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Sample records for neutron-group constants library

  1. The TENDL neutron data library and the TEND1038 38-group neutron constant system

    International Nuclear Information System (INIS)

    Abramovich, S.N.; Gorelov, V.P.; Gorshikhin, A.A.; Grebennikov, A.N.; Il'in, V.N.; Krut'ko, N.A.; Farafontov, G.G.

    2002-01-01

    The library contains neutron data for 103 nuclei - i.e. for 38 actinide nuclei (from 232 Th to 249 Cm), 26 fission fragment nuclei and 39 nuclei in structural and technological materials. The 38-group constants were obtained from TENDL. The high-energy group boundary is 20 MeV. The energy range below 1.2 eV contains 11 groups. Temperature and resonance effects were taken into account. The delayed neutron parameters for 6 groups and the yields of 40 fission fragments were obtained (light and heavy, stable and non-stable). The fast neutron features of spherical critical assemblies were calculated using constants from TEND1038. (author)

  2. A program for calculating group constants on the basis of libraries of evaluated neutron data

    International Nuclear Information System (INIS)

    Sinitsa, V.V.

    1987-01-01

    The GRUKON program is designed for processing libraries of evaluated neutron data into group and fine-group (having some 300 groups) microscopic constants. In structure it is a package of applications programs with three basic components: a monitor, a command language and a library of functional modules. The first operative version of the package was restricted to obtaining mid-group non-block cross-sections from evaluated neutron data libraries in the ENDF/B format. This was then used to process other libraries. In the next two versions, cross-section table conversion modules and self-shielding factor calculation modules, respectively, were added to the functions already in the package. Currently, a fourth version of the GRUKON applications program package, for calculation of sub-group parameters, is under preparation. (author)

  3. A library of neutron data for calculating group constants

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  4. Integrated system for production of neutronics and photonics calculational constants. Volume XVI. Tabular and graphical presentation of 175 neutron group constants derived from the LLL evaluated neutron data library (ENDL)

    International Nuclear Information System (INIS)

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

    1975-01-01

    As of February 3, 1975, 175 neutron group constants had been derived from the Evaluated Nuclear Data Library (ENDL) at LLL. In this volume, tables and graphs of the constants are presented along with the conventions used in their preparation. (U.S.)

  5. 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)

  6. The FOND-2.2 evaluated neutron data library (Russian library of evaluated neutron data files for generating sets of constants in the ABBN constants system)

    International Nuclear Information System (INIS)

    Koshcheev, V.N.; Nikolaev, M.N.; Korchagina, Zh.A.; Savoskina, G.V.

    2001-01-01

    A short description is given of the Russian evaluated neutron data library FOND-2.2. The main purpose of FOND-2.2 is to provide sets of constants for the ABBN constants system. A history of its compilation and the sources of the neutron data are given. The contents of FOND-2.2 are presented with brief comments. (author)

  7. 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)

  8. One-group constant libraries for nuclear equilibrium state

    Energy Technology Data Exchange (ETDEWEB)

    Mizutani, Akihiko; Sekimoto, Hiroshi [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors

    1997-03-01

    One-group constant libraries for the nuclear equilibrium state were generated for both liquid sodium cooled MOX fuel type fast reactor and PWR type thermal reactor with Equilibrium Cell Iterative Calculation System (ECICS) using JENDL-3.2, -3, -2 and ENDF/B-VI nuclear data libraries. ECICS produced one-group constant sets for 129 heavy metal nuclides and 1238 fission products. (author)

  9. Application of the variational method for calculation of neutron spectra and group constants - Master thesis

    International Nuclear Information System (INIS)

    Milosevic, M.

    1979-01-01

    One-dimensional variational method for cylindrical configuration was applied for calculating group constants, together with effects of elastic slowing down, anisotropic elastic scattering, inelastic scattering, heterogeneous resonance absorption with the aim to include the presence of a number of different isotopes and effects of neutron leakage from the reactor core. Neutron flux shape P 3 and adjoint function are proposed in order to enable calculation of smaller size reactors and inclusion of heterogeneity effects by cell calculations. Microscopic multigroup constants were prepared based on the UKNDL data library. Analytical-numerical approach was applied for solving the equations of the P 3 approximation to obtain neutron flux moments and adjoint functions

  10. GRUCAL: a program system for the calculation of macroscopic group constants

    International Nuclear Information System (INIS)

    Woll, D.

    1984-01-01

    Nuclear reactor calculations require material- and composition-dependent, energy-averaged neutron physical data in order to decribe the interaction between neutrons and isotopes. The multigroup cross section code GRUCAL calculates these macroscopic group constants for given material compositions from the material-dependent data of the group constant library GRUBA. The instructions for calculating group constants are not fixed in the program, but are read in from an instruction file. This makes it possible to adapt GRUCAL to various problems or different group constant concepts

  11. Macro testing for group constant library TPLIB-95

    International Nuclear Information System (INIS)

    Yao Dong; Zeng Daogui; Liu Jingbo; Wang Yingming; Li Huiyun

    1996-04-01

    A macro test of the group constant library TPLIB-95 was introduced. The TPLIB-95 is an updated group constant library created by China Nuclear Data Center for LWR fuel assembly calculation program package TPFAP based on the JENDL-3.1 evaluation nuclear data library. The calculations and analyses were carried out by using five thermal reactor benchmark issues, a set of PWR zero-power critical experiments, the first cycle reactor core of 300 MW Qinshan NPP as well as the first cycle reactor core of 900 MW Daya Bay NPP. The calculation results for the thermal reactor benchmark issues showed that the maximum deviation between the calculated and measured values for spectrum indexes is large, like 6.7% for ρ 28 of BAPL-2. However, the maximum deviation for k eff is only 0.29% for TRX-2. The calculation results for zero-power critical experiments showed that the calculated value of k eff obtained by using TPLIB-95 is closer to the measured value compared with the one obtained by using the original library TPLIB. The agreement between the calculated and measured values for critical boron concentration in the first cycle reactor cores in Qinshan NPP and Daya Bay NPP is quite good. The maximum deviation for the critical boron concentration is only 15 x 10 -6 /L. (8 figs., 5 tabs.)

  12. GRUCAL, a computer program for calculating macroscopic group constants

    International Nuclear Information System (INIS)

    Woll, D.

    1975-06-01

    Nuclear reactor calculations require material- and composition-dependent, energy averaged nuclear data to describe the interaction of neutrons with individual isotopes in material compositions of reactor zones. The code GRUCAL calculates these macroscopic group constants for given compositions from the material-dependent data of the group constant library GRUBA. The instructions for calculating group constants are not fixed in the program, but will be read at the actual execution time from a separate instruction file. This allows to accomodate GRUCAL to various problems or different group constant concepts. (orig.) [de

  13. Library of files of evaluated neutron data

    International Nuclear Information System (INIS)

    Blokhin, A.I.; Ignatyuk, A.V.; Koshcheev, V.N.; Kuz'minov, B.D.; Manokhin, V.N.; Manturov, G.N.; Nikolaev, M.N.

    1988-01-01

    It is reported about development of the evaluated neutron data files library which was recommended by the GKAE Nuclear Data Commission as the base of improving constant systems in neutron engeneering calculations. A short description of the library content is given and status of the library is pointed out

  14. Comparison of neutron fluxes obtained by 2-D and 3-D geometry with different shielding libraries in biological shield of the TRIGA MARK II reactor

    International Nuclear Information System (INIS)

    Bozic, M.; Zagar, T.; Ravnik, M.

    2003-01-01

    Neutron fluxes in different spatial locations in biological shield are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Libraries used with TORT code were BUGLE-96 library (coupled library with 47 neutron groups and 20 gamma groups) and VITAMIN-B6 library (coupled library with 199 neutron groups and 42 gamma groups). BUGLE-96 library is derived from VITAMIN-B6 library. 2-D and 3-D models for homogeneous type of problem (without inserted beam port 4) and problem with asymmetry (non-homogeneous problem; inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. The main purpose is to verify the possibility for using 2-D approximation model instead of large 3-D model in some calculations. Another purpose of this paper was to compare neutron spectral constants obtained from neutron fluxes (3-D model) determined with smaller BUGLE-96 library with new constants obtained from fluxes calculated with bigger VITAMIN-B6 library. These neutron spectral constants are used in isotopic calculation with SCALE code package (ORIGEN-S). In past only neutron spectral constants determined by neutron fluxes from BUGLE-96 library were used. Experimental results used for isotopic composition comparison are available from irradiation experiment with selected type of concrete and other materials in beam port 4 (irradiation channel 4) in TRIGA Mark II reactor. These experimental results were used as a benchmark in this paper. (author)

  15. 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

  16. The universal library of fission products and delayed neutron group yields

    International Nuclear Information System (INIS)

    Koldobskiy, A.B.; Zhivun, V.M.

    1997-01-01

    A new fission product yield library based on the Semiempirical method for the estimation of their mass and charge distribution is described. Contrary to other compilations, this library can be used with all possible excitation energies of fissionable actinides. The library of delayed neutron group yields, based on the fission product yield compilation, is described as well. (author). 15 refs, 4 tabs

  17. 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

  18. Thermal neutron group constants in monoatomic-gas approximation

    Energy Technology Data Exchange (ETDEWEB)

    Matausek, M V; Bosevski, T [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-12-15

    To solve the problem of space-energy neutron distribution in an elementary reactor cell, a combination of the multigroup procedure and the P{sub 3} approximation of the spherical harmonics method was chosen. The calculation was divided into two independent parts: the first part was to provide multigroup constants which serve as input data for the second part - the determination of the slow neutron spectra. In the present report only the first part of the problem will be discussed. The velocity dependence of cross-sections and scattering function in thermal range was interpreted by the monoatomic-gas model. A digital computer program was developed for the evaluation of the group values for these quantities (author00.

  19. Development of a common nuclear group constants library system: JSSTDL-295n-104γ based on JENDL-3 nuclear data library

    International Nuclear Information System (INIS)

    Hasegawa, A.

