01

    In this work, we present a derivation of cross sections for single ionization and electron capture processes within the Classical Trajectory Monte Carlo (CTMC) approach. Specifically, we have used a potential stemming from an ab initio calculation in Green et al.’s framework to describe the dynamics of the water molecule system. Proposing a modified version of the Classical Over-Barrier (COB) potential, we have found that a cut-off of roughly 28 a.u. on the initial distance of the projectile produced a reasonable accuracy. A global agreement has been obtained in our calculations compared to experimental and other theoretical results for C{sup 6+} ion energies ranging from 10 keV/u to 10 MeV/u.

  11. Sampling procedures using optical-data and partial wave cross sections in a Monte Carlo code for simulating kilovolt electron and positron transport in solids

    International Nuclear Information System (INIS)

    Fernandez-Varea, J.M.; Salvat, F.; Liljequist, D.

    1994-09-01

    The details of a Monte Carlo code for computing the penetration and energy loss of electrons and positrons in solids are described. The code, intended for electrons and positrons with energies from ∼ 100 eV to ∼ 100 keV, is based on the simulation of individual elastic and inelastic collisions. Elastic collisions are simulated using differential cross sections computed by the relativistic partial wave method applied to a muffin-tin Dirac-Hartree-Fock-Slater potential. Inelastic collisions are simulated by means of a model based on optical and photoelectric data, which are extended to the non-zero momentum transfer region by means of somewhat different algorithms for valence electron excitations and inner-shell excitations. This report focuses on the description of detailed formulae and sampling methods. 10 refs, 3 figs, 8 tabs

  12. Application of a new cross section library based on ENDF/B-IV to reactor core analysis

    International Nuclear Information System (INIS)

    Lima Bezerra, J. de.

    1991-04-01

    The use of the ENDF/B-IV library in the LEOPARD code for the Angra-1 reactor simulation is presented. The results are compared to those obtained using the ENDF/B-II library and show better values for the power distribution but an underestimated global reactivity as compared to experimental results. (F.E.). 1 ref, 55 figs, 1 tab

  13. Generation and Testing of the ENDF/B-VI Continuous-Energy Cross-Section Library for Use with Continuous-Energy Versions of KENO

    International Nuclear Information System (INIS)

    Goluoglu, Sedat; Dunn, Michael E.; Greene, Norman Maurice; Petrie Jr, Lester M.; Hollenbach, Daniel F.

    2007-01-01

    KENO V.a and KENO-VI are Monte Carlo codes that solve the multigroup form of the Boltzmann transport equation. These codes are part of the SCALE system of codes and are used for performing criticality calculations of systems with fissionable material. In general, continuous-energy Monte Carlo methods are preferred because such an approach avoids many of the assumptions inherent in the multigroup treatment. On the other hand, continuous-energy treatment is much more demanding in terms of computer storage space for data, memory requirements, and calculation speed. Continuous-energy versions of KENO V.a and KENO-VI have been created and are being extensively tested. Generation of ENDF/B-VI continuous-energy cross sections is explained, and the results of the validation and verification of the codes and the data are presented

  14. 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.

  15. 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

  16. Neutron Thermal Cross Sections, Westcott Factors, Resonance Integrals, Maxwellian Averaged Cross Sections and Astrophysical Reaction Rates Calculated from the ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0, ROSFOND-2010, CENDL-3.1 and EAF-2010 Evaluated Data Libraries

    Science.gov (United States)

    Pritychenko, B.; Mughabghab, S. F.

    2012-12-01

    We present calculations of neutron thermal cross sections, Westcott factors, resonance integrals, Maxwellian-averaged cross sections and astrophysical reaction rates for 843 ENDF materials using data from the major evaluated nuclear libraries and European activation file. Extensive analysis of newly-evaluated neutron reaction cross sections, neutron covariances, and improvements in data processing techniques motivated us to calculate nuclear industry and neutron physics quantities, produce s-process Maxwellian-averaged cross sections and astrophysical reaction rates, systematically calculate uncertainties, and provide additional insights on currently available neutron-induced reaction data. Nuclear reaction calculations are discussed and new results are presented. Due to space limitations, the present paper contains only calculated Maxwellian-averaged cross sections and their uncertainties. The complete data sets for all results are published in the Brookhaven National Laboratory report.

  17. Condensation and homogenization of cross sections for the deterministic transport codes with Monte Carlo method: Application to the GEN IV fast neutron reactors

    International Nuclear Information System (INIS)

    Cai, Li

    2014-01-01

    In the framework of the Generation IV reactors neutronic research, new core calculation tools are implemented in the code system APOLLO3 for the deterministic part. These calculation methods are based on the discretization concept of nuclear energy data (named multi-group and are generally produced by deterministic codes) and should be validated and qualified with respect to some Monte-Carlo reference calculations. This thesis aims to develop an alternative technique of producing multi-group nuclear properties by a Monte-Carlo code (TRIPOLI-4). At first, after having tested the existing homogenization and condensation functionalities with better precision obtained nowadays, some inconsistencies are revealed. Several new multi-group parameters estimators are developed and validated for TRIPOLI-4 code with the aid of itself, since it has the possibility to use the multi-group constants in a core calculation. Secondly, the scattering anisotropy effect which is necessary for handling neutron leakage case is studied. A correction technique concerning the diagonal line of the first order moment of the scattering matrix is proposed. This is named the IGSC technique and is based on the usage of an approximate current which is introduced by Todorova. An improvement of this IGSC technique is then presented for the geometries which hold an important heterogeneity property. This improvement uses a more accurate current quantity which is the projection on the abscissa X. The later current can represent the real situation better but is limited to 1D geometries. Finally, a B1 leakage model is implemented in the TRIPOLI-4 code for generating multi-group cross sections with a fundamental mode based critical spectrum. This leakage model is analyzed and validated rigorously by the comparison with other codes: Serpent and ECCO, as well as an analytical case.The whole development work introduced in TRIPOLI-4 code allows producing multi-group constants which can then be used in the core

  18. View-CXS neutron and photon cross-sections viewer

    International Nuclear Information System (INIS)

    Subbaiah, K.V.; Sunil Sunny, C.

    2004-01-01

    A graphical user-friendly interface is developed in Visual Basic (VB)-6 to view the variation of neutron and photon interaction cross-sections of different isotopes as a function of energy. VB subroutines developed read the binary data files of cross-sections created in MCNP-ACE (Briesmeister, J.F., 1993. MCNP - a general purpose Monte Carlo N-Particle Transport code. Version 4A. LANL, USA), ANISN-DLC (Engle W.W. Jr., 1967, A User's Manual for ANISN, K-1693; ORNL, 1974. 100 group neutron cross section data based on ENDF/B-III. Oak Ridge National Laboratory, USA) and KENO-AMPX (Petrie, L.M., Landers, N.F., 1984 KENO-Va- An Improved Monte Carlo Criticality Program with Super Grouping. RSICC-CCC-548, USA) formats using LAHEY-77 Fortran Compiler. The information on isotopes present in each library will be displayed with the help of database files prepared using Micro-Soft ACESS. The cross-section data can be viewed in different presentation styles namely, line graphs, bar graphs, histograms etc., with different color and symbol options. The cross-section plots generated can be saved as Bit-Map file to embed in any other text files. This software enables inter comparison of cross-sections from different type of libraries for isotopes as well as mixtures. Provision is made to view the cross-sections for nuclear reactions such as (n,γ), (n,f), (n,α), etc. The software can be obtained from Radiation Safety Information and Computational Centre (RSICC), ORNL, USA with the code package identification number PSR-514. The software package needs a hard disk space of about 80 MB when installed and works in WINDOWS-95/98/2000 operating systems

  19. 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.

  20. 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

  1. Recommended activation detector cross sections (RNDL-82)

    International Nuclear Information System (INIS)

    Bondars, Kh.Ya.; Lapenas, A.A.

    1984-01-01

    The results of the comparison between measured and calculated average cross sections in 5 benchmark experiments are presented. Calculations have been based on the data from 10 libraries of evaluated cross sections. The recommended library (RNDL-82) of the activation detector cross sections has been created on the basis of the comparison. RNDL-82, including 26 reactions, and the basic characteristics of the detectors are presented. (author)

  2. Tables and graphs of electron-interaction cross sections from 10 eV to 100 GeV derived from the LLNL Evaluated Electron Data Library (EEDL), Z = 1--100

    Energy Technology Data Exchange (ETDEWEB)

    Perkins, S.T.; Cullen, D.E. (Lawrence Livermore National Lab., CA (United States)); Seltzer, S.M. (National Inst. of Standards and Technology (NML), Gaithersburg, MD (United States). Center for Radiation Research)

    1991-11-12

    Energy-dependent evaluated electron interaction cross sections and related parameters are presented for elements H through Fm (Z = 1 to 100). Data are given over the energy range from 10 eV to 100 GeV. Cross sections and average energy deposits are presented in tabulated and graphic form. In addition, ionization cross sections and average energy deposits for each shell are presented in graphic form. This information is derived from the Livermore Evaluated Electron Data Library (EEDL) as of July, 1991.

  3. FEMA DFIRM Cross Sections

    Data.gov (United States)

    Minnesota Department of Natural Resources — FEMA Cross Sections are required for any Digital Flood Insurance Rate Map database where cross sections are shown on the Flood Insurance Rate Map (FIRM). Normally...

  4. Low Energy Neutrino Cross Sections

    International Nuclear Information System (INIS)

    Zeller, G.P.

    2004-01-01

    Present atmospheric and accelerator based neutrino oscillation experiments operate at low neutrino energies (Ev ∼ 1 GeV) to access the relevant regions of oscillation parameter space. As such, they require precise knowledge of the cross sections for neutrino-nucleon interactions in the sub-to-few GeV range. At these energies, neutrinos predominantly interact via quasi-elastic (QE) or single pion production processes, which historically have not been as well studied as the deep inelastic scattering reactions that dominate at higher energies.Data on low energy neutrino cross sections come mainly from bubble chamber, spark chamber, and emulsion experiments that collected their data decades ago. Despite relatively poor statistics and large neutrino flux uncertainties, these measurements provide an important and necessary constraint on Monte Carlo models in present use. The following sections discuss the current status of QE, resonant single pion, coherent pion, and single kaon production cross section measurements at low energy

  5. The PSIMECX medium-energy neutron activation cross-section library. Part II: Calculational methods for light to medium mass nuclei

    International Nuclear Information System (INIS)

    Atchison, F.

    1998-09-01

    The PSIMECX library contains calculated nuclide production cross-sections from neutron-induced reactions in the energy range about 2 to 800 MeV in the following 72 stable isotopes of 24 elements: 12 C, 13 C, 16 O, 17 O, 18 O, 23 Na, 24 Mg, 25 Mg, 26 Mg, 27 Al, 28 Si, 29 Si, 30 Si, 31 P, 32 S, 33 S, 34 S, 36 S, 35 Cl, 37 Cl, 39 K, 40 K, 41 K, 40 Ca, 42 Ca, 43 Ca, 44 Ca, 46 Ca, 48 Ca, 46 Ti, 47 Ti, 48 Ti, 49 Ti, 50 Ti, 50 V, 51 V, 50 Cr, 52 Cr, 53 Cr, 54 Cr, 55 Mn, 54 Fe, 56 Fe, 57 Fe, 58 Fe, 58 Ni, 60 Ni, 61 Ni, 62 Ni, 64 Ni, 63 Cu, 65 Cu, 64 Zn, 66 Zn, 67 Zn, 68 Zn, 70 Zn, 92 Mo, 94 Mo, 95 Mo, 96 Mo, 97 Mo, 98 Mo, 100 Mo, 121 Sb, 123 Sb, 204 Pb, 206 Pb, 207 Pb, 208 Pb, 232 Th and 238 U. The energy range covers essentially all transmutation channels other than capture. The majority of the selected elements are principal constituents of normal materials of construction used in and around accelerator facilities and the library is, first and foremost, designed to be a tool for the estimation of their activation in wide-band neutron fields. This second report, of a series of three, describes and discusses the calculational methods used for the stable isotopes up to and including 123 Sb. The library itself has been described in the first report of the series and the treatment for the heavy nuclei is given in the third. (author)

  6. 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.

  7. 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.

  8. Status of neutron dosimetry cross sections

    International Nuclear Information System (INIS)

    Griffin, P.J.; Kelly, J.G.

    1992-01-01

    Several new cross section libraries, such as ENDF/B-VI(release 2), IRDF-90,JEF-2.2, and JENDL-3 Dosimetry, have recently been made available to the dosimetry community. the Sandia National Laboratories (SNL) Radiation Metrology Laboratory (RML) has worked with these libraries since pre-release versions were available. this paper summarizes the results of the intercomparison and testing of dosimetry cross sections. As a result of this analysis, a compendium of the best dosimetry cross sections was assembled from the available libraries for use within the SNL RML. this library, referred to as the SNLRML Library, contains 66 general dosimetry sensors and 3 special dosimeters unique to the RML sensor inventory. The SNLRML cross sections have been put into a format compatible with commonly used spectrum determination codes

  9. The PSIMECX medium-energy neutron activation cross-section library. Part II: Calculational methods for light to medium mass nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Atchison, F.

    1998-09-01

    The PSIMECX library contains calculated nuclide production cross-sections from neutron-induced reactions in the energy range about 2 to 800 MeV in the following 72 stable isotopes of 24 elements: {sup 12}C, {sup 13}C, {sup 16}O, {sup 17}O, {sup 18}O, {sup 23}Na, {sup 24}Mg, {sup 25}Mg, {sup 26}Mg, {sup 27}Al, {sup 28}Si, {sup 29}Si, {sup 30}Si, {sup 31}P, {sup 32}S, {sup 33}S, {sup 34}S, {sup 36}S, {sup 35}Cl, {sup 37}Cl, {sup 39}K, {sup 40}K, {sup 41}K, {sup 40}Ca, {sup 42}Ca, {sup 43}Ca, {sup 44}Ca, {sup 46}Ca, {sup 48}Ca, {sup 46}Ti, {sup 47}Ti, {sup 48}Ti, {sup 49}Ti, {sup 50}Ti, {sup 50}V, {sup 51}V, {sup 50}Cr, {sup 52}Cr, {sup 53}Cr, {sup 54}Cr, {sup 55}Mn, {sup 54}Fe, {sup 56}Fe, {sup 57}Fe, {sup 58}Fe, {sup 58}Ni, {sup 60}Ni, {sup 61}Ni, {sup 62}Ni, {sup 64}Ni, {sup 63}Cu, {sup 65}Cu, {sup 64}Zn, {sup 66}Zn, {sup 67}Zn, {sup 68}Zn, {sup 70}Zn, {sup 92}Mo, {sup 94}Mo, {sup 95}Mo, {sup 96}Mo, {sup 97}Mo, {sup 98}Mo, {sup 100}Mo, {sup 121}Sb, {sup 123}Sb, {sup 204}Pb, {sup 206}Pb, {sup 207}Pb, {sup 208}Pb, {sup 232}Th and {sup 238}U. The energy range covers essentially all transmutation channels other than capture. The majority of the selected elements are principal constituents of normal materials of construction used in and around accelerator facilities and the library is, first and foremost, designed to be a tool for the estimation of their activation in wide-band neutron fields. This second report, of a series of three, describes and discusses the calculational methods used for the stable isotopes up to and including {sup 123}Sb. The library itself has been described in the first report of the series and the treatment for the heavy nuclei is given in the third. (author)

  10. Jet inclusive cross sections

    International Nuclear Information System (INIS)

    Del Duca, V.

    1992-11-01

    Minijet production in jet inclusive cross sections at hadron colliders, with large rapidity intervals between the tagged jets, is evaluated by using the BFKL pomeron. We describe the jet inclusive cross section for an arbitrary number of tagged jets, and show that it behaves like a system of coupled pomerons

  11. Curves and tables of neutron cross sections

    International Nuclear Information System (INIS)

    Nakagawa, Tsuneo; Asami, Tetsuo; Yoshida, Tadashi

    1990-07-01

    Neutron cross-section curves from the Japanese Evaluated Nuclear Data Library version 3, JENDL-3, are presented in both graphical and tabular form for users in a wide range of application areas in the nuclear energy field. The contents cover cross sections for all the main reactions induced by neutrons with an energy below 20 MeV including; total, elastic scattering, capture, and fission, (n,n'), (n,2n), (n,3n), (n,α), (n,p) reactions. The 2200 m/s cross-section values, resonance integrals, and Maxwellian- and fission-spectrum averaged cross sections are also tabulated. (author)

  12. Performance of JEF2.2 based continuous energy cross sections in predicting the multiplication factor of critical systems

    International Nuclear Information System (INIS)

    John, T.M.; de Leege, P.F.A.; Hoogenboom, J.E.

    1996-01-01

    The continuous energy representation of cross sections for neutronics calculations avoids the requirement of resonance self shielding and the assumptions about the neutron spectrum used for weighing cross sections, required in the preparation of a multigroup cross sections library. The cross sections library prepared for a particular temperature of the nuclide is valid irrespective of the environment of the nuclide and can be used in calculations for many types of reactors. It is comparatively easier to incorporate them in Monte Carlo simulation of neutron transport. The Monte Carlo code MCNP is capable of using a continuous energy representation of nuclear cross sections in simulation of neutron or photon transport. The ACER module of NJOY is able to generate the continuous energy cross section of any nuclide in a format that can be used by MCNP, from any evaluated data file in ENDF/B format. Continuous energy cross sections prepared from the evaluated data file JEF2.2 was used to analyse some standard critical benchmarks and also the critical configuration of the HOR, a 2 MW research reactor at Delft, the Netherlands. Results show that continuous energy cross sections prepared from JEF2.2 evaluated file predicts the multiplication factor of critical systems very close to unity. (author). 6 refs., 2 tabs., 1 fig

  13. Validation of MCNP6 Version 1.0 with the ENDF/B-VII.1 Cross Section Library for Uranium Metal, Oxide, and Solution Systems on the High Performance Computing Platform Moonlight

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, Bryan Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); MacQuigg, Michael Robert [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wysong, Andrew Russell [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-04-21

    In this document, the code MCNP is validated with ENDF/B-VII.1 cross section data under the purview of ANSI/ANS-8.24-2007, for use with uranium systems. MCNP is a computer code based on Monte Carlo transport methods. While MCNP has wide reading capability in nuclear transport simulation, this validation is limited to the functionality related to neutron transport and calculation of criticality parameters such as keff.

  14. Validation of MCNP6 Version 1.0 with the ENDF/B-VII.1 Cross Section Library for Uranium Metal, Oxide, and Solution Systems on the High Performance Computing Platform Moonlight

    International Nuclear Information System (INIS)

    Chapman, Bryan Scott; MacQuigg, Michael Robert; Wysong, Andrew Russell

    2016-01-01

    In this document, the code MCNP is validated with ENDF/B-VII.1 cross section data under the purview of ANSI/ANS-8.24-2007, for use with uranium systems. MCNP is a computer code based on Monte Carlo transport methods. While MCNP has wide reading capability in nuclear transport simulation, this validation is limited to the functionality related to neutron transport and calculation of criticality parameters such as k eff .

  15. Floodplain Cross Section Lines

    Data.gov (United States)

    Department of Homeland Security — This table is required for any Digital Flood Insurance Rate Map database where cross sections are shown on the Flood Insurance Rate Map (FIRM). Normally any FIRM...

  16. Multitrajectory eikonal cross sections

    International Nuclear Information System (INIS)

    Turner, R.E.

    1983-01-01

    With the use of reference and distorted transition operators, a time-correlation-function representation of the inelastic differential cross section has recently been used to obtain distorted eikonal cross sections. These cross sections involve straight-line and reference classical translational trajectories that are unaffected by any internal-state changes which have occurred during the collision. This distorted eikonal theory is now extended to include effects of internal-state changes on the translational motion. In particular, a different classical trajectory is associated with each pair of internal states. Expressions for these inelastic cross sections are obtained in terms of time-ordered cosine and sine memory functions using the Zwanzig-Feshbach projection-operator method. Explicit formulas are obtained in the time-disordered perturbation approximation

  17. Influence of the ab initio n–d cross sections in the critical heavy-water benchmarks

    International Nuclear Information System (INIS)

    Morillon, B.; Lazauskas, R.; Carbonell, J.

    2013-01-01

    Highlights: ► We solve the three nucleon problem using different NN potential (MT, AV18 and INOY) to calculate the Neutron–deuteron cross sections. ► These cross sections are compared to the existing experimental data and to international libraries. ► We describe the different sets of heavy water benchmarks for which the Monte Carlo simulations have been performed including our new Neutron–deuteron cross sections. ► The results obtained by the ab initio INOY potential have been compared with the calculations based on the international library cross sections and are found to be of the same quality. - Abstract: The n–d elastic and breakup cross sections are computed by solving the three-body Faddeev equations for realistic and semi-realistic nucleon–nucleon potentials. These cross sections are inserted in the Monte Carlo simulation of the nuclear processes considered in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). The results obtained using thes ab initio n–d cross sections are compared with those provided by the most renown international libraries

  18. BUGJEFF311.BOLIB (JEFF-3.1.1) and BUGENDF70.BOLIB (ENDF/B-VII.0) - Generation Methodology and Preliminary Testing of two ENEA-Bologna Group Cross Section Libraries for LWR Shielding and Pressure Vessel Dosimetry

    Science.gov (United States)

    Pescarini, Massimo; Sinitsa, Valentin; Orsi, Roberto; Frisoni, Manuela

    2016-02-01

    Two broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format, dedicated to LWR shielding and pressure vessel dosimetry applications, were generated following the methodology recommended by the US ANSI/ANS-6.1.2-1999 (R2009) standard. These libraries, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, are respectively based on JEFF-3.1.1 and ENDF/B-VII.0 nuclear data and adopt the same broad-group energy structure (47 n + 20 γ) of the ORNL BUGLE-96 similar library. They were respectively obtained from the ENEA-Bologna VITJEFF311.BOLIB and VITENDF70.BOLIB libraries in AMPX format for nuclear fission applications through problem-dependent cross section collapsing with the ENEA-Bologna 2007 revision of the ORNL SCAMPI nuclear data processing system. Both previous libraries are based on the Bondarenko self-shielding factor method and have the same AMPX format and fine-group energy structure (199 n + 42 γ) as the ORNL VITAMIN-B6 similar library from which BUGLE-96 was obtained at ORNL. A synthesis of a preliminary validation of the cited BUGLE-type libraries, performed through 3D fixed source transport calculations with the ORNL TORT-3.2 SN code, is included. The calculations were dedicated to the PCA-Replica 12/13 and VENUS-3 engineering neutron shielding benchmark experiments, specifically conceived to test the accuracy of nuclear data and transport codes in LWR shielding and radiation damage analyses.

  19. The MCLIB library: Monte Carlo simulation of neutron scattering instruments

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, P.A.

    1995-09-01

    Monte Carlo is a method to integrate over a large number of variables. Random numbers are used to select a value for each variable, and the integrand is evaluated. The process is repeated a large number of times and the resulting values are averaged. For a neutron transport problem, first select a neutron from the source distribution, and project it through the instrument using either deterministic or probabilistic algorithms to describe its interaction whenever it hits something, and then (if it hits the detector) tally it in a histogram representing where and when it was detected. This is intended to simulate the process of running an actual experiment (but it is much slower). This report describes the philosophy and structure of MCLIB, a Fortran library of Monte Carlo subroutines which has been developed for design of neutron scattering instruments. A pair of programs (LQDGEOM and MC{_}RUN) which use the library are shown as an example.

  20. The MCLIB library: Monte Carlo simulation of neutron scattering instruments

    International Nuclear Information System (INIS)

    Seeger, P.A.

    1995-01-01

    Monte Carlo is a method to integrate over a large number of variables. Random numbers are used to select a value for each variable, and the integrand is evaluated. The process is repeated a large number of times and the resulting values are averaged. For a neutron transport problem, first select a neutron from the source distribution, and project it through the instrument using either deterministic or probabilistic algorithms to describe its interaction whenever it hits something, and then (if it hits the detector) tally it in a histogram representing where and when it was detected. This is intended to simulate the process of running an actual experiment (but it is much slower). This report describes the philosophy and structure of MCLIB, a Fortran library of Monte Carlo subroutines which has been developed for design of neutron scattering instruments. A pair of programs (LQDGEOM and MC RUN) which use the library are shown as an example

  1. Implementation of the rapid cross section adjustment approach at General Electric

    International Nuclear Information System (INIS)

    Cowan, C.L.; Kujawski, E.; Protsik, R.

    1978-01-01

    The General Electric rapid cross section adjustment approach was developed to use the shielding factor method for formulating multigroup cross sections. In this approach, space- and composition-dependent cross sections for a particular reactor or shield design are prepared from a generalized cross section library by the use of resonance self-shielding factors, and by the adjustment of elastic scattering cross sections for the local neutron flux spectra. The principal tool in the cross section adjustment package is the data processing code TDOWN. This code was specified to give the user a high degree of flexibility in the analysis of advanced reactor designs. Of particular interest in the analysis of critical experiments is the ability to carry out cell heterogeneity self-shielding calculations using a multiregion equivalence relationship, and the homogenization of the cross sections over the specified cell with the flux weighting obtained from transport theory calculations. Extensive testing of the rapid cross section adjustment approach, including comparisons with Monte Carlo methods, indicated that this approach can be utilized with a high degree of confidence in the design analysis of complex fast reactor systems. 2 figures, 1 table

  2. Validation of MCNP6 Version 1.0 with the ENDF/B-VII.1 Cross Section Library for Plutonium Metals, Oxides, and Solutions on the High Performance Computing Platform Moonlight

    Energy Technology Data Exchange (ETDEWEB)

    Chapman, Bryan Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gough, Sean T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-12-05

    This report documents a validation of the MCNP6 Version 1.0 computer code on the high performance computing platform Moonlight, for operations at Los Alamos National Laboratory (LANL) that involve plutonium metals, oxides, and solutions. The validation is conducted using the ENDF/B-VII.1 continuous energy group cross section library at room temperature. The results are for use by nuclear criticality safety personnel in performing analysis and evaluation of various facility activities involving plutonium materials.

  3. Discrete ordinates cross-sections generation in parallel plane geometry -- 1: Concept

    International Nuclear Information System (INIS)

    Yavuz, M.

    1998-01-01

    Cross-section formulations derived from the linear Boltzman transport equation have been the subjects of several studies. In these studies, theoretical foundations and concepts are provided, and the solution techniques are derived. The author presents new methods for generating cross-section sets for transport problems, with an arbitrary scattering anisotropy of order L (L ≤ N - 1), approximated by the S N (and P N-1 ) methods. The formulations require knowledge of the eigensolutions, which may be determined by a recent eigenvalue equation found in Yavuz. The motivation for this study is to generate few-group cross sections for pin cells (and/or assemblies) using a Monte Carlo code, for example, MCNP, with a continuous-energy cross-section library. However, this work is a first step, and it describes a new concept to perform inverse transport calculations, provided that the surface Green's functions over desired angular and energy intervals are known

  4. Recommended evaluation procedure for photonuclear cross section

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young-Ouk; Chang, Jonghwa; Fukahori, Tokio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    In order to generate photonuclear cross section library for the necessary applications, data evaluation is combined with theoretical evaluation, since photonuclear cross sections measured cannot provide all necessary data. This report recommends a procedure consisting of four steps: (1) analysis of experimental data, (2) data evaluation, (3) theoretical evaluation and, if necessary, (4) modification of results. In the stage of analysis, data obtained by different measurements are reprocessed through the analysis of their discrepancies to a representative data set. In the data evaluation, photonuclear absorption cross sections are evaluated via giant dipole resonance and quasi-deutron mechanism. With photoabsorption cross sections from the data evaluation, theoretical evaluation is applied to determine various decay channel cross sections and emission spectra using equilibrium and preequilibrium mechanism. After this, the calculated results are compared with measured data, and in some cases the results are modified to better describe measurements. (author)

  5. Measurements and Monte Carlo calculations of neutron production cross-sections at 180o for the 140 MeV proton incident reactions on carbon, iron, and gold

    International Nuclear Information System (INIS)

    Iwamoto, Yosuke; Satoh, Daiki; Hagiwara, Masayuki; Yashima, Hiroshi; Nakane, Yoshihiro; Tamii, Atsushi; Iwase, Hiroshi; Endo, Akira; Nakashima, Hiroshi; Sakamoto, Yukio; Hatanaka, Kichiji; Niita, Koji

    2010-01-01

    The neutron production cross-sections of carbon, iron, and gold targets with 140 MeV protons at 180 o were measured at the RCNP cyclotron facility. The time-of-flight technique was used to obtain the neutron energy spectra in the energy range above 1 MeV. The carbon and iron target results were compared with the experimental data from 113 MeV (p,xn) reactions at 150 o reported by Meier et al. Our data agreed well with them in spite of different incident energies and angles. Calculations were then performed using different intra-nuclear cascade models (Bertini, ISOBAR, and JQMD) implemented with PHITS code. The results calculated using the ISOBAR and JQMD models roughly agreed with the experimental iron and gold target data, but the Bertini could not reproduce the high-energy neutrons above 10 MeV.

  6. Validation of the 3D transport monte carlo code TRIPOLI-4.3 for moderated and unmoderated metallic fissile media configurations with JEF2.2 and ENDF/B-VI.4 cross section evaluations

    International Nuclear Information System (INIS)

    Gagnier, E.; Lee, Y.K.; Aguiar, L.; Vedrenne, N.

    2003-01-01

    This paper presents an extended validation of TRIPOLI-4.3 covering all the metallic fissile media configurations present in the CEA facilities. More than 300 ICSBEP benchmarks have been calculated with TRIPOLI-4.3 and compared to the experimental results. These benchmarks include high-enriched uranium fissile media and plutonium fissile media with a low content of plutonium 240. The configurations are calculated with continuous-energy cross-section libraries JEF2.2 and ENDF/B-VI.4 and compared to MCNP or SCALE results presented in the ICSBEP reports. (author)

  7. Electron collision cross sections of mercury

    International Nuclear Information System (INIS)

    Suzuki, Susumu; Kuzuma, Kiyotaka; Itoh, Haruo

    2006-01-01

    In this paper, we propose a new collision cross section set for mercury which revises the original set summarized by Hayashi in 1989. Hanne reported three excitation collision cross sections (6 3 P 0 , 6 3 P 1 , 6 3 P 2 ) determined from an electron beam experiment in 1988. As a matter for regret, no attentive consideration was given to combining these three excitation cross sections with the cross section set of Hayashi. Therefore we propose a new set where these three excitation cross sections are included. In this study, other two excitation cross sections (6 1 P 1 , 6 3 D 3 ) except for the three excitation collision cross sections (6 3 P 0 , 6 3 P 1 , 6 3 P 2 ) are taken from the original set of Hayashi. The momentum transfer cross section and the ionization collision cross section are also taken from Hayashi. A Monte Carlo Simulation (MCS) technique is applied for evaluating our new cross section set. The present results of the electron drift velocity and the ionization coefficient are compared to experimental values. Agreement is secured in relation to the electron drift velocity for 1.5 Td 2 ) is the reduced electric field, E (V/cm) is the electric field, N (1/cm 3 ) is the number density of mercury atoms at 0degC, 1 Torr, E/N is also equal to 2.828 x 10 -17 E/p 0 from the relation of the ideal gas equation, p 0 (Torr) is gas pressure at 0degC, 1 Torr=1.33322 x 10 -2 N/cm -2 and 10 -17 V/cm 2 is called 1 Td. Thus it is ensured that our new cross section set is reasonable enough to be used up to 100 eV when considering with the electron drift velocity and the ionization coefficient. (author)

  8. Testing neutron cross-section files from the BROND-2 and ENDF/B-6 libraries in benchmark experiments on neutron transmission through spherical layers

    International Nuclear Information System (INIS)

    Androsenko, A.A.; Androsenko, P.A.; Blokhin, A.I.; Kulagin, N.T.; Pronyaev, V.G.; Simakov, S.P.

    1997-01-01

    The effect of angular anisotropy in inelastic secondary neutron scattering on neutron leakage spectra from the surface of spherical specimens is investigated. It is shown how inadequate representation of the cross-section structure in the neutron energy resonance region can affect the neutron leakage spectrum. (author). 19 refs, 5 figs, 6 tabs

  9. Testing neutron cross-section files from the BROND-2 and ENDF/B-6 libraries in benchmark experiments on neutron transmission through spherical layers

    Energy Technology Data Exchange (ETDEWEB)

    Androsenko, A A; Androsenko, P A; Blokhin, A I; Kulagin, N T; Pronyaev, V G; Simakov, S P [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1997-06-01

    The effect of angular anisotropy in inelastic secondary neutron scattering on neutron leakage spectra from the surface of spherical specimens is investigated. It is shown how inadequate representation of the cross-section structure in the neutron energy resonance region can affect the neutron leakage spectrum. (author). 19 refs, 5 figs, 6 tabs.

  10. 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.

  11. 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

  12. The impact of different Monte Carlo models on the cross section measurement of top-pair production at 7 TeV proton-proton collisions

    International Nuclear Information System (INIS)

    Krause, Claudius

    2010-09-01

    High energy proton-proton collisions lead to a large amount of secondary particles to be measured in a detector. A final state containing top quarks is of particular interest. But top quarks are only produced in a small fraction of the collisions. Hence, criteria must be defined to separate events containing top quarks from the background. From detectors, we record signals, for example hits in the tracker system or deposits in the calorimeters. In order to obtain the momentum of the particles, we apply algorithms to reconstruct tracks in space. More sophisticated algorithms are needed to identify the flavour of quarks, such as b-tagging. Several steps are needed to test these algorithms. Collision products of proton-proton events are generated using Monte Carlo techniques and their passage through the detector is simulated. After that, the algorithms are applied and the signal efficiency and the mistagging rate can be obtained. There are, however, many different approaches and algorithms realized in programs, so the question arises if the choice of the Monte Carlo generator influences the measured quantities. In this thesis, two commonly used Monte Carlo generators, SHERPA and MadGraph/MadEvent, are compared and the differences in the selection efficiency of semimuonic tt events are estimated. In addition, the distributions of kinematic variables are shown. A special chapter about the matching of matrix elements with parton showers is included. The main algorithms, CKKW for SHERPA and MLM for MadGraph/MadEvent, are introduced.

  13. The impact of different Monte Carlo models on the cross section measurement of top-pair production at 7 TeV proton-proton collisions

    Energy Technology Data Exchange (ETDEWEB)

    Krause, Claudius

    2012-04-15

    High energy proton-proton collisions lead to a large amount of secondary particles to be measured in a detector. A final state containing top quarks is of particular interest. But top quarks are only produced in a small fraction of the collisions. Hence, criteria must be defined to separate events containing top quarks from the background. From detectors, we record signals, for example hits in the tracker system or deposits in the calorimeters. In order to obtain the momentum of the particles, we apply algorithms to reconstruct tracks in space. More sophisticated algorithms are needed to identify the flavour of quarks, such as b-tagging. Several steps are needed to test these algorithms. Collision products of proton-proton events are generated using Monte Carlo techniques and their passage through the detector is simulated. After that, the algorithms are applied and the signal efficiency and the mistagging rate can be obtained. There are, however, many different approaches and algorithms realized in programs, so the question arises if the choice of the Monte Carlo generator influences the measured quantities. In this thesis, two commonly used Monte Carlo generators, SHERPA and MadGraph/MadEvent, are compared and the differences in the selection efficiency of semimuonic tt events are estimated. In addition, the distributions of kinematic variables are shown. A special chapter about the matching of matrix elements with parton showers is included. The main algorithms, CKKW for SHERPA and MLM for MadGraph/MadEvent, are introduced.

  14. HYDRA: a Java library for Markov Chain Monte Carlo

    Directory of Open Access Journals (Sweden)

    Gregory R. Warnes

    2002-03-01

    Full Text Available Hydra is an open-source, platform-neutral library for performing Markov Chain Monte Carlo. It implements the logic of standard MCMC samplers within a framework designed to be easy to use, extend, and integrate with other software tools. In this paper, we describe the problem that motivated our work, outline our goals for the Hydra pro ject, and describe the current features of the Hydra library. We then provide a step-by-step example of using Hydra to simulate from a mixture model drawn from cancer genetics, first using a variable-at-a-time Metropolis sampler and then a Normal Kernel Coupler. We conclude with a discussion of future directions for Hydra.

  15. The neutron instrument Monte Carlo library MCLIB: Recent developments

    International Nuclear Information System (INIS)

    Seeger, P.A.; Daemen, L.L.; Hjelm, R.P. Jr.; Thelliez, T.G.

    1998-01-01

    A brief review is given of the developments since the ICANS-XIII meeting made in the neutron instrument design codes using the Monte Carlo library MCLIB. Much of the effort has been to assure that the library and the executing code MC RUN connect efficiently with the World Wide Web application MC-WEB as part of the Los Alamos Neutron Instrument Simulation Package (NISP). Since one of the most important features of MCLIB is its open structure and capability to incorporate any possible neutron transport or scattering algorithm, this document describes the current procedure that would be used by an outside user to add a feature to MCLIB. Details of the calling sequence of the core subroutine OPERATE are discussed, and questions of style are considered and additional guidelines given. Suggestions for standardization are solicited, as well as code for new algorithms

  16. 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.

  17. 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

  18. 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

  19. Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

    International Nuclear Information System (INIS)

    White, Morgan C.

    2000-01-01

    The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V and V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to

  20. Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    White, Morgan C. [Univ. of Florida, Gainesville, FL (United States)

    2000-07-01

    The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second

  1. Radar cross section

    CERN Document Server

    Knott, Gene; Tuley, Michael

    2004-01-01

    This is the second edition of the first and foremost book on this subject for self-study, training, and course work. Radar cross section (RCS) is a comparison of two radar signal strengths. One is the strength of the radar beam sweeping over a target, the other is the strength of the reflected echo sensed by the receiver. This book shows how the RCS ?gauge? can be predicted for theoretical objects and how it can be measured for real targets. Predicting RCS is not easy, even for simple objects like spheres or cylinders, but this book explains the two ?exact? forms of theory so well that even a

  2. Three-Dimensional (X,Y,Z) Deterministic Analysis of the PCA-Replica Neutron Shielding Benchmark Experiment using the TORT-3.2 Code and Group Cross Section Libraries for LWR Shielding and Pressure Vessel Dosimetry

    OpenAIRE

    Pescarini Massimo; Orsi Roberto; Frisoni Manuela

    2016-01-01

    The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the ORNL TORT-3.2 3D SN code. PCA-Replica, specifically conceived to test the accuracy of nuclear data and transport codes employed in LWR shielding and radiation damage calculations, reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a PWR pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with ...

  3. 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)

  4. 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.

  5. Monte Carlo simulations of the pulsed thermal neutron flux in two-region hydrogenous systems (using standard MCNP data libraries)

    International Nuclear Information System (INIS)

    Wiacek, U.; Krynicka, E.

    2005-02-01

    Monte Carlo simulations of the pulsed neutron experiment in two- region systems (two concentric spheres and two coaxial finite cylinders) are presented. The MCNP code is used. Aqueous solutions of H 3 BO 3 or KCl are used in the inner region. The outer region is the moderator of Plexiglas. Standard data libraries of the thermal neutron scattering cross-sections of hydrogen in hydrogenous substances are used. The time-dependent thermal neutron transport is simulated when the inner region has a constant size and the external size of the surrounding outer region is variable. The time decay constant of the thermal neutron flux in the system is found in each simulation. The results of the simulations are compared with results of real pulsed neutron experiments on the corresponding systems. (author)

  6. ENDF/B-VIII.0: The 8th Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards and Thermal Scattering Data

    Science.gov (United States)

    Brown, D. A.; Chadwick, M. B.; Capote, R.; Kahler, A. C.; Trkov, A.; Herman, M. W.; Sonzogni, A. A.; Danon, Y.; Carlson, A. D.; Dunn, M.; Smith, D. L.; Hale, G. M.; Arbanas, G.; Arcilla, R.; Bates, C. R.; Beck, B.; Becker, B.; Brown, F.; Casperson, R. J.; Conlin, J.; Cullen, D. E.; Descalle, M.-A.; Firestone, R.; Gaines, T.; Guber, K. H.; Hawari, A. I.; Holmes, J.; Johnson, T. D.; Kawano, T.; Kiedrowski, B. C.; Koning, A. J.; Kopecky, S.; Leal, L.; Lestone, J. P.; Lubitz, C.; Márquez Damián, J. I.; Mattoon, C. M.; McCutchan, E. A.; Mughabghab, S.; Navratil, P.; Neudecker, D.; Nobre, G. P. A.; Noguere, G.; Paris, M.; Pigni, M. T.; Plompen, A. J.; Pritychenko, B.; Pronyaev, V. G.; Roubtsov, D.; Rochman, D.; Romano, P.; Schillebeeckx, P.; Simakov, S.; Sin, M.; Sirakov, I.; Sleaford, B.; Sobes, V.; Soukhovitskii, E. S.; Stetcu, I.; Talou, P.; Thompson, I.; van der Marck, S.; Welser-Sherrill, L.; Wiarda, D.; White, M.; Wormald, J. L.; Wright, R. Q.; Zerkle, M.; Žerovnik, G.; Zhu, Y.

    2018-02-01

    We describe the new ENDF/B-VIII.0 evaluated nuclear reaction data library. ENDF/B-VIII.0 fully incorporates the new IAEA standards, includes improved thermal neutron scattering data and uses new evaluated data from the CIELO project for neutron reactions on 1H, 16O, 56Fe, 235U, 238U and 239Pu described in companion papers in the present issue of Nuclear Data Sheets. The evaluations benefit from recent experimental data obtained in the U.S. and Europe, and improvements in theory and simulation. Notable advances include updated evaluated data for light nuclei, structural materials, actinides, fission energy release, prompt fission neutron and γ-ray spectra, thermal neutron scattering data, and charged-particle reactions. Integral validation testing is shown for a wide range of criticality, reaction rate, and neutron transmission benchmarks. In general, integral validation performance of the library is improved relative to the previous ENDF/B-VII.1 library.

  7. ENDF/B-VIII.0: The 8 th Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards and Thermal Scattering Data

    Energy Technology Data Exchange (ETDEWEB)

    Brown, D. A.; Chadwick, M. B.; Capote, R.; Kahler, A. C.; Trkov, A.; Herman, M. W.; Sonzogni, A. A.; Danon, Y.; Carlson, A. D.; Dunn, M.; Smith, D. L.; Hale, G. M.; Arbanas, G.; Arcilla, R.; Bates, C. R.; Beck, B.; Becker, B.; Brown, F.; Casperson, R. J.; Conlin, J.; Cullen, D. E.; Descalle, M. -A.; Firestone, R.; Gaines, T.; Guber, K. H.; Hawari, A. I.; Holmes, J.; Johnson, T. D.; Kawano, T.; Kiedrowski, B. C.; Koning, A. J.; Kopecky, S.; Leal, L.; Lestone, J. P.; Lubitz, C.; Márquez Damián, J. I.; Mattoon, C. M.; McCutchan, E. A.; Mughabghab, S.; Navratil, P.; Neudecker, D.; Nobre, G. P. A.; Noguere, G.; Paris, M.; Pigni, M. T.; Plompen, A. J.; Pritychenko, B.; Pronyaev, V. G.; Roubtsov, D.; Rochman, D.; Romano, P.; Schillebeeckx, P.; Simakov, S.; Sin, M.; Sirakov, I.; Sleaford, B.; Sobes, V.; Soukhovitskii, E. S.; Stetcu, I.; Talou, P.; Thompson, I.; van der Marck, S.; Welser-Sherrill, L.; Wiarda, D.; White, M.; Wormald, J. L.; Wright, R. Q.; Zerkle, M.; Žerovnik, G.; Zhu, Y.

    2018-02-01

    We describe the new ENDF/B-VIII.0 evaluated nuclear reaction data library. ENDF/B-VIII.0 fully incorporates the new IAEA standards, includes improved thermal neutron scattering data and uses new evaluated data from the CIELO project for neutron reactions on 1H, 16O, 56Fe, 235U, 238U and 239Pu described in companion papers in the present issue of Nuclear Data Sheets. The evaluations benefit from recent experimental data obtained in the U.S. and Europe, and improvements in theory and simulation. Notable advances include updated evaluated data for light nuclei, structural materials, actinides, fission energy release, prompt fission neutron and γ-ray spectra, thermal neutron scattering data, and charged-particle reactions. Integral validation testing is shown for a wide range of criticality, reaction rate, and neutron transmission benchmarks. In general, integral validation performance of the library is improved relative to the previous ENDF/B-VII.1 library.

  8. Research on Fast-Doppler-Broadening of neutron cross sections

    International Nuclear Information System (INIS)

    Li, S.; Wang, K.; Yu, G.

    2012-01-01

    A Fast-Doppler-Broadening method is developed in this work to broaden Continuous Energy neutron cross-sections for Monte Carlo calculations. Gauss integration algorithm and parallel computing are implemented in this method, which is unprecedented in the history of cross section processing. Compared to the traditional code (NJOY, SIGMA1, etc.), the new Fast-Doppler-Broadening method shows a remarkable speedup with keeping accuracy. The purpose of using Gauss integration is to avoid complex derivation of traditional broadening formula and heavy load of computing complementary error function that slows down the Doppler broadening process. The OpenMP environment is utilized in parallel computing which can take full advantage of modern multi-processor computers. Combination of the two can reduce processing time of main actinides (such as 238 U, 235 U) to an order of magnitude of 1∼2 seconds. This new method is fast enough to be applied to Online Doppler broadening. It can be combined or coupled with Monte Carlo transport code to solve temperature dependent problems and neutronics-thermal hydraulics coupled scheme which is a big challenge for the conventional NJOY-MCNP system. Examples are shown to determine the efficiency and relative errors compared with the NJOY results. A Godiva Benchmark is also used in order to test the ACE libraries produced by the new method. (authors)

  9. Standard cross-section data

    International Nuclear Information System (INIS)

    Carlson, A.D.

    1984-01-01

    The accuracy of neutron cross-section measurement is limited by the uncertainty in the standard cross-section and the errors associated with using it. Any improvement in the standard immediately improves all cross-section measurements which have been made relative to that standard. Light element, capture and fission standards are discussed. (U.K.)

  10. Validation of SCALE 4.0 -- CSAS25 module and the 27-group ENDF/B-IV cross-section library for low-enriched uranium systems

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, W.C.

    1993-02-01

    A version of KENO V.a and the 27-group library in SCALE-4.0 were validated for use in evaluating the nuclear criticality safety of low-enriched uranium systems. A total of 59 critical systems were analyzed. A statistical analysis of the results was performed, and subcritical acceptanced criteria are established.

  11. Validation of SCALE 4. 0 -- CSAS25 module and the 27-group ENDF/B-IV cross-section library for low-enriched uranium systems

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, W.C.

    1993-02-01

    A version of KENO V.a and the 27-group library in SCALE-4.0 were validated for use in evaluating the nuclear criticality safety of low-enriched uranium systems. A total of 59 critical systems were analyzed. A statistical analysis of the results was performed, and subcritical acceptanced criteria are established.

  12. 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

  13. Cross sections for atmospheric corrections

    International Nuclear Information System (INIS)

    Meyer, J.P.; Casse, M.; Westergaard, N.

    1975-01-01

    A set of cross sections for spallation of relativistic nuclei is proposed based on (i) the best available proton cross sections, (ii) an extrapolation to heavier nuclei of the dependence on the number of nucleons lost of the 'target factor' observed for C 12 and O 16 by Lindstrom et al. (1975), in analogy with Rudstam's formalism, and (iii) on a normalization of all cross sections to the total cross sections for production of fragments with Asub(f) >= 6. The obtained cross sections for peripheral interactions are not inconsistent with simple geometrical considerations. (orig.) [de

  14. Comparison of threshold reaction cross sections for the Ti, V, Cr, Fe, Ni, Cu, and Zn isotopes from evaluated data libraries

    Energy Technology Data Exchange (ETDEWEB)

    Blokhin, A I; Manokhin, V N; Nasyrova, S M [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1998-09-01

    Evaluated excitation functions for various threshold reactions on Ti, V, Cr, Ge, Ni, Cu and Zn isotopes are compared to reveal discrepancies between different nuclear data libraries. The recommended excitation functions for (n,p), (n,np), (N,{alpha}) and (n,2n) reactions, evaluated on the basis of empirical systematics are given for comparison to facilitate selection of a more reliable data. The available experimental data are also plotted. (author) 10 refs, 70 figs

  15. Top quark production cross-section measurements

    CERN Document Server

    Chen, Ye; The ATLAS collaboration

    2017-01-01

    Measurements of the inclusive and differential cross-sections for top-quark pair and single top production cross sections in proton-proton collisions with the ATLAS detector at the Large Hadron Collider are presented at center-of-mass energies of 8 TeV and 13 TeV. The inclusive measurements reach high precision and are compared to the best available theoretical calculations. These measurements, including results using boosted tops, probe our understanding of top-pair production in the TeV regime. The results are compared to Monte Carlo generators implementing LO and NLO matrix elements matched with parton showers and NLO QCD calculations. For the t-channel single top measurement, the single top-quark and anti-top-quark total production cross-sections, their ratio, as well as differential cross sections are also presented. A measurement of the production cross-section of a single top quark in association with a W boson, the second largest single-top production mode, is also presented. Finally, measurements of ...

  16. A new approach to make collapsed cross section for burnup calculation of subcritical system

    International Nuclear Information System (INIS)

    Matsunaka, Masayuki; Kondo, Keitaro; Miyamaru, Hiroyuki; Murata, Isao

    2008-01-01

    A general-purpose transport and burnup code system for precise analysis of subcritical reactors like a fusion-fission (FF) hybrid reactor was developed and used for analyzing their performance. The FF hybrid reactor is a subcritical system, which has a concept of fusion reactor with a blanket region containing nuclear fuel and has been under discussion by author's group for years because the present burnup calculation system mainly consists of a general-purpose Monte Carlo code MCNP-4B, a point burnup code ORIGEN2. JENDL-3.3 pointwise cross section library and JENDL Activation Cross Section File 96 were used as base cross section libraries to make group constant for burnup calculation. A new method has been proposed to make group constant for the burnup calculation as accurate as possible directly using output data of the neutron transport calculation by MCNP and evaluated nuclear data libraries. This method is strict and a general procedure to make one group cross sections in Monte Carlo calculations, while it takes very long computation time. Some speed-up techniques were discussed for the present group constant making process so as to decrease calculation time. Adoption of postprocessing to make group constant improved the calculation accuracy because of increasing number of cross sections to be updated in each burnup cycle. The present calculation system is capable of performing neutronics analysis of subcritical reactors more precise than our previous one. However, at the moment, it still takes long computation time to make group constants. Further speed-up techniques are now under investigation so as to apply the present system to neutronics design analysis for various subcritical systems. (author)

  17. 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)

  18. Validation of evaluated neutron standard cross sections

    International Nuclear Information System (INIS)

    Badikov, S.; Golashvili, T.

    2008-01-01

    Some steps of the validation and verification of the new version of the evaluated neutron standard cross sections were carried out. In particular: -) the evaluated covariance data was checked for physical consistency, -) energy-dependent evaluated cross-sections were tested in most important neutron benchmark field - 252 Cf spontaneous fission neutron field, -) a procedure of folding differential standard neutron data in group representation for preparation of specialized libraries of the neutron standards was verified. The results of the validation and verification of the neutron standards can be summarized as follows: a) the covariance data of the evaluated neutron standards is physically consistent since all the covariance matrices of the evaluated cross sections are positive definite, b) the 252 Cf spectrum averaged standard cross-sections are in agreement with the evaluated integral data (except for 197 Au(n,γ) reaction), c) a procedure of folding differential standard neutron data in group representation was tested, as a result a specialized library of neutron standards in the ABBN 28-group structure was prepared for use in reactor applications. (authors)

  19. Tables and graphs of photon-interaction cross sections from 0.1 keV to 100 MeV derived from the LLL evaluated-nuclear-data library

    International Nuclear Information System (INIS)

    Plechaty, E.F.; Cullen, D.E.; Howerton, R.J.

    1981-01-01

    Energy-dependent evaluated photon interaction cross sections and related parameters are presented for elements H through Cf(Z = 1 to 98). Data are given over the energy range from 0.1 keV to 100 MeV. The related parameters include form factors and average energy deposits per collision (with and without fluorescence). Fluorescence information is given for all atomic shells that can emit a photon with a kinetic energy of 0.1 keV or more. In addition, the following macroscopic properties are given: total mean free path and energy deposit per centimeter. This information is derived from the Livermore Evaluated-Nuclear-Data Library (ENDL) as of October 1978

  20. Tables and graphs of photon-interaction cross sections from 0.1 keV to 100 MeV derived from the LLL Evaluated-Nuclear-Data Library

    International Nuclear Information System (INIS)

    Plechaty, E.F.; Cullen, D.E.; Howerton, R.J.

    1978-01-01

    Energy-dependent evaluated photon interaction cross sections and related parameters are presented for elements H through Cf (Z = 1 to 98). Data are given over the energy range from 0.1 keV to 100 MeV. The related parameters include form factors and average energy deposits per collision (with and without fluorescence). Fluorescence information is given for all atomic shells that can emit a photon with a kinetic energy of 0.1 keV or more. In addition, the following macroscopic properties are given: total mean free path and energy deposit per centimeter. This information is derived from the Livermore Evaluated-Nuclear-Data Library (ENDL) as of October 1978

  1. Establishment of an international reference data library of nuclear activation cross sections. Summary report of the first research co-ordination meeting held in Debrecen, Hungary, from 4 to 7 October 1994

    International Nuclear Information System (INIS)

    Pashchenko, A.B.

    1995-02-01

    The report contains the Summary of the First IAEA Research Co-ordination Meeting (RCM) of the new Co-ordinated Research Programme (CRP) on ''Establishment of an International Reference Data Library of Nuclear Activation Cross Sections''. The meeting was organized by the IAEA Nuclear Data Section with co-operation and assistance of local organizers from the Institute of Experimental Physics and held in Debrecen, Hungary, from 4 to 7 October 1994. The purpose of the RCM was to discuss the scope and goals of the CRP, to report and evaluate the first results of the research carried out by each participating laboratory, to review the current tasks, identify further actions of participants and agree on the coordination of work under this CRP. The detailed agenda, the list of participants, conclusions and recommendations of the meeting are presented in the summary report. (author)

  2. Simplified polynomial representation of cross sections for reactor calculation

    International Nuclear Information System (INIS)

    Dias, A.M.; Sakai, M.

    1985-01-01

    It is shown a simplified representation of a cross section library generated by transport theory using the cell model of Wigner-Seitz for typical PWR fuel elements. The effect of burnup evolution through tables of reference cross sections and the effect of the variation of the reactor operation parameters considered by adjusted polynomials are presented. (M.C.K.) [pt

  3. Nuclear characteristics of Pu fueled LWR and cross section sensitivities

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Toshikazu [Osaka Univ., Suita (Japan). Faculty of Engineering

    1998-03-01

    The present status of Pu utilization to thermal reactors in Japan, nuclear characteristics and topics and cross section sensitivities for analysis of Pu fueled thermal reactors are described. As topics we will discuss the spatial self-shielding effect on the Doppler reactivity effect and the cross section sensitivities with the JENDL-3.1 and 3.2 libraries. (author)

  4. A Cross-Section Adjustment Method for Double Heterogeneity Problem in VHTGR Analysis

    International Nuclear Information System (INIS)

    Yun, Sung Hwan; Cho, Nam Zin

    2011-01-01

    Very High Temperature Gas-Cooled Reactors (VHTGRs) draw strong interest as candidates for a Gen-IV reactor concept, in which TRISO (tristructuralisotropic) fuel is employed to enhance the fuel performance. However, randomly dispersed TRISO fuel particles in a graphite matrix induce the so-called double heterogeneity problem. For design and analysis of such reactors with the double heterogeneity problem, the Monte Carlo method is widely used due to its complex geometry and continuous-energy capabilities. However, its huge computational burden, even in the modern high computing power, is still problematic to perform wholecore analysis in reactor design procedure. To address the double heterogeneity problem using conventional lattice codes, the RPT (Reactivityequivalent Physical Transformation) method considers a homogenized fuel region that is geometrically transformed to provide equivalent self-shielding effect. Another method is the coupled Monte Carlo/Collision Probability method, in which the absorption and nu-fission resonance cross-section libraries in the deterministic CPM3 lattice code are modified group-wise by the double heterogeneity factors determined by Monte Carlo results. In this paper, a new two-step Monte Carlo homogenization method is described as an alternative to those methods above. In the new method, a single cross-section adjustment factor is introduced to provide self-shielding effect equivalent to the self-shielding in heterogeneous geometry for a unit cell of compact fuel. Then, the homogenized fuel compact material with the equivalent cross-section adjustment factor is used in continuous-energy Monte Carlo calculation for various types of fuel blocks (or assemblies). The procedure of cross-section adjustment is implemented in the MCNP5 code

  5. Validation of a new 39 neutron group self-shielded library based on the nucleonics analysis of the Lotus fusion-fission hybrid test facility performed with the Monte Carlo code

    International Nuclear Information System (INIS)

    Pelloni, S.; Cheng, E.T.

    1985-02-01

    The Swiss LOTUS fusion-fission hybrid test facility was used to investigate the influence of the self-shielding of resonance cross sections on the tritium breeding and on the thorium ratios. Nucleonic analyses were performed using the discrete-ordinates transport codes ANISN and ONEDANT, the surface-flux code SURCU, and the version 3 of the MCNP code for the Li 2 CO 3 and the Li 2 O blanket designs with lead, thorium and beryllium multipliers. Except for the MCNP calculation which bases on the ENDF/B-V files, all nuclear data are generated from the ENDF/B-IV basic library. For the deterministic methods three NJOY group libraries were considered. The first, a 39 neutron group self-shielded library, was generated at EIR. The second bases on the same group structure as the first does and consists of infinitely diluted cross sections. Finally the third library was processed at LANL and consists of coupled 30+12 neutron and gamma groups; these cross sections are not self-shielded. The Monte Carlo analysis bases on a continuous and on a discrete 262 group library from the ENDF/B-V evaluation. It is shown that the results agree well within 3% between the unshielded libraries and between the different transport codes and theories. The self-shielding of resonance cross sections results in a decrease of the thorium capture rate and in an increase of the tritium breeding of about 6%. The remaining computed ratios are not affected by the self-shielding of cross sections. (Auth.)

  6. XCOM: Photon Cross Sections Database

    Science.gov (United States)

    SRD 8 XCOM: Photon Cross Sections Database (Web, free access)   A web database is provided which can be used to calculate photon cross sections for scattering, photoelectric absorption and pair production, as well as total attenuation coefficients, for any element, compound or mixture (Z <= 100) at energies from 1 keV to 100 GeV.

  7. Doppler broadening of cross sections

    International Nuclear Information System (INIS)

    Buckler, P.A.C.; Pull, I.C.