    1992-01-01

    JSSTDL 295n-104γ: A common group cross-section library system has been developed in JAERI to be used in fairly wide range of applications in nuclear industry. This system is composed of a common 295n-104γ group cross-section library based on JENDL-3 nuclear data file and its utility codes. Target of this system is focused to the criticality or shielding calculations in fast and fusion reactors using ANISN, DOT, or MORSE code. Specifications of the common group constants were decided responding to the request from various nuclear data users, particularly from nuclear design group in Japan. Group structure is decided so as to cover almost all group structures currently used in our country. This library includes self-shielding factor tables for primary reactions. A routine for generating macro-scopic cross-section using the self-shielding factor table is also provided. Neutron cross-sections and photon production cross-sections are processed by Prof. GROUCH-G/B code system and γ ray transport cross-sections are generated by GAMLEG-JR. In this paper, outline and present status of the JSSTDL library system is described along with two examples adopted in JENDL-3 benchmark test. One is for shielding calculation, where effects of self-shielding factor (f-table) is shown in conjunction with the analysis of the ASPIS natural iron deep penetration experiment. Without considering resonance self-shielding effect in resonance energy region for resonant nuclides like iron, the results is completely missled in the attenuation profile calculation in the shields. The other example is fast rector criticality calculations of very small critical assemblies with very high enrichment fuel materials where some basic characteristics of this library is presented. (orig.)

  20. 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

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

    International Nuclear Information System (INIS)

    Yang Shunhai; Sang Xinzeng

    1993-01-01

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

  2. The Group Neutron Data Library (GNDL)

    International Nuclear Information System (INIS)

    Voronkov, A.V.; Zhuravlev, V.I.; Natrusova, E.G.

    1987-01-01

    The paper describes the structure, organization and basic data representation formats of the GNDL, which was developed at the M.V. Keldysh Institute of Applied Mathematics of the USSR Academy of Sciences for the purpose of neutron data storage and retrieval. A simple method for linking up applications programs with the library is proposed. (author)

  3. Application of the variational method for calculation of neutron spectra and group constants - Master thesis; Primena varijacione metode na odredjivanje spektra neutrona i grupnih konstanti - Magistarski rad

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

    1979-07-01

    One-dimensional variational method for cylindrical configuration was applied for calculating group constants, together with effects of elastic slowing down, anisotropic elastic scattering, inelastic scattering, heterogeneous resonance absorption with the aim to include the presence of a number of different isotopes and effects of neutron leakage from the reactor core. Neutron flux shape P{sub 3} and adjoint function are proposed in order to enable calculation of smaller size reactors and inclusion of heterogeneity effects by cell calculations. Microscopic multigroup constants were prepared based on the UKNDL data library. Analytical-numerical approach was applied for solving the equations of the P{sub 3} approximation to obtain neutron flux moments and adjoint functions.

  4. 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

  5. Translation of selected papers published in Nuclear Constants 5(59), 1984

    International Nuclear Information System (INIS)

    1987-06-01

    The papers selected for this issue of the publication deal with the following topics: The Neutron Physics Constants Bank of the I.V. Kurchatov Institute of Atomic Energy - its structure and contents such as libraries, programs and data preparation codes for reactor calculations. A new version of the unified constant system package (called OKS) has been developed for access to constant systems, such as ARAMAKO-2F, in calculating radiation transport. Input language and performance are described. The group neutron data library GNDL is described in terms of structure, organization and basic data representation formats. The ARMAN'YAK code is described. Its calculation time, special features, and present state are briefly mentioned. Use of the code (for the preparation of constants for calculations and for compiling a library of files of nuclear concentrations) is indicated. A library of neutron data for calculating group constants - the FOND library - is described. The computerized library includes data on the 67 most important nuclear reactor and radiation shielding materials. Under the title ''The INDEhkS program and machine system'' a set of programs for the comparative analysis of calculated and experimental data from integral and macroscopic experiments is presented. The present status of the ARAMAKO multigroup constant calculation system for solving neutron and gamma quantum transport equations is reviewed. A method and a program for automatic preparation of few-group constants for reactor calculations in three-dimensional hexagonal geometry is proposed. A program (GRUKON) for calculating group constants on the basis of libraries of evaluated neutron data is presented. Evaluation of the methodical error in 26-group approximation is discussed. The accuracy of calculation of linear and bilinear functionals using a 26-group approximation is evaluated. A description is given of a five-group system of constants along with a status report on its development

  6. FOND-2.2-evaluated nuclear data library for constants sets generation at ABBN constants providing system

    International Nuclear Information System (INIS)

    Koscheev, V.N.; Nikolaev, M.N.; Tsiboulia, A.M.

    2002-01-01

    The library FOND-2.2 of evaluated nuclear data files, which was created at the ABBN laboratory of IPPE, is described. FOND-2 library is the basic nuclear data source used for the preparation of group data sets with different energy structures. ABBN-93.1 group data set was retrieved from the FOND-2 data library and nowadays it is widely used in different applications, in neutronics calculations of different nuclear energetic installations with different kinds of neutron spectra, in radiation shielding calculation, and so on. (author)

  7. 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

  8. The generation, validation and testing of a coupled 219-group neutron 36-group gamma ray AMPX-II library

    International Nuclear Information System (INIS)

    Panini, G.C.; Siciliano, F.; Lioi, A.

    1987-01-01

    The main characteristics of a P 3 coupled 219-group neutron 36-group gamma-ray library in the AMPX-II Master Interface Format obtained processing ENDF/B-IV data by means of various AMPX-II System modules are presented in this note both for the more reprocessing aspects and features of the generated component files-neutrons, photon and secondary gamma-ray production cross sections. As far as the neutron data are concerned there is the avaibility of 186 data sets regarding most significant fission products. Results of the additional validation of the neutron data pertaining to eighteen benchmark experiments are also given. Some calculational tests on both neutron and coupled data emphasize the important role of the secondary gamma-ray data in nuclear criticality safety calculations

  9. 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)

  10. 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)

  11. 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)

  12. Migros-3: a code for the generation of group constants for reactor calculations from neutron nuclear data in KEDAK format

    International Nuclear Information System (INIS)

    Broeders, I.; Krieg, B.

    1977-01-01

    The code MIGROS-3 was developed from MIGROS-2. The main advantage of MIGROS-3 is its compatibility with the new conventions of the latest version of the Karlsruhe nuclear data library, KEDAK-3. Moreover, to some extent refined physical models were used and numerical methods were improved. MIGROS-3 allows the calculation of microscopic group cross sections of the ABBN type from isotopic neutron data given in KEDAK-format. All group constants, necessary for diffusion-, consistent P 1 - and Ssub(N)-calculations can be generated. Anisotropy of elastic scattering can be taken into account up to P 5 . A description of the code and the underlying theory is given. The input and output description, a sample problem and the program lists are provided. (orig.) [de

  13. The neutron physics constants bank of the I.V. Kurchatov Institute of Atomic Energy

    International Nuclear Information System (INIS)

    Yudkevich, M.S.

    1987-01-01

    This paper describes the structure and contents of a neutron physics constants bank consisting of libraries, service programs and data preparation codes for reactor calculations. Use of the bank makes the constants fully accessible to users. (author)

  14. Qualification test of few group constants generated from an MC method by the two-step neutronics analysis system McCARD/MASTER

    International Nuclear Information System (INIS)

    Park, Ho Jin; Shim, Hyung Jin; Joo, Han Gyu; Kim, Chang Hyo

    2011-01-01

    The purpose of this paper is to examine the qualification of few group constants estimated by the Seoul National University Monte Carlo particle transport analysis code McCARD in terms of core neutronics analyses and thus to validate the McCARD method as a few group constant generator. The two- step core neutronics analyses are conducted for a mini and a realistic PWR by the McCARD/MASTER code system in which McCARD is used as an MC group constant generation code and MASTER as a diffusion core analysis code. The two-step calculations for the effective multiplication factors and assembly power distributions of the two PWR cores by McCARD/MASTER are compared with the reference McCARD calculations. By showing excellent agreements between McCARD/MASTER and the reference MC core neutronics analyses for the two PWRs, it is concluded that the MC method implemented in McCARD can generate few group constants which are well qualified for high-accuracy two-step core neutronics calculations. (author)

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

    International Nuclear Information System (INIS)

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

    1977-07-01

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

  16. The program of group constants creation (SMOK) on basis libraries of evaluated nuclear data in ENDE/B format for physical module FORTUN-88

    International Nuclear Information System (INIS)

    Borisov, A.A.

    1991-01-01

    The SMOK program for creation of group microconstants in the FORTUN-88 physical module format providing for calculations of neutron transport by the Monte Carlo method is described. The program processes files of evaluated neutron nuclear data in the ENDF-4 format. The constant structure gives an apportunity to simulate the process of neutron collisions with matter in details. The program service capabilities provide for graphical constant comparison. 11 refs

  17. Reactor group constants and benchmark test

    Energy Technology Data Exchange (ETDEWEB)

    Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

    The evaluated nuclear data files such as JENDL, ENDF/B-VI and JEF-2 are validated by analyzing critical mock-up experiments for various type reactors and assessing applicability for nuclear characteristics such as criticality, reaction rates, reactivities, etc. This is called Benchmark Testing. In the nuclear calculations, the diffusion and transport codes use the group constant library which is generated by processing the nuclear data files. In this paper, the calculation methods of the reactor group constants and benchmark test are described. Finally, a new group constants scheme is proposed. (author)

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

    International Nuclear Information System (INIS)

    Alpan, F. Arzu; Haghighat, Alireza

    2005-01-01

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

  19. 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

  20. 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

  1. ZZ DECNET-GENDF, Fusion Damage Library of 175 Neutron and 42 Photon VITAMIN-J Groups

    International Nuclear Information System (INIS)

    1997-01-01

    1 - Description of program or function: DECNET is a library for fusion damage computations of 175 neutron + 42 photon VITAMIN-J energy group with the standard weighting function: Maxwellian (at the temperature to which the material is referenced) + 1/E + Fission Spectrum + 1/E + Fusion Peak + 1/E; it includes neutron kerma and gamma-ray production data from radioactive nuclei at 3 temperatures with the same materials of ZZ-GEFF-2-GENDF (see below) from 1-H-1 to Bi-209, mostly taken from EFF-2 with some nuclides from JEF-2.2 - Ag-107, Ag-109, Cd, the 6 Hf isotopes and the 4 W isotopes; however the list of the materials disagrees with that of GEFF-2 in that all elemental nuclides have been split into the components isotopes to follow the respective decay chain and not all materials of GEFF-2 produces nuclei which disintegrate. The library has been produced by the DECKER code which has been developed for this purpose. The format of the library is GENDF. 2 - Method of solution: The library has been produced by the DECKER code developed at ENEA Bologna for this purpose. The code reads the nuclide(s) for which decay kerma and photon production are requested and looks for the necessary data on the RDD (Radioactive Decay Data) file from JEF-2.2

  2. 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

  3. Development of the CANDU 66-group SN transport library

    International Nuclear Information System (INIS)

    Tsang, K.T.