    1962-12-01

    Expressions for temperature dependent cross-sections in terms of resonance parameters are obtained, involving generalisations of the conventional Doppler functions, ψ and φ. Descriptions of Fortran sub-routines, which calculate broadened cross-sections in accordance with the derived formulae, are included. (author)

  8. Positive Scattering Cross Sections using Constrained Least Squares

    International Nuclear Information System (INIS)

    Dahl, J.A.; Ganapol, B.D.; Morel, J.E.

    1999-01-01

    A method which creates a positive Legendre expansion from truncated Legendre cross section libraries is presented. The cross section moments of order two and greater are modified by a constrained least squares algorithm, subject to the constraints that the zeroth and first moments remain constant, and that the standard discrete ordinate scattering matrix is positive. A method using the maximum entropy representation of the cross section which reduces the error of these modified moments is also presented. These methods are implemented in PARTISN, and numerical results from a transport calculation using highly anisotropic scattering cross sections with the exponential discontinuous spatial scheme is presented

  9. Measurements and Monte Carlo calculations of forward-angle secondary-neutron-production cross-sections for 137 and 200 MeV proton-induced reactions in carbon

    Science.gov (United States)

    Iwamoto, Yosuke; Hagiwara, Masayuki; Matsumoto, Tetsuro; Masuda, Akihiko; Iwase, Hiroshi; Yashima, Hiroshi; Shima, Tatsushi; Tamii, Atsushi; Nakamura, Takashi

    2012-10-01

    Secondary neutron-production double-differential cross-sections (DDXs) have been measured from interactions of 137 MeV and 200 MeV protons in a natural carbon target. The data were measured between 0° and 25° in the laboratory. DDXs were obtained with high energy resolution in the energy region from 3 MeV up to the maximum energy. The experimental data of 137 MeV protons at 10° and 25° were in good agreement with that of 113 MeV protons at 7.5° and 30° at LANSCE/WNR in the energy region below 80 MeV. Benchmark calculations were carried out with the PHITS code using the evaluated nuclear data files of JENDL/HE-2007 and ENDF/B-VII, and the theoretical models of Bertini-GEM and ISOBAR-GEM. For the 137 MeV proton incidence, calculations using JENDL/HE-2007 generally reproduced the shape and the intensity of experimental spectra well including the ground state of the 12N state produced by the 12C(p,n)12N reaction. For the 200 MeV proton incidence, all calculated results underestimated the experimental data by the factor of two except for the calculated result using ISOBAR model. ISOBAR predicts the nucleon emission to the forward angles qualitatively better than the Bertini model. These experimental data will be useful to evaluate the carbon data and as benchmark data for investigating the validity of the Monte Carlo simulation for the shielding design of accelerator facilities.

  10. ZZ SNLRML, Dosimetry Cross-Section Recommendations

    International Nuclear Information System (INIS)

    1996-01-01

    Description of program or function: Format: SAND-II; Number of groups: 640 group SAND-II group structure. Nuclides: Cd, B, Au, S, Ni, Li, F, Na, Mg, Al, Si, P, Sc, Ti, Mn, Fe, Co, Cu, Zn, Zr, Nb, Mo, Rh, Ag, In, I, Th, U, Np, Pu, Am. Origin: ENDF/B-VI, ENDF/B-V, IRDF-90, JENDL-3, JEF 2.2 and GLUCS data with special modifications from private communications. Weighting spectrum: flat. SNLRML is a reactor dosimetry library that draws upon all available evaluated cross section libraries and selects the best evaluation for application to research reactor spectrum determinations. Many of the components of the SNLRML come from the ENDF/B-VI and IRDF-90 (DLC-0161) libraries. The library format was selected for easy interface with spectrum determination codes such as SAND-II (CCC-0112 and LSL-M2 (PSR-233) and the new PSR-0345/SNL/SAND-II has been enhanced to interface with SNLRML. The data is recommended for spectrum determination applications and for the prediction of neutron activation of typical radiation sensor materials. The library has been tested for consistency of the cross section in wide variety of neutron environments. The results and cautions from this testing have been documented. The data has been interfaced with radiation transport codes, such as TWODANT-SYS (CCC-0547) and MCNP (CCC-0200), in order to compare calculated and measured activities for benchmark reactor experiments

  11. Measuring the top anti-t Production Cross-Section in the Electron + Jets Channel in Proton - Anti-proton Collisions at s**(1/2) = 1.96-TeV with the D0 Detector at the Tevatron: A Monte Carlo Study

    Energy Technology Data Exchange (ETDEWEB)

    Park, Su-Jung; /Bonn U.

    2004-02-01

    The measurement of the t{bar t} production cross section at {radical}s = 1.96 TeV using the final state with an electron and jets is studied with Monte Carlo event samples. All methods used in the real data analysis to measure efficiencies and to estimate the background contributions are examined. The studies focus on measuring the electron reconstruction efficiencies as well as on improving the electron identification and background suppression. With a generated input cross section of 7 pb the following result is obtained: {sigma}{sub t{bar t}} = (7 {+-} 1.63(stat){sub -1.14}{sup +0.94} (syst)) pb.

  12. Photon-splitting cross sections

    International Nuclear Information System (INIS)

    Johannessen, A.M.; Mork, K.J.; Overbo, I.

    1980-01-01

    The differential cross section for photon splitting (scattering of one photon into two photons) in a Coulomb field, obtained earlier by Shima, has been integrated numerically to yield various differential cross sections. Energy spectra differential with respect to the energy of one of the outgoing photons are presented for several values of the primary photon energy. Selected examples of recoil momentum distributions and some interesting doubly or multiply differential cross sections are also given. Values for the total cross section are obtained essentially for all energies. The screening effect caused by atomic electrons is also taken into account, and is found to be important for high energies, as in e + e - pair production. Comparisons with various approximate results obtained by previous authors mostly show fair agreement. We also discuss the possibilities for experimental detection and find the most promising candidate to be a measurement of both photons, and their energies, at a moderately high energy

  13. Neutron Cross Sections for Aluminium

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, Leif

    1963-08-15

    Total, elastic, inelastic, (n, 2n), (n, {alpha}), (n, p), and (n, {gamma}) cross sections for aluminium have been compiled from thermal to 100 MeV based upon literature search and theoretical interpolations and estimates. Differential elastic cross sections in the centre of mass system are represented by the Legendre coefficients. This method was chosen in order to obtain the best description of the energy dependence of the anisotropy.

  14. Accurate Cross Sections for Microanalysis

    OpenAIRE

    Rez, Peter

    2002-01-01

    To calculate the intensity of x-ray emission in electron beam microanalysis requires a knowledge of the energy distribution of the electrons in the solid, the energy variation of the ionization cross section of the relevant subshell, the fraction of ionizations events producing x rays of interest and the absorption coefficient of the x rays on the path to the detector. The theoretical predictions and experimental data available for ionization cross sections are limited mainly to K shells of a...

  15. 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.

  16. Electron collision cross section sets of TMS and TEOS vapours

    Science.gov (United States)

    Kawaguchi, S.; Takahashi, K.; Satoh, K.; Itoh, H.

    2017-05-01

    Reliable and detailed sets of electron collision cross sections for tetramethylsilane [TMS, Si(CH3)4] and tetraethoxysilane [TEOS, Si(OC2H5)4] vapours are proposed. The cross section sets of TMS and TEOS vapours include 16 and 20 kinds of partial ionization cross sections, respectively. Electron transport coefficients, such as electron drift velocity, ionization coefficient, and longitudinal diffusion coefficient, in those vapours are calculated by Monte Carlo simulations using the proposed cross section sets, and the validity of the sets is confirmed by comparing the calculated values of those transport coefficients with measured data. Furthermore, the calculated values of the ionization coefficient in TEOS/O2 mixtures are compared with measured data to confirm the validity of the proposed cross section set.

  17. Influence of nuclear cross section data at efficiency calculation of the 3He semiconductor neutron spectrometer

    International Nuclear Information System (INIS)

    Avdic, S.; Pesic, M.

    1992-01-01

    The ORTEC 580 Neutron Spectrometer system contains a detector unit in diode coincidence arrangement for measurement of fast neutron spectrum in the energy range from 1 MeV to 14 MeV. Numerical code HE3 for computation of semiconductor 3 He detector efficiency in a collimated neutron beam is based on analytical method in infinite diode approximation and Monte Carlo method for real spectrometer geometry. Calculations are performed in the first collision approximation in the detector active volume including evaluation of correction factors. Accuracy of relative detector efficiency calculation is improved by using neutron cross section from nuclear library ENDF/B-6. (author)

  18. Quality Quantification of Evaluated Cross Section Covariances

    International Nuclear Information System (INIS)

    Varet, S.; Dossantos-Uzarralde, P.; Vayatis, N.

    2015-01-01

    Presently, several methods are used to estimate the covariance matrix of evaluated nuclear cross sections. Because the resulting covariance matrices can be different according to the method used and according to the assumptions of the method, we propose a general and objective approach to quantify the quality of the covariance estimation for evaluated cross sections. The first step consists in defining an objective criterion. The second step is computation of the criterion. In this paper the Kullback-Leibler distance is proposed for the quality quantification of a covariance matrix estimation and its inverse. It is based on the distance to the true covariance matrix. A method based on the bootstrap is presented for the estimation of this criterion, which can be applied with most methods for covariance matrix estimation and without the knowledge of the true covariance matrix. The full approach is illustrated on the 85 Rb nucleus evaluations and the results are then used for a discussion on scoring and Monte Carlo approaches for covariance matrix estimation of the cross section evaluations

  19. Validation of the WIMSD4M cross-section generation code with benchmark results

    International Nuclear Information System (INIS)

    Deen, J.R.; Woodruff, W.L.; Leal, L.E.

    1995-01-01

    The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section libraries for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D 2 O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented

  20. Validation of the WIMSD4M cross-section generation code with benchmark results

    Energy Technology Data Exchange (ETDEWEB)

    Deen, J.R.; Woodruff, W.L. [Argonne National Lab., IL (United States); Leal, L.E. [Oak Ridge National Lab., TN (United States)

    1995-01-01

    The WIMSD4 code has been adopted for cross-section generation in support of the Reduced Enrichment Research and Test Reactor (RERTR) program at Argonne National Laboratory (ANL). Subsequently, the code has undergone several updates, and significant improvements have been achieved. The capability of generating group-collapsed micro- or macroscopic cross sections from the ENDF/B-V library and the more recent evaluation, ENDF/B-VI, in the ISOTXS format makes the modified version of the WIMSD4 code, WIMSD4M, very attractive, not only for the RERTR program, but also for the reactor physics community. The intent of the present paper is to validate the WIMSD4M cross-section libraries for reactor modeling of fresh water moderated cores. The results of calculations performed with multigroup cross-section data generated with the WIMSD4M code will be compared against experimental results. These results correspond to calculations carried out with thermal reactor benchmarks of the Oak Ridge National Laboratory (ORNL) unreflected HEU critical spheres, the TRX LEU critical experiments, and calculations of a modified Los Alamos HEU D{sub 2}O moderated benchmark critical system. The benchmark calculations were performed with the discrete-ordinates transport code, TWODANT, using WIMSD4M cross-section data. Transport calculations using the XSDRNPM module of the SCALE code system are also included. In addition to transport calculations, diffusion calculations with the DIF3D code were also carried out, since the DIF3D code is used in the RERTR program for reactor analysis and design. For completeness, Monte Carlo results of calculations performed with the VIM and MCNP codes are also presented.

  1. Contribution to the study of the unresolved resonance range of the neutrons cross sections

    International Nuclear Information System (INIS)

    Noguere, Gilles

    2014-01-01

    This document presents the statistical description of neutron cross sections in the unresolved resonance range. The modeling of the total cross section and of the 'shape - elastic' cross section is based on the 'average R-Matrix' formalism. The partial cross sections describing the radiative capture, elastic scattering, inelastic scattering and fission process are calculated using the Hauser-Feshbach formalism with width fluctuation corrections. In the unresolved resonance range, these models depend on the average resonance parameters (neutron strength function Sc, mean level spacing D c , average partial reaction widths Γ c , channel radius a c , effective radius R' and distant level parameter R-bar c ∞ ). The codes (NJOY, CALENDF...) dedicated to the processing of nuclear data libraries (JEFF, ENDF/B, JENDL, CENDL, BROND... ) use the average parameters to take into account the self-shielding phenomenon for the simulation of the neutron transport in Monte-Carlo (MCNP, TRIPOLI... ) and deterministic (APOLLO, ERANOS...) codes. The evaluation work consists in establishing a consistent set of average parameters as a function of the total angular momentum J of the system and of the orbital moment of the incident neutron l. The work presented in this paper aims to describe the links between the S-Matrix and the 'average R-Matrix' formalism for the calculation of Sc, R-bar c ∞ , ac and R'. (author) [fr

  2. ENDF/B-5 fission product cross section evaluations

    International Nuclear Information System (INIS)

    Schenter, R.E.; England, T.R.

    1979-12-01

    Cross section evaluations were made for the 196 fission product nuclides on the ENDF/B-5 data files. Most of the evaluations involve updating the capture cross sections of the important absorbers for fast and thermal reactor systems. This included updating thermal values, resonance integrals, resonance parameter sets, and fast capture cross sections. For the fast capture results generalized least-squares calculations were made with the computer code FERRET. Input for these cross section adjustments included nuclear models calculations and both integral and differential experimental data results. The differential cross sections and their uncertainties were obtained from the CSIRS library. Integral measurement results came from CFRMF and STEK Assemblies 500, 1000, 2000, 3000, 4000. Comparisons of these evaluations with recent capture measurements are shown. 15 figures, 10 tables

  3. [Absolute fission cross sections in the 14 MeV energy region]. Progress report, July 1982-June 1983

    International Nuclear Information System (INIS)

    1983-01-01

    Progress is reported on the following studies: thermal neutron absorption cross section of sulfur and the 252 Cf nu bar dilemma, the sigma (H)/sigma (Mn) cross section ratio, the sigma (H)/sigma (B) cross section ratio, 14 MeV neutron cross section measurements, beryllium-based pulsed neutron detector, and testing charged particle transport and Monte Carlo codes

  4. 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%

  5. Converting point-wise nuclear cross sections to pole representation using regularized vector fitting

    Science.gov (United States)

    Peng, Xingjie; Ducru, Pablo; Liu, Shichang; Forget, Benoit; Liang, Jingang; Smith, Kord

    2018-03-01

    Direct Doppler broadening of nuclear cross sections in Monte Carlo codes has been widely sought for coupled reactor simulations. One recent approach proposed analytical broadening using a pole representation of the commonly used resonance models and the introduction of a local windowing scheme to improve performance (Hwang, 1987; Forget et al., 2014; Josey et al., 2015, 2016). This pole representation has been achieved in the past by converting resonance parameters in the evaluation nuclear data library into poles and residues. However, cross sections of some isotopes are only provided as point-wise data in ENDF/B-VII.1 library. To convert these isotopes to pole representation, a recent approach has been proposed using the relaxed vector fitting (RVF) algorithm (Gustavsen and Semlyen, 1999; Gustavsen, 2006; Liu et al., 2018). This approach however needs to specify ahead of time the number of poles. This article addresses this issue by adding a poles and residues filtering step to the RVF procedure. This regularized VF (ReV-Fit) algorithm is shown to efficiently converge the poles close to the physical ones, eliminating most of the superfluous poles, and thus enabling the conversion of point-wise nuclear cross sections.

  6. Covariance matrices for nuclear cross sections derived from nuclear model calculations

    International Nuclear Information System (INIS)

    Smith, D. L.

    2005-01-01

    The growing need for covariance information to accompany the evaluated cross section data libraries utilized in contemporary nuclear applications is spurring the development of new methods to provide this information. Many of the current general purpose libraries of evaluated nuclear data used in applications are derived either almost entirely from nuclear model calculations or from nuclear model calculations benchmarked by available experimental data. Consequently, a consistent method for generating covariance information under these circumstances is required. This report discusses a new approach to producing covariance matrices for cross sections calculated using nuclear models. The present method involves establishing uncertainty information for the underlying parameters of nuclear models used in the calculations and then propagating these uncertainties through to the derived cross sections and related nuclear quantities by means of a Monte Carlo technique rather than the more conventional matrix error propagation approach used in some alternative methods. The formalism to be used in such analyses is discussed in this report along with various issues and caveats that need to be considered in order to proceed with a practical implementation of the methodology

  7. Inelastic neutron spectra and cross sections for 238 U

    International Nuclear Information System (INIS)

    Kornilov, N.V.; Kagalenko, A.V.

    1994-01-01

    The report discusses the experimental facilities of IPPE, results of spectra and cross sections investigations. The problems of existing data libraries were highlighted. Some of these problems for example, inelastic spectra at high energy may be solved by correct theoretical calculation. Others like level cross sections at E > 2 MeV and the possible structure of excitation function for group levels between 0.5 to 0.85 MeV demand new experimental efforts. 21 refs., 11 figs., 5 tabs

  8. Cross Sections for Inner-Shell Ionization by Electron Impact

    Energy Technology Data Exchange (ETDEWEB)

    Llovet, Xavier, E-mail: xavier@ccit.ub.edu [Centres Científics i Tecnològics, Universitat de Barcelona, Lluís Solé i Sabarís 1-3, 08028 Barcelona (Spain); Powell, Cedric J. [Materials Measurement Science Division, National Institute of Standards and Technology, Gaithersburg, Maryland 20899-8370 (United States); Salvat, Francesc [Facultat de Física (ECM and ICC), Universitat de Barcelona, Diagonal 645, 08028 Barcelona (Spain); Jablonski, Aleksander [Institute of Physical Chemistry, Polish Academy of Sciences, ul. Kasprzaka 44/52, 01-224 Warsaw (Poland)

    2014-03-15

    An analysis is presented of measured and calculated cross sections for inner-shell ionization by electron impact. We describe the essentials of classical and semiclassical models and of quantum approximations for computing ionization cross sections. The emphasis is on the recent formulation of the distorted-wave Born approximation by Bote and Salvat [Phys. Rev. A 77, 042701 (2008)] that has been used to generate an extensive database of cross sections for the ionization of the K shell and the L and M subshells of all elements from hydrogen to einsteinium (Z = 1 to Z = 99) by electrons and positrons with kinetic energies up to 1 GeV. We describe a systematic method for evaluating cross sections for emission of x rays and Auger electrons based on atomic transition probabilities from the Evaluated Atomic Data Library of Perkins et al. [Lawrence Livermore National Laboratory, UCRL-ID-50400, 1991]. We made an extensive comparison of measured K-shell, L-subshell, and M-subshell ionization cross sections and of Lα x-ray production cross sections with the corresponding calculated cross sections. We identified elements for which there were at least three (for K shells) or two (for L and M subshells) mutually consistent sets of cross-section measurements and for which the cross sections varied with energy as expected by theory. The overall average root-mean-square deviation between the measured and calculated cross sections was 10.9% and the overall average deviation was −2.5%. This degree of agreement between measured and calculated ionization and x-ray production cross sections was considered to be very satisfactory given the difficulties of these measurements.

  9. Terahertz radar cross section measurements

    DEFF Research Database (Denmark)

    Iwaszczuk, Krzysztof; Heiselberg, Henning; Jepsen, Peter Uhd

    2010-01-01

    We perform angle- and frequency-resolved radar cross section (RCS) measurements on objects at terahertz frequencies. Our RCS measurements are performed on a scale model aircraft of size 5-10 cm in polar and azimuthal configurations, and correspond closely to RCS measurements with conventional radar...

  10. Pion-nucleus cross sections

    International Nuclear Information System (INIS)

    Barashenkov, V.S.

    1990-01-01

    The tables of inelastic and total cross sections of π ± mesons interactions with nuclei 4 He- 238 U are presented. The tables are obtained by theoretical analysis of known experimental data for energies higher some tens of MeV. 1 ref.; 1 tab

  11. Parametric equations for calculation of macroscopic cross sections

    International Nuclear Information System (INIS)

    Botelho, Mario Hugo; Carvalho, Fernando

    2015-01-01

    Neutronic calculations of the core of a nuclear reactor is one thing necessary and important for the design and management of a nuclear reactor in order to prevent accidents and control the reactor efficiently as possible. To perform these calculations a library of nuclear data, including cross sections is required. Currently, to obtain a cross section computer codes are used, which require a large amount of processing time and computer memory. This paper proposes the calculation of macroscopic cross section through the development of parametric equations. The paper illustrates the proposal for the case of macroscopic cross sections of absorption (Σa), which was chosen due to its greater complexity among other cross sections. Parametric equations created enable, quick and dynamic way, the determination of absorption cross sections, enabling the use of them in calculations of reactors. The results show efficient when compared with the absorption cross sections obtained by the ALPHA 8.8.1 code. The differences between the cross sections are less than 2% for group 2 and less than 0.60% for group 1. (author)

  12. Preparation of next generation set of group cross sections. A task report to the Japan Nuclear Cycle Development Institute

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    2000-03-01

    The SLAROM code, performing fast reactor cell calculation based on a deterministic methodology, has been revised by adding the universal module PEACO of generating Ultra-fine group neutron spectra. The revised SLAROM, then, was utilized for evaluating reaction rate distributions in ZPPR-13A simulated by a 2-dim RZ homogeneous model, although actually ZPPR-13A composed of radially heterogeneous cells. The reaction rate distributions of ZPPR-13A were also calculated by the code MVP, that is a continuous energy Monte Carlo calculation code based on a probabilistic methodology. By comparing both results, it was concluded that the module PEACO has excellent capability for evaluating highly accurate effective cross sections. Also it was proved that the use of a new fine group cross section library set (next generation set), reflecting behavior of cross sections of structural materials, such as Fe and 0, in the fast neutron energy region, is indispensable for attaining a better agreement within 1% between both calculation methods. Also, for production of a next generation set of group cross sections, the code NJOY97.V107 was added to the group cross section production system and both front and end processing parts were prepared. This system was utilized to produce the new 70 group JFS-3 library using the evaluated nuclear data library JENDL-3.2. Furthermore, to confirm the capability of this new group cross section production system, the above new JFS-3 library was applied to core performance analysis of ZPPR-9 core with a 2-dim RZ homogeneous model and analysis of heterogeneous cells of ZPPR-9 core by using the deterministic method. Also the analysis using the code MVP was performed. By comparison of both results the following conclusion has been derived; the deterministic method, with the PEACO module for resonance cross sections, contributes to improve accuracy of predicting reaction rate distributions and Na void reactivity in fast reactor cores. And it becomes clear

  13. Monte-Carlo Method Python Library for dose distribution Calculation in Brachytherapy

    Energy Technology Data Exchange (ETDEWEB)

    Randriantsizafy, R D; Ramanandraibe, M J [Madagascar Institut National des Sciences et Techniques Nucleaires, Antananarivo (Madagascar); Raboanary, R [Institut of astro and High-Energy Physics Madagascar, University of Antananarivo, Antananarivo (Madagascar)

    2007-07-01

    The Cs-137 Brachytherapy treatment is performed in Madagascar since 2005. Time treatment calculation for prescribed dose is made manually. Monte-Carlo Method Python library written at Madagascar INSTN is experimentally used to calculate the dose distribution on the tumour and around it. The first validation of the code was done by comparing the library curves with the Nucletron company curves. To reduce the duration of the calculation, a Grid of PC's is set up with listner patch run on each PC. The library will be used to modelize the dose distribution in the CT scan patient picture for individual and better accuracy time calculation for a prescribed dose.

  14. Monte-Carlo Method Python Library for dose distribution Calculation in Brachytherapy

    International Nuclear Information System (INIS)

    Randriantsizafy, R.D.; Ramanandraibe, M.J.; Raboanary, R.

    2007-01-01

    The Cs-137 Brachytherapy treatment is performed in Madagascar since 2005. Time treatment calculation for prescribed dose is made manually. Monte-Carlo Method Python library written at Madagascar INSTN is experimentally used to calculate the dose distribution on the tumour and around it. The first validation of the code was done by comparing the library curves with the Nucletron company curves. To reduce the duration of the calculation, a Grid of PC's is set up with listner patch run on each PC. The library will be used to modelize the dose distribution in the CT scan patient picture for individual and better accuracy time calculation for a prescribed dose.

  15. Utilization of KENO-IV computer code with HANSEN-ROACH library

    International Nuclear Information System (INIS)

    Lima Barros, M. de; Vellozo, S.O.

    1982-01-01

    Several analysis with KENO-IV computer code, which is based in the Monte Carlo method, and the cross section library HANSEN-ROACH, were done, aiming to present the more convenient form to execute criticality calculations with this computer code and this cross sections. (E.G.) [pt

  16. Integrated system for production of neutronics and photonics calculational constants. Volume 21, Part C, Program SIGMAL (version 79-1): Doppler-broaden evaluated cross sections in the Livermore-Evaluated Nuclear Data Library (ENDL) format

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1979-01-01

    A code, SIGMAL, to Doppler-broaden evaluated cross sections in the ENDL format was designed. This code can Doppler-broaden cross sections that result from neutrons, protons, deuterons, tritons, 3 He, or alpha particles incident on any target nuclei. The code allows broadening to up to 100 final temperatures, either directly from the initial temperature or by bootstrapping to successively higher temperatures. 6 figures, 2 tables

  17. Handbook of LHC Higgs Cross Sections: 2. Differential Distributions

    CERN Document Server

    Dittmaier, S; Passarino, G; Tanaka, R; Alekhin, S; Alwall, J; Bagnaschi, E A; Banfi, A; Blumlein, J; Bolognesi, S; Chanon, N; Cheng, T; Cieri, L; Cooper-Sarkar, A M; Cutajar, M; Dawson, S; Davies, G; De Filippis, N; Degrassi, G; Denner, A; D'Enterria, D; Diglio, S; Di Micco, B; Di Nardo, R; Ellis, R K; Farilla, A; Farrington, S; Felcini, M; Ferrera, G; Flechl, M; de Florian, D; Forte, S; Ganjour, S; Garzelli, M V; Gascon-Shotkin, S; Glazov, S; Goria, S; Grazzini, M; Guillet, J -Ph; Hackstein, C; Hamilton, K; Harlander, R; Hauru, M; Heinemeyer, S; Hoche, S; Huston, J; Jackson, C; Jimenez-Delgado, P; Jorgensen, M D; Kado, M; Kallweit, S; Kardos, A; Kauer, N; Kim, H; Kovac, M; Kramer, M; Krauss, F; Kuo, C -M; Lehti, S; Li, Q; Lorenzo, N; Maltoni, F; Mellado, B; Moch, S O; Muck, A; Muhlleitner, M; Nadolsky, P; Nason, P; Neu, C; Nikitenko, A; Oleari, C; Olsen, J; Palmer, S; Paganis, S; Papadopoulos, C G; Petersen, T C; Petriello, F; Petrucci, F; Piacquadio, G; Pilon, E; Potter, C T; Price, J; Puljak, I; Quayle, W; Radescu, V; Rebuzzi, D; Reina, L; Rojo, J; Rosco, D; Salam, G P; Sapronov, A; Schaarschmidt, J; Schonherr, M; Schumacher, M; Siegert, F; Slavich, P; Spira, M; Stewart, I W; Stirling, W J; Stockli, F; Sturm, C; Tackmann, F J; Thorne, R S; Tommasini, D; Torrielli, P; Tramontano, F; Trocsanyi, Z; Ubiali, M; Uccirati, S; Acosta, M Vazquez; Vickey, T; Vicini, A; Waalewijn, W J; Wackeroth, D; Warsinsky, M; Weber, M; Wiesemann, M; Weiglein, G; Yu, J; Zanderighi, G

    2012-01-01

    This Report summarises the results of the second year's activities of the LHC Higgs Cross Section Working Group. The main goal of the working group was to present the state of the art of Higgs Physics at the LHC, integrating all new results that have appeared in the last few years. The first working group report Handbook of LHC Higgs Cross Sections: 1. Inclusive Observables (CERN-2011-002) focuses on predictions (central values and errors) for total Higgs production cross sections and Higgs branching ratios in the Standard Model and its minimal supersymmetric extension, covering also related issues such as Monte Carlo generators, parton distribution functions, and pseudo-observables. This second Report represents the next natural step towards realistic predictions upon providing results on cross sections with benchmark cuts, differential distributions, details of specific decay channels, and further recent developments.

  18. Highlights of top quark cross-section measurements at ATLAS

    Directory of Open Access Journals (Sweden)

    Berta Peter

    2017-01-01

    Full Text Available The highlights of the measurements of top quark production in proton-proton collisions at the Large Hadron Collider with the ATLAS detector are presented. The inclusive measurements of the top-pair production cross section have reached high precision and are compared to the best available theoretical calculations. The differential cross section measurements, including results using boosted top quarks, probe our understanding of top-pair production in the TeV regime. The results are compared to Monte Carlo generators implementing LO and NLO matrix elements matched with parton showers. Measurements of the single top quark production cross section are presented in the t-channel and s-channel, and with associated production with a W boson. For the t-channel production, results on the ratio between top quark and antitop quark production cross sections and differential measurements are also included.

  19. Neutron cross sections for fusion

    International Nuclear Information System (INIS)

    Haight, R.C.

    1979-10-01

    First generation fusion reactors will most likely be based on the 3 H(d,n) 4 He reaction, which produces 14-MeV neutrons. In these reactors, both the number of neutrons and the average neutron energy will be significantly higher than for fission reactors of the same power. Accurate neutron cross section data are therefore of great importance. They are needed in present conceptual designs to calculate neutron transport, energy deposition, nuclear transmutation including tritium breeding and activation, and radiation damage. They are also needed for the interpretation of radiation damage experiments, some of which use neutrons up to 40 MeV. In addition, certain diagnostic measurements of plasma experiments require nuclear cross sections. The quality of currently available data for these applications will be reviewed and current experimental programs will be outlined. The utility of nuclear models to provide these data also will be discussed. 65 references

  20. Negative ion detachment cross sections

    International Nuclear Information System (INIS)

    Champion, R.L.; Doverspike, L.D.

    1992-10-01

    The authors have measured absolute cross sections for electron detachment and charge exchange for collision of O and S with atomic hydrogen, have investigated the sputtering and photodesorption of negative ions from gas covered surfaces, and have begun an investigation of photon-induced field emission of electrons from exotic structures. Brief descriptions of these activities as well as future plans for these projects are given below

  1. Consistent evaluation of neutron cross sections for the 242-244Cm isotopes

    International Nuclear Information System (INIS)

    Ignatyuk, A.V.; Maslov, V.M.

    1989-01-01

    The knowledge of neutron cross-sections for Curium isotopes is necessary for solving the problems of the external fuel cycle. Experimental information on the cross-sections is very meager and does not satisfy requirements and existing evaluations in different libraries differ substantially for fission and (n,2n) reaction cross-sections. This situation requires a critical review of the entire set of evaluations of the neutron cross-sections for Curium. 17 refs, 3 figs

  2. Microscopic cross sections: An utopia?

    Energy Technology Data Exchange (ETDEWEB)

    Hilaire, S. [CEA Bruyeres-le-Chatel, DIF 91 (France); Koning, A.J. [Nuclear Research and Consultancy Group, PO Box 25, 1755 ZG Petten (Netherlands); Goriely, S. [Institut d' Astronomie et d' Astrophysique, Universite Libre de Bruxelles, Campus de la Plaine, CP 226, 1050 Brussels (Belgium)

    2010-07-01

    The increasing need for cross sections far from the valley of stability poses a challenge for nuclear reaction models. So far, predictions of cross sections have relied on more or less phenomenological approaches, depending on parameters adjusted to available experimental data or deduced from systematical relations. While such predictions are expected to be reliable for nuclei not too far from the experimentally known regions, it is clearly preferable to use more fundamental approaches, based on sound physical bases, when dealing with very exotic nuclei. Thanks to the high computer power available today, all major ingredients required to model a nuclear reaction can now be (and have been) microscopically (or semi-microscopically) determined starting from the information provided by a nucleon-nucleon effective interaction. We have implemented all these microscopic ingredients in the TALYS nuclear reaction code, and we are now almost able to perform fully microscopic cross section calculations. The quality of these ingredients and the impact of using them instead of the usually adopted phenomenological parameters will be discussed. (authors)

  3. Microscopic cross sections: An utopia?

    International Nuclear Information System (INIS)

    Hilaire, S.; Koning, A.J.; Goriely, S.

    2010-01-01

    The increasing need for cross sections far from the valley of stability poses a challenge for nuclear reaction models. So far, predictions of cross sections have relied on more or less phenomenological approaches, depending on parameters adjusted to available experimental data or deduced from systematical relations.While such predictions are expected to be reliable for nuclei not too far from the experimentally known regions, it is clearly preferable to use more fundamental approaches, based on sound physical bases, when dealing with very exotic nuclei. Thanks to the high computer power available today, all major ingredients required to model a nuclear reaction can now be (and have been) microscopically (or semi-microscopically) determined starting from the information provided by a nucleon-nucleon effective interaction. We have implemented all these microscopic ingredients in the TALYS nuclear reaction code, and we are now almost able to perform fully microscopic cross section calculations. The quality of these ingredients and the impact of using them instead of the usually adopted phenomenological parameters will be discussed. (authors)

  4. Porosity effects in the neutron total cross section of graphite

    International Nuclear Information System (INIS)

    Santisteban, J. R; Dawidowski, J; Petriw, S. N

    2009-01-01

    Graphite has been used in nuclear reactors since the birth of the nuclear industry due to its good performance as a neutron moderator material. Graphite is still an option as moderator for generation IV reactors due to its good mechanical and thermal properties at high operation temperatures. So, there has been renewed interest in a revision of the computer libraries used to describe the neutron cross section of graphite. For sub-thermal neutron energies, polycrystalline graphite shows a larger total cross section (between 4 and 8 barns) than predicted by existing theoretical models (0.2 barns). In order to investigate the origin of this discrepancy we measured the total cross section of graphite samples of three different origins, in the energy range from 0.001 eV to 10 eV. Different experimental arrangements and sample treatments were explored, to identify the effect of various experimental parameters on the total cross section measurement. The experiments showed that the increase in total cross section is due to neutrons scattered around the forward direction. We associate these small-angle scattered neutrons (SANS) to the porous structure of graphite, and formulate a very simple model to compute its contribution to the total cross section of the material. This results in an analytic expression that explicitly depends on the density and mean size of the pores, which can be easily incorporated in nuclear library codes. [es

  5. Neutron standard cross sections in reactor physics - Need and status

    International Nuclear Information System (INIS)

    Carlson, A.D.

    1990-01-01

    The design and improvement of nuclear reactors require detailed neutronics calculations. These calculations depend on comprehensive libraries of evaluated nuclear cross sections. Most of the cross sections that form the data base for these evaluations have been measured relative to neutron cross-section standards. The use of these standards can often simplify the measurement process by eliminating the need for a direct measurement of the neutron fluence. The standards are not known perfectly, however; thus the accuracy of a cross-section measurement is limited by the uncertainty in the standard cross section relative to which it is measured. Improvements in a standard cause all cross sections measured relative to that standard to be improved. This is the reason for the emphasis on improving the neutron cross-section standards. The continual process of measurement and evaluation has led to improvements in the accuracy and range of applicability of the standards. Though these improvements have been substantial, this process must continue in order to obtain the high-quality standards needed by the user community

  6. Poster - 18: New features in EGSnrc for photon cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Ali, Elsayed; Mainegra-Hing, Ernesto; Rogers, David W.O. [The Ottawa Hospital Cancer Centre, National Research Council Canada, Carleton University (Canada)

    2016-08-15

    Purpose: To implement two new features in the EGSnrc Monte Carlo system. The first is an option to account for photonuclear attenuation, which can contribute a few percent to the total cross section at the higher end of the energy range of interest to medical physics. The second is an option to use exact NIST XCOM photon cross sections. Methods: For the first feature, the photonuclear total cross sections are generated from the IAEA evaluated data. In the current, first-order implementation, after a photonuclear event, there is no energy deposition or secondary particle generation. The implementation is validated against deterministic calculations and experimental measurements of transmission signals. For the second feature, before this work, if the user explicitly requested XCOM photon cross sections, EGSnrc still used its own internal incoherent scattering cross sections. These differ by up to 2% from XCOM data between 30 keV and 40 MeV. After this work, exact XCOM incoherent scattering cross sections are an available option. Minor interpolation artifacts in pair and triplet XCOM cross sections are also addressed. The default for photon cross section in EGSnrc is XCOM except for the new incoherent scattering cross sections, which have to be explicitly requested. The photonuclear, incoherent, pair and triplet data from this work are available for elements and compounds for photon energies from 1 keV to 100 GeV. Results: Both features are implemented and validated in EGSnrc.Conclusions: The two features are part of the standard EGSnrc distribution as of version 4.2.3.2.

  7. ZZ DLC-11 RITTS, 121-Group Coupled Cross-Section for ANISN, DOT, MORSE

    International Nuclear Information System (INIS)

    1970-01-01

    A - Nature of physical problem solved: Format: ANISN, DTF-4, DOT and MORSE. Number of groups: 100 neutron energy groups (14.92 MeV to thermal) 21 gamma-ray energy groups (14.0 to 0.01 MeV) Nuclides: H, C, O, N, Na, Mg, P, S, Cl, K, and Ca, (microscopic cross sections) and 9 organic materials including 11-element standard man, 4-element standard man, skin, bone, tissue, brain, lung, red marrow, and muscle (macroscopic cross sections). Origin: ENDF/B for H, C, N, O, Na, and Mg; O5R library for Ca, S, and K; GAM-2 library for Cl; Evaluation by J.J. Ritts for P. Weighting spectrum: 1/E for the top 99 groups and Maxwellian for the thermal group values. DLC-11 data is suitable for neutron, gamma-ray, or coupled neutron and gamma-ray transport calculations. It is intended for use in multigroup discrete ordinates or Monte Carlo transport codes which treat anisotropic scattering by Legendre expansion up to order P3. DLC-11 is a collection of multigroup cross section data which were compiled by J. J. Ritts for use in depth-dose calculations in anthropomorphic phantoms. For convenience the data are grouped as follows - 1. A coupled 121-group (100 neutron, 21 gamma-ray) set of data for the 11 elements H, C, O, N. Na, Mg, P, S, Cl, K, and Ca. This set includes P3 coupled 121-group microscopic cross sections plus 121-group kerma factors for the 11 elements. 2. A 100-group set of neutron cross sections for the 11 elements. 3. A coupled 121-group set of macroscopic cross sections for 9 organic materials including 11-element standard man, 4-element standard man, skin, bone, tissue, brain, lung, red marrow, and muscle. B - Method of solution: The basic data sources were ENDF/B for H, C, N, O, Na, and Mg, the O5R library for Ca, S, and K, the GAM-2 library for Cl and an evaluation by Ritts for P. A 1/E spectrum was assumed for averaging the top 99 groups and a Maxwellian for averaging the thermal group values. The gamma-ray cross sections were computed from DLC-3/HPIC using MUG. The

  8. Wind Turbine Radar Cross Section

    Directory of Open Access Journals (Sweden)

    David Jenn

    2012-01-01

    Full Text Available The radar cross section (RCS of a wind turbine is a figure of merit for assessing its effect on the performance of electronic systems. In this paper, the fundamental equations for estimating the wind turbine clutter signal in radar and communication systems are presented. Methods of RCS prediction are summarized, citing their advantages and disadvantages. Bistatic and monostatic RCS patterns for two wind turbine configurations, a horizontal axis three-blade design and a vertical axis helical design, are shown. The unique electromagnetic scattering features, the effect of materials, and methods of mitigating wind turbine clutter are also discussed.

  9. New Tools to Prepare ACE Cross-section Files for MCNP Analytic Test Problems

    International Nuclear Information System (INIS)

    Brown, Forrest B.

    2016-01-01

    Monte Carlo calculations using one-group cross sections, multigroup cross sections, or simple continuous energy cross sections are often used to: (1) verify production codes against known analytical solutions, (2) verify new methods and algorithms that do not involve detailed collision physics, (3) compare Monte Carlo calculation methods with deterministic methods, and (4) teach fundamentals to students. In this work we describe 2 new tools for preparing the ACE cross-section files to be used by MCNP ® for these analytic test problems, simple a ce.pl and simple a ce m g.pl.

  10. ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kahler, A.C.; Herman, M.; Kahler,A.C.; MacFarlane,R.E.; Mosteller,R.D.; Kiedrowski,B.C.; Frankle,S.C.; Chadwick,M.B.; McKnight,R.D.; Lell,R.M.; Palmiotti,G.; Hiruta,H.; Herman,M.; Arcilla,R.; Mughabghab,S.F.; Sublet,J.C.; Trkov,A.; Trumbull,T.H.; Dunn,M.

    2011-12-01

    The ENDF/B-VII.1 library is the latest revision to the United States Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [M. B. Chadwick et al., 'ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data,' Nuclear Data Sheets, 112, 2887 (2011)]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected {sup 235}U and {sup 239}Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also

  11. ACTIV87 Fast neutron activation cross section file 1987

    International Nuclear Information System (INIS)

    Manokhin, V.N.; Pashchenko, A.B.; Plyaskin, V.I.; Bychkov, V.M.; Pronyaev, V.G.; Schwerer, O.

    1989-10-01

    This document summarizes the content of the Fast Neutron Activation Cross Section File based on data from different evaluated data libraries and individual evaluations in ENDF/B-5 format. The entire file or selective retrievals from it are available on magnetic tape, free of charge, from the IAEA Nuclear Data Section. (author)

  12. Generation of neutron scattering cross sections for silicon dioxide

    International Nuclear Information System (INIS)

    Ramos, R; Marquez Damian, J.I; Granada, J.R.; Cantargi, F

    2009-01-01

    A set of neutron scattering cross sections for silicon and oxygen bound in silicon dioxide were generated and validated. The cross sections were generated in the ACE format for MCNP using the nuclear data processing system NJOY, and the validation was done with published experimental data. This cross section library was applied to the calculation of five critical configurations published in the benchmark Critical Experiments with Heterogeneous Compositions of Highly Enriched Uranium, Silicon Dioxide and Polyethylene. The original calculations did not use the thermal scattering libraries generated in this work and presented significant differences with the experimental results. For this reason, the newly generated library was added to the input and the multiplication factor for each configuration was recomputed. The utilization of the thermal scattering libraries did not result in an improvement of the computational results. Based on this we conclude that integral experiments to validate this type of thermal cross sections need to be designed with a higher influence of thermal scattering in the measured result, and the experiments have to be performed under more controlled conditions. [es

  13. 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

  14. K+ nucleus total cross sections

    International Nuclear Information System (INIS)

    Sawafta, R.

    1990-01-01

    The scattering of K + mesons from nuclei has attracted considerable interest in the last few years. The K + holds a very special position as the weakest of all strongly interaction probes. The average cross section is not larger than about 10 mb at lab momenta below 800 MeV/c, corresponding to a mean free path in the nucleus larger than 5 fm. Thus the K + is capable of probing the entire volume of the nucleus. Single scattering of the K + with a nucleon in the nucleus dominates the nuclear scattering, and only small and calculable higher order corrections are needed. The nucleon is a dynamical entity and its internal structure can, in principle, be altered by its surrounding nuclear environment. This work reports an experiment in which the K + is used to compare the nucleon in the nucleus with a free nucleon

  15. Terahertz radar cross section measurements.

    Science.gov (United States)

    Iwaszczuk, Krzysztof; Heiselberg, Henning; Jepsen, Peter Uhd

    2010-12-06

    We perform angle- and frequency-resolved radar cross section (RCS) measurements on objects at terahertz frequencies. Our RCS measurements are performed on a scale model aircraft of size 5-10 cm in polar and azimuthal configurations, and correspond closely to RCS measurements with conventional radar on full-size objects. The measurements are performed in a terahertz time-domain system with freely propagating terahertz pulses generated by tilted pulse front excitation of lithium niobate crystals and measured with sub-picosecond time resolution. The application of a time domain system provides ranging information and also allows for identification of scattering points such as weaponry attached to the aircraft. The shapes of the models and positions of reflecting parts are retrieved by the filtered back projection algorithm.

  16. [Fast neutron cross section measurements

    International Nuclear Information System (INIS)

    1991-01-01

    In the 14 MeV Neutron Laboratory, we have continued the development of a facility that is now the only one of its kind in operation in the United States. We have refined the klystron bunching system described in last year's report to the point that 1.2 nanosecond pulses have been directly measured. We have tested the pulse shape discrimination capability of our primary NE 213 neutron detector. We have converted the RF sweeper section of the beamline to a frequency of 1 MHz to replace the function of the high voltage pulser described in last year's report which proved to be difficult to maintain and unreliable in its operation. We have also overcome several other significant experimental difficulties, including a major problem with a vacuum leak in the main accelerator column. We have completed additional testing to prove the remainder of the generation and measurement systems, but overcoming some of these experimental difficulties has delayed the start of actual data taking. We are now in a position to begin our first series of ring geometry elastic scattering measurements, and these will be underway before the end of the current contract year. As part of our longer term planning, we are continuing the conceptual analysis of several schemes to improve the intensity of our current pulsed beam. These include the provision of a duoplasmatron ion source and/or the provision of preacceleration bunching. Additional details are given later in this report. A series of measurements were carried out at the Tandem Dynamatron Facility involving the irradiation of a series of yttrium foils and the determination of activation cross sections using absolute counting techniques. The experimental work has been completed, and final analysis of the cross section data will be completed within several months

  17. Measurements and Monte Carlo calculations of neutron production cross-sections at 180{sup o} for the 140 MeV proton incident reactions on carbon, iron, and gold

    Energy Technology Data Exchange (ETDEWEB)

    Iwamoto, Yosuke, E-mail: iwamoto.yosuke@jaea.go.j [Japan Atomic Energy Agency, 2-4, Shirakatashirane, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Satoh, Daiki [Japan Atomic Energy Agency, 2-4, Shirakatashirane, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Hagiwara, Masayuki [KEK (Japan); Yashima, Hiroshi [Kyoto University (Japan); Nakane, Yoshihiro [Japan Atomic Energy Agency, 2-4, Shirakatashirane, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Tamii, Atsushi [Research Center for Nuclear Physics, Osaka University (Japan); Iwase, Hiroshi [KEK (Japan); Endo, Akira; Nakashima, Hiroshi; Sakamoto, Yukio [Japan Atomic Energy Agency, 2-4, Shirakatashirane, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Hatanaka, Kichiji [Research Center for Nuclear Physics, Osaka University (Japan); Niita, Koji [Research Organization for Information Science and Technology (Japan)

    2010-08-21

    The neutron production cross-sections of carbon, iron, and gold targets with 140 MeV protons at 180{sup o} were measured at the RCNP cyclotron facility. The time-of-flight technique was used to obtain the neutron energy spectra in the energy range above 1 MeV. The carbon and iron target results were compared with the experimental data from 113 MeV (p,xn) reactions at 150{sup o} reported by Meier et al. Our data agreed well with them in spite of different incident energies and angles. Calculations were then performed using different intra-nuclear cascade models (Bertini, ISOBAR, and JQMD) implemented with PHITS code. The results calculated using the ISOBAR and JQMD models roughly agreed with the experimental iron and gold target data, but the Bertini could not reproduce the high-energy neutrons above 10 MeV.

  18. Neutron cross sections: Book of curves

    International Nuclear Information System (INIS)

    McLane, V.; Dunford, C.L.; Rose, P.F.

    1988-01-01

    Neuton Cross Sections: Book of Curves represents the fourth edition of what was previously known as BNL-325, Neutron Cross Sections, Volume 2, CURVES. Data is presented only for (i.e., intergrated) reaction cross sections (and related fission parameters) as a function of incident-neutron energy for the energy range 0.01 eV to 200 MeV. For the first time, isometric state production cross sections have been included. 11 refs., 4 figs

  19. Nuclear Forensics and Radiochemistry: Cross Sections

    Energy Technology Data Exchange (ETDEWEB)

    Rundberg, Robert S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-11-08

    The neutron activation of components in a nuclear device can provide useful signatures of weapon design or sophistication. This lecture will cover some of the basics of neutron reaction cross sections. Nuclear reactor cross sections will also be presented to illustrate the complexity of convolving neutron energy spectra with nuclear excitation functions to calculate useful effective reactor cross sections. Deficiencies in the nuclear database will be discussed along with tools available at Los Alamos to provide new neutron cross section data.

  20. JENDL gas-production cross section file

    International Nuclear Information System (INIS)

    Nakagawa, Tsuneo; Narita, Tsutomu

    1992-05-01

    The JENDL gas-production cross section file was compiled by taking cross-section data from JENDL-3 and by using the ENDF-5 format. The data were given to 23 nuclei or elements in light nuclei and structural materials. Graphs of the cross sections and brief description on their evaluation methods are given in this report. (author)

  1. Integral nucleus-nucleus cross sections

    International Nuclear Information System (INIS)

    Barashenkov, V.S.; Kumawat, H.

    2003-01-01

    Expressions approximating the experimental integral cross sections for elastic and inelastic interactions of light and heavy nuclei at the energies up to several GeV/nucleon are presented. The calculated cross sections are inside the corridor of experimental errors or very close to it. Described in detail FORTRAN code and a numerical example of the cross section approximation are also presented

  2. Double-differential heavy-ion production cross sections

    International Nuclear Information System (INIS)

    Miller, T. M.; Townsend, L. W.

    2004-01-01

    Current computational tools used for space or accelerator shielding studies transport energetic heavy ions either using a one-dimensional straight-ahead approximation or by dissociating the nuclei into protons and neutrons and then performing neutron and proton transport using Monte Carlo techniques. Although the heavy secondary particles generally travel close to the beam direction, a proper treatment of the light ions produced in these reactions requires that double-differential cross sections should be utilised. Unfortunately, no fundamental nuclear model capable of serving as an event generator to provide these cross sections for all ions and energies of interest exists currently. Herein, we present a model for producing double-differential heavy-ion production cross sections that uses heavy-ion fragmentation yields produced by the NUCFRG2 fragmentation code coupled with a model of energy degradation in nucleus-nucleus collisions and systematics of momentum distributions to provide energy and angular dependences of the heavy-ion production. (authors)

  3. From cutting-edge pointwise cross-section to groupwise reaction rate: A primer

    Science.gov (United States)

    Sublet, Jean-Christophe; Fleming, Michael; Gilbert, Mark R.

    2017-09-01

    The nuclear research and development community has a history of using both integral and differential experiments to support accurate lattice-reactor, nuclear reactor criticality and shielding simulations, as well as verification and validation efforts of cross sections and emitted particle spectra. An important aspect to this type of analysis is the proper consideration of the contribution of the neutron spectrum in its entirety, with correct propagation of uncertainties and standard deviations derived from Monte Carlo simulations, to the local and total uncertainty in the simulated reactions rates (RRs), which usually only apply to one application at a time. This paper identifies deficiencies in the traditional treatment, and discusses correct handling of the RR uncertainty quantification and propagation, including details of the cross section components in the RR uncertainty estimates, which are verified for relevant applications. The methodology that rigorously captures the spectral shift and cross section contributions to the uncertainty in the RR are discussed with quantified examples that demonstrate the importance of the proper treatment of the spectrum profile and cross section contributions to the uncertainty in the RR and subsequent response functions. The recently developed inventory code FISPACT-II, when connected to the processed nuclear data libraries TENDL-2015, ENDF/B-VII.1, JENDL-4.0u or JEFF-3.2, forms an enhanced multi-physics platform providing a wide variety of advanced simulation methods for modelling activation, transmutation, burnup protocols and simulating radiation damage sources terms. The system has extended cutting-edge nuclear data forms, uncertainty quantification and propagation methods, which have been the subject of recent integral and differential, fission, fusion and accelerators validation efforts. The simulation system is used to accurately and predictively probe, understand and underpin a modern and sustainable understanding

  4. 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)

  5. New SCALE-4 features related to cross-section processing

    International Nuclear Information System (INIS)

    Petrie, L.M.; Landers, N.F.; Greene, N.M.; Parks, C.V.

    1991-01-01

    The SCALE code system has a standardized scheme for processing problem-dependent cross section from problem-independent waste libraries. Some improvements and new capabilities in the processing scheme have been incorporated into the new Version 4 release of the SCALE system. The new features include the capability to consider annular cylindrical and spherical unit cells, and improved Dancoff factor formulation, and changes to the NITAWL-II module to perform resonance self-shielding with reference to infinite dilute values. A review of these major changes in the cross-section processing scheme for SCALE-4 is presented in this paper

  6. Uncertainties and correlations for the 56Fe damage cross sections and spectra averaged quantities based on TENDL-TMC

    International Nuclear Information System (INIS)

    Simakov, S.P.; Konobeyev, A.Yu.; Koning, A.

    2016-01-01

    The goal of this work is a calculation of the covariance matrices for the physical quantities used to characterize the neutron induced radiation damage in the materials. Such quantities usually encompass: the charged particles kinetic energy deposition KERMA (locally deposited nuclear heating), damage energy (to calculate then the number of displaced atoms) and gas production cross sections [(n,xα), (n,xt), (n,xp) … to calculate then transmuting of target nuclei to gases]. The uncertainties and energy-energy or reaction-reaction correlations for such quantities were not assessed so far, whereas the covariances for many underlying cross sections are often presented in the evaluated data libraries. Due to the dependence of damage quantities on many reactions channels, on both total and differential cross sections, and in particular on the energy distribution of reaction recoils, the evaluation of uncertainty is not straightforward. To reach a goal, we used the method based on idea of Total Monte Carlo application to the Nuclear Data. This report summarises the current results for evaluation, validation and representation in the ENDF-6 format of the radiation damage covariances for n + 56 Fe from thermal energy up to 20 MeV. This study was motivated by the IAEA Coordinated Research Project ''Primary Radiation Damage Cross Sections'' and by present dedicated Technical Meeting “Nuclear Reaction Data and Uncertainties for Radiation Damage”

  7. 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)

  8. 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

  9. 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

  10. [Fast neutron cross section measurements

    International Nuclear Information System (INIS)

    Knoll, G.F.

    1992-01-01

    From its inception, the Nuclear Data Project at the University of Michigan has concentrated on two major objectives: (1) to carry out carefully controlled nuclear measurements of the highest possible reliability in support of the national nuclear data program, and (2) to provide an educational opportunity for students with interests in experimental nuclear science. The project has undergone a successful transition from a primary dependence on our photoneutron laboratory to one in which our current research is entirely based on a unique pulsed 14 MeV fast neutron facility. The new experimental facility is unique in its ability to provide nanosecond bursts of 14 MeV neutrons under conditions that are ''clean'' and as scatter-free as possible, and is the only one of its type currently in operation in the United States. It has been designed and put into operation primarily by graduate students, and has met or exceeded all of its important initial performance goals. We have reached the point of its routine operation, and most of the data are now in hand that will serve as the basis for the first two doctoral dissertations to be written by participating graduate students. Our initial results on double differential neutron cross sections will be presented at the May 1993 Fusion Reactor Technology Workshop. We are pleased to report that, after investing several years in equipment assembly and optimization, the project has now entered its ''data production'' phase

  11. Relativistic photon-Maxwellian electron cross sections

    International Nuclear Information System (INIS)

    Wienke, B.R.; Lathrop, B.L.; Devaney, J.J.

    1986-01-01

    Temperature corrected cross sections, complementing the Klein-Nishina set, are developed for astrophysical, plasma, and transport applications. The set is obtained from a nonlinear least squares fit to the exact photon-Maxwellian electron cross sections, using the static formula as the asymptotic basis. Two parameters are sufficient (two decimal places) to fit the exact cross sections over a range of 0-100 keV in electron temperature, and 0-1 MeV in incident photon energy. The fit is made to the total cross sections, yet the parameters predict both total and differential scattering cross sections well. Corresponding differential energy cross sections are less accurate. An extended fit to (just) the total cross sections, over the temperature and energy range 0-5 MeV, is also described. (author)

  12. EDDIX--a database of ionisation double differential cross sections.

    Science.gov (United States)

    MacGibbon, J H; Emerson, S; Liamsuwan, T; Nikjoo, H

    2011-02-01

    The use of Monte Carlo track structure is a choice method in biophysical modelling and calculations. To precisely model 3D and 4D tracks, the cross section for the ionisation by an incoming ion, double differential in the outgoing electron energy and angle, is required. However, the double differential cross section cannot be theoretically modelled over the full range of parameters. To address this issue, a database of all available experimental data has been constructed. Currently, the database of Experimental Double Differential Ionisation Cross sections (EDDIX) contains over 1200 digitalised experimentally measured datasets from the 1960s to present date, covering all available ion species (hydrogen to uranium) and all available target species. Double differential cross sections are also presented with the aid of an eight parameter functions fitted to the cross sections. The parameters include projectile species and charge, target nuclear charge and atomic mass, projectile atomic mass and energy, electron energy and deflection angle. It is planned to freely distribute EDDIX and make it available to the radiation research community for use in the analytical and numerical modelling of track structure.

  13. 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

  14. 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.

  15. ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Kahler, A. [Los Alamos National Laboratory (LANL); Macfarlane, R E [Los Alamos National Laboratory (LANL); Mosteller, R D [Los Alamos National Laboratory (LANL); Kiedrowski, B C [Los Alamos National Laboratory (LANL); Frankle, S C [Los Alamos National Laboratory (LANL); Chadwick, M. B. [Los Alamos National Laboratory (LANL); Mcknight, R D [Argonne National Laboratory (ANL); Lell, R M [Argonne National Laboratory (ANL); Palmiotti, G [Idaho National Laboratory (INL); Hiruta, h [Idaho National Laboratory (INL); Herman, Micheal W [Brookhaven National Laboratory (BNL); Arcilla, r [Brookhaven National Laboratory (BNL); Mughabghab, S F [Brookhaven National Laboratory (BNL); Sublet, J C [Culham Science Center, Abington, UK; Trkov, A. [Jozef Stefan Institute, Slovenia; Trumbull, T H [Knolls Atomic Power Laboratory; Dunn, Michael E [ORNL

    2011-01-01

    The ENDF/B-VII.1 library is the latest revision to the United States' Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [1]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unrnoderated and uranium reflected (235)U and (239)Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as (236)U; (238,242)Pu and (241,243)Am capture in fast systems. Other deficiencies, such as the overprediction of Pu solution system critical

  16. Evaluation of fission product neutron cross sections for JENDL

    International Nuclear Information System (INIS)

    1984-01-01

    The recent activities on the evaluation of fission product (FP) neutron cross sections for JENDL (Japanese Evaluated Nuclear Data Library) are presented briefly. The integral test of JENDL-1 FP cross section file was performed using the CFRMF sample activation data and the STEK sample reactivity data, and the ratio of experiment to calculation was nearly constant for all the samples in the STEK measurement. Therefore, a tentative analysis was performed by applying the correction to the calculated scattering reactivity component. Better agreement with the experiment was obtained after applying this correction in most cases. The evaluation work on the JENDL-2 FP neutron cross sections is now in progress. The improvement of the data evaluation is presented in an itemized form. The JENDL-2 FP file will contain the evaluated data for 100 nuclides from Kr to Tb. The improvement and the future scope of the integral test for JENDL-2 FP data are summarized. (Asami, T.)