    2001-01-01

    The design of the shield configuration around a nuclear reactor is strongly dependent on the neutron and photon spatial and energy distributions. The nuclear heat deposition and material damage in and surrounding the reactor core are also a function of the neutron and photon distributions. Therefore, to ensure a suitable configuration of materials for shielding or heat transfer, an accurate calculation of the particle fluxes in the reactor systems is essential. The CANDU 66-group library was developed to update the cross sections that are needed to assess the performance of CANDU bulk shields. Since about 1980, shielding analysts at Atomic Energy of Canada Limited (AECL) and Ontario Power Generation Inc. (OPGI) have been using a 38-group CANDU-specific library to perform S N transport calculations. In 1994, a new CANDU 67-group cross-section library was developed. The 67-group cross-section library was developed to provide radiation-physics analysts with up-to-date nuclear data to correct deficiencies with documentation of the old library. Although there were improvements over the 38-group library, initial use showed there were some deficiencies in the 67-group library. To correct these deficiencies, the CANDU 66-group S N transport cross-section library was developed. The 66-group library is based on the 241-group cross-section library VITAMIN-B6. Collapsing and weighting of the 241-group cross sections into 66 groups were performed using the modular code system SCALE 4.4. This paper describes how the modules in the SCALE system were applied to generate the 66-group library. The CANDU 66-group library includes both core-weighted and lattice-weighted cross sections of 235 U, 238 U, and 239 Pu with, and without, delayed fission-product photons. In addition, the 66-group library contains more response functions than did the 67-group library. Finally, the CANDU 66-group library has been validated against one-dimensional benchmark problems. The results generated with

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

    International Nuclear Information System (INIS)

    Mi Aijun; Li Junjie

    2010-01-01

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

  5. 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

  6. 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)

  7. 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

  8. Directional effects in transitional resonance spectra and group constants

    International Nuclear Information System (INIS)

    Hill, R.N.; Oh, K.O.; Rhodes, J.D.

    1989-01-01

    Analytical exploratory investigations indicate that transition effects such as streaming cause a considerable spatial variation in the neutron spectra across resonances; streaming leads to opposite effects in the forward and backward directions. The neglect of this coupled spatial/angular variations of the transitory resonance spectra is an approximation that is common to all current group constant generation methodologies. This paper presents a description of the spatial/angular coupling of the neutron flux across isolated resonances. It appears to be necessary to differentiate between forward-and backward-directed neutron flux components or even to consider components in narrower angular cones. The effects are illustrated for an isolated actinide resonance in a simplified fast reactor blanket problem. The resonance spectra of the directional flux components φ + and φ - , and even more so the 90-deg cone components, are shown to deviate significantly from the infinite medium approximation, and the differences increase with penetration. The charges in φ + lead to a decreasing scattering group constant that enhances neutron transmission; the changes in φ - lead to an increasing group constant inhibiting backward scattering. Therefore, the changes in the forward-and backward-directed spectra both lead to increased neutron transmission. Conversely, the flux (φ = φ + +φ - ) is shown to agree closely with the infinite medium approximation both in the analytical formulas and in the numerical solution. The directional effect cancel in the summation. The forward-and backward-directed flux components are used as weighting spectra to illustrate the group constant changes for a single resonance

  9. Design criteria for the 218-group criticality safety reference library

    International Nuclear Information System (INIS)

    Westfall, R.M.; Ford, W.E. III; Webster, C.C.

    1978-01-01

    The generation of a 218-group neutron cross-section library from ENDF/B-IV data is described. Experience in selecting broad-group subsets and applying them in the analysis of critical experiments is related. Recommendations on the use of the 218-group library are made. 3 figures, 5 tables

  10. Current status of Russian Evaluated Neutron Data Libraries

    International Nuclear Information System (INIS)

    Blokhin, A.I.; Ignatyuk, A.V.; Manokhin, V.N.; Nikolaev, M.N.

    1996-01-01

    The status of Russian Evaluated Data Libraries is discussed. The last modifications of the BROND-2 files and their relations to the additional files of the FOND library and the ABBN-90 group constants are considered. The main characteristics of new libraries for the photoneutron data, dosimetry and activation reaction cross sections and transmutation cross sections for intermediate energies are described briefly. (author)

  11. Neutron Library (ENDL82) in the transmittal format

    International Nuclear Information System (INIS)

    Howerton, R.J.; Dye, R.E.; Perkins, S.T.

    1982-01-01

    There are four main libraries of data included within the system described. They are ENDL (Evaluated Neutron Data Library), ECPL (Evaluated Charged-Particle Data Library), ACTL (Evaluated Neutron-Induced Activation Cross-Section Library), and EGDL (Evaluated Photon Interaction Data Library). The first three deal with nuclear processes induced by neutrons or light charged particles (Z less than or equal to 2, A less than or equal to 4). The fourth (EGDL) contains the data appropriate to photons with energies between 100 eV and 100 MeV that interact with atoms of the elements in their ground state, i.e., cold targets. EGDL does not contain data for photonuclear reactions

  12. Decay constants of subcritical system by diffusion theory for two groups

    International Nuclear Information System (INIS)

    Moura Neto, C. de.

    1977-01-01

    The effects of a neutronic pulse applied to a subcritical multiplicative medium are analysed on the basis of the diffusion theory for one and two groups. The decay constants of the system for various values of geometric buckling were determined from the experimental data. A natural uranium-light water lattice was pulsed employing a Texas Nuclear 9905 neutron generator. The least square method was employed in the data reduction procedures to determine the decay constants. The separation of the decay constants associated with thermal and epithermal fluxes is attempted through two groups formulation. (author)

  13. Decay constants of a subcritical system by two-group diffusion theory

    International Nuclear Information System (INIS)

    Moura Neto, C. de.

    1979-08-01

    The effects of a neutronic pulse applied to a subcritical multiplicative medium are analyzed on the basis of the diffusion theory for one and two groups. The decay constants of the system were determined from the experimental data, for various values geometric buckling. A natural uranium light-water configuration was pulsed employing a Texas Nuclear 9905 neutron generator. The least square method was employed in the data reduction procedures to determine the decay constants. The separation of the decay constants associated with thermal and epithermal fluxes are verified through two groups formulation. (Author) [pt

  14. International evaluated neutron nuclear data libraries

    International Nuclear Information System (INIS)

    Liu Tingjin

    2001-01-01

    The current status of five major evaluated neutron nuclear data libraries in the world are introduced. They are ENDF/B-6 (U. S. A.), JENDL-3.2 (Japan), JEF-2.2 (Europe), CENDL-2.1 (China), BROND-2 (Russia). The developing trend of the international neutron evaluated nuclear data library is discussed. How to get and use these data for the domestic users is given

  15. Development of a 1200 fine group nuclear data library for advanced nuclear systems

    Institute of Scientific and Technical Information of China (English)

    Jun Zou; Lei-Ming Shang; Fang Wang; Li-Juan Hao

    2017-01-01

    Accurate and reliable nuclear data libraries are essential for calculation and design of advanced nuclear systems.A 1200 fine group nuclear data library Hybrid Evaluated Nuclear Data Library/Fine Group (HENDL/FG) with neutrons of up to 150 MeV has been developed to improve the accuracy of neutronics calculations and analysis.Corrections of Doppler,resonance self-shielding,and thermal upscatter effects were done for HENDL/FG.Shielding and critical safety benchmarks were performed to test the accuracy and reliability of the library.The discrepancy between calculated and measured nuclear parameters fell into a reasonable range.

  16. 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

  17. The computer library of experimental neutron data

    International Nuclear Information System (INIS)

    Bychkov, V.M.; Manokhin, V.N.; Surgutanov, V.V.

    1976-05-01

    The paper describes the computer library of experimental neutron data at the Obninsk Nuclear Data Centre. The format of the library (EXFOR) and the system of programmes for supplying the library are briefly described. (author)

  18. 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

  19. 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

  20. Production of neutronic discrete equations for a cylindrical geometry in one group energy and benchmark the results with MCNP-4B code with one group energy library

    International Nuclear Information System (INIS)

    Salehi, A. A.; Vosoughi, N.; Shahriari, M.

    2002-01-01

    In reactor core neutronic calculations, we usually choose a control volume and investigate about the input, output, production and absorption inside it. Finally, we derive neutron transport equation. This equation is not easy to solve for simple and symmetrical geometry. The objective of this paper is to introduce a new direct method for neutronic calculations. This method is based on physics of problem and with meshing of the desired geometry, writing the balance equation for each mesh intervals and with notice to the conjunction between these mesh intervals, produce the final discrete equation series without production of neutron transport differential equation and mandatory passing form differential equation bridge. This method, which is named Direct Discrete Method, was applied in static state, for a cylindrical geometry in one group energy. The validity of the results from this new method are tested with MCNP-4B code with a one group energy library. One energy group direct discrete equation produces excellent results, which can be compared with the results of MCNP-4B

  1. Validation of the BUGJEFF311.BOLIB, BUGENDF70.BOLIB and BUGLE-B7 broad-group libraries on the PCA-Replica (H2O/Fe neutron shielding benchmark experiment

    Directory of Open Access Journals (Sweden)

    Pescarini Massimo

    2016-01-01

    Full Text Available The PCA-Replica 12/13 (H2O/Fe neutron shielding benchmark experiment was analysed using the TORT-3.2 3D SN code. PCA-Replica reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with the same energy group structure (47 n + 20 γ and based on different nuclear data were alternatively used: the ENEA BUGJEFF311.BOLIB (JEFF-3.1.1 and UGENDF70.BOLIB (ENDF/B-VII.0 libraries and the ORNL BUGLE-B7 (ENDF/B-VII.0 library. Dosimeter cross sections derived from the IAEA IRDF-2002 dosimetry file were employed. The calculated reaction rates for the Rh-103(n,n′Rh-103m, In-115(n,n′In-115m and S-32(n,pP-32 threshold activation dosimeters and the calculated neutron spectra are compared with the corresponding experimental results.