  17. Reaction cross section calculation of some alkaline earth elements

    Science.gov (United States)

    Tel, Eyyup; Kavun, Yusuf; Sarpün, Ismail Hakki

    2017-09-01

    Reaction cross section knowledge is crucial to application nuclear physics such as medical imaging, radiation shielding and material evaluations. Nuclear reaction codes can be used if the experimental data are unavailable or are improbably to be produced because of the experimental trouble. In this study, there action cross sections of some target alkaline earth elements have been calculated by using pre-equilibrium and equilibrium nuclear reaction models for nucleon induced reactions. While these calculations, the Hybrid Model, the Geometry Dependent Hybrid Model, the Full Exciton Model, the Cascade Exciton Model for pre-equilibrium reactions and the Weisskopf-Ewing Model for equilibrium reactions have been used. The calculated cross sections have been discussed and compared with the experimental data taken from Experimental Nuclear Reaction Data library.

  18. Reaction cross section calculation of some alkaline earth elements

    Directory of Open Access Journals (Sweden)

    Tel Eyyup

    2017-01-01

    Full Text Available Reaction cross section knowledge is crucial to application nuclear physics such as medical imaging, radiation shielding and material evaluations. Nuclear reaction codes can be used if the experimental data are unavailable or are improbably to be produced because of the experimental trouble. In this study, there action cross sections of some target alkaline earth elements have been calculated by using pre-equilibrium and equilibrium nuclear reaction models for nucleon induced reactions. While these calculations, the Hybrid Model, the Geometry Dependent Hybrid Model, the Full Exciton Model, the Cascade Exciton Model for pre-equilibrium reactions and the Weisskopf-Ewing Model for equilibrium reactions have been used. The calculated cross sections have been discussed and compared with the experimental data taken from Experimental Nuclear Reaction Data library.

  19. Documentation for WIMSD-formatted libraries based on ENDF/B-VII.1 evaluated nuclear data files with extended actinide burn-up chains and cross section data up to 2000 K for fuel materials

    International Nuclear Information System (INIS)

    López Aldama, Daniel

    2014-11-01

    In the frame of WIMS Library Update Project the WIMSD-IAEA-69 and WIMSD-IAEA-172 libraries were prepared and made available at the Nuclear Data Section (NDS) of the International Atomic Energy Agency (IAEA). The main libraries were prepared from different sources of evaluated nuclear data that were available before December 2003. Also others WIMSD libraries were prepared from the major evaluated nuclear data libraries and made available at http://www-nds.iaea.org/wimsd. During the last ten years new libraries have been prepared every time that a major version of an evaluated nuclear data library has been released, namely JEFF-3.1 and ENDF/B-VII.0. Recently, end-users have requested to extend the temperature ranges of fuel materials included in the libraries and also to extend the burn-up chains to higher actinides up to Cf-254. The inclusion of new structural materials, like bismuth, has been also considered. Therefore, new WIMSD-formatted libraries in the 69- and 172-energy structure have been prepared with more materials, extended actinides burn-up chains and higher temperatures in thermal and resonance range

  20. On the treatment of ill-conditioned cases in the Monte Carlo library least-squares approach for inverse radiation analyzers

    Science.gov (United States)

    Meric, Ilker; Johansen, Geir A.; Holstad, Marie B.; Mattingly, John; Gardner, Robin P.

    2012-05-01

    Prompt gamma-ray neutron activation analysis (PGNAA) has been and still is one of the major methods of choice for the elemental analysis of various bulk samples. This is mostly due to the fact that PGNAA offers a rapid, non-destructive and on-line means of sample interrogation. The quantitative analysis of the prompt gamma-ray data could, on the other hand, be performed either through the single peak analysis or the so-called Monte Carlo library least-squares (MCLLS) approach, of which the latter has been shown to be more sensitive and more accurate than the former. The MCLLS approach is based on the assumption that the total prompt gamma-ray spectrum of any sample is a linear combination of the contributions from the individual constituents or libraries. This assumption leads to, through the minimization of the chi-square value, a set of linear equations which has to be solved to obtain the library multipliers, a process that involves the inversion of the covariance matrix. The least-squares solution may be extremely uncertain due to the ill-conditioning of the covariance matrix. The covariance matrix will become ill-conditioned whenever, in the subsequent calculations, two or more libraries are highly correlated. The ill-conditioning will also be unavoidable whenever the sample contains trace amounts of certain elements or elements with significantly low thermal neutron capture cross-sections. In this work, a new iterative approach, which can handle the ill-conditioning of the covariance matrix, is proposed and applied to a hydrocarbon multiphase flow problem in which the parameters of interest are the separate amounts of the oil, gas, water and salt phases. The results of the proposed method are also compared with the results obtained through the implementation of a well-known regularization method, the truncated singular value decomposition. Final calculations indicate that the proposed approach would be able to treat ill-conditioned cases appropriately.

  1. On the treatment of ill-conditioned cases in the Monte Carlo library least-squares approach for inverse radiation analyzers

    International Nuclear Information System (INIS)

    Meric, Ilker; Johansen, Geir A; Holstad, Marie B; Mattingly, John; Gardner, Robin P

    2012-01-01

    Prompt gamma-ray neutron activation analysis (PGNAA) has been and still is one of the major methods of choice for the elemental analysis of various bulk samples. This is mostly due to the fact that PGNAA offers a rapid, non-destructive and on-line means of sample interrogation. The quantitative analysis of the prompt gamma-ray data could, on the other hand, be performed either through the single peak analysis or the so-called Monte Carlo library least-squares (MCLLS) approach, of which the latter has been shown to be more sensitive and more accurate than the former. The MCLLS approach is based on the assumption that the total prompt gamma-ray spectrum of any sample is a linear combination of the contributions from the individual constituents or libraries. This assumption leads to, through the minimization of the chi-square value, a set of linear equations which has to be solved to obtain the library multipliers, a process that involves the inversion of the covariance matrix. The least-squares solution may be extremely uncertain due to the ill-conditioning of the covariance matrix. The covariance matrix will become ill-conditioned whenever, in the subsequent calculations, two or more libraries are highly correlated. The ill-conditioning will also be unavoidable whenever the sample contains trace amounts of certain elements or elements with significantly low thermal neutron capture cross-sections. In this work, a new iterative approach, which can handle the ill-conditioning of the covariance matrix, is proposed and applied to a hydrocarbon multiphase flow problem in which the parameters of interest are the separate amounts of the oil, gas, water and salt phases. The results of the proposed method are also compared with the results obtained through the implementation of a well-known regularization method, the truncated singular value decomposition. Final calculations indicate that the proposed approach would be able to treat ill-conditioned cases appropriately. (paper)

  2. Thermal neutron scattering cross sections of beryllium and magnesium oxides

    International Nuclear Information System (INIS)

    Al-Qasir, Iyad; Jisrawi, Najeh; Gillette, Victor; Qteish, Abdallah

    2016-01-01

    Highlights: • Neutron thermalization in BeO and MgO was studied using Ab initio lattice dynamics. • The BeO phonon density of states used to generate the current ENDF library has issues. • The BeO cross sections can provide a more accurate ENDF library than the current one. • For MgO an ENDF library is lacking: a new accurate one can be built from our results. • BeO is a better filter than MgO, especially when cooled down to 77 K. - Abstract: Alkaline-earth beryllium and magnesium oxides are fundamental materials in nuclear industry and thermal neutron scattering applications. The calculation of the thermal neutron scattering cross sections requires a detailed knowledge of the lattice dynamics of the scattering medium. The vibrational properties of BeO and MgO are studied using first-principles calculations within the frame work of the density functional perturbation theory. Excellent agreement between the calculated phonon dispersion relations and the experimental data have been obtained. The phonon densities of states are utilized to calculate the scattering laws using the incoherent approximation. For BeO, there are concerns about the accuracy of the phonon density of states used to generate the current ENDF/B-VII.1 libraries. These concerns are identified, and their influences on the scattering law and inelastic scattering cross section are analyzed. For MgO, no up to date thermal neutron scattering cross section ENDF library is available, and our results represent a potential one for use in different applications. Moreover, the BeO and MgO efficiencies as neutron filters at different temperatures are investigated. BeO is found to be a better filter than MgO, especially when cooled down, and cooling MgO below 77 K does not significantly improve the filter’s efficiency.

  3. 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

  4. Verification of the cross-section and depletion chain processing module of DRAGON 3.06

    International Nuclear Information System (INIS)

    Chambon, R.; Marleau, G.; Zkiek, A.

    2008-01-01

    In this paper we present a verification of the module of the lattice code DRAGON 3.06 used for processing microscopic cross-section libraries, including their associated depletion chain. This verification is performed by reprogramming the capabilities of DRAGON in another language (MATLAB) and testing them on different problems typical of the CANDU reactor. The verification procedure consists in first programming MATLAB m-files to read the different cross section libraries in ASCII format and to compute the reference cross-sections and depletion chains. The same information is also recovered from the output files of DRAGON (using different m-files) and the resulting cross sections and depletion chain are compared with the reference library, the differences being evaluated and tabulated. The results show that the cross-section calculations and the depletion chains are correctly processed in version 3.06 of DRAGON. (author)

  5. On-the-fly doppler broadening for Monte Carlo codes

    International Nuclear Information System (INIS)

    Yesilyurt, G.; Martin, W. R.; Brown, F. B.

    2009-01-01

    A methodology to allow on-the-fly Doppler broadening of neutron cross sections for use in Monte Carlo codes has been developed. The Monte Carlo code only needs to store 0 K cross sections for each isotope and the method will broaden the 0 K cross sections for any isotope in the library to any temperature in the range 77 K-3200 K. The methodology is based on a combination of Taylor series expansions and asymptotic series expansions. The type of series representation was determined by investigating the temperature dependence of U3o8 resonance cross sections in three regions: near the resonance peaks, mid-resonance, and the resonance wings. The coefficients for these series expansions were determined by a regression over the energy and temperature range of interest. Since the resonance parameters are a function of the neutron energy and target nuclide, the ψ and χ functions in the Adler-Adler multi-level resonance model can be represented by series expansions in temperature only, allowing the least number of terms to approximate the temperature dependent cross sections within a given accuracy. The comparison of the broadened cross sections using this methodology with the NJOY cross sections was excellent over the entire temperature range (77 K-3200 K) and energy range. A Monte Carlo code was implemented to apply the combined regression model and used to estimate the additional computing cost which was found to be less than <1%. (authors)

  6. Background-cross-section-dependent subgroup parameters

    International Nuclear Information System (INIS)

    Yamamoto, Toshihisa

    2003-01-01

    A new set of subgroup parameters was derived that can reproduce the self-shielded cross section against a wide range of background cross sections. The subgroup parameters are expressed with a rational equation which numerator and denominator are expressed as the expansion series of background cross section, so that the background cross section dependence is exactly taken into account in the parameters. The advantage of the new subgroup parameters is that they can reproduce the self-shielded effect not only by group basis but also by subgroup basis. Then an adaptive method is also proposed which uses fitting procedure to evaluate the background-cross-section-dependence of the parameters. One of the simple fitting formula was able to reproduce the self-shielded subgroup cross section by less than 1% error from the precise evaluation. (author)

  7. Scattering cross section for various potential systems

    Directory of Open Access Journals (Sweden)

    Myagmarjav Odsuren

    2017-08-01

    Full Text Available We discuss the problems of scattering in this framework, and show that the applied method is very useful in the investigation of the effect of the resonance in the observed scattering cross sections. In this study, not only the scattering cross sections but also the decomposition of the scattering cross sections was computed for the α–α system. To obtain the decomposition of scattering cross sections into resonance and residual continuum terms, the complex scaled orthogonality condition model and the extended completeness relation are used. Applying the present method to the α–α and α–n systems, we obtained good reproduction of the observed phase shifts and cross sections. The decomposition into resonance and continuum terms makes clear that resonance contributions are dominant but continuum terms and their interference are not negligible. To understand the behavior of observed phase shifts and the shape of the cross sections, both resonance and continuum terms are calculated.

  8. Scattering cross section for various potential systems

    Energy Technology Data Exchange (ETDEWEB)

    Odsuren, Myagmarjav; Khuukhenkhuu, Gonchigdorj; Davaa, Suren [Nuclear Research Center, School of Engineering and Applied Sciences, National University of Mongolia, Ulaanbaatar (Mongolia); Kato, Kiyoshi [Nuclear Reaction Data Centre, Faculty of Science, Hokkaido University, Sapporo (Japan)

    2017-08-15

    We discuss the problems of scattering in this framework, and show that the applied method is very useful in the investigation of the effect of the resonance in the observed scattering cross sections. In this study, not only the scattering cross sections but also the decomposition of the scattering cross sections was computed for the α–α system. To obtain the decomposition of scattering cross sections into resonance and residual continuum terms, the complex scaled orthogonality condition model and the extended completeness relation are used. Applying the present method to the α–α and α–n systems, we obtained good reproduction of the observed phase shifts and cross sections. The decomposition into resonance and continuum terms makes clear that resonance contributions are dominant but continuum terms and their interference are not negligible. To understand the behavior of observed phase shifts and the shape of the cross sections, both resonance and continuum terms are calculated.

  9. New software library of geometrical primitives for modelling of solids used in Monte Carlo detector simulations

    CERN Multimedia

    CERN. Geneva

    2012-01-01

    We present our effort for the creation of a new software library of geometrical primitives, which are used for solid modelling in Monte Carlo detector simulations. We plan to replace and unify current geometrical primitive classes in the CERN software projects Geant4 and ROOT with this library. Each solid is represented by a C++ class with methods suited for measuring distances of particles from the surface of a solid and for determination as to whether the particles are located inside, outside or on the surface of the solid. We use numerical tolerance for determining whether the particles are located on the surface. The class methods also contain basic support for visualization. We use dedicated test suites for validation of the shape codes. These include also special performance and numerical value comparison tests for help with analysis of possible candidates of class methods as well as to verify that our new implementation proposals were designed and implemented properly. Currently, bridge classes are u...

  10. Cross-section methodology in SIMMER

    International Nuclear Information System (INIS)

    Soran, P.D.

    1975-11-01

    The cross-section methodology incorporated in the SIMMER code is described. Data base for all cross sections is the ENDF/B system with various progressing computer codes to group collapse and modify the group constants which are used in SIMMER. Either infinitely dilute cross sections or the Bondarenko formalism can be used in SIMMER. Presently only a microscopic treatment is considered, but preliminary macroscopic algorithms have been investigated

  11. Cross-section methodology in SIMMER

    International Nuclear Information System (INIS)

    Soran, P.D.

    1976-05-01

    The cross-section methodology incorporated in the SIMMER code is described. Data base for all cross sections is the ENDF/B system with various progressing computer codes to group collapse and modify the group constants which are used in SIMMER. Either infinitely dilute cross sections or the Bondarenko formalism can be used in SIMMER. Presently only a microscopic treatment is considered, but preliminary macroscopic algorithms have been investigated

  12. High ET jet cross sections at CDF

    International Nuclear Information System (INIS)

    Flaugher, B.

    1996-08-01

    The inclusive jet cross section for p anti p collisions at √s = 1.8 TeV as measured by the CDF collaboration will be presented. Preliminary CDF measurements of the Σ E T cross section at √s = 1.8 TeV and the central inclusive jet cross section at √s = 0.630 TeV will also be shown

  13. Measurements of neutron capture cross sections

    International Nuclear Information System (INIS)

    Nakajima, Yutaka

    1984-01-01

    A review of measurement techniques for the neutron capture cross sections is presented. Sell transmission method, activation method, and prompt gamma-ray detection method are described using examples of capture cross section measurements. The capture cross section of 238 U measured by three different prompt gamma-ray detection methods (large liquid scintillator, Moxon-Rae detector, and pulse height weighting method) are compared and their discrepancies are resolved. A method how to derive the covariance is described. (author)

  14. Total neutron cross section of lead

    International Nuclear Information System (INIS)

    Kanda, K.; Aizawa, O.

    1976-01-01

    The total thermal-neutron cross section of natural lead under various physical conditions was measured by the transmission method. It became clear that the total cross section at room temperature previously reported is lower than the present data. The total cross section at 400, 500, and 600 0 C, above the melting point of lead, 327 0 C, was also measured, and the changes in the cross section as a function of temperature were examined, especially near and below the melting point. The data obtained for the randomly oriented polycrystalline state at room temperature were in reasonable agreement with the theoretical values calculated by the THRUSH and UNCLE-TOM codes

  15. 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)

  16. Neutrino-nucleus cross sections for oscillation experiments

    Science.gov (United States)

    Katori, Teppei; Martini, Marco

    2018-01-01

    Neutrino oscillations physics is entering an era of high precision. In this context, accelerator-based neutrino experiments need a reduction in systematic errors to the level of a few percent. Today, one of the most important sources of systematic errors are neutrino-nucleus cross sections which, in the energy region of hundreds of MeV to a few GeV, are known to a precision not exceeding 20%. In this article we review the present experimental and theoretical knowledge of neutrino-nucleus interaction physics. After introducing neutrino-oscillation physics and accelerator-based neutrino experiments, we give an overview of general aspects of neutrino-nucleus cross sections, from both the theoretical and experimental point of view. Then, we focus on these cross sections in different reaction channels. We start with the quasi-elastic and quasi-elastic-like cross section, placing a special emphasis on the multinucleon emission channel, which has attracted a lot of attention in the last few years. We review the main aspects of the different microscopic models for this channel by discussing analogies and the differences among them. The discussion is always driven by a comparison with the experimental data. We then consider the one-pion production channel where agreement between data and theory remains highly unsatisfactory. We describe how to interpret pion data, and then analyze, in particular, the puzzle related to the difficulty of theoretical models and Monte Carlo to simultaneously describe MiniBooNE and MINERvA experimental results. Inclusive cross sections are also discussed, as well as the comparison between the {ν }μ and {ν }e cross sections, relevant for the charge-conjugation-parity violation experiments. The impact of nuclear effects on the reconstruction of neutrino energy and on the determination of the neutrino-oscillation parameters is also reviewed. Finally, we look to the future by discussing projects and efforts in relation to future detectors, beams

  17. Single Top quark production cross section using the ATLAS detector at the LHC

    CERN Document Server

    Hirschbuehl, D; The ATLAS collaboration

    2014-01-01

    Measurements of single top-quark production cross section in proton proton collisions at 7 and 8 TeV are presented. In the leading order process, a W boson is exchanged in the t-channel. For this process for the first time a fiducial cross section measured within the detector acceptance is presented and the modelling uncertainty when extrapolating to the total inclusive cross section is assessed with a large number of different Monte Carlo generators. The result is in good agreement with the most up-to-date theory predictions. Furthermore, the single top-quark and anti-top total production cross sections, their ratio, as well as a measurement of the inclusive production cross section is presented. Differential cross sections are measured as a function of the transverse momentum and the absolute value of the rapidity of top and anti-top quarks.

  18. Measurement of top-quark pair production cross sections and calibration of the top-quark Monte-Carlo mass using LHC run I proton-proton collision data at √(s) = 7 and 8 TeV with the CMS experiment

    International Nuclear Information System (INIS)

    Kieseler, Jan

    2015-12-01

    In this thesis, measurements of the production cross sections for top-quark pairs and the determination of the top-quark mass are presented. Dileptonic decays of top-quark pairs (t anti t) with two opposite-charged lepton (electron and muon) candidates in the final state are considered. The studied data samples are collected in proton-proton collisions at the CERN Large Hadron Collider with the CMS detector and correspond to integrated luminosities of 5.0 fb -1 and 19.7 fb -1 at center-of-mass energies of √(s) = 7 TeV and √(s) = 8 TeV, respectively. The cross sections, σ t anti t , are measured in the fiducial detector volume (visible phase space), defined by the kinematics of the top-quark decay products, and are extrapolated to the full phase space. The visible cross sections are extracted in a simultaneous binned-likelihood fit to multi-differential distributions of final-state observables, categorized according to the multiplicity of jets associated to b quarks (b jets) and other jets in each event. The fit is performed with emphasis on a consistent treatment of correlations between systematic uncertainties and taking into account features of the t anti t event topology. By comparison with predictions from the Standard Model at next-to-next-to leading order (NNLO) accuracy, the top-quark pole mass, m t pole , is extracted from the measured cross sections for different state-of-the-art PDF sets. Furthermore, the top-quark mass parameter used in Monte-Carlo simulations, m t MC , is determined using the distribution of the invariant mass of a lepton candidate and the leading b jet in the event, m lb . Being defined by the kinematics of the top-quark decay, this observable is unaffected by the description of the top-quark production mechanism. Events are selected from the data collected at √(s) = 8 TeV that contain at least two jets and one b jet in addition to the lepton candidate pair. A novel technique is presented, in which fixed-order calculations in

  19. 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

  20. First measurement of the Rayleigh cross section

    NARCIS (Netherlands)

    Naus, H.; Ubachs, W.

    2000-01-01

    Rayleigh cross section for N2, Ar and SF6 was performed using the technique of cavity ring-down spectroscopy (CRDS). The experiment was based on the assumption that scattering cross section is equal to the extinction in the absence of absorption. The theory explains the molecular origin of

  1. Total cross section of highly excited strings

    International Nuclear Information System (INIS)

    Lizzi, F.; Senda, I.

    1990-01-01

    The unpolarized total cross section for the joining of two highly excited strings is calculated. The calculation is performed by taking the average overall states in the given excitation levels of the initial strings. We find that the total cross section grows with the energy and momentum of the initial states. (author). 8 refs, 1 fig

  2. Compilation of cross-sections. Pt. 1

    International Nuclear Information System (INIS)

    Flaminio, V.; Moorhead, W.G.; Morrison, D.R.O.; Rivoire, N.

    1983-01-01

    A compilation of integral cross-sections for hadronic reactions is presented. This is an updated version of CERN/HERA 79-1, 79-2, 79-3. It contains all data published up to the beginning of 1982, but some more recent data have also been included. Plots of the cross-sections versus incident laboratory momentum are also given. (orig.)

  3. Total cross section results for deuterium electrodisintegration

    International Nuclear Information System (INIS)

    Skopik, D.M.; Murphy, J.J. II; Shin, Y.M.

    1976-01-01

    Theoretical total cross sections for deuterium electrodisintegration are presented as a function of incident electron energy. The cross section has been calculated using virtual photon theory with Partovi's photodisintegration calculation for E/subx/ > 10 MeV and effective range theory for E/subx/ 2 H(e, n) reaction in Tokamak reactors

  4. Model cross section calculations using LAHET

    International Nuclear Information System (INIS)

    Prael, R.E.

    1992-01-01

    The current status of LAHET is discussed. The effect of a multistage preequilibrium exciton model following the INC is examined for neutron emission benchmark calculations, as is the use of a Fermi breakup model for light nuclei rather than an evaporation model. Comparisons are made also for recent fission cross section experiments, and a discussion of helium production cross sections is presented

  5. Vibrational enhancement of total breakup cross sections

    International Nuclear Information System (INIS)

    Haftel, M.I.; Lim, T.K.

    1984-01-01

    This paper considers the role of multi-two-body bound states, namely vibrational excitations, on total three-body breakup cross-sections. Total cross-sections are usually easy to measure, and they play a fundamental role in chemical kinetics. (orig.)

  6. Interference analysis of fission cross section

    International Nuclear Information System (INIS)

    Toshkov, S.A.; Yaneva, N.B.

    1976-01-01

    The formula for the reaction cross-section based on the R-matrix formalism considering the interference between the two neighbouring resonances, referred to the same value of total momentum was used for the analysis of the cross-section of resonance neutron induced fission of 230Pu. The experimental resolution and thermal motion of the target nuclei were accounted for numerical integration

  7. Compilation of cross-sections. Pt. 4

    International Nuclear Information System (INIS)

    Alekhin, S.I.; Ezhela, V.V.; Lugovsky, S.B.; Tolstenkov, A.N.; Yushchenko, O.P.; Baldini, A.; Cobal, M.; Flaminio, V.; Capiluppi, P.; Giacomelli, G.; Mandrioli, G.; Rossi, A.M.; Serra, P.; Moorhead, W.G.; Morrison, D.R.O.; Rivoire, N.

    1987-01-01

    This is the fourth volume in our series of data compilations on integrated cross-sections for weak, electromagnetic, and strong interaction processes. This volume covers data on reactions induced by photons, neutrinos, hyperons, and K L 0 . It contains all data published up to June 1986. Plots of the cross-sections versus incident laboratory momentum are also given. (orig.)

  8. Comparative analysis among several cross section sets

    International Nuclear Information System (INIS)

    Caldeira, A.D.

    1983-01-01

    Critical parameters were calculated using the one dimensional multigroup transport theory for several cross section sets. Calculations have been performed for water mixtures of uranium metal, plutonium metal and uranium-thorium oxide, and for metallics systems, to determine the critical dimensions of geometries (sphere and cylinder). For this aim, the following cross section sets were employed: 1) multigroup cross section sets obtained from the GAMTEC-II code; 2) the HANSEN-ROACH cross section sets; 3) cross section sets from the ENDF/B-IV, processed by the NJOY code. Finally, we have also calculated the corresponding critical radius using the one dimensional multigroup transport DTF-IV code. The numerical results agree within a few percent with the critical values obtained in the literature (where the greatest discrepancy occured in the critical dimensions of water mixtures calculated with the values generated by the NJOY code), a very good results in comparison with similar works. (Author) [pt

  9. Activation cross section data file, (1)

    International Nuclear Information System (INIS)

    Yamamuro, Nobuhiro; Iijima, Shungo.

    1989-09-01

    To evaluate the radioisotope productions due to the neutron irradiation in fission of fusion reactors, the data for the activation cross sections ought to be provided. It is planning to file more than 2000 activation cross sections at final. In the current year, the neutron cross sections for 14 elements from Ni to W have been calculated and evaluated in the energy range 10 -5 to 20 MeV. The calculations with a simplified-input nuclear cross section calculation system SINCROS were described, and another method of evaluation which is consistent with the JENDL-3 were also mentioned. The results of cross section calculation are in good agreement with experimental data and they were stored in the file 8, 9 and 10 of ENDF/B format. (author)

  10. SCALE system cross-section validation for criticality safety analysis

    International Nuclear Information System (INIS)

    Hathout, A.M.; Westfall, R.M.; Dodds, H.L. Jr.

    1980-01-01

    The purpose of this study is to test selected data from three cross-section libraries for use in the criticality safety analysis of UO 2 fuel rod lattices. The libraries, which are distributed with the SCALE system, are used to analyze potential criticality problems which could arise in the industrial fuel cycle for PWR and BWR reactors. Fuel lattice criticality problems could occur in pool storage, dry storage with accidental moderation, shearing and dissolution of irradiated elements, and in fuel transport and storage due to inadequate packing and shipping cask design. The data were tested by using the SCALE system to analyze 25 recently performed critical experiments

  11. Monte Carlo analysis of KRITZ-2 critical benchmarks on the reactivity temperature coefficient using ENDF/B-VII.1 and JENDL-4.0 nuclear data libraries

    International Nuclear Information System (INIS)

    El Ouahdani, S.; Boukhal, H.; Erradi, L.; Chakir, E.; El Bardouni, T.; Hajjaji, O.; Boulaich, Y.; Benaalilou, K.; Kaddour, M.

    2016-01-01

    Highlights: • A set of KRITZ-2 experiments with UO 2 and MOX LWR lattices, at room and elevated temperatures, have been analysed using the MCNP6.1 code with the libraries: JENDL-4 and ENDF/B-VII.1. • The detailed comparisons of the calculations and measurements demonstrate a good agreement between calculations and measurements. • To investigate better the influence of cross sections differences on the reactivity temperature coefficient, we break it down into its components using a pin cell model. - Abstract: A set of KRITZ-2 experiments light water moderated lattices with uranium oxide and mixed-oxide fuel rods, at room and elevated temperatures, performed in the early 1970’s have been assessed. Using the MCNP6.1 code with the most recent cross section libraries: JENDL-4 and ENDF/B-VII.1, the critical experiments KRITZ: 2-1, KRITZ: 2-13, and KRITZ: 2-19 achieved in the Sweden reactor KRITZ were analyzed. We have used the ENDF/B-VII.1 data provided with the MCNP6.1.1 version in ACE format and the Makxsf utility to handle the data in the specific temperatures not available in the MCNP6.1.1 original data. The JENDL-4 evaluations were processed using NJOY99 (update 364) to the temperatures of interest. The detailed comparisons of the calculated and measured (Benchmark, 2005) effective multiplication factors and pin power distributions for UO2 and MOX fuelled cores presented in this work demonstrate a good agreement between calculation and measurements. The maximum deviation of the calculation from the experimental data for k eff , is 0.58% (absolute value) obtained for the KRITZ 2:1 at 248.5 °C using ENDF/B-VII.1 data. To investigate better the influence of cross sections differences on the reactivity and temperature coefficient, we break down the infinite multiplication factor into its components using a pin cell model. Using this simple model we evaluated the temperature effect on the infinite multiplication factor and the effect on its components. We have

  12. Status of software for PGNAA bulk analysis by the Monte Carlo - Library Least-Squares (MCLLS) approach

    International Nuclear Information System (INIS)

    Gardner, R.P.; Zhang, W.; Metwally, W.A.

    2005-01-01

    The Center for Engineering Applications of Radioisotopes (CEAR) has been working for about ten years on the Monte Carlo - Library Least-Squares (MCLLS) approach for treating the nonlinear inverse analysis problem for PGNAA bulk analysis. This approach consists essentially of using Monte Carlo simulation to generate the libraries of all the elements to be analyzed plus any other required libraries. These libraries are then used in the linear Library Least-Squares (LLS) approach with unknown sample spectra to analyze for all elements in the sample. The other libraries include all sources of background which includes: (1) gamma-rays emitted by the neutron source, (2) prompt gamma-rays produced in the analyzer construction materials, (3) natural gamma-rays from K-40 and the uranium and thorium decay chains, and (4) prompt and decay gamma-rays produced in the NaI detector by neutron activation. A number of unforeseen problems have arisen in pursuing this approach including: (1) the neutron activation of the most common detector (NaI) used in bulk analysis PGNAA systems, (2) the nonlinearity of this detector, and (3) difficulties in obtaining detector response functions for this (and other) detectors. These problems have been addressed by CEAR recently and have either been solved or are almost solved at the present time. Development of Monte Carlo simulation for all of the libraries has been finished except the prompt gamma-ray library from the activation of the NaI detector. Treatment for the coincidence schemes for Na and particularly I must be first determined to complete the Monte Carlo simulation of this last library. (author)

  13. Is the quasielastic pion cross section really bigger than the pion-nucleus reaction cross section

    International Nuclear Information System (INIS)

    Silbar, R.R.

    1979-01-01

    It is shown that soft pion charge exchanges may increase the inclusive (π + ,π 0 ') cross section, relative to the total quasielastic (π + ,π + ') cross section, by as much as a factor of two. 4 references

  14. Evaluation of computational models and cross sections used by MCNP6 for simulation of electron backscattering

    Energy Technology Data Exchange (ETDEWEB)

    Poškus, Andrius, E-mail: andrius.poskus@ff.vu.lt

    2016-02-01

    This work evaluates the accuracy of the single-event (SE) and condensed-history (CH) models of electron transport in Monte Carlo simulations of electron backscattering from thick layers of Be, C, Al, Cu, Ag, Au and U at incident electron energies from 200 eV to 15 MeV. The CH method is used in simulations performed with MCNP6.1, and the SE method is used in simulations performed with an open-source single-event code MCNelectron written by the author of this paper. Both MCNP6.1 and MCNelectron use mainly ENDF/B-VI.8 library data, but MCNelectron allows replacing cross sections of certain types of interactions by alternative datasets from other sources. The SE method is evaluated both using only ENDF/B-VI.8 cross sections (the “SE-ENDF/B method”, which is equivalent to using MCNP6.1 in SE mode) and with an alternative set of elastic scattering cross sections obtained from relativistic (Dirac) partial-wave (DPW) calculations (the “SE-DPW method”). It is shown that at energies from 200 eV to 300 keV the estimates of the backscattering coefficients obtained using the SE-DPW method are typically within 10% of the experimental data, which is approximately the same accuracy that is achieved using MCNP6.1 in CH mode. At energies below 1 keV and above 300 keV, the SE-DPW method is much more accurate than the SE-ENDF/B method due to lack of angular distribution data in the ENDF/B library in those energy ranges. At energies from 500 keV to 15 MeV, the CH approximation is roughly twice more accurate than the SE-DPW method, with the average relative errors equal 7% and 14%, respectively. The energy probability density functions (PDFs) of backscattered electrons for Al and Cu, calculated using the SE method with DPW cross sections when energy of incident electrons is 20 keV, have an average absolute error as low as 4% of the average PDF. This error is approximately twice less than the error of the corresponding PDF calculated using the CH approximation. It is concluded

  15. Status of the Monte Carlo library least-squares (MCLLS) approach for non-linear radiation analyzer problems

    Science.gov (United States)

    Gardner, Robin P.; Xu, Libai

    2009-10-01

    The Center for Engineering Applications of Radioisotopes (CEAR) has been working for over a decade on the Monte Carlo library least-squares (MCLLS) approach for treating non-linear radiation analyzer problems including: (1) prompt gamma-ray neutron activation analysis (PGNAA) for bulk analysis, (2) energy-dispersive X-ray fluorescence (EDXRF) analyzers, and (3) carbon/oxygen tool analysis in oil well logging. This approach essentially consists of using Monte Carlo simulation to generate the libraries of all the elements to be analyzed plus any other required background libraries. These libraries are then used in the linear library least-squares (LLS) approach with unknown sample spectra to analyze for all elements in the sample. Iterations of this are used until the LLS values agree with the composition used to generate the libraries. The current status of the methods (and topics) necessary to implement the MCLLS approach is reported. This includes: (1) the Monte Carlo codes such as CEARXRF, CEARCPG, and CEARCO for forward generation of the necessary elemental library spectra for the LLS calculation for X-ray fluorescence, neutron capture prompt gamma-ray analyzers, and carbon/oxygen tools; (2) the correction of spectral pulse pile-up (PPU) distortion by Monte Carlo simulation with the code CEARIPPU; (3) generation of detector response functions (DRF) for detectors with linear and non-linear responses for Monte Carlo simulation of pulse-height spectra; and (4) the use of the differential operator (DO) technique to make the necessary iterations for non-linear responses practical. In addition to commonly analyzed single spectra, coincidence spectra or even two-dimensional (2-D) coincidence spectra can also be used in the MCLLS approach and may provide more accurate results.

  16. Measurement of MA fission cross sections at YAYOI

    Energy Technology Data Exchange (ETDEWEB)

    Ohkawachi, Yasushi; Ohki, Shigeo; Wakabayashi, Toshio [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-03-01

    Fission cross section ratios of minor actinide nuclides (Am-241, Am-243) relative to U-235 in the fast neutron energy region have been measured using a back-to-back (BTB) fission chamber at YAYOI fast neutron source reactor. A small BTB fission chamber was developed to measure the fission cross section ratios in the center of the core at YAYOI reactor. Dependence of the fission cross section ratios on neutron spectra was investigated by changing the position of the detector in the reactor core. The measurement results were compared with the fission cross sections in the JENDL-3.2, ENDF/B-VI and JEF-2.2 libraries. It was found that calculated values of Am-241 using the JENDL-3.2, ENDF/B-VI and JEF-2.2 data are lower by about 15% than the measured value in the center of the core (the neutron average energy is 1.44E+6(eV)). And, good agreement can be seen the measured value and calculated value of Am-243 using the JENDL-3.2 data in the center of the core (the neutron average energy is 1.44E+6)(eV), but calculated values of Am-243 using the ENDF/B-VI and JEF-2.2 data are lower by 11% and 13% than the measured value. (author)

  17. Fully hadronic ttbar cross section measurement with ATLAS detector

    CERN Document Server

    Bertella, C; The ATLAS collaboration

    2011-01-01

    The top quark pair production cross section in the fully hadronic final state is characterized by a six jet topology, two of which could be identified as originating from a b-quark using ATLAS b-tagging algorithms. Compared to other decay channels, this final state presents an advantageous larger branching ratio; on the other hand it suffers from a very large QCD multi-jet background, generally difficult to estimate from Monte Carlo simulation and therefore evaluated using data-driven techniques. It is essential, to efficiently trigger on these fully hadronic ttbar events, to use dedicated triggers. An overview of the analysis for ttbar production cross section measurement in the fully hadronic final state and the state-of-the-art of the b-jet trigger performance estimation are presented in this contribution.

  18. Jet cross sections in polarized photon-hadron collisions

    CERN Document Server

    de Florian, Daniel

    1999-01-01

    We present a computation of one- and two-jet cross sections in polarized photon-hadron collisions, which is accurate to next-to-leading order in QCD. Our results can be used to compute photoproduction cross sections in electron-proton scattering. To this purpose, we investigate the structure of the polarized Weizsaecker-Williams function, where we include a universal, non-logarithmic term, neglected in the literature. We construct a Monte Carlo code, within the framework of the subtraction method, and we use it to study the phenomenology of jet production in the energy range relevant to HERA. In particular, we investigate the perturbative stability of our results, and we discuss the possibility of constraining polarized parton densities of the proton and the photon using jet data.

  19. Inclusive photon cross section at 13 TeV

    CERN Document Server

    Cueto , Ana; The ATLAS collaboration

    2017-01-01

    Measurements of prompt photon production in proton-proton collisions provide a testing ground for perturbative QCD with a hard colourless probe. This kind of measurements were recently used to investigate novel approaches to the description of parton radiation, showcasing the importance of resummation of threshold logarithms in QCD and of electroweak corrections, further they are also sensitive at leading order to the gluon density in the proton. ATLAS has made public very recently the first measurements of inclusive photon production in the new kinematic regime opened by the 13 TeV collisions with an integrated luminosity of 3.2 fb^{-1}. The inclusive photo-production cross section is measured as a function of the photon transverse energy above 125 GeV in different regions of photon pseudorapidity. Next-to-leading-order perturbative QCD and Monte Carlo event-generator predictions are compared to the measured cross section and provide an adequate description of the data.

  20. Electron-impact cross sections of Ne

    International Nuclear Information System (INIS)

    Tsurubuchi, S.; Arakawa, K.; Kinokuni, S.; Motohashi, K.

    2000-01-01

    Electron-impact absolute emission cross sections were measured for the 3p→3s transitions of Ne. Excitation functions of the 3s→2p first resonance lines were measured in the energy range from the threshold to 1000 eV by a polarization-free optical method and relative cross sections were normalized to the absolute values, (41.0±5.4)x10 -19 cm 2 for the 73.6 nm line and (7.1±1.0)x10 -19 cm 2 for the 74.4 nm line, which were determined at 500 eV. The integrated level-excitation cross sections of Suzuki et al for the 1s 2 and 1s 4 levels were combined with the corresponding 3p→3s cascade cross sections obtained in this paper to give absolute emission cross sections for the resonance lines. The level-excitation cross sections of the 1s 2 and 1s 4 states in Paschen notation were determined from the threshold to 1000 eV by subtracting 3p→3s cascade cross sections from the corresponding 3s→2p emission cross sections of the resonance lines. A large cascade contribution is found in the emission cross section of the resonance lines. It is 28.5% for the 73.6 nm line and 49.6% for the 74.4 nm line at 40 eV, and 17.0 and 61.8%, respectively, at 300 eV. (author)

  1. The correction of pebble bed reactor nodal cross sections for the effects of leakage and depletion history

    Science.gov (United States)

    Hudson, Nathanael Harrison

    An accurate and computationally fast method to generate nodal cross sections for the Pebble Bed Reactor (PBR) was presented. In this method, named Spectral History Correction (SHC), a set of fine group microscopic cross section libraries, pre-computed at specified depletion and moderation states, was coupled with the nodal nuclide densities and group bucklings to compute the new fine group spectrum for each node. The relevant fine group cross-section library was then recollapsed to the local broad group cross-section structure with this new fine group spectrum. This library set was tracked in terms of fuel isotopic densities. Fine group modulation factors (to correct the homogeneous flux for heterogeneous effects) and fission spectra were also stored with the cross section library. As the PBR simulation converges to a steady state fuel cycle, the initial nodal cross section library becomes inaccurate due to the burnup of the fuel and the neutron leakage into and out of the node. Because of the recirculation of discharged fuel pebbles with fresh fuel pebbles, a node can consist of a collection of pebbles at various burnup stages. To account for the nodal burnup, the microscopic cross sections were combined with nodal averaged atom densities to approximate the fine group macroscopic cross-sections for that node. These constructed, homogeneous macroscopic cross sections within the node were used to calculate a numerical solution for the fine group spectrum with B1 theory. This new fine spectrum was used to collapse the pre-computed microscopic cross section library to the broad group structure employed by the fuel cycle code. This SHC technique was developed and practically implemented as a subroutine within the PBR fuel cycle code PEBBED. The SHC subroutine was called to recalculate the broad group cross sections during the code convergence. The result was a fast method that compared favorably to the benchmark scheme of cross section calculation with the lattice

  2. Cross-sectional anatomy for computed tomography

    International Nuclear Information System (INIS)

    Farkas, M.L.

    1988-01-01

    This self-study guide recognizes that evaluation and interpretation of CT-images demands a firm understanding of both cross-sectional anatomy and the principles of computed tomography. The objectives of this book are: to discuss the basic principles of CT, to stress the importance of cross-sectional anatomy to CT through study of selected cardinal transverse sections of head, neck, and trunk, to explain orientation and interpretation of CT-images with the aid of corresponding cross-sectional preparations

  3. NNLO jet cross sections by subtraction

    International Nuclear Information System (INIS)

    Somogyi, G.; Bolzoni, P.; Trocsanyi, Z.

    2010-06-01

    We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of an earlier NNLO subtraction scheme over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state. (orig.)

  4. Differential Top Cross-section Measurements

    CERN Document Server

    Fenton, Michael James; The ATLAS collaboration

    2017-01-01

    The top quark is the heaviest known fundamental particle. The measurement of the differential top-quark pair production cross-section provides a stringent test of advanced perturbative QCD calculations. The ATLAS collaboration has performed detailed measurements of those differential cross sections at a centre-of-mass energy of 13 TeV. This talk focuses on differential cross-section measurements in the lepton+jets final state, including using boosted top quarks to probe our understanding of top quark production in the TeV regime.

  5. Hardon cross sections at ultra high energies

    International Nuclear Information System (INIS)

    Yodh, G.B.

    1987-01-01

    A review of results on total hadronic cross sections at ultra high energies obtained from a study of longitudinal development of cosmic ray air showers is given. The experimental observations show that proton-air inelastic cross section increases from 275 mb to over 500 mb as the collision energy in the center of mass increases from 20 GeV to 20 TeV. The proton-air inelastic cross section, obtained from cosmic ray data at √s = 30 TeV, is compared with calculations using various different models for the energy variation of the parameters of the elementary proton-proton interaction. Three conclusions are derived

  6. NNLO jet cross sections by subtraction

    Energy Technology Data Exchange (ETDEWEB)

    Somogyi, G.; Bolzoni, P. [Deutsches Elektronen-Synchrotron (DESY), Zeuthen (Germany); Trocsanyi, Z. [CERN, Geneva (Switzerland)

    2010-06-15

    We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of an earlier NNLO subtraction scheme over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state. (orig.)

  7. NNLO jet cross sections by subtraction

    CERN Document Server

    Somogyi, Gabor; Trocsanyi, Zoltan

    2010-01-01

    We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of the NNLO subtraction scheme of [1-4], over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state.

  8. Calculation of the resonance cross section functions

    International Nuclear Information System (INIS)

    Slipicevic, K.F.

    1967-11-01

    This paper includes the procedure for calculating the Doppler broadened line shape functions ψ and χ which are needed for calculation of resonance cross section functions. The obtained values are given in tables

  9. Fission cross section measurements at intermediate energies

    International Nuclear Information System (INIS)

    Laptev, Alexander

    2005-01-01

    The activity in intermediate energy particle induced fission cross-section measurements of Pu, U isotopes, minor actinides and sub-actinides in PNPI of Russia is reviewed. The neutron-induced fission cross-section measurements are under way in the wide energy range of incident neutrons from 0.5 MeV to 200 MeV at the GNEIS facility. In number of experiments at the GNEIS facility, the neutron-induced fission cross sections were obtained for many nuclei. In another group of experiments the proton-induced fission cross-section have been measured for proton energies ranging from 200 to 1000 MeV at 100 MeV intervals using the proton beam of PNPI synchrocyclotron. (author)

  10. Measurement of multinucleon transfer cross-sections

    Indian Academy of Sciences (India)

    Keywords. Ni(C, ), Fe(C, ), =C, C, B, B, Be, Be, Be, Be, Li, Li; = 60 MeV; measured reaction cross-section; elastic scattering angular distribution; deduced transfer probabilities and enhancement factors.

  11. Capture cross sections on unstable nuclei

    Science.gov (United States)

    Tonchev, A. P.; Escher, J. E.; Scielzo, N.; Bedrossian, P.; Ilieva, R. S.; Humby, P.; Cooper, N.; Goddard, P. M.; Werner, V.; Tornow, W.; Rusev, G.; Kelley, J. H.; Pietralla, N.; Scheck, M.; Savran, D.; Löher, B.; Yates, S. W.; Crider, B. P.; Peters, E. E.; Tsoneva, N.; Goriely, S.

    2017-09-01

    Accurate neutron-capture cross sections on unstable nuclei near the line of beta stability are crucial for understanding the s-process nucleosynthesis. However, neutron-capture cross sections for short-lived radionuclides are difficult to measure due to the fact that the measurements require both highly radioactive samples and intense neutron sources. Essential ingredients for describing the γ decays following neutron capture are the γ-ray strength function and level densities. We will compare different indirect approaches for obtaining the most relevant observables that can constrain Hauser-Feshbach statistical-model calculations of capture cross sections. Specifically, we will consider photon scattering using monoenergetic and 100% linearly polarized photon beams. Challenges that exist on the path to obtaining neutron-capture cross sections for reactions on isotopes near and far from stability will be discussed.

  12. Calculation of the resonance cross section functions

    Energy Technology Data Exchange (ETDEWEB)

    Slipicevic, K F [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1967-11-15

    This paper includes the procedure for calculating the Doppler broadened line shape functions {psi} and {chi} which are needed for calculation of resonance cross section functions. The obtained values are given in tables.

  13. Pion-nucleus cross sections approximation

    International Nuclear Information System (INIS)

    Barashenkov, V.S.; Polanski, A.; Sosnin, A.N.

    1990-01-01

    Analytical approximation of pion-nucleus elastic and inelastic interaction cross-section is suggested, with could be applied in the energy range exceeding several dozens of MeV for nuclei heavier than beryllium. 3 refs.; 4 tabs

  14. Tachyonic ionization cross sections of hydrogenic systems

    Energy Technology Data Exchange (ETDEWEB)

    Tomaschitz, Roman [Department of Physics, Hiroshima University, 1-3-1 Kagami-yama, Higashi-Hiroshima 739-8526 (Japan)

    2005-03-11

    Transition rates for induced and spontaneous tachyon radiation in hydrogenic systems as well as the transversal and longitudinal ionization cross sections are derived. We investigate the interaction of the superluminal radiation field with matter in atomic bound-bound and bound-free transitions. Estimates are given for Ly-{alpha} transitions effected by superluminal quanta in hydrogen-like ions. The tachyonic photoelectric effect is scrutinized, in the Born approximation and at the ionization threshold. The angular maxima occur at different scattering angles in the transversal and longitudinal cross sections, which can be used to sift out longitudinal tachyonic quanta in a photon flux. We calculate the tachyonic ionization and recombination cross sections for Rydberg states and study their asymptotic scaling with respect to the principal quantum number. At the ionization threshold of highly excited states of order n {approx} 10{sup 4}, the longitudinal cross section starts to compete with photoionization, in recombination even at lower levels.

  15. a cross-sectional analytic study 2014

    African Journals Online (AJOL)

    Assessment of HIV/AIDS comprehensive correct knowledge among Sudanese university: a cross-sectional analytic study 2014. ... There are limited studies on this topic in Sudan. In this study we investigated the Comprehensive correct ...

  16. Capture cross sections on unstable nuclei

    Directory of Open Access Journals (Sweden)

    Tonchev A.P.

    2017-01-01

    Full Text Available Accurate neutron-capture cross sections on unstable nuclei near the line of beta stability are crucial for understanding the s-process nucleosynthesis. However, neutron-capture cross sections for short-lived radionuclides are difficult to measure due to the fact that the measurements require both highly radioactive samples and intense neutron sources. Essential ingredients for describing the γ decays following neutron capture are the γ-ray strength function and level densities. We will compare different indirect approaches for obtaining the most relevant observables that can constrain Hauser-Feshbach statistical-model calculations of capture cross sections. Specifically, we will consider photon scattering using monoenergetic and 100% linearly polarized photon beams. Challenges that exist on the path to obtaining neutron-capture cross sections for reactions on isotopes near and far from stability will be discussed.

  17. Cross sections, benchmarks, etc.: What is data testing all about

    International Nuclear Information System (INIS)

    Wagschal, J.; Yeivin, Y.

    1985-01-01

    In order to determine the consistency of two distinct measurements of a physical quantity, the discrepancy d between the two should be compared with its own standard deviation, σ = √(σ/sub 1//sup 2/+σ/sub 2//sup 2/). To properly test a given cross-section library by a set of benchmark (integral) measurements, the quantity corresponding to (d/σ)/sup 2/ is the quadratic d/sup dagger/C/sup -1/d. Here d is the vector of which the components are the discrepancies between the calculated values of the integral parameters and their corresponding measured values, and C is the uncertainty matrix of these discrepancies. This quadratic form is the only true measure of the joint consistency of the library and benchmarks. On the other hand, the very matrix C is essentially all one needs to adjust the library by the benchmarks. Therefore, any argument against adjustment simultaneously disqualifies all serious attempts to test cross-section libraries against integral benchmarks

  18. Methods for calculating anisotropic transfer cross sections

    International Nuclear Information System (INIS)

    Cai, Shaohui; Zhang, Yixin.

    1985-01-01

    The Legendre moments of the group transfer cross section, which are widely used in the numerical solution of the transport calculation can be efficiently and accurately constructed from low-order (K = 1--2) successive partial range moments. This is convenient for the generation of group constants. In addition, a technique to obtain group-angle correlation transfer cross section without Legendre expansion is presented. (author)

  19. Heisenberg rise of total cross sections

    International Nuclear Information System (INIS)

    Ezhela, V.V.; Yushchenko, O.P.

    1988-01-01

    It is shown that on the basis of the original idea of Heisenberg on the quasiclassical picture of extended particle interactions one can construct a satisfactory description of the total cross sections, elastic cross sections, elastic diffractive slopes and mean charged multiplicities in the cm energy range from 5 to 900 GeV, and produce reasonable extrapolations up to several tens of TeV. 14 refs.; 7 figs.; 2 tabs

  20. Evaluation methods for neutron cross section standards

    International Nuclear Information System (INIS)

    Bhat, M.R.

    1980-01-01

    Methods used to evaluate the neutron cross section standards are reviewed and their relative merits, assessed. These include phase-shift analysis, R-matrix fit, and a number of other methods by Poenitz, Bhat, Kon'shin and the Bayesian or generalized least-squares procedures. The problems involved in adopting these methods for future cross section standards evaluations are considered, and the prospects for their use, discussed. 115 references, 5 figures, 3 tables

  1. Experimental validation of lead cross sections for scale and MCNP

    International Nuclear Information System (INIS)

    Henrikson, D.J.

    1995-01-01

    Moving spent nuclear fuel between facilities often requires the use of lead-shielded casks. Criticality safety that is based upon calculations requires experimental validation of the fuel matrix and lead cross section libraries. A series of critical experiments using a high-enriched uranium-aluminum fuel element with a variety of reflectors, including lead, has been identified. Twenty-one configurations were evaluated in this study. The fuel element was modelled for KENO V.a and MCNP 4a using various cross section sets. The experiments addressed in this report can be used to validate lead-reflected calculations. Factors influencing calculated k eff which require further study include diameters of styrofoam inserts and homogenization

  2. On Micro VAX farms and shower libraries: Monte Carlo techniques developed for the D0 detector

    International Nuclear Information System (INIS)

    Raja, R.

    1988-01-01

    In order to predict correctly the effects of cracks and dead material in a nearly hermetic calorimeter, hadronic and electromagnetic showers need to be simulated accurately on a particle by particle basis. Tracking all the particles of all showers in the calorimeter leads to very large CPU times (typically 5 hours on a VAX780) for events at √(s) = 2TeV. Parametrizing the energy deposition of electromagnetic particles in showers with energy below 200 MeV results in event times of the order of 1 hour on a VAX780. This is still unacceptably large. The D0 collaboration then employed a farm of 16 MicroVax II's to get acceptable throughputs. The calorimeter hit patterns of each individual track was output, to be summed up by a later job. These individual hit patterns were entered into a random access shower library file, which was then used for subsequent Monte Carlo simulations. This shower library technique results in further speed-ups of a factor of 60 without degrading the quality of simulation significantly

  3. A Pebble Bed Reactor cross section methodology

    International Nuclear Information System (INIS)

    Hudson, Nathanael H.; Ougouag, Abderrafi M.; Rahnema, Farzad; Gougar, Hans

    2009-01-01

    A method is presented for the evaluation of microscopic cross sections for the Pebble Bed Reactor (PBR) neutron diffusion computational models during convergence to an equilibrium (asymptotic) fuel cycle. This method considers the isotopics within a core spectral zone and the leakages from such a zone as they arise during reactor operation. The randomness of the spatial distribution of fuel grains within the fuel pebbles and that of the fuel and moderator pebbles within the core, the double heterogeneity of the fuel, and the indeterminate burnup of the spectral zones all pose a unique challenge for the computation of the local microscopic cross sections. As prior knowledge of the equilibrium composition and leakage is not available, it is necessary to repeatedly re-compute the group constants with updated zone information. A method is presented to account for local spectral zone composition and leakage effects without resorting to frequent spectrum code calls. Fine group data are pre-computed for a range of isotopic states. Microscopic cross sections and zone nuclide number densities are used to construct fine group macroscopic cross sections, which, together with fission spectra, flux modulation factors, and zone buckling, are used in the solution of the slowing down balance to generate a new or updated spectrum. The microscopic cross-sections are then re-collapsed with the new spectrum for the local spectral zone. This technique is named the Spectral History Correction (SHC) method. It is found that this method accurately recalculates local broad group microscopic cross sections. Significant improvement in the core eigenvalue, flux, and power peaking factor is observed when the local cross sections are corrected for the effects of the spectral zone composition and leakage in two-dimensional PBR test problems.

  4. Cross Sections and Transport Properties of BR- Ions in AR

    Science.gov (United States)

    Jovanovic, Jasmina; Stojanovic, Vladimir; Raspopovic, Zoran; Petrovic, Zoran

    2014-10-01

    We have used a combination of a simple semi-analytic theory - Momentum Transfer Theory (MTT) and exact Monte Carlo (MC) simulations to develop Br- in Ar momentum transfer cross section based on the available data for reduced mobility at the temperature T = 300 K over the range 10 Td higher energies based on behavior of similar ions in similar gases and by the addition of the total detachment cross section that was used from the threshold around 7.7 eV. Relatively complete set was derived which can be used in modeling of plasmas by both hybrid, particle in cell (PIC) and fluid codes. A good agreement between calculated and measured ion mobilities and longitudinal diffusion coefficients is an independent proof of the validity of the cross sections that were derived for the negative ion mobility data. In addition to transport coefficients we have also calculated the net rate coefficients of elastic scattering and detachment. Author acknowledge Ministry of Education, Science and Technology, Proj. Nos. 171037 and 410011.

  5. Fully hadronic ttbar cross section measurement with ATLAS detector

    CERN Document Server

    Bertella, Claudia

    2011-01-01

    The top quark pair production cross section in the fully hadronic final state is characterized by a six jet topology, two of which could be identified as originating from a b-quark using ATLAS b-tagging algorithms. Compared to other decay channels, this final state presents an advantageous larger branching ratio; on the other hand it suffers from a very large QCD multi-jet background, generally difficult to estimate from Monte Carlo simulation and therefore evaluated using data-driven techniques. The analysis is performed using 36pb-1 of pp collisions produced at the LHC with a center-of-mass energy of 7 TeV. The observed upper limit is set at 261 pb at 95% confidence level, where the expected Standard Model cross-section for the ttbar process is 165+11-16 pb. In the future, when the LHC luminosity increases, it is essential, to efficiently trigger on these fully hadronic ttbar events, to use dedicated triggers. An overview of the analysis for ttbar production cross section measurement in the fully hadronic f...

  6. Sensitivity of neutron air transport to nitrogen cross section uncertainties

    International Nuclear Information System (INIS)

    Niiler, A.; Beverly, W.B.; Banks, N.E.

    1975-01-01

    The sensitivity of the transport of 14-MeV neutrons in sea level air to uncertainties in the ENDF/B-III values of the various Nitrogen cross sections has been calculated using the correlated sampling Monte Carlo neutron transport code SAMCEP. The source consisted of a 14.0- to 14.9-MeV band of isotropic neutrons and the fluences (0.5 to 15.0 MeV) were calculated at radii from 50 to 1500 metres. The maximum perturbations, assigned to the ENDF/B-III or base cross section set in the 6.0- to 14.5-MeV energy range were; (1) 2 percent to the total, (2) 10 percent to the total elastic, (3) 40 percent to the inelastic and absorption and (4) 20 percent to the first Legendre coefficient and 10 percent to the second Legendre coefficient of the elastic angular distribtuions. Transport calculations were carried out using various physically realistic sets of perturbed cross sections, bounded by evaluator-assigned uncertainties, as well as the base set. Results show that in some energy intervals at 1500 metres, the differential fluence level with a perturbed set differed by almost a factor of two from the differential fluence level with the base set. 5 figures

  7. 7Li neutron-induced elastic scattering cross section measurement using a slowing-down spectrometer

    Directory of Open Access Journals (Sweden)

    Heusch M.

    2010-10-01

    Full Text Available A new integral measurement of the 7Li neutron induced elastic scattering cross section was determined in a wide neutron energy range. The measurement was performed on the LPSC-PEREN experimental facility using a heterogeneous graphite-LiF slowing-down time spectrometer coupled with an intense pulsed neutron generator (GENEPI-2. This method allows the measurement of the integral elastic scattering cross section in a slowing-down neutron spectrum. A Bayesian approach coupled to Monte Carlo calculations was applied to extract naturalC, 19F and 7Li elastic scattering cross sections.