  2. Validation of the BUGJEFF311.BOLIB, BUGENDF70.BOLIB and BUGLE-B7 broad-group libraries on the PCA-Replica (H2O/Fe) neutron shielding benchmark experiment

    Science.gov (United States)

    Pescarini, Massimo; Orsi, Roberto; Frisoni, Manuela

    2016-03-01

    The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the TORT-3.2 3D SN code. PCA-Replica reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with the same energy group structure (47 n + 20 γ) and based on different nuclear data were alternatively used: the ENEA BUGJEFF311.BOLIB (JEFF-3.1.1) and UGENDF70.BOLIB (ENDF/B-VII.0) libraries and the ORNL BUGLE-B7 (ENDF/B-VII.0) library. Dosimeter cross sections derived from the IAEA IRDF-2002 dosimetry file were employed. The calculated reaction rates for the Rh-103(n,n')Rh-103m, In-115(n,n')In-115m and S-32(n,p)P-32 threshold activation dosimeters and the calculated neutron spectra are compared with the corresponding experimental results.

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  4. 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.

  5. Neutron Capture Gamma-Ray Libraries for Nuclear Applications

    International Nuclear Information System (INIS)

    Sleaford, B. W.; Summers, N.; Escher, J.; Firestone, R. B.; Basunia, S.; Hurst, A.; Krticka, M.; Molnar, G.; Belgya, T.; Revay, Z.; Choi, H. D.

    2011-01-01

    The neutron capture reaction is useful in identifying and analyzing the gamma-ray spectrum from an unknown assembly as it gives unambiguous information on its composition. This can be done passively or actively where an external neutron source is used to probe an unknown assembly. There are known capture gamma-ray data gaps in the ENDF libraries used by transport codes for various nuclear applications. The Evaluated Gamma-ray Activation file (EGAF) is a new thermal neutron capture database of discrete line spectra and cross sections for over 260 isotopes that was developed as part of an IAEA Coordinated Research Project. EGAF is being used to improve the capture gamma production in ENDF libraries. For medium to heavy nuclei the quasi continuum contribution to the gamma cascades is not experimentally resolved. The continuum contains up to 90% of all the decay energy and is modeled here with the statistical nuclear structure code DICEBOX. This code also provides a consistency check of the level scheme nuclear structure evaluation. The calculated continuum is of sufficient accuracy to include in the ENDF libraries. This analysis also determines new total thermal capture cross sections and provides an improved RIPL database. For higher energy neutron capture there is less experimental data available making benchmarking of the modeling codes more difficult. We are investigating the capture spectra from higher energy neutrons experimentally using surrogate reactions and modeling this with Hauser-Feshbach codes. This can then be used to benchmark CASINO, a version of DICEBOX modified for neutron capture at higher energy. This can be used to simulate spectra from neutron capture at incident neutron energies up to 20 MeV to improve the gamma-ray spectrum in neutron data libraries used for transport modeling of unknown assemblies.

  6. Neutron Capture Gamma-Ray Libraries for Nuclear Applications

    International Nuclear Information System (INIS)

    Sleaford, B.W.; Firestone, R.B.; Summers, N.; Escher, J.; Hurst, A.; Krticka, M.; Basunia, S.; Molnar, G.; Belgya, T.; Revay, Z.; Choi, H.D.

    2010-01-01

    The neutron capture reaction is useful in identifying and analyzing the gamma-ray spectrum from an unknown assembly as it gives unambiguous information on its composition. this can be done passively or actively where an external neutron source is used to probe an unknown assembly. There are known capture gamma-ray data gaps in the ENDF libraries used by transport codes for various nuclear applications. The Evaluated Gamma-ray Activation file (EGAF) is a new thermal neutron capture database of discrete line spectra and cross sections for over 260 isotopes that was developed as part of an IAEA Coordinated Research project. EGAF is being used to improve the capture gamma production in ENDF libraries. For medium to heavy nuclei the quasi continuum contribution to the gamma cascades is not experimentally resolved. The continuum contains up to 90% of all the decay energy and is modeled here with the statistical nuclear structure code DICEBOX. This code also provides a consistency check of the level scheme nuclear structure evaluation. The calculated continuum is of sufficient accuracy to include in the ENDF libraries. This analysis also determines new total thermal capture cross sections and provides an improved RIPL database. For higher energy neutron capture there is less experimental data available making benchmarking of the modeling codes more difficult. They are investigating the capture spectra from higher energy neutrons experimentally using surrogate reactions and modeling this with Hauser-Feshbach codes. This can then be used to benchmark CASINO, a version of DICEBOX modified for neutron capture at higher energy. This can be used to simulate spectra from neutron capture at incident neutron energies up to 20 MeV to improve the gamma-ray spectrum in neutron data libraries used for transport modeling of unknown assemblies.

  7. Nuclear data evaluation and group constant generation for reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Gil, Choong Sup; Min, Byung Joo; Lee, Jong Tai [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1993-01-01

    In nuclear or shielding design analysis for reactors or other facilities, nuclear data are one of the primary importances. Research project for nuclear data evaluation and their effective applications has been continuously performed. The objectives of this project are (1) to compile the latest evaluated nuclear data files, (2) to establish their processing code systems, and (3) to evaluate the multi-group constant library using the newly compiled data files and the code systems. As the results of this project, the latest version of NJOY nuclear data processing system, NJOY91.38 which is capable of processing data in ENDF-6 format, was compiled and installed in Cyber 960-31(OS : NOS/VE) and HP710 workstation. A 50-group constant library for fast reactor was generated with NJOY91.38 using evaluated data from JEF-1 and benchmark test of this library was performed. The newly generated library has been found to do an excellent job of calculating integral quantities for fast critical assemblies and is expected to be positively used to develop fast reactors. (Author).

  8. 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)

  9. BROND: USSR Evaluated Neutron Data Library

    International Nuclear Information System (INIS)

    Manokhin, V.N.; Lemmel, H.D.; McLaughlin, P.K.

    1989-10-01

    BROND, the USSR computerized data library for evaluated neutron reaction data was released in 1987/1989. Upon request it is available on magnetic tape, costfree, from the IAEA Nuclear Data Section. This document describes the contents of this library in the versions BROND-NDS1 and BROND-NDS2. (author). 16 refs

  10. 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.)

  11. TIMS-1: a processing code for production of group constants of heavy resonant nuclei

    International Nuclear Information System (INIS)

    Takano, Hideki; Ishiguro, Yukio; Matsui, Yasushi.

    1980-09-01

    The TIMS-1 code calculates the infinitely dilute group cross sections and the temperature dependent self-shielding factors for arbitrary values of σ 0 and R, where σ 0 is the effective background cross section of potential scattering and R the ratio of the atomic number densities for two resonant nuclei if any. This code is specifically programmed to use the evaluated nuclear data file of ENDF/B or JENDL as input data. In the unresolved resonance region, the resonance parameters and the level spacings are generated by using Monte Carlo method from the Porter-Thomas and Wigner distributions respectively. The Doppler broadened cross sections are calculated on the ultra-fine lethargy meshes of about 10 -3 -- 10 -5 using the generated and resolved resonance parameters. The effective group constants are calculated by solving the neutron slowing down equation with the use of the recurrence formula for the neutron slowing down source. The output of the calculated results is given in a format being consistent with the JAERI-Fast set (JFS) or the Standard Reactor Analysis Code (SRAC) library. Both FACOM 230/75 and M200 versions of TIMS-1 are available. (author)

  12. 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)

  13. BROND. USSR Evaluated Neutron Data Library

    International Nuclear Information System (INIS)

    Manokhin, V.N.

    1990-09-01

    BROND, the USSR computerized data library for evaluated neutron reaction data was released in 1987/1990. Upon request it is available on magnetic tape, costfree, from the IAEA Nuclear Data Section. This document describes the contents of this library in the versions BROND-NDS1 and BROND-NDS2, BROND-NDS3, BROND-NDS4. (author). 16 refs

  14. BROND: USSR evaluated neutron data library

    International Nuclear Information System (INIS)

    Manokhin, V.N.

    1991-03-01

    BROND, the USSR computerized data library for evaluated neutron reaction data was released in 1987/1990. Upon request it is available on magnetic tape, costfree, from the IAEA Nuclear Data Section. This document describes the contents of this library in the four supplements BROND-NDS1, BROND-NDS2, BROND-NDS3, BROND-NDS4. (author). 16 refs

  15. 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

  16. 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

  17. 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

    materials and fission products, and 20 actinides. Covariances are given in 33-energy groups, from 10?5 eV to 19.6 MeV, obtained by processing with LANL processing code NJOY using 1/E flux. In addition to these 110 files, the library contains 20 files with nu-bar covariances, 3 files with covariances of prompt fission neutron spectra (238,239,240-Pu), and 2 files with mu-bar covariances (23-Na, 56-Fe). Over the period of three years several working versions of the library have been released and tested by ANL and INL reactor analysts. Useful feedback has been collected allowing gradual improvements of the library. In addition, QA system was developed to check basic properties and features of the whole library, allowing visual inspection of uncertainty and correlations plots, inspection of uncertainties of integral quantities with independent databases, and dispersion of cross sections between major evaluated libraries. The COMMARA-2.0 beta version of the library was released to ANL and INL reactor analysts in October 2010. The final version, described in the present report, was released in March 2011.

  18. 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

    structural materials and fission products, and 20 actinides. Covariances are given in 33-energy groups, from 10?5 eV to 19.6 MeV, obtained by processing with LANL processing code NJOY using 1/E flux. In addition to these 110 files, the library contains 20 files with nu-bar covariances, 3 files with covariances of prompt fission neutron spectra (238,239,240-Pu), and 2 files with mu-bar covariances (23-Na, 56-Fe). Over the period of three years several working versions of the library have been released and tested by ANL and INL reactor analysts. Useful feedback has been collected allowing gradual improvements of the library. In addition, QA system was developed to check basic properties and features of the whole library, allowing visual inspection of uncertainty and correlations plots, inspection of uncertainties of integral quantities with independent databases, and dispersion of cross sections between major evaluated libraries. The COMMARA-2.0 beta version of the library was released to ANL and INL reactor analysts in October 2010. The final version, described in the present report, was released in March 2011.

  19. 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)

  20. Reference neutron activation library

    International Nuclear Information System (INIS)

    2002-04-01

    Many scientific endeavors require accurate nuclear data. Examples include studies of environmental protection connected with the running of a nuclear installation, the conceptual designs of fusion energy producing devices, astrophysics and the production of medical isotopes. In response to this need, many national and international data libraries have evolved over the years. Initially nuclear data work concentrated on materials relevant to the commercial power industry which is based on the fission of actinides, but recently the topic of activation has become of increasing importance. Activation of materials occurs in fission devices, but is generally overshadowed by the primary fission process. In fusion devices, high energy (14 MeV) neutrons produced in the D-T fusion reaction cause activation of the structure, and (with the exception of the tritium fuel) is the dominant source of activity. Astrophysics requires cross-sections (generally describing neutron capture) or its studies of nucleosynthesis. Many analytical techniques require activation analysis. For example, borehole logging uses the detection of gamma rays from irradiated materials to determine the various components of rocks. To provide data for these applications, various specialized data libraries have been produced. The most comprehensive of these have been developed for fusion studies, since it has been appreciated that impurities are of the greatest importance in determining the overall activity, and thus data on all elements are required. These libraries contain information on a wide range of reactions: (n,γ), (n,2n), (n,α), (n,p), (n,d), (n,t), (n, 3 He)and (n,n')over the energy range from 10 -5 eV to 15 or 20 MeV. It should be noted that the production of various isomeric states have to be treated in detail in these libraries,and that the range of targets must include long-lived radioactive nuclides in addition to stable nuclides. These comprehensive libraries thus contain almost all the

  1. Reference neutron activation library

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-04-01

    Many scientific endeavors require accurate nuclear data. Examples include studies of environmental protection connected with the running of a nuclear installation, the conceptual designs of fusion energy producing devices, astrophysics and the production of medical isotopes. In response to this need, many national and international data libraries have evolved over the years. Initially nuclear data work concentrated on materials relevant to the commercial power industry which is based on the fission of actinides, but recently the topic of activation has become of increasing importance. Activation of materials occurs in fission devices, but is generally overshadowed by the primary fission process. In fusion devices, high energy (14 MeV) neutrons produced in the D-T fusion reaction cause activation of the structure, and (with the exception of the tritium fuel) is the dominant source of activity. Astrophysics requires cross-sections (generally describing neutron capture) or its studies of nucleosynthesis. Many analytical techniques require activation analysis. For example, borehole logging uses the detection of gamma rays from irradiated materials to determine the various components of rocks. To provide data for these applications, various specialized data libraries have been produced. The most comprehensive of these have been developed for fusion studies, since it has been appreciated that impurities are of the greatest importance in determining the overall activity, and thus data on all elements are required. These libraries contain information on a wide range of reactions: (n,{gamma}), (n,2n), (n,{alpha}), (n,p), (n,d), (n,t), (n,{sup 3}He)and (n,n')over the energy range from 10{sup -5} eV to 15 or 20 MeV. It should be noted that the production of various isomeric states have to be treated in detail in these libraries,and that the range of targets must include long-lived radioactive nuclides in addition to stable nuclides. These comprehensive libraries thus contain

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    Kim, Jung Do; Gil, Choong Sup.

    1997-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-03-01

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

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

    International Nuclear Information System (INIS)

    Suzuki, Tomoo

    1971-01-01

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

  6. 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.)

  7. 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.

  8. Neutron data library for transactinides at energies up to 100 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Korovin, Y.A.; Artisyuk, V.V.; Konobeyev, A.Y. [Obninsk Institute of Nuclear Power Engineering (Russian Federation)

    1995-10-01

    New neutron data library for transactinides is briefly described. The library includes evaluated cross-sections for fission and threshold neutron induced reactions for isotopes of U, Np and Pu at energies 0-100 MeV.

  9. BROND-2. USSR Evaluated Neutron Data Library

    International Nuclear Information System (INIS)

    Manokhin, V.N.

    1992-01-01

    BROND-2, the USSR computerized data library for evaluated neutron reaction data was released in 1992. Upon request it is available on magnetic tape, costfree, from the IAEA Nuclear Data Section. The Library BROND-2 reported in this document supersedes the earlier versions ''BROND-NDS1'' plus three supplements ''BROND-NDS2'', ''BROND-NDS3'', ''BROND-NDS4''. 16 refs

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

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Moraes, Marisa de

    1985-01-01

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

  11. Decay of the pulsed thermal neutron flux in two-zone hydrogenous systems - Monte Carlo simulations using MCNP standard data libraries

    International Nuclear Information System (INIS)

    Wiacek, Urszula; Krynicka, Ewa

    2006-01-01

    Pulsed neutron experiments in two-zone spherical and cylindrical geometry has been simulated using the MCNP code. The systems are built of hydrogenous materials. The inner zone is filled with aqueous solutions of absorbers (H 3 BO 3 or KCl). It is surrounded by the outer zone built of Plexiglas. The system is irradiated with the pulsed thermal neutron flux and the thermal neutron decay in time is observed. Standard data libraries of the thermal neutron scattering cross-sections of hydrogen in hydrogenous substances have been used to simulate the neutron transport. The time decay constant of the fundamental mode of the thermal neutron flux determined in each simulation has been compared with the corresponding result of the real pulsed neutron experiment

  12. Development code for group constant processing

    International Nuclear Information System (INIS)

    Su'ud, Z.

    1997-01-01

    In this paper methods, formalism and algorithm related to group constant processing problem from basic library such as ENDF/B VI will be described. Basically the problems can be grouped as follows; the treatment of resolved resonance using NR approximation, the treatment of unresolved resonance using statistical method, the treatment of low lying resonance using intermediate resonance approximation, the treatment of thermal energy regions, and the treatment group transfer matrices cross sections. it is necessary to treat interference between resonance properly especially in the unresolved region. in this paper the resonance problems are treated based on Breit-wigner method, and doppler function is treated using Pade approximation for calculation efficiency. finally, some samples of calculational result for some nuclei, mainly for comparison between many methods are discussed in this paper

  13. Integrated system for production of neutronics and photonics calculational constants. Volume 15, Part C. The LLL Evaluated Nuclear Data Library (ENDL): translation of ENDL neutron-induced interaction data into the ENDF/B format

    International Nuclear Information System (INIS)

    Howerton, R.J.

    1976-01-01

    The LLL evaluated nuclear data library (ENDL) has been translated into the evaluated neutron data file/version B (ENDF/B) format. This translation is for the convenience of those who wish to use ENDL data but who are more familiar with ENDF/B formats and procedures. Only that portion of ENDL dealing with neutron-induced interactions (including photon production from neutron-induced reactions) has been translated

  14. 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)

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  16. WIMS-IJSO - An extended version of the WIMS group constant library

    International Nuclear Information System (INIS)

    Trkov, A.; Perdan, A.

    1982-11-01

    WIMS-IJSO is a preliminary extended version of the WIMS library as supplied with the CDC version of the S-WIMS-code. It is a result of the feasibility study of the possibility to update and extend the WIMS library. In the course of its preparation valuable experience was gained in understanding the definitions, the conventions and the structure of the WIMS library. This experience will be used in the preparation of an extended WIMS library for some materials from carefully chosen evaluated data. The library as supplied with the WIMS-D4 package will be used as the basis. The materials added to the library had been tested in reactor calculations and their performance was found to be satisfactory. The aim of releasing the preliminary version of the WIMS library is to allow different users to apply the data to various problems

  17. Validation of the BUGJEFF311.BOLIB, BUGENDF70.BOLIB and BUGLE-B7 broad-group libraries on the PCA-Replica (H2O/Fe) neutron shielding benchmark experiment

    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 TORT-3.2 3D SN code. PCA-Replica reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with the same energy group structure (47 n + 20 γ) and based on different nuclear data were alternatively used: the ENEA BUGJEFF311.BOLIB (JEFF-3.1.1) and U...

  18. BROND: USSR recommended evaluated neutron data library

    International Nuclear Information System (INIS)

    Manokhin, V.N.

    1988-03-01

    BROND is the recommended evaluated data library of the USSR for neutron induced nuclear reactions. It is a computer library recorded on magnetic tape presented in the internationally recommended format ENDF-5. It contains 65 files with recommended data for 65 elements or isotopes. For each file the present report gives a summary documentation on the contents, the evaluation methods and the originators of the files. (author). Refs

  19. LLNL nuclear data libraries used for fusion calculations

    International Nuclear Information System (INIS)

    Howerton, R.J.

    1984-01-01

    The Physical Data Group of the Computational Physics Division of the Lawrence Livermore National Laboratory has as its principal responsibility the development and maintenance of those data that are related to nuclear reaction processes and are needed for Laboratory programs. Among these are the Magnetic Fusion Energy and the Inertial Confinement Fusion programs. To this end, we have developed and maintain a collection of data files or libraries. These include: files of experimental data of neutron induced reactions; an annotated bibliography of literature related to charged particle induced reactions with light nuclei; and four main libraries of evaluated data. We also maintain files of calculational constants developed from the evaluated libraries for use by Laboratory computer codes. The data used for fusion calculations are usually these calculational constants, but since they are derived by prescribed manipulation of evaluated data this discussion will describe the evaluated libraries

  20. Correction factor for the experimental prompt neutron decay constant

    International Nuclear Information System (INIS)

    Talamo, Alberto; Gohar, Y.; Sadovich, S.; Kiyavitskaya, H.; Bournos, V.; Fokov, Y.; Routkovskaya, C.

    2013-01-01

    Highlights: • Definition of a spatial correction factor for the experimental prompt neutron decay constant. • Introduction of a MCNP6 calculation methodology to simulate Rossi-alpha distribution for pulsed neutron sources. • Comparison of MCNP6 results with experimental data for count rate, Rossi-alpha, and Feynman-alpha distributions. • Improvement of the comparison between numerical and experimental results by taking into account the dead-time effect. - Abstract: This study introduces a new correction factor to obtain the experimental effective multiplication factor of subcritical assemblies by the point kinetics formulation. The correction factor is defined as the ratio between the MCNP6 prompt neutron decay constant obtained in criticality mode and the one obtained in source mode. The correction factor mainly takes into account the longer neutron lifetime in the reflector region and the effects of the external neutron source. For the YALINA Thermal facility, the comparison between the experimental and computational effective multiplication factors noticeably improves after the application of the correction factor. The accuracy of the MCNP6 computational model of the YALINA Thermal subcritical assembly has been verified by reproducing the neutron count rate, Rossi-α, and Feynman-α distributions obtained from the experimental data

  1. A nuclear data library for activity determinations of selected nuclides

    International Nuclear Information System (INIS)

    Baard, J.H.

    1991-11-01

    This report describes the GAMLIB 1-5 library, which is used in the calculation of the activity of radionuclides present in the gamma-ray spectra of irradiated neutron fluence detectors. The library contains all constants needed to calculate the activity for reactions normally applied in neutron fluence determinations, performed in irradiation experiments in the HFR. It also contains the nuclide constants for the activity calculation of gamma-ray measurements of U and Pu samples. The library consists of two kinds of tables, the first containing gamma-ray energies and gamma-ray emission probabilities with their uncertainties and the nuclide code, the other the nuclide code, decay constant, gamma -ray energies and gamma-ray emission probabilities. No cross-section data are stored in this library. All the relevant dat of the Nuclear Data Guide (Dordrecht, Kluwer 1989) have been used as base for this library. Other data have been obtained from recent literature. This library comprises 155 nuclides and 1115 gamma-ray energies. (author). 9 refs

  2. ORLIB: a computer code that produces one-energy group, time- and spatially-averaged neutron cross sections

    International Nuclear Information System (INIS)

    Blink, J.A.; Dye, R.E.; Kimlinger, J.R.

    1981-12-01

    Calculation of neutron activation of proposed fusion reactors requires a library of neutron-activation cross sections. One such library is ACTL, which is being updated and expanded by Howerton. If the energy-dependent neutron flux is also known as a function of location and time, the buildup and decay of activation products can be calculated. In practice, hand calculation is impractical without energy-averaged cross sections because of the large number of energy groups. A widely used activation computer code, ORIGEN2, also requires energy-averaged cross sections. Accordingly, we wrote the ORLIB code to collapse the ACTL library, using the flux as a weighting function. The ORLIB code runs on the LLNL Cray computer network. We have also modified ORIGEN2 to accept the expanded activation libraries produced by ORLIB

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

    International Nuclear Information System (INIS)

    Muir, D.W.

    1997-01-01

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

  4. 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

  5. Determination of prompt neutron decay constant of the AP-600 reactor core

    International Nuclear Information System (INIS)

    Surbakti, T.

    1998-01-01

    Determination of prompt neutron decay constant of the AP-600 reactor core has been performed using combination of two codes WIMS/D4 and Batan-2DIFF. The calculation was done at beginning of cycle and all of control rods pulled out. Cell generation from various kinds of core materials was done with 4 neutron energy group in 1-D transport code (WIMS/D4). The cell is considered for 1/4 fuel assembly in cluster model with square pitch arrange and then, the dimension of its unit cell is calculated. The unit cell consist of a fuel and moderator unit. The unit cell dimension as input data of WIMS/D4 code, called it annulus, is obtained from the equivalent unit cell. Macroscopic cross sections as output was used as input on neutron diffusion code Batan-2DIFF for core calculation as appropriate with three enrichment regions of the fuel of AP-600 core, namely 2, 2.5, and 3%. From result of diffusion code ( Batan-2DIFF) is obtained the value of delayed neutron fraction of 6.932E-03 and average prompt neutron life-time of 26.38 μs, so that the value of prompt neutron decay constant is 262.8 s-1. If it is compared the calculation result with the design value, the deviation are, for the design value of delayed neutron fraction is 7.5E-03, about 8% and the design value of average prompt neutron life time is 19.6 μs, about 34% respectively. The deviation because there are still unknown several core components of AP-600, so it didn't include in calculation yet

  6. 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

  7. Four energy group neutron flux distribution in the Syrian miniature neutron source reactor using the WIMSD4 and CITATION code

    International Nuclear Information System (INIS)

    Khattab, K.; Omar, H.; Ghazi, N.

    2009-01-01

    A 3-D (R, θ , Z) neutronic model for the Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis. The group constants for all the reactor components were generated using the WIMSD4 code. The reactor excess reactivity and the four group neutron flux distributions were calculated using the CITATION code. This model is used in this paper to calculate the point wise four energy group neutron flux distributions in the MNSR versus the radius, angle and reactor axial directions. Good agreement is noticed between the measured and the calculated thermal neutron flux in the inner and the outer irradiation site with relative difference less than 7% and 5% respectively. (author)

  8. 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)

  9. 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)

  10. Extension of the AUS reactor neutronics system for application to fusion blanket neutronics

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1984-03-01

    The AUS modular code scheme for reactor neutronics computations has been extended to apply to fusion blanket neutronics. A new group cross-section library with 200 neutron groups, 37 photon groups and kerma factor data has been generated from ENDF/B-IV. The library includes neutron resonance subgroup parameters and temperature-dependent data for thermal neutron scattering matrices. The validity of the overall calculation system for fusion applications has been checked by comparison with a number of published conceptual design studies

  11. 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

  12. The pulsed neutron method applied to the determination of the nuclear constants of graphite (1961); La methode des neutrons pulses appliquee a la determination des constantes nucleaires du graphite (1961)

    Energy Technology Data Exchange (ETDEWEB)

    Lalande, J [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1961-07-01

    A method for determining the nuclear constants {sigma}{sub a} and {lambda}{sub t} of a moderator is described. The disappearance of a burst of neutrons introduced into a finite medium is studied as a function of time. This decrease in the thermal neutron density is the product of two exponentials; one representing the absorption, the other the leakage. By varying one or other of these factors, the constants of the factor left unvaried can be determined, and from this the nuclear constant values are deduced. (author) [French] On decrit une methode permettant de determiner les constantes nucleaires {sigma}{sub a} et {lambda}{sub t} d'un moderateur. On etudie la decroissance dans le temps d'une bouffee de neutrons introduite dans un milieu fini. Cette decroissance de la densite en neutrons thermiques est le produit de deux exponentielles; l'une represente l'absorption, l'autre les fuites. Par variation de l'un ou l'autre de ces facteurs, on determine les constantes du facteur laisse invariant dont on deduit les valeurs des constantes nucleaires. (auteur)

  13. 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)

  14. 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)

  15. 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).

  16. Use of one delayed-neutron precursor group in transient analysis

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1983-01-01

    In most reactor dynamics calculations six groups of delayed-neutron precursors are usually accounted for. However, under certain circumstances it may be advantageous to simplify the calculation and utilize a single delayed-neutron group. The motivation for going to one precursor group is economy. For LWR transient codes that use point kinetics the equations are solved very rapidly and six precursor groups should always be used. However, codes with spatially dependent neutron kinetics are very long running and the use of one precursor group may save computer costs and not impair the accuracy of the results significantly. Furthermore, in some codes, the elimation of five presursor groups makes additional memory available which may be used to give a net increase in the accuracy of the calculations, e.g., by allowing for an increase in mesh density. In order to use one delayed neutron precursor group it is necessary to derive a single decay constant, 6 lambda-, which, along with the total (or one group) delayed neutron fraction β = Σ/sub i = 1/β/sub i/, will adequately describe the transeint precursor behavior. The present summary explains how a recommendation for lambda- was derived

  17. Two level calculation of assembly neutronic data libraries; Schema de calcul de bibliotheques a deux niveaux

    Energy Technology Data Exchange (ETDEWEB)

    Benomar, M

    1998-09-01

    The neutronic modeling of a nuclear reactor core requires 2 steps. The first step that is called transport calculation, is an accurate modeling of each type of assemblies put in a simple configuration. APOLLO2, a French neutronic code is used. This step allows the constitution of assembly data libraries. The second step represents the computing of the whole core by the diffusion theory and by using the data libraries defined in the first step. This work is dedicated to the improvement of the first step by allowing both a 172 group energy meshing and a two-dimension spatial processing. (A.C.) 7 refs.

  18. Analysis of neutron leakage effect in the determination of macrogroup constants

    International Nuclear Information System (INIS)

    Martinez, A.S.; Vieira, H.D.

    1986-01-01

    A method to include the neutron leakage in the macrogroup constants calculation is presented. The method leads to independent equations for neutron flux and neutron current density. The results that have been gotten with the present method are very precise despite its simplicity. (Author) [pt

  19. Research on amplification multiple of source neutron number for ADS

    International Nuclear Information System (INIS)

    Liu Guisheng; Zhao Zhixiang; Zhang Baocheng; Shen Qingbiao; Ding Dazhao

    1998-01-01

    NJOY-91.91 and MILER code systems was applied to process and generate 44 group cross sections in AMPX master library format from CENDL-2 and ENDF/B-6. It is important an ADS (Accelerator-Driven System) assembly spectrum is used as the weighting spectrum for generating multi-group constants. Amplification multiples of source neutron number for several fast assemblies were calculated

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

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description: These cross-section libraries are released by the Diagnostics Applications Group, X-5, at Los Alamos National Laboratory for use with the MCNP Monte Carlo code package. This release includes all of the X-5 distributed neutron data libraries, the photon libraries MCPLIB1 and MCPLIB02, the electron libraries EL1 and EL03, an updated XSDIR file, and information files Readme.txt and Readme e ndf60.txt. This release is intended to completely replace previous RSICC releases DLC-105, DLC-181, and DLC-189 as well as the cross sections previously included with CCC-200/MCNP4A, and will be updated as new libraries become available. The README file provides information regarding each data library of this release. Additional documentation for some of the individual libraries and example SPECS files for use with MAKXSF are also provided. The XSDIR file is specific to this release and may not work with previous packages. Currently the neutron data library ENDF60 (based on ENDF/B-VI, up through and including release 2) is the default library for continuous-energy neutron transport. Additionally, the libraries MCPLIB02 and EL03 are the default libraries for photon and electron transport respectively. More information on the data libraries contained in this release is available in Appendix G of the MCNP4C manual. 2 - Description of program or function: ZZ-MCB-DLC200 contains the same cross section tables as the DLC-0200/03 package for the MCNP-4C code, except that the installation procedures are adapted to the MCB1C code system (NEA 1643/01). 3 - Application of the data: DLC-200/MCNPDATA is for use with Version 4C and later of the MCNP transport code. This data library provides a comprehensive set of cross sections for a wide range of radiation transport applications using the Monte Carlo code package CCC-700/MCNP4C. See Appendix G of the MCNP report LA-13709-M for information on the libraries and how to select specific nuclides for use in MCNP. 4 - Source and scope

  1. Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries

    International Nuclear Information System (INIS)

    Uddin, M.N.; Sarker, M.M.; Khan, M.J.H.; Islam, S.M.A.

    2010-01-01

    The aim of this study is to analyze the neutronic parameters of TRIGA Mark-II research reactor using the chain of NJOY-WIMS-CITATION computer codes based on evaluated nuclear data libraries CENDL-2.2 and JEFF-3.1.1. The nuclear data processing code NJOY99.0 has been employed to generate the 69 group WIMS library for the isotopes of TRIGA core. The cell code WIMSD-5B was used to generate the cross sections in CITATION format and then 3-dimensional diffusion code CITTATION was used to calculate the neutronic parameters of the TRIGA Mark-II research reactor. All the analyses were performed using the 7-group macroscopic cross section library. The CITATION test-runs using different cross section sets based on different models applied in WIMS calculations have shown a strong influence of those models on the final integral parameters. Some of the cells were specially treated with PRIZE options available in WIMSD-5B to take into account the fine structure of the flux gradient in the fuel-reflector interface region. It was observed that two basic parameters, the effective multiplication factor, k eff and the thermal neutron flux, were in good agreement among the calculated results with each other as well as the measured values. The maximum power densities at the hot spot were 1.0446E02 W/cc and 1.0426E02 W/cc for the libraries CENDL-2.2 and JEFF-3.1.1 respectively. The calculated total peaking factors 5.793 and 5.745 were compared to the original SAR value of 5.6325 as well as MCNP result. Consequently, this analysis will be helpful to enhance the neutronic calculations and also be used for the further thermal-hydraulics study of the TRIGA core.

  2. DWARF, 1-D Few-Group Neutron Diffusion with Thermal Feedback for Burnup and Xe Oscillation

    International Nuclear Information System (INIS)

    Anderson, E.C.; Putnam, G.E.

    1975-01-01

    1 - Description of problem or function: DWARF allows one-dimensional simulation of reactor burnup and xenon oscillation problems in slab, cylindrical, or spherical geometry using a few-group diffusion theory model. 2 - Method of solution: The few-group, neutron diffusion theory equations are reduced to a system of finite-difference equations that are solved for each group by the Gauss method at each time point. Fission neutron source iteration can be accelerated with Chebyshev extrapolation. A thermal feedback iterative loop is used to obtain consistent solutions for the distributions of reactor power, neutron flux, and fuel and coolant properties with the neutron group constants functions of the latter. Solutions for the new nuclide concentrations of a time-point are made with the flux assumed constant in the time interval. 3 - Restrictions on the complexity of the problem - Maxima of: 4 groups; 40 regions; 50 macroscopic materials (Only 10 are functions of the feedback variables); 50 nuclides per region; 250 mesh points

  3. 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)

  4. 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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  6. 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

  7. NULIF: neutron spectrum generator, few-group constant calculator, and fuel depletion code

    International Nuclear Information System (INIS)

    Wittkopf, W.A.; Tilford, J.M.; Andrews, J.B. II; Kirschner, G.; Hassan, N.M.; Colpo, P.N.

    1977-02-01

    The NULIF code generates a microgroup neutron spectrum and calculates spectrum-weighted few-group parameters for use in a spatial diffusion code. A wide variety of fuel cells, non-fuel cells, and fuel lattices, typical of PWR (or BWR) lattices, are treated. A fuel depletion routine and change card capability allow a broad range of problems to be studied. Coefficient variation with fuel burnup, fuel temperature change, moderator temperature change, soluble boron concentration change, burnable poison variation, and control rod insertion are readily obtained. Heterogeneous effects, including resonance shielding and thermal flux depressions, are treated. Coefficients are obtained for one thermal group and up to three epithermal groups. A special output routine writes the few-group coefficient data in specified format on an output tape for automated fitting in the PDQ07-HARMONY system of spatial diffusion-depletion codes

  8. New nuclear data group constant sets for fusion reactor nuclear analyses based on JENDL-4.0 and FENDL-3.0

    International Nuclear Information System (INIS)

    Konno, Chikara; Ohta, Masayuki; Kwon, Saerom; Ochiai, Kentaro; Sato, Satoshi

    2015-01-01

    We have produced new nuclear data group constant sets from JENDL-4.0 and FENDL-3.0 for fusion reactor nuclear analyses; FUSION-J40-175, FUSION-F30-175 (40 materials, neutron 175 groups, gamma 42 groups), FUSION-J40-42 and FUSION-F30-42 (40 materials, neutron 42 groups, gamma 21 groups). MATXS files of JENDL-4.0 and FENDL-3.0 were newly produced with the NJOY2012 code. FUSION-J40-175, FUSION-J40-42, FUSION-F30-175 and FUSION-F30-42 were produced with the TRANSX code. KERMA factors, DPA and gas production cross-section data were also prepared from the MATXS files with TRANSX. Test calculations were carried out in order to validate these nuclear group constant sets. They suggested that these group constant sets had no problem. (author)

  9. New evaluation of thermal neutron scattering libraries for light and heavy water

    Directory of Open Access Journals (Sweden)

    Marquez Damian Jose Ignacio

    2017-01-01

    Full Text Available In order to improve the design and safety of thermal nuclear reactors and for verification of criticality safety conditions on systems with significant amount of fissile materials and water, it is necessary to perform high-precision neutron transport calculations and estimate uncertainties of the results. These calculations are based on neutron interaction data distributed in evaluated nuclear data libraries. To improve the evaluations of thermal scattering sub-libraries, we developed a set of thermal neutron scattering cross sections (scattering kernels for hydrogen bound in light water, and deuterium and oxygen bound in heavy water, in the ENDF-6 format from room temperature up to the critical temperatures of molecular liquids. The new evaluations were generated and processable with NJOY99 and also with NJOY-2012 with minor modifications (updates, and with the new version of NJOY-2016. The new TSL libraries are based on molecular dynamics simulations with GROMACS and recent experimental data, and result in an improvement of the calculation of single neutron scattering quantities. In this work, we discuss the importance of taking into account self-diffusion in liquids to accurately describe the neutron scattering at low neutron energies (quasi-elastic peak problem. To improve modeling of heavy water, it is important to take into account temperature-dependent static structure factors and apply Sköld approximation to the coherent inelastic components of the scattering matrix. The usage of the new set of scattering matrices and cross-sections improves the calculation of thermal critical systems moderated and/or reflected with light/heavy water obtained from the International Criticality Safety Benchmark Evaluation Project (ICSBEP handbook. For example, the use of the new thermal scattering library for heavy water, combined with the ROSFOND-2010 evaluation of the cross sections for deuterium, results in an improvement of the C/E ratio in 48 out of

  10. New evaluation of thermal neutron scattering libraries for light and heavy water

    Science.gov (United States)

    Marquez Damian, Jose Ignacio; Granada, Jose Rolando; Cantargi, Florencia; Roubtsov, Danila

    2017-09-01

    In order to improve the design and safety of thermal nuclear reactors and for verification of criticality safety conditions on systems with significant amount of fissile materials and water, it is necessary to perform high-precision neutron transport calculations and estimate uncertainties of the results. These calculations are based on neutron interaction data distributed in evaluated nuclear data libraries. To improve the evaluations of thermal scattering sub-libraries, we developed a set of thermal neutron scattering cross sections (scattering kernels) for hydrogen bound in light water, and deuterium and oxygen bound in heavy water, in the ENDF-6 format from room temperature up to the critical temperatures of molecular liquids. The new evaluations were generated and processable with NJOY99 and also with NJOY-2012 with minor modifications (updates), and with the new version of NJOY-2016. The new TSL libraries are based on molecular dynamics simulations with GROMACS and recent experimental data, and result in an improvement of the calculation of single neutron scattering quantities. In this work, we discuss the importance of taking into account self-diffusion in liquids to accurately describe the neutron scattering at low neutron energies (quasi-elastic peak problem). To improve modeling of heavy water, it is important to take into account temperature-dependent static structure factors and apply Sköld approximation to the coherent inelastic components of the scattering matrix. The usage of the new set of scattering matrices and cross-sections improves the calculation of thermal critical systems moderated and/or reflected with light/heavy water obtained from the International Criticality Safety Benchmark Evaluation Project (ICSBEP) handbook. For example, the use of the new thermal scattering library for heavy water, combined with the ROSFOND-2010 evaluation of the cross sections for deuterium, results in an improvement of the C/E ratio in 48 out of 65

  11. Preparation of small group constants for calculation of shielding

    International Nuclear Information System (INIS)

    Khokhlov, V.F.; Shejno, I.N.; Tkachev, V.D.

    1979-01-01

    Studied is the effect of the shielding calculation error connected with neglect of the angular and spatial neutron flux dependences while determining the small-group constants on the basis of the many-group ones. The economical method allowing for dependences is proposed. The spatial dependence is substituted by the average value according to the zones singled out in the limits of the zones of the same content; the angular cross section dependence is substituted by the average values in the half-ranges of the angular variable. To solve the transfer equation the ALGOL-ROSA-M program using the method of characteristic interpolation and trial run method is developed. The program regards correctly for nonscattered and single scattered radiations. Compared are the calculation results of neutron transmission (10.5 MeV-0.01 eV) in the 21-group approximation with the 3-group calculations for water (the layer thickness is 30 cm) and 5-group calculations for heterogeneous shielding of alternating stainless steel layers (3 layers, each of the 16 cm thickness) and graphite layers (2 layers, each of the 20 cm thickness). The analysis shows that the method proposed permits to obtain rather accurate results in the course of preparation of the small-group cross sections, decreasing considerably the number of the groups (from 21 to 3-5) and saving the machine time

  12. Integrated system for production of neutronics and photonics calculational constants. Neutron-induced interactions: index of experimental data

    International Nuclear Information System (INIS)

    MacGregor, M.H.; Cullen, D.E.; Howerton, R.J.; Perkins, S.T.

    1976-01-01

    Indexes to the neutron-induced interaction data in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976 are tabulated. The tabulation has two arrangements: isotope (ZA) order and reaction-number order

  13. Integrated system for production of neutronics and photonics calculational constants. Neutron-induced interactions: index of experimental data

    Energy Technology Data Exchange (ETDEWEB)

    MacGregor, M.H.; Cullen, D.E.; Howerton, R.J.; Perkins, S.T.

    1976-07-04

    Indexes to the neutron-induced interaction data in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976 are tabulated. The tabulation has two arrangements: isotope (ZA) order and reaction-number order.

  14. COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    Energy Technology Data Exchange (ETDEWEB)

    Woo Y. Yoon; David W. Nigg

    2008-09-01

    COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete

  15. COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    International Nuclear Information System (INIS)

    Yoon, Woo Y.; Nigg, David W.

    2008-01-01

    COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete

  16. 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)

  17. New WIMS library generation from ENDF/B6 and effect of resonance group structure on cell parameters

    International Nuclear Information System (INIS)

    Pazirandeh, Ali; Tabesh, Alireza

    2002-01-01

    Due to inaccessibility to NJOY, steps were taken to create WIMS library, which can be extracted from ENDF/B6 without using NJOY. In addition to using preprocessing codes few programs were written to calculate integral resonance, slowing down power per unit lethargy, potential scattering, and differential scattering cross section, scattering matrices. For neutrons with energy above 4 eV, isotropic elastic scattering was assumed. For neutrons below 4 eV the free gas model was used, except for light elements, which tabulated values of S(α,β) in ENDF/B6 used. The Goldstein-Cohen factors are taken from WIMKAL88.Lib. The integral resonance with self absorption per unit lethargy was obtained from GROUPIE output. The P 1 scattering matrices are calculated only for four elements, namely H, D, C and O at 300 K. In order to examine the created libraries, k eff , δ 28 , ρ 28 , ρ 25 and CR are calculated using new WIMS library, WIMKAL88.Lib and NEA329.Lib. The results showed general agreement. The controversial issue of WIMS library group structure, particularly in resonance region has raised the question of whether the number of resonance group i.e., 13 is optimized. We generated different WIMS libraries consisting of 5, 8, 13, 18 and 23 resonance groups. The main aim was to examine the effect to resonance group structure on calculated core parameters, mainly, k eff , δ 28 , ρ 28 , ρ 25 and CR. These parameters are also calculated and compared with those obtained using WIMKAL88, and NEA329 libraries. (author)

  18. The LAW library

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  19. New Neutron, Proton, and S(α,β) MCNP Data Libraries Based on ENDF/B-VII

    International Nuclear Information System (INIS)

    Little, Robert C.; Trellue, Holly R.; MacFarlane, Robert E.; Kahler, A.C.; Lee, Mary Beth; White, Morgan C.

    2008-01-01

    The general-purpose Evaluated Nuclear Data File ENDF/B-VII.0 was released in December 2006. A number of sub-libraries were included in ENDF/B-VII.0 such that data were provided for incident neutrons, photons, and charged particles. This paper describes the creation of MCNP data libraries at Los Alamos National Laboratory based on three ENDF/B-VII.0 sub-libraries: neutrons, protons, and thermal scattering. An ACE-formatted continuous-energy neutron data library called ENDF70 for MCNP has been produced. This library provides data for 390 materials at five temperatures: 293.6, 600, 900, 1200, and 2500 K. The library was processed primarily with Version 248 of NJOY99. Extensive checking and quality-assurance tests were applied to the data. Improvements to the processing code were made and certain evaluations were modified as a result of these tests. ENDF/B-VII.0 included proton evaluations for 48 target materials. Forty-seven proton evaluations (all except for 13 C) were processed at room temperature and combined into the MCNP library ENDF70PROT. Neutron thermal S(α,β) scattering data exist for twenty different materials in ENDF/B-VII.0. All twenty of these evaluations were processed at all applicable temperatures (these vary for each evaluation), and combined into the MCNP library ENDF70SAB. All of these ENDF/B-VII.0 based MCNP libraries (ENDF70, ENDF70PROT, and ENDF70SAB) are available as part of the MCNP5 1.50 release. (authors)

  20. 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.

  1. 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

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  3. On the use of SERPENT Monte Carlo code to generate few group diffusion constants

    Energy Technology Data Exchange (ETDEWEB)

    Piovezan, Pamela, E-mail: pamela.piovezan@ctmsp.mar.mil.b [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Carluccio, Thiago; Domingos, Douglas Borges; Rossi, Pedro Russo; Mura, Luiz Felipe, E-mail: fermium@cietec.org.b, E-mail: thiagoc@ipen.b [Fermium Tecnologia Nuclear, Sao Paulo, SP (Brazil); Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The accuracy of diffusion reactor codes strongly depends on the quality of the groups constants processing. For many years, the generation of such constants was based on 1-D infinity cell transport calculations. Some developments using collision probability or the method of characteristics allow, nowadays, 2-D assembly group constants calculations. However, these 1-D and 2-D codes how some limitations as , for example, on complex geometries and in the neighborhood of heavy absorbers. On the other hand, since Monte Carlos (MC) codes provide accurate neutro flux distributions, the possibility of using these solutions to provide group constants to full-core reactor diffusion simulators has been recently investigated, especially for the cases in which the geometry and reactor types are beyond the capability of the conventional deterministic lattice codes. The two greatest difficulties on the use of MC codes to group constant generation are the computational costs and the methodological incompatibility between analog MC particle transport simulation and deterministic transport methods based in several approximations. The SERPENT code is a 3-D continuous energy MC transport code with built-in burnup capability that was specially optimized to generate these group constants. In this work, we present the preliminary results of using the SERPENT MC code to generate 3-D two-group diffusion constants for a PWR like assembly. These constants were used in the CITATION diffusion code to investigate the effects of the MC group constants determination on the neutron multiplication factor diffusion estimate. (author)

  4. 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

  5. Validation of neutron data libraries by backscattered spectra of Pu-Be Neutrons

    CERN Document Server

    El-Agib, I

    1999-01-01

    Elastically backscattered spectra of Pu-Be neutrons have been measured for SiO sub 2 , water, graphite, paraffin oil and Al slabs using a proton recoil spectrometer. The results were compared with the calculated spectra obtained by the three-dimensional Monte-Carlo transport code MCNP-4B and point-wise cross sections from the ENDF/B-V, ENDF/B-VI, JENDL-3.1 and BROND-2 data libraries. The good agreement between the measured and calculated results indicates that this procedure can be used for validation of different data libraries. This simple method renders possible the detection of oxygen, carbon and hydrogen in bulk samples. (author)

  6. An accurate solution of point reactor neutron kinetics equations of multi-group of delayed neutrons

    International Nuclear Information System (INIS)

    Yamoah, S.; Akaho, E.H.K.; Nyarko, B.J.B.

    2013-01-01

    Highlights: ► Analytical solution is proposed to solve the point reactor kinetics equations (PRKE). ► The method is based on formulating a coefficient matrix of the PRKE. ► The method was applied to solve the PRKE for six groups of delayed neutrons. ► Results shows good agreement with other traditional methods in literature. ► The method is accurate and efficient for solving the point reactor kinetics equations. - Abstract: The understanding of the time-dependent behaviour of the neutron population in a nuclear reactor in response to either a planned or unplanned change in the reactor conditions is of great importance to the safe and reliable operation of the reactor. In this study, an accurate analytical solution of point reactor kinetics equations with multi-group of delayed neutrons for specified reactivity changes is proposed to calculate the change in neutron density. The method is based on formulating a coefficient matrix of the homogenous differential equations of the point reactor kinetics equations and calculating the eigenvalues and the corresponding eigenvectors of the coefficient matrix. A small time interval is chosen within which reactivity relatively stays constant. The analytical method was applied to solve the point reactor kinetics equations with six-groups delayed neutrons for a representative thermal reactor. The problems of step, ramp and temperature feedback reactivities are computed and the results compared with other traditional methods. The comparison shows that the method presented in this study is accurate and efficient for solving the point reactor kinetics equations of multi-group of delayed neutrons

  7. 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)

  8. Determination of the fine-structure constant {alpha} by measuring the quotient of the Planck constant and the neutron mass

    Energy Technology Data Exchange (ETDEWEB)

    Krueger, E; Nistler, W; Weirauch, W [Physikalisch-Technische Bundesanstalt, Braunschweig (Germany)

    1997-04-01

    Using a special high-precision apparatus at ILL the quotient h/m{sub n} (h Planck constant, m{sub n} neutron mass) has been measured. The value measured for h/m{sub n} leads to {alpha}{sup -1} = 137.03601082(524) (relative uncertainty: 3.9{center_dot}10{sup -8}) It was the first time that this fundamental constant has been determined by means of neutrons. The experiment, which had been running since 1981 in a preliminary version and since 1987 in the final version, which was finished in December 1996, is described. (author).

  9. Development and verification of a 281-group WIMS-D library based on ENDF/B-VII.1

    International Nuclear Information System (INIS)

    Dong, Zhengyun; Wu, Jun; Ma, Xubo; Yu, Hui; Chen, Yixue

    2016-01-01

    Highlights: • A new WIMS-D library based on SHEM 281 energy structures is developed. • The method for calculating the lambda factor is illustrated and parameters are discussed. • The results show the improvements of this library compared with other libraries. - Abstract: The WIMS-D library based on WIMS 69 or XMAS 172 energy group structures is widely used in thermal reactor research. Otherwise, the resonance overlap effect is not taken into account in the two energy group structure, which limits the accuracy of resonance treatment. The SHEM 281 group structure is designed by the French to avoid the resonance overlap effect. In this study, a new WIMS-D library with SHEM 281 mesh is developed by using the NJOY nuclear data processing system based on the latest Evaluated Nuclear Data Library ENDF/B-VII.1. The parameters such as the thermal cut-off energy and lambda factor that depend on group structure are discussed. The lambda factor is calculated by Neutron Resonance Spectrum Calculation System and the effect of this factor is analyzed. The new library is verified through the analysis of various criticality benchmarks by using DRAGON code. The values of multiplication factor are consistent with the experiment data and the results also are improved in comparison with other WIMS libraries.

  10. 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)

  11. Selected articles translated from Jadernye Konstanty (Nuclear Constants) volume 1, 1996

    International Nuclear Information System (INIS)

    1997-08-01

    This report contains selected articles translated from Jedernye Konstanty (Nuclear Constants). Eight papers are included and each one is separately indexed. Nuclear data libraries, Neutron Reactions, Low energy Photofission etc. are dealt with. Refs, figs, tabs

  12. 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 ...

  13. Neutron-proton analyzing power at 12 MeV and charged πNN coupling