  8. From cutting-edge pointwise cross-section to groupwise reaction rate: A primer

    Directory of Open Access Journals (Sweden)

    Sublet Jean-Christophe

    2017-01-01

    Full Text Available The nuclear research and development community has a history of using both integral and differential experiments to support accurate lattice-reactor, nuclear reactor criticality and shielding simulations, as well as verification and validation efforts of cross sections and emitted particle spectra. An important aspect to this type of analysis is the proper consideration of the contribution of the neutron spectrum in its entirety, with correct propagation of uncertainties and standard deviations derived from Monte Carlo simulations, to the local and total uncertainty in the simulated reactions rates (RRs, which usually only apply to one application at a time. This paper identifies deficiencies in the traditional treatment, and discusses correct handling of the RR uncertainty quantification and propagation, including details of the cross section components in the RR uncertainty estimates, which are verified for relevant applications. The methodology that rigorously captures the spectral shift and cross section contributions to the uncertainty in the RR are discussed with quantified examples that demonstrate the importance of the proper treatment of the spectrum profile and cross section contributions to the uncertainty in the RR and subsequent response functions. The recently developed inventory code FISPACT-II, when connected to the processed nuclear data libraries TENDL-2015, ENDF/B-VII.1, JENDL-4.0u or JEFF-3.2, forms an enhanced multi-physics platform providing a wide variety of advanced simulation methods for modelling activation, transmutation, burnup protocols and simulating radiation damage sources terms. The system has extended cutting-edge nuclear data forms, uncertainty quantification and propagation methods, which have been the subject of recent integral and differential, fission, fusion and accelerators validation efforts. The simulation system is used to accurately and predictively probe, understand and underpin a modern and

  9. Improved calculation of displacements per atom cross section in solids by gamma and electron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Piñera, Ibrahin, E-mail: ipinera@ceaden.edu.cu [Centro de Aplicaciones Tecnológicas y Desarrollo Nuclear, CEADEN, 30 St. 502, Playa 11300, Havana (Cuba); Cruz, Carlos M.; Leyva, Antonio; Abreu, Yamiel; Cabal, Ana E. [Centro de Aplicaciones Tecnológicas y Desarrollo Nuclear, CEADEN, 30 St. 502, Playa 11300, Havana (Cuba); Espen, Piet Van; Remortel, Nick Van [University of Antwerp, CGB, Groenenborgerlaan 171, 2020 Antwerpen (Belgium)

    2014-11-15

    Highlights: • We present a calculation procedure for dpa cross section in solids under irradiation. • Improvement about 10–90% for the gamma irradiation induced dpa cross section. • Improvement about 5–50% for the electron irradiation induced dpa cross section. • More precise results (20–70%) for thin samples irradiated with electrons. - Abstract: Several authors had estimated the displacements per atom cross sections under different approximations and models, including most of the main gamma- and electron-material interaction processes. These previous works used numerical approximation formulas which are applicable for limited energy ranges. We proposed the Monte Carlo assisted Classical Method (MCCM), which relates the established theories about atom displacements to the electron and positron secondary fluence distributions calculated from the Monte Carlo simulation. In this study the MCCM procedure is adapted in order to estimate the displacements per atom cross sections for gamma and electron irradiation. The results obtained through this procedure are compared with previous theoretical calculations. An improvement in about 10–90% for the gamma irradiation induced dpa cross section is observed in our results on regard to the previous evaluations for the studied incident energies. On the other hand, the dpa cross section values produced by irradiation with electrons are improved by our calculations in about 5–50% when compared with the theoretical approximations. When thin samples are irradiated with electrons, more precise results are obtained through the MCCM (in about 20–70%) with respect to the previous studies.

  10. Partial cross sections near the higher resonances

    International Nuclear Information System (INIS)

    Falk-Vairant, P.; Valladas, G.

    1961-07-01

    As a continuation of the report given at the 10. Rochester Conference, recent measurements of charge-exchange cross section and π 0 production in π - -p interactions are presented here. Section 1 gives a summary of the known results for the elastic, inelastic, and charge-exchange cross sections. Section 2 presents the behavior of the cross sections in the T=1/2 state, in order to discuss the resonances at 600 and 890 MeV. Section 3 discusses the charge-exchange scattering and the interference term between the T=1/2 and T=3/2 states. Section 4 presents some comments on inelastic processes. This report is reprinted from 'Reviews of Modern Physics', Vol. 33, No. 3, 362-367, July, 1961

  11. Photoproton cross section for /sup 19/F

    Energy Technology Data Exchange (ETDEWEB)

    Tsubota, H [Tohoku Univ., Sendai (Japan). Coll. of General Education; Kawamura, N; Oikawa, S; Uegaki, J I

    1975-02-01

    Proton energy spectra have been measured at 90/sup 0/ for the /sup 19/F(e,e'p)/sup 18/O reaction in the giant resonance region. The (..gamma..,p/sub 0/) and (..gamma..,p/sub 1/) differential cross sections are extracted from the proton energy spectra by using virtual-photon spectra. The integrated differential cross section of the (..gamma..,p/sub 0/) and (..gamma..,p/sub 1/) reactions are 1.80+-0.27 and 0.50+-0.45 MeV-mb/sr, respectively. The results are discussed with the shell model theory by comparing with the (..gamma..,p/sub 0/) cross section of the neighboring 4n-nucleus /sup 20/Ne. A significant increase of the proton yield leaving the non-ground states is found at 25 MeV of the incident electron energy. This is discussed in terms of the core excitation effect.

  12. Prospects for Precision Neutrino Cross Section Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Harris, Deborah A. [Fermilab

    2016-01-28

    The need for precision cross section measurements is more urgent now than ever before, given the central role neutrino oscillation measurements play in the field of particle physics. The definition of precision is something worth considering, however. In order to build the best model for an oscillation experiment, cross section measurements should span a broad range of energies, neutrino interaction channels, and target nuclei. Precision might better be defined not in the final uncertainty associated with any one measurement but rather with the breadth of measurements that are available to constrain models. Current experience shows that models are better constrained by 10 measurements across different processes and energies with 10% uncertainties than by one measurement of one process on one nucleus with a 1% uncertainty. This article describes the current status of and future prospects for the field of precision cross section measurements considering the metric of how many processes, energies, and nuclei have been studied.

  13. NNLO jet cross sections by subtraction

    Energy Technology Data Exchange (ETDEWEB)

    Somogyi, G.; Bolzoni, P. [DESY, Platanenallee 6, D-15738 Zeuthen (Germany); Trocsanyi, Z. [CERN PH-TH, on leave from University of Debrecen and Institute of Nuclear Research of HAS, H-4001 P.O.Box 51 (Hungary)

    2010-08-15

    We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of the NNLO subtraction scheme of Refs. [G. Somogyi, Z. Trocsanyi, and V. Del Duca, JHEP 06, 024 (2005), (arXiv:hep-ph/0502226); G. Somogyi and Z. Trocsanyi, (2006), (arXiv:hep-ph/0609041); G. Somogyi, Z. Trocsanyi, and V. Del Duca, JHEP 01, 070 (2007), (arXiv:hep-ph/0609042); G. Somogyi and Z. Trocsanyi, JHEP 01, 052 (2007), (arXiv:hep-ph/0609043)] over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state.

  14. NNLO jet cross sections by subtraction

    International Nuclear Information System (INIS)

    Somogyi, G.; Bolzoni, P.; Trocsanyi, Z.

    2010-01-01

    We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of the NNLO subtraction scheme of Refs. [G. Somogyi, Z. Trocsanyi, and V. Del Duca, JHEP 06, 024 (2005), (arXiv:hep-ph/0502226); G. Somogyi and Z. Trocsanyi, (2006), (arXiv:hep-ph/0609041); G. Somogyi, Z. Trocsanyi, and V. Del Duca, JHEP 01, 070 (2007), (arXiv:hep-ph/0609042); G. Somogyi and Z. Trocsanyi, JHEP 01, 052 (2007), (arXiv:hep-ph/0609043)] over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state.

  15. Average cross sections calculated in various neutron fields

    International Nuclear Information System (INIS)

    Shibata, Keiichi

    2002-01-01

    Average cross sections have been calculated for the reactions contained in the dosimetry files, JENDL/D-99, IRDF-90V2, and RRDF-98 in order to select the best data for the new library IRDF-2002. The neutron spectra used in the calculations are as follows: 1) 252 Cf spontaneous fission spectrum (NBS evaluation), 2) 235 U thermal fission spectrum (NBS evaluation), 3) Intermediate-energy Standard Neutron Field (ISNF), 4) Coupled Fast Reactivity Measurement Facility (CFRMF), 5) Coupled thermal/fast uranium and boron carbide spherical assembly (ΣΣ), 6) Fast neutron source reactor (YAYOI), 7) Experimental fast reactor (JOYO), 8) Japan Material Testing Reactor (JMTR), 9) d-Li neutron spectrum with a 2-MeV deuteron beam. The items 3)-7) represent fast neutron spectra, while JMTR is a light water reactor. The Q-value for the d-Li reaction mentioned above is 15.02 MeV. Therefore, neutrons with energies up to 17 MeV can be produced in the d-Li reaction. The calculated average cross sections were compared with the measurements. Figures 1-9 show the ratios of the calculations to the experimental data which are given. It is found from these figures that the 58 Fe(n, γ) cross section in JENDL/D-99 reproduces the measurements in the thermal and fast reactor spectra better than that in IRDF-90V2. (author)

  16. MEKS: A program for computation of inclusive jet cross sections at hadron colliders

    Science.gov (United States)

    Gao, Jun; Liang, Zhihua; Soper, Davison E.; Lai, Hung-Liang; Nadolsky, Pavel M.; Yuan, C.-P.

    2013-06-01

    EKS is a numerical program that predicts differential cross sections for production of single-inclusive hadronic jets and jet pairs at next-to-leading order (NLO) accuracy in a perturbative QCD calculation. We describe MEKS 1.0, an upgraded EKS program with increased numerical precision, suitable for comparisons to the latest experimental data from the Large Hadron Collider and Tevatron. The program integrates the regularized patron-level matrix elements over the kinematical phase space for production of two and three partons using the VEGAS algorithm. It stores the generated weighted events in finely binned two-dimensional histograms for fast offline analysis. A user interface allows one to customize computation of inclusive jet observables. Results of a benchmark comparison of the MEKS program and the commonly used FastNLO program are also documented. Program SummaryProgram title: MEKS 1.0 Catalogue identifier: AEOX_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEOX_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland. Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 9234 No. of bytes in distributed program, including test data, etc.: 51997 Distribution format: tar.gz Programming language: Fortran (main program), C (CUBA library and analysis program). Computer: All. Operating system: Any UNIX-like system. RAM: ˜300 MB Classification: 11.1. External routines: LHAPDF (https://lhapdf.hepforge.org/) Nature of problem: Computation of differential cross sections for inclusive production of single hadronic jets and jet pairs at next-to-leading order accuracy in perturbative quantum chromodynamics. Solution method: Upon subtraction of infrared singularities, the hard-scattering matrix elements are integrated over available phase space using an optimized VEGAS algorithm. Weighted events are generated and filled

  17. Optical Model and Cross Section Uncertainties

    Energy Technology Data Exchange (ETDEWEB)

    Herman,M.W.; Pigni, M.T.; Dietrich, F.S.; Oblozinsky, P.

    2009-10-05

    Distinct minima and maxima in the neutron total cross section uncertainties were observed in model calculations using spherical optical potential. We found this oscillating structure to be a general feature of quantum mechanical wave scattering. Specifically, we analyzed neutron interaction with 56Fe from 1 keV up to 65 MeV, and investigated physical origin of the minima.We discuss their potential importance for practical applications as well as the implications for the uncertainties in total and absorption cross sections.

  18. Cross sections for charm production by neutrinos

    Energy Technology Data Exchange (ETDEWEB)

    Ushida, N [Aichi Univ. of Education, Kariya (Japan). Dept. of Physics; Kondo, T [Fermi National Accelerator Lab., Batavia, IL (USA); Fujioka, G; Fukushima, J; Takahashi, Y; Tatsumi, S; Yokoyama, C [Kobe Univ. (Japan). Dept. of Physics; Homma, Y; Tsuzuki, Y [Kobe Univ. (Japan). Coll. of Liberal Arts; Bahk, S

    1983-02-03

    The production of charmed particles has been measured using a hybrid emulsion spectrometer in the Fermilab wide-band neutrino beam. The relative cross section for charged current charmed particle production is sigma(v -> ..mu../sup -/c)/sigma(v -> ..mu../sup -/) = 6.5 +- 1.9/1.8%, and the energy dependence of the cross section is presented. One event with charm pair production was observed. A limit of sigma(v -> ..mu..canti c)/sigma(v -> ..mu..c) < 6% (90% CL) is found for the ratio of charged current pair and single charm production.

  19. Covariance Evaluation Methodology for Neutron Cross Sections

    Energy Technology Data Exchange (ETDEWEB)

    Herman,M.; Arcilla, R.; Mattoon, C.M.; Mughabghab, S.F.; Oblozinsky, P.; Pigni, M.; Pritychenko, b.; Songzoni, A.A.

    2008-09-01

    We present the NNDC-BNL methodology for estimating neutron cross section covariances in thermal, resolved resonance, unresolved resonance and fast neutron regions. The three key elements of the methodology are Atlas of Neutron Resonances, nuclear reaction code EMPIRE, and the Bayesian code implementing Kalman filter concept. The covariance data processing, visualization and distribution capabilities are integral components of the NNDC methodology. We illustrate its application on examples including relatively detailed evaluation of covariances for two individual nuclei and massive production of simple covariance estimates for 307 materials. Certain peculiarities regarding evaluation of covariances for resolved resonances and the consistency between resonance parameter uncertainties and thermal cross section uncertainties are also discussed.

  20. New remarks on KERMA factors and DPA cross section data in ACE files

    International Nuclear Information System (INIS)

    Konno, Chikara; Sato, Satoshi; Ohta, Masayuki; Kwon, Saerom; Ochiai, Kentaro

    2016-01-01

    KERMA factors and DPA cross section data are essential for nuclear heating and material damage estimation in fusion reactor designs. Recently we compared KERMA factors and DPA cross section data in the latest official ACE files of JENDL-4.0, ENDF/B-VII.1, JEFF-3.2 and FENDL-3.0 and it was found out that the KERMA factors and DPA cross section data of a lot of nuclei did not always agree among the nuclear data libraries. We investigated the nuclear data libraries and the nuclear data processing code NJOY and specified new reasons for the discrepancies; (1) incorrect nuclear data and NJOY bugs, (2) huge helium production cross section data, (3) gamma production data format in the nuclear data, (4) no detailed secondary particle data (energy–angular distribution data). These problems should be resolved based on this study.

  1. New remarks on KERMA factors and DPA cross section data in ACE files

    Energy Technology Data Exchange (ETDEWEB)

    Konno, Chikara, E-mail: konno.chikara@jaea.go.jp; Sato, Satoshi; Ohta, Masayuki; Kwon, Saerom; Ochiai, Kentaro

    2016-11-01

    KERMA factors and DPA cross section data are essential for nuclear heating and material damage estimation in fusion reactor designs. Recently we compared KERMA factors and DPA cross section data in the latest official ACE files of JENDL-4.0, ENDF/B-VII.1, JEFF-3.2 and FENDL-3.0 and it was found out that the KERMA factors and DPA cross section data of a lot of nuclei did not always agree among the nuclear data libraries. We investigated the nuclear data libraries and the nuclear data processing code NJOY and specified new reasons for the discrepancies; (1) incorrect nuclear data and NJOY bugs, (2) huge helium production cross section data, (3) gamma production data format in the nuclear data, (4) no detailed secondary particle data (energy–angular distribution data). These problems should be resolved based on this study.

  2. Cross section sensitivity study for fusion blankets incorporating lead neutron multiplier

    International Nuclear Information System (INIS)

    Pelloni, S.; Cheng, E.T.

    1983-01-01

    In the recent European INTOR design, lead has been considered for incorporation in the blanket as either an explicit or implicit neutron multiplier. The blanket employs either Li 2 SiO 3 or Li 17 Pb 83 as tritium breeding material. Nucleonic analysis was performed for this blanket using the DLC37 and DLC41 cross section libraries. The reaction rates were estimated using the reaction cross sections provided with both libraries. In addition to that, they were estimated using the MACKLIB-IV response library. The calculated tritium breeding ratio was found to be 5% less and 15% more in the calculations with DLC41 and DLC41 plus MACKLIB-IV libraries, respectively, than in the calculation with the DLC37 library. The Fe, Pb, and Li cross sections given by the ENDF/B-IV and V were reviewed. A sensitivity study of these cross section uncertainties shows that the tritium breeding ratio is relatively insensitive to the above mentioned partial cross sections. The calculated tritium breeding ratio can be known within +-2%. (Auth.)

  3. Approximation of the cross-sections for charged-particle emission reactions near the threshold

    International Nuclear Information System (INIS)

    Badikov, S.A.; Pashchenko, A.B.

    1990-01-01

    We perform an analytical approximation of the energy dependence of the cross-sections for the reactions (n,p) and (n,γ) from the BOSPOR library, correct them for the latest differential and integral experimental data using the common features, characteristic of the energy dependence of the threshold reaction cross-section and making some physical assumptions. 19 refs, 1 fig., 1 tab

  4. The fortran programme for the calculation of the absorption and double scattering corrections in cross-section measurements with fast neutrons using the monte Carlo method (1963); Programme fortran pour le calcul des corrections d'absorption et de double diffusion dans les mesures de sections efficaces pour les neutrons rapides par la methode de monte-carlo (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, B [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    A calculation for double scattering and absorption corrections in fast neutron scattering experiments using Monte-Carlo method is given. Application to cylindrical target is presented in FORTRAN symbolic language. (author) [French] Un calcul des corrections de double diffusion et d'absorption dans les experiences de diffusion de neutrons rapides par la methode de Monte-Carlo est presente. L'application au cas d'une cible cylindrique est traitee en langage symbolique FORTRAN. (auteur)

  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. 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)

  7. Graphs of neutron cross sections in JSD1000 for radiation shielding safety analysis

    International Nuclear Information System (INIS)

    Yamano, Naoki

    1984-03-01

    Graphs of neutron cross sections and self-shielding factors in the JSD1000 library are presented for radiation shielding safety analysis. The compilation contains various reaction cross sections for 42 nuclides from 1 H to 241 Am in the energy range from 3.51 x 10 -4 eV to 16.5 MeV. The Bondarenko-type self-shielding factors of each reaction are given by the background cross sections from σ 0 = 0 to σ 0 = 10000. (author)

  8. Ecological Panel Inference from Repeated Cross Sections

    NARCIS (Netherlands)

    Pelzer, Ben; Eisinga, Rob; Franses, Philip Hans

    2004-01-01

    This chapter presents a Markov chain model for the estimation of individual-level binary transitions from a time series of independent repeated cross-sectional (RCS) samples. Although RCS samples lack direct information on individual turnover, it is demonstrated here that it is possible with these

  9. Stability of tokamaks with elongated cross section

    International Nuclear Information System (INIS)

    An, C.H.; Bateman, G.

    1978-08-01

    Fixed boundary n = 1 MHD instabilities are studied computationally as a function of diamagnetism (β/sub pol/) and current profile in elongated toroidal equilibria (1 2) or a diamagnetic plasma (β/sub pol/ > 1) with only a mildly elongated cross section

  10. Modelisation of the fission cross section

    International Nuclear Information System (INIS)

    Morariu, Claudia

    2013-03-01

    The neutron cross sections of four nuclear systems (n+ 235 U, n+ 233 U, n+ 241 Am and n+ 237 Np) are studied in the present document. The target nuclei of the first case, like 235 U and 239 Pu, have a large fission cross section after the absorption of thermal neutrons. These nuclei are called 'fissile' nuclei. The other type of nuclei, like 237 Np and 241 Am, fission mostly with fast neutrons, which exceed the fission threshold energy. These types of nuclei are called 'fertile'. The compound nuclei of the fertile nuclei have a binding energy higher than the fission barrier, while for the fissile nuclei the binding energy is lower than the fission barrier. In this work, the neutron induced cross sections for both types of nuclei are evaluated in the fast energy range. The total, reaction and shape-elastic cross sections are calculated by the coupled channel method of the optical model code ECIS, while the compound nucleus mechanism are treated by the statistical models implemented in the codes STATIS, GNASH and TALYS. The STATIS code includes a refined model of the fission process. Results from the theoretical calculations are compared with data retrieved from the experimental data base EXFOR. (author) [fr

  11. Photoelectric absorption cross sections with variable abundances

    Science.gov (United States)

    Balucinska-Church, Monika; Mccammon, Dan

    1992-01-01

    Polynomial fit coefficients have been obtained for the energy dependences of the photoelectric absorption cross sections of 17 astrophysically important elements. These results allow the calculation of X-ray absorption in the energy range 0.03-10 keV in material with noncosmic abundances.

  12. (, 3) Differential cross section of He

    Indian Academy of Sciences (India)

    The angular distribution of the five-fold differential cross section for the electron impact double ionization of He (21 ) and He (23 ) has been studied. The kinematical conditions for maxima/minima in the angular distribution for the two cases have been compared. The two-step process for the double ionization is found to ...

  13. Precise relative cross sections for np scattering

    International Nuclear Information System (INIS)

    Goetz, J.; Brogli-Gysin, C.; Hammans, M.; Haffter, P.; Henneck, R.; Jourdan, J.; Masson, G.; Qin, L.M.; Robinson, S.; Sick, I.; Tuccillo, M.

    1994-01-01

    We present data on the differential cross section for neutron-proton scattering for an incident neutron energy of 67 MeV. These data allow a precise determination of the 1 P 1 phase which, in phase-shift analyses, is strongly correlated with the S-D amplitude which we are measuring via different observables. ((orig.))

  14. Symmetric charge transfer cross section of uranium

    International Nuclear Information System (INIS)

    Shibata, Takemasa; Ogura, Koichi

    1995-03-01

    Symmetric charge transfer cross section of uranium was calculated under consideration of reaction paths. In the charge transfer reaction a d 3/2 electron in the U atom transfers into the d-electron site of U + ( 4 I 9/2 ) ion. The J value of the U atom produced after the reaction is 6, 5, 4 or 3, at impact energy below several tens eV, only resonant charge transfer in which the product atom is ground state (J=6) takes place. Therefore, the cross section is very small (4-5 x 10 -15 cm 2 ) compared with that considered so far. In the energy range of 100-1000eV the cross section increases with the impact energy because near resonant charge transfer in which an s-electron in the U atom transfers into the d-electron site of U + ion. Charge transfer cross section between U + in the first excited state (289 cm -1 ) and U in the ground state was also obtained. (author)

  15. LAMBDA p total cross-section measurement

    CERN Multimedia

    CERN PhotoLab

    1970-01-01

    A view of the apparatus used for the LAMBDA p total cross-section measurement at the time of its installation. The hyperons decaying into a proton and a pion in the conical tank in front were detected in the magnet spectrometer in the upper half of the picture. A novel detection technique using exclusively multiwire proportional chambers was employed.

  16. 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.

  17. Comparative studies of ENDF/B-6.8, JEF-2.2 and JENDL-3.2 data libraries by monte carlo modeling of high temperature reactors on plutonium based fuel cycles

    International Nuclear Information System (INIS)

    Talamo, Alberto; Gudowski, Waclaw; Cetnar, Jerzy

    2004-01-01

    We performed a numerical comparative analysis of the burnup capability of the Gas Turbine-Modular Helium Reactor (GT-MHR) by the Monte Carlo Continuous Energy Burnup Code (MCB). The MCB code is an extension of MCNP that includes the burnup implementation; it adopts continuous energy cross sections and it evaluates the transmutation trajectories for over 2,400 decaying nuclides. We equipped the MCB code with three different nuclear data libraries: JENDL-3.2, JEF-2.2 and ENDF/B-6.8 processed for temperatures from 300 to 1,800K. The GT-MHR model studied in this paper is fueled by actinides coming from the Light Water Reactors waste, converted into two different types of fuel: Driver Fuel and Transmutation Fuel. The Driver Fuel supplies the fissile nuclides needed to maintain the criticality of the reactor, whereas the Transmutation Fuel depletes non-fissile isotopes and controls reactivity excess. We set the refueling and shuffling period to one year and the in-core fuel residency time to three years. The comparative analysis of the MCB code consists of accuracy and precision studies. In the accuracy studies, we performed the burnup calculation with different nuclear data libraries during the year at which the refueling and shuffling schedule set the equilibrium of the fuel composition. In the precision studies, we repeated the same simulations 20 times with a different pseudorandom number stride and the same nuclear data library. (author)

  18. Measurement cross sections for radioisotopes production

    International Nuclear Information System (INIS)

    Garrido, E.

    2011-01-01

    New radioactive isotopes for nuclear medicine can be produced using particle accelerators. This is one goal of Arronax, a high energy - 70 MeV - high intensity - 2*350 μA - cyclotron set up in Nantes. A priority list was established containing β - - 47 Sc, 67 Cu - β + - 44 Sc, 64 Cu, 82 Sr/ 82 Rb, 68 Ge/ 68 Ga - and α emitters - 211 At. Among these radioisotopes, the Scandium 47 and the Copper 67 have a strong interest in targeted therapy. The optimization of their productions required a good knowledge of their cross-sections but also of all the contaminants created during irradiation. We launched on Arronax a program to measure these production cross-sections using the Stacked-Foils' technique. It consists in irradiating several groups of foils - target, monitor and degrader foils - and in measuring the produced isotopes by γ-spectrometry. The monitor - nat Cu or nat Ni - is used to correct beam loss whereas degrader foils are used to lower beam energy. We chose to study the nat Ti(p,X) 47 Sc and 68 Zn(p,2p) 67 Cu reactions. Targets are respectively natural Titanium foil - bought from Goodfellow - and enriched Zinc 68 deposited on Silver. In the latter case, Zn targets were prepared in-house - electroplating of 68 Zn - and a chemical separation between Copper and Gallium isotopes has to be made before γ counting. Cross-section values for more than 40 different reactions cross-sections have been obtained from 18 MeV to 68 MeV. A comparison with the Talys code is systematically done. Several parameters of theoretical models have been studied and we found that is not possible to reproduce faithfully all the cross-sections with a given set of parameters. (author)

  19. MCBooster: a library for fast Monte Carlo generation of phase-space decays on massively parallel platforms.

    Science.gov (United States)

    Alves Júnior, A. A.; Sokoloff, M. D.

    2017-10-01

    MCBooster is a header-only, C++11-compliant library that provides routines to generate and perform calculations on large samples of phase space Monte Carlo events. To achieve superior performance, MCBooster is capable to perform most of its calculations in parallel using CUDA- and OpenMP-enabled devices. MCBooster is built on top of the Thrust library and runs on Linux systems. This contribution summarizes the main features of MCBooster. A basic description of the user interface and some examples of applications are provided, along with measurements of performance in a variety of environments

  20. 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)

  1. Monte Carlo simulation on nuclear energy study. Annual report of Nuclear Code Evaluation Committee

    International Nuclear Information System (INIS)

    Sakurai, Kiyoshi; Yamamoto, Toshihiro

    1999-03-01

    In this report, research results discussed in 1998 fiscal year at Nuclear Code Evaluation Special Committee of Nuclear Code Committee were summarised. Present status of Monte Carlo calculation in high energy region investigated / discussed at Monte Carlo simulation working-group and automatic compilation system for MCNP cross sections developed at MCNP high temperature library compilation working-group were described. The 6 papers are indexed individually. (J.P.N.)

  2. Production of application libraries for WIMS

    International Nuclear Information System (INIS)

    Dean, C.J.

    1996-01-01

    The WIMS codes have been developed over a period of 30 years to include sophisticated modelling of many thermal reactor lattices. WIMS6 is currently being developed to include the flexibility of WIMSE together with the case of use seen in LWRWIMS. A new cross-section library is to be released with WIMS6. For the first time since the original WIMSD was written, the library is being generated from completely new differential data based mainly on JEF2.2. Quality Assurance into the XMAS library is achieved in three levels; through general procedures during assembly, by validation studies and by verification. The main general procedure involves forming overall physics quantities (Thermal average cross section, resonance integrals and fission spectrum average values) from the point evaluations and comparing them with similar quantities from the XMAS library. Examples of validation are studies optimizing the background cross-sections at which resonance integrals are tabulated and comparing reaction rates from homogeneous calculations with those formed in Monte Carlo and deterministic codes which represent energy variation of cross-section on a hyperfine energy mesh. Verification involves performing benchmark calculations and comparing results with experiment and calculations with the Apollo code. The current XMAS library contains 140 nuclides based on JEF2.2 evaluations. (author)

  3. An analysis of MCNP cross-sections and tally methods for low-energy photon emitters.

    Science.gov (United States)

    Demarco, John J; Wallace, Robert E; Boedeker, Kirsten

    2002-04-21

    Monte Carlo calculations are frequently used to analyse a variety of radiological science applications using low-energy (10-1000 keV) photon sources. This study seeks to create a low-energy benchmark for the MCNP Monte Carlo code by simulating the absolute dose rate in water and the air-kerma rate for monoenergetic point sources with energies between 10 keV and 1 MeV. The analysis compares four cross-section datasets as well as the tally method for collision kerma versus absorbed dose. The total photon attenuation coefficient cross-section for low atomic number elements has changed significantly as cross-section data have changed between 1967 and 1989. Differences of up to 10% are observed in the photoelectric cross-section for water at 30 keV between the standard MCNP cross-section dataset (DLC-200) and the most recent XCOM/NIST tabulation. At 30 keV, the absolute dose rate in water at 1.0 cm from the source increases by 7.8% after replacing the DLC-200 photoelectric cross-sections for water with those from the XCOM/NIST tabulation. The differences in the absolute dose rate are analysed when calculated with either the MCNP absorbed dose tally or the collision kerma tally. Significant differences between the collision kerma tally and the absorbed dose tally can occur when using the DLC-200 attenuation coefficients in conjunction with a modern tabulation of mass energy-absorption coefficients.

  4. POLIDENT: A Module for Generating Continuous-Energy Cross Sections from ENDF Resonance Data

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, M.E.; Greene, N.M.

    2000-12-01

    POLIDENT (Point Libraries of Data from ENDF/B Tapes) is an AMPX module that accesses the resonance parameters from File 2 of an ENDF/B library and constructs the continuous-energy cross sections in the resonance energy region. The cross sections in the resonance range are subsequently combined with the File 3 background data to construct the cross-section representation over the complete energy range. POLIDENT has the capability to process all resonance reactions that are identified in File 2 of the ENDF/B library. In addition, the code has the capability to process the single- and multi-level Breit-Wigner, Reich-Moore and Adler-Adler resonance formalisms that are identified in File 2. POLIDENT uses a robust energy-mesh-generation scheme that determines the minimum, maximum and points of inflection in the cross-section function in the resolved-resonance region. Furthermore, POLIDENT processes all continuous-energy cross-section reactions that are identified in File 3 of the ENDF/B library and outputs all reactions in an ENDF/B TAB1 format that can be accessed by other AMPX modules.

  5. (n,{alpha}) cross section measurement of gaseous sample using gridded ionization chamber. Cross section determination

    Energy Technology Data Exchange (ETDEWEB)

    Sanami, Toshiya; Baba, Mamoru; Saito, Keiichiro; Ibara, Yasutaka; Hirakawa, Naohiro [Tohoku Univ., Sendai (Japan). Faculty of Engineering

    1997-03-01

    We are developing a method of (n,{alpha}) cross section measurement using gaseous samples in a gridded ionization chamber (GIC). This method enables cross section measurements in large solid angle without the distortion by the energy loss in a sample, but requires a method to estimate the detection efficiency. We solve this problem by using GIC signals and a tight neutron collimation. The validity of this method was confirmed through the {sup 12}C(n,{alpha}{sub 0}){sup 9}Be measurement. We applied this method to the {sup 16}O(n,{alpha}){sup 13}C cross section around 14.1 MeV. (author)

  6. single-top quark production cross section using the ATLAS detector

    CERN Document Server

    Feng, Cunfeng; The ATLAS collaboration

    2014-01-01

    Measurements of single top-quark production cross section in proton proton collisions at 7 and 8 TeV are presented. In the leading order process, a W boson is exchanged in the t-channel. For this process, for the first time a fiducial cross section measured within the detector acceptance is presented and the modelling uncertainty when extrapolating to the total inclusive cross section is assessed with a large number of different Monte Carlo generators. The result is in good agreement with the most up-to-date theory predictions. Furthermore, the single top-quark and anti-top total production cross sections, their ratio, as well as a measurement of the inclusive production cross section is presented. Differential cross sections are measured as a function of the transverse momentum and the absolute value of the rapidity of top and anti-top quarks. In addition, a measurement of the production cross section of a single top quark in association with a W boson is presented. The s-channel production is explored and l...

  7. Rotational averaging of multiphoton absorption cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Friese, Daniel H., E-mail: daniel.h.friese@uit.no; Beerepoot, Maarten T. P.; Ruud, Kenneth [Centre for Theoretical and Computational Chemistry, University of Tromsø — The Arctic University of Norway, N-9037 Tromsø (Norway)

    2014-11-28

    Rotational averaging of tensors is a crucial step in the calculation of molecular properties in isotropic media. We present a scheme for the rotational averaging of multiphoton absorption cross sections. We extend existing literature on rotational averaging to even-rank tensors of arbitrary order and derive equations that require only the number of photons as input. In particular, we derive the first explicit expressions for the rotational average of five-, six-, and seven-photon absorption cross sections. This work is one of the required steps in making the calculation of these higher-order absorption properties possible. The results can be applied to any even-rank tensor provided linearly polarized light is used.

  8. Measurements of neutron spallation cross section. 2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, E.; Nakamura, T. [Tohoku Univ., Sendai (Japan). Cyclotron and Radioisotope Center; Imamura, M.; Nakao, N.; Shibata, S.; Uwamino, Y.; Nakanishi, N.; Tanaka, Su.

    1997-03-01

    Neutron spallation cross section of {sup 59}Co(n,xn){sup 60-x}Co, {sup nat}Cu(n,sp){sup 56}Mn, {sup nat}Cu(n,sp){sup 58}Co, {sup nat}Cu(n,xn){sup 60}Cu, {sup nat}Cu(n,xn){sup 61}Cu and {sup nat}Cu(n,sp){sup 65}Ni was measured in the quasi-monoenergetic p-Li neutron fields in the energy range above 40 MeV which have been established at three AVF cyclotron facilities of (1) INS of Univ. of Tokyo, (2) TIARA of JAERI and (3) RIKEN. Our experimental data were compared with the ENDF/B-VI high energy file data by Fukahori and the calculated cross section data by Odano. (author)

  9. Structured ion impact: Doubly differential cross sections

    International Nuclear Information System (INIS)

    DuBois, R.D.

    1987-01-01

    The electron emission in coincidence with a projectile that has been ionized has been measured, thus making it possible to separate and identify electrons resulting from these various mechanisms. In 1985, coincidence doubly differential cross sections were measured for 400 to 750 keV/atomic mass unit (amu) He + impact on He, Ne, Ar, Kr, and H 2 O. Cross sections were measured for selected angles and for electron energies ranging from 10 to 1000 eV. Because of the coincidence mode of measurement, the total electron emission was subdivided into its target emission and its projectile emission components. The most interesting findings were that target ionization does not account for the electron emission spectrum at lower electron energies. A sizable percentage of these low-energy electrons were shown to originate as a result of simultaneous projectile/target ionizations. Similar features were observed for all targets and impact energies that were studied

  10. Electron-collision cross sections for iodine

    International Nuclear Information System (INIS)

    Zatsarinny, O.; Bartschat, K.; Garcia, G.; Blanco, F.; Hargreaves, L.R.; Jones, D.B.; Murrie, R.; Brunton, J.R.; Brunger, M.J.; Hoshino, M.; Buckman, S.J.

    2011-01-01

    We present results from a joint experimental and theoretical study of elastic electron scattering from atomic iodine. The experimental results were obtained by subtracting known cross sections from the measured data obtained with a pyrolyzed mixed beam containing a variety of atomic and molecular species. The calculations were performed using both a fully relativistic Dirac B-spline R-matrix (close-coupling) method and an optical model potential approach. Given the difficulty of the problem, the agreement between the two sets of theoretical predictions and the experimental data for the angle-differential and the angle-integrated elastic cross sections at 40 eV and 50 eV is satisfactory.

  11. Absolute partial photoionization cross sections of ethylene

    Science.gov (United States)

    Grimm, F. A.; Whitley, T. A.; Keller, P. R.; Taylor, J. W.

    1991-07-01

    Absolute partial photoionization cross sections for ionization out of the first four valence orbitals to the X 2B 3u, A 2B 3g, B 2A g and C 2B 2u states of the C 2H 4+ ion are presented as a function of photon energy over the energy range from 12 to 26 eV. The experimental results have been compared to previously published relative partial cross sections for the first two bands at 18, 21 and 24 eV. Comparison of the experimental data with continuum multiple scattering Xα calculations provides evidence for extensive autoionization to the X 2B 3u state and confirms the predicted shape resonances in ionization to the A 2B 3g and B 2A g states. Identification of possible transitions for the autoionizing resonances have been made using multiple scattering transition state calculations on Rydberg excited states.

  12. Radar cross section measurements using terahertz waves

    DEFF Research Database (Denmark)

    Iwaszczuk, Krzysztof; Heiselberg, Henning; Jepsen, Peter Uhd

    2010-01-01

    Radar cross sections at terahertz frequencies are measured on scale models of aircrafts. A time domain broadband THz system generates freely propagating THz pulses measured with sub-picosecond time resolution. The THz radiation is generated using fs laser pulses by optical rectification...... in order to measure realistic radar cross sections. RCS polar and azimuthal angle plots of F-16 and F-35 are presented....... in a lithium niobate crystal with application of the tilted wave front method, resulting in high electric field THz pulses with a broad band spectrum from 100 GHz up to 4 THz. The corresponding wave lengths are two orders of magnitude smaller than normal radars and we therefore use scale models of size 5-10 cm...

  13. Electron Capture Cross Sections for Stellar Nucleosynthesis

    Directory of Open Access Journals (Sweden)

    P. G. Giannaka

    2015-01-01

    Full Text Available In the first stage of this work, we perform detailed calculations for the cross sections of the electron capture on nuclei under laboratory conditions. Towards this aim we exploit the advantages of a refined version of the proton-neutron quasiparticle random-phase approximation (pn-QRPA and carry out state-by-state evaluations of the rates of exclusive processes that lead to any of the accessible transitions within the chosen model space. In the second stage of our present study, we translate the abovementioned e--capture cross sections to the stellar environment ones by inserting the temperature dependence through a Maxwell-Boltzmann distribution describing the stellar electron gas. As a concrete nuclear target we use the 66Zn isotope, which belongs to the iron group nuclei and plays prominent role in stellar nucleosynthesis at core collapse supernovae environment.

  14. Test of RIPL-2 cross section calculations

    International Nuclear Information System (INIS)

    Herman, M.

    2002-01-01

    The new levels and optical segments and microscopic HF-BCS level densities (part of the density segment) were tested in practical calculations of cross sections for neutron induced reactions on 22 targets (40-Ca, 47-Ti, 52-Cr, 55-Mn, 58-Ni, 63-Cu, 71-Ga, 80-Se, 92-Mo, 93-Nb, 100-Mo, 109-Ag, 114-Cd, 124-Sn, 127-I, 133-Cs, 140-Ce, 153-Eu, 169-Tm, 186-W, 197-Au, 208-Pb). For each target all reactions involving up to 3 neutron, 1 proton and 1 α-particle emissions (subject to actual reaction thresholds) were considered in the incident energy range from 1 keV up to 20 MeV (in some cases up to 27 MeV). In addition, total, elastic, and neutron capture cross sections were calculated

  15. Double differential cross sections of ethane molecule

    Science.gov (United States)

    Kumar, Rajeev

    2018-05-01

    Partial and total double differential cross sections corresponding to various cations C2H6+, C2H4+, C2H5+, C2H3+, C2H2+, CH3+, H+, CH2+, C2H+, H2+, CH+, H3+, C2+ and C+ produced during the direct and dissociative electron ionization of Ethane (C2H6) molecule have been calculated at fixed impinging electron energies 200 and 500eV by using modified Jain-Khare semi empirical approach. The calculation for double differential cross sections is made as a function of energy loss suffered by primary electron and angle of incident. To the best of my knowledge no other data is available for the comparison.

  16. Cross sections required for FMIT dosimetry

    International Nuclear Information System (INIS)

    Gold, R.; McElroy, W.N.; Lippincott, E.P.; Mann, F.M.; Oberg, D.L.; Roberts, J.H.; Ruddy, F.H.

    1980-01-01

    The Fusion Materials Irradiation Test (FMIT) facility, currently under construction, is designed to produce a high flux of high energy neutrons for irradiation effects experiments on fusion reactor materials. Characterization of the flux-fluence-spectrum in this rapidly varying neutron field requires adaptation and extension of currently available dosimetry techniques. This characterization will be carried out by a combination of active, passive, and calculational dosimetry. The goal is to provide the experimenter with accurate neutron flux-fluence-spectra at all positions in the test cell. Plans have been completed for a number of experimental dosimetry stations and provision for these facilities has been incorporated into the FMIT design. Overall needs of the FMIT irradiation damage program delineate goal accuracies for dosimetry that, in turn, create new requirements for high energy neutron cross section data. Recommendations based on these needs have been derived for required cross section data and accuracies

  17. Elliptical cross section fuel rod study II

    International Nuclear Information System (INIS)

    Taboada, H.; Marajofsky, A.

    1996-01-01

    In this paper it is continued the behavior analysis and comparison between cylindrical fuel rods of circular and elliptical cross sections. Taking into account the accepted models in the literature, the fission gas swelling and release were studied. An analytical comparison between both kinds of rod reveals a sensible gas release reduction in the elliptical case, a 50% swelling reduction due to intragranular bubble coalescence mechanism and an important swelling increase due to migration bubble mechanism. From the safety operation point of view, for the same linear power, an elliptical cross section rod is favored by lower central temperatures, lower gas release rates, greater gas store in ceramic matrix and lower stored energy rates. (author). 6 refs., 8 figs., 1 tab

  18. Electron collision cross sections and radiation chemistry

    International Nuclear Information System (INIS)

    Hatano, Y.

    1983-01-01

    A survey is given of the cross section data needs in radiation chemistry, and of the recent progress in electron impact studies on dissociative excitation of molecules. In the former some of the important target species, processes, and collision energies are presented, while in the latter it is demonstrated that radiation chemistry is a source of new ideas and information in atomic collision research. 37 references, 4 figures

  19. Absolute photoionization cross sections of atomic oxygen

    Science.gov (United States)

    Samson, J. A. R.; Pareek, P. N.

    1985-01-01

    The absolute values of photoionization cross sections of atomic oxygen were measured from the ionization threshold to 120 A. An auto-ionizing resonance belonging to the 2S2P4(4P)3P(3Do, 3So) transition was observed at 479.43 A and another line at 389.97 A. The experimental data is in excellent agreement with rigorous close-coupling calculations that include electron correlations in both the initial and final states.

  20. Total dissociation cross section of halo nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Formanek, J. [Karlova Univ., Prague (Czech Republic). Fakulta Matematicko-Fyzikalni; Lombard, R.J. [Paris-11 Univ., 91 - Orsay (France). Inst. de Physique Nucleaire

    1996-10-01

    Calculations of the total dissociation cross section is performed in the impact parameter representation. The case of {sup 11}Be and {sup 11}Li loosing one and two neutron(s), respectively, by collision on a {sup 12}C target, which remains in its ground state are discussed. The results are found to depend essentially on the rms radius of the halo wave function. (author). 12 refs.

  1. Measurements of Fission Cross Sections of Actinides

    CERN Multimedia

    Wiescher, M; Cox, J; Dahlfors, M

    2002-01-01

    A measurement of the neutron induced fission cross sections of $^{237}$Np, $^{241},{243}$Am and of $^{245}$Cm is proposed for the n_TOF neutron beam. Two sets of fission detectors will be used: one based on PPAC counters and another based on a fast ionization chamber (FIC). A total of 5x10$^{18}$ protons are requested for the entire fission measurement campaign.

  2. Cross section of the CMS solenoid

    CERN Multimedia

    Tejinder S. Virdee, CERN

    2005-01-01

    The pictures show a cross section of the CMS solenoid. One can see four layers of the superconducting coil, each of which contains the superconductor (central part, copper coloured - niobium-titanium strands in a copper coating, made into a "Rutherford cable"), surrounded by an ultra-pure aluminium as a magnetic stabilizer, then an aluminium alloy as a mechanical stabilizer. Besides the four layers there is an aluminium mechanical piece that includes pipes that transport the liquid helium.

  3. Cross sections for multistep direct reactions

    International Nuclear Information System (INIS)

    Demetriou, Paraskevi; Marcinkowski, Andrzej; Marianski, Bohdan

    2002-01-01

    Inelastic scattering and charge-exchange reactions have been analysed at energies ranging from 14 to 27 MeV using the modified multistep direct reaction theory (MSD) of Feshbach, Kerman and Koonin. The modified theory considers the non-DWBA matrix elements in the MSD cross section formulae and includes both incoherent particle-hole excitations and coherent collective excitations in the continuum, according to the prescriptions. The results show important contributions from multistep processes at all energies considered. (author)

  4. Capture cross sections for very heavy systems

    International Nuclear Information System (INIS)

    Rowley, N.; Grar, N.; Ntshangase, S.S.

    2006-01-01

    In intermediate-mass systems, collective excitations of the target and projectile can greatly enhance the sub-barrier capture cross section σ cap by giving rise to a distribution of Coulomb barriers. For such systems, capture essentially leads directly to fusion (formation of a compound nucleus (CN)), which then decays through the emission of light particles (neutrons, protons, and alpha particles). Thus the evaporation-residue (ER) cross section is essentially equal to σ cap . For heavier systems the experimental situation is significantly more complicated due to the presence of quasifission (QF) (rapid separation into two fragments before the CN is formed) and by fusion-fission (FF) of the CN itself. Thus three cross sections need to be measured in order to evaluate σ cap . Although the ER essentially recoil along the beam direction. QF and FF fragments are scattered to all angles and require the measurement of angular distribution in order to obtain the excitation function and barrier distribution for capture. Two other approaches to this problem exist. If QF is not important, one can still measure just the ER cross section and try to reconstruct the corresponding σ cap through use of an evaporation-model code that takes account of the FF degree of freedom. Some earlier results on σ cap obtained in this way will be re-analyzed with detail coupled-channels calculations, and the extra-push phenomenon discussed. One may also try to obtain σ cap by exploiting unitarity, that is, by measuring instead the flux of particles corresponding to quasielastic (QE) scattering from the Coulomb barrier. Some new QE results obtained for the 86 Kr + 208 Pb system at iThemba LABS in South Africa will be presented [ru

  5. Inclusive jet cross section at D0

    International Nuclear Information System (INIS)

    Bhattacharjee, M.

    1996-09-01

    Preliminary measurement of the central (|η| ≤ 0.5) inclusive jet cross sections for jet cone sizes of 1.0, 0.7, and 0.5 at D null based on the 1992-1993 (13.7 pb -1 ) and 1994-1995 (90 pb -1 ) data samples are presented. Comparisons to Next-to-Leading Order (NLO) Quantum Chromodynamics (QCD) calculations are made

  6. Fully double-logarithm-resummed cross sections

    International Nuclear Information System (INIS)

    Albino, S.; Bolzoni, P.; Kniehl, B.A.; Kotikov, A.

    2011-01-01

    We calculate the complete double logarithmic contribution to cross sections for semi-inclusive hadron production in the modified minimal-subtraction (MS-bar) scheme by applying dimensional regularization to the double logarithm approximation. The full double logarithmic contribution to the coefficient functions for inclusive hadron production in e + e - annihilation is obtained in this scheme for the first time. Our result agrees with all fixed order results in the literature, which extend to next-to-next-to-leading order.

  7. Atomic-process cross section data, 1

    International Nuclear Information System (INIS)

    1974-12-01

    Compiled by the Data Study Group, the data are intended for fusion plasma physics research. Cross sections of the latest experimental and theoretic studies cover the processes involving H,D,T as principal plasma materials as well as photons and electrons: emission and absorption of electromagnetic wave, electron collision, ion collision, recombination, neutral atom mutual collision, etc. Edition is so made to enable the future renewal by users. (J.P.N.)

  8. Neutron-induced Fission Cross Sections of Am and Cm isotopes (Final Report of Research Contract 14485). Resonance and Fast Neutron Induced Fission Cross Sections of Americium and Curium Nuclides (Third-year Progress Report of Research Contract 14485)

    International Nuclear Information System (INIS)

    Alekseev, A.A.; Bergman, A.A.; Berlev, A.I.; Koptelov, E.A.; Egorov, A.S.; Samylin, B.F.; Trufanov, A.M.; Fursov, B.I.; Shorin, V.S.

    2012-01-01

    The neutron induced fission cross sections of Am and Cm isotopes were measured relative to 239 Pu in the neutron energy range from 1 eV to 20 keV at the INR RAS lead slowing down spectrometer LSDS-100. The fission resonance integrals were also estimated using the measured cross section data. The results have been compared with the available experimental and evaluated data. This analysis has shown the present status of the measured fission cross sections and the necessity to revise the evaluated cross sections libraries for the minor actinides. (author)

  9. Neutron capture cross section of ^243Am

    Science.gov (United States)

    Jandel, M.

    2009-10-01

    The Detector for Advanced Neutron Capture Experiments (DANCE) at Los Alamos National Laboratory (LANL) was used for neutron capture cross section measurement on ^243Am. The high granularity of DANCE (160 BaF2 detectors in a 4π geometry) enables the efficient detection of prompt gamma-rays following neutron capture. DANCE is located on the 20.26 m neutron flight path 14 (FP14) at the Manuel Lujan Jr. Neutron Scattering Center at the Los Alamos Neutron Science Center (LANSCE). The methods and techniques established in [1] were used for the determination of the ^243Am neutron capture cross section. The cross sections were obtained in the range of neutron energies from 0.02 eV to 400 keV. The resonance region was analyzed using SAMMY7 and resonance parameters were extracted. The results will be compared to existing evaluations and calculations. Work was performed under the auspices of the U.S. Department of Energy at Los Alamos National Laboratory by the Los Alamos National Security, LLC under Contract No. DE-AC52-06NA25396 and at Lawrence Livermore National Laboratory by the Lawrence Livermore National Security, LLC under Contract No. DE-AC52-07NA27344. [4pt] [1] M. Jandel et al., Phys. Rev. C78, 034609 (2008)

  10. MXS cross-section preprocessor user's manual

    International Nuclear Information System (INIS)

    Parker, F.; Ishikawa, M.; Luck, L.

    1987-03-01

    The MXS preprocessor has been designed to reduce the execution time of programs using isotopic cross-section data and to both reduce the execution time and improve the accuracy of shielding-factor interpolation in the SIMMER-II accident analysis program. MXS is a dual-purpose preprocessing code to: (1) mix isotopes into materials and (2) fit analytic functions to the shelf-shielding data. The program uses the isotope microscopic neutron cross-section data from the CCCC standard interface file ISOTXS and the isotope Bondarenko self-shielding data from the CCCC standard interface file BRKOXS to generate cross-section and self-shielding data for materials. The materials may be a mixture of several isotopes. The self-shielding data for the materials may be the actual shielding factors or a set of coefficients for functions representing the background dependence of the shielding factors. A set of additional data is given to describe the functions necessary to interpolate the shielding factors over temperature

  11. Neutron capture cross sections of Kr

    Directory of Open Access Journals (Sweden)

    Fiebiger Stefan

    2017-01-01

    Full Text Available Neutron capture and β− -decay are competing branches of the s-process nucleosynthesis path at 85Kr [1], which makes it an important branching point. The knowledge of its neutron capture cross section is therefore essential to constrain stellar models of nucleosynthesis. Despite its importance for different fields, no direct measurement of the cross section of 85Kr in the keV-regime has been performed. The currently reported uncertainties are still in the order of 50% [2, 3]. Neutron capture cross section measurements on a 4% enriched 85Kr gas enclosed in a stainless steel cylinder were performed at Los Alamos National Laboratory (LANL using the Detector for Advanced Neutron Capture Experiments (DANCE. 85Kr is radioactive isotope with a half life of 10.8 years. As this was a low-enrichment sample, the main contaminants, the stable krypton isotopes 83Kr and 86Kr, were also investigated. The material was highly enriched and contained in pressurized stainless steel spheres.

  12. Measurement of actinide neutron cross sections

    International Nuclear Information System (INIS)

    Firestone, Richard B.; Nitsche, Heino; Leung, Ka-Ngo; Perry, DaleL.; English, Gerald

    2003-01-01

    The maintenance of strong scientific expertise is critical to the U.S. nuclear attribution community. It is particularly important to train students in actinide chemistry and physics. Neutron cross-section data are vital components to strategies for detecting explosives and fissile materials, and these measurements require expertise in chemical separations, actinide target preparation, nuclear spectroscopy, and analytical chemistry. At the University of California, Berkeley and the Lawrence Berkeley National Laboratory we have trained students in actinide chemistry for many years. LBNL is a leader in nuclear data and has published the Table of Isotopes for over 60 years. Recently, LBNL led an international collaboration to measure thermal neutron capture radiative cross sections and prepared the Evaluated Gamma-ray Activation File (EGAF) in collaboration with the IAEA. This file of 35, 000 prompt and delayed gamma ray cross-sections for all elements from Z=1-92 is essential for the neutron interrogation of nuclear materials. LBNL has also developed new, high flux neutron generators and recently opened a 1010 n/s D+D neutron generator experimental facility

  13. NNLO jet cross sections by subtraction

    Science.gov (United States)

    Somogyi, G.; Bolzoni, P.; Trócsányi, Z.

    2010-08-01

    We report on the computation of a class of integrals that appear when integrating the so-called iterated singly-unresolved approximate cross section of the NNLO subtraction scheme of Refs. [G. Somogyi, Z. Trócsányi, and V. Del Duca, JHEP 06, 024 (2005), arXiv:hep-ph/0502226; G. Somogyi and Z. Trócsányi, (2006), arXiv:hep-ph/0609041; G. Somogyi, Z. Trócsányi, and V. Del Duca, JHEP 01, 070 (2007), arXiv:hep-ph/0609042; G. Somogyi and Z. Trócsányi, JHEP 01, 052 (2007), arXiv:hep-ph/0609043] over the factorised phase space of unresolved partons. The integrated approximate cross section itself can be written as the product of an insertion operator (in colour space) times the Born cross section. We give selected results for the insertion operator for processes with two and three hard partons in the final state.

  14. Neutron cross section and covariance data evaluation of experimental data for {sup 27}Al

    Energy Technology Data Exchange (ETDEWEB)

    Chunjuan, Li; Jianfeng, Liu [Physics Department , Zhengzhou Univ., Zhengzhou (China); Tingjin, Liu [China Nuclear Data Center, China Inst. of Atomic Energy, Beijing (China)

    2006-07-15

    The evaluation of neutron cross section and covariance data for {sup 27}Al in the energy range from 210 keV to 20 MeV was carried out on the basis of the experimental data mainly taken from EXFOR library. After the experimental data and their errors were analyzed, selected and corrected, SPCC code was used to fit the data and merge the covariance matrix. The evaluated neutron cross section data and covariance matrix for {sup 27}Al given can be collected for the evaluated library and also can be used as the basis of theoretical calculation concerned. (authors)

  15. Neutron cross section and covariance data evaluation of experimental data for 27Al

    International Nuclear Information System (INIS)

    Li Chunjuan; Liu Jianfeng; Liu Tingjin

    2006-01-01

    The evaluation of neutron cross section and covariance data for 27 Al in the energy range from 210 keV to 20 MeV was carried out on the basis of the experimental data mainly taken from EXFOR library. After the experimental data and their errors were analyzed, selected and corrected, SPCC code was used to fit the data and merge the covariance matrix. The evaluated neutron cross section data and covariance matrix for 27 Al given can be collected for the evaluated library and also can be used as the basis of theoretical calculation concerned. (authors)

  16. Preliminary performance evaluation of on-the-fly doppler broadening capability for Monte Carlo simulation in MCS

    International Nuclear Information System (INIS)

    Khassenov, A.; Choi, S.; Lee, H.; Zhang, P.; Zheng, Y.; Lee, D.

    2015-01-01

    This paper examines multipole representation of the cross section and its further Doppler broadening at the resolved resonance region. At the first step, a conversion is performed from nuclear data file resonance parameters to multipoles, which are corresponding poles and residues. The application of multipole representation allows generating the cross section at the target temperature, without pre-generation of 0K cross section libraries. In order to reduce the computational time for cross section generation, window energy concept was implemented and tested. On-the-fly Doppler broadening module based on multipole and windowed multipole representations were implemented into Monte Carlo code, and a pin cell problem was simulated. Simulation time and multiplication factors for different cases were compared with original Monte Carlo simulation results. (author)

  17. Recent joint developments in cross-section uncertainty analysis at Los Alamos and EIR [Eidgenoessiches Institut fuer Reaktorforschung

    International Nuclear Information System (INIS)

    Davidson, J.W.; Dudziak, D.J.; Muir, D.W.; Stepanek, J.; Higgs, C.E.

    1986-01-01

    This paper discusses recent developments and future plans for the SENSIBL code (the successor to the SENSIT[6] and SENSIT-2D[7] codes), along with associated covariance data and cross section libraries. 34 refs

  18. Neutronic calculation and cross section sensitivity analysis of the Livermore mirror fusion/fission hybrid reactor blanket

    International Nuclear Information System (INIS)

    Ku, L.P.; Price, W.G. Jr.

    1977-08-01

    The neutronic calculation for the Livermore mirror fusion/fission hybrid reactor blanket was performed using the PPPL cross section library. Significant differences were found in the tritium breeding and plutonium production in comparison to the results of the LLL calculation. The cross section sensitivity study for tritium breeding indicates that the response is sensitive to the cross section of 238 U in the neighborhood of 14 MeV and 1 MeV. The response is also sensitive to the cross sections of iron in the vicinity of 14 MeV near the first wall. Neutron transport in the resonance region is not important in this reactor model

  19. 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)

  20. Improved algorithms for the calculation of resolved resonance cross sections with applications to the structural Doppler effect in fast reactors

    International Nuclear Information System (INIS)

    Hwang, R.N.; Toppel, B.J.; Henryson, H. II.

    1980-10-01

    Motivated by a need for an economical yet rigorous tool which can address the computation of the structural material Doppler effect, an extremely efficient improved RABANL capability has been developed utilizing the fact that the Doppler broadened line shape functions become essentially identical to the natural line shape functions or Lorentzian limits beyond about 100 Doppler widths from the resonance energy, or when the natural width exceeds about 200 Doppler widths. The computational efficiency has been further enhanced by preprocessing or screening a significant number of selected resonances during library preparation into composition and temperature independent smooth background cross sections. The resonances which are suitable for such pre-processing are those which are either very broad or those which are very weak. The former contribute very little to the Doppler effect and their self-shielding effect can readily be averaged into slowly varying background cross section data, while the latter contribute very little to either the Doppler or to self-shielding effects. To illustrate the accuracy and efficiency of the improved RABANL algorithms and resonance screening techniques, calculations have been performed for two systems, the first with a composition typical of the STF converter region and the second typical of an LMFBR core composition. Excellent agreement has been found for RABANL compared to the reference Monte Carlo solution obtained using the code VIM, and improved results have also been obtained for the narrow resonance approximation in the ultra-fine-group option of MC 2 -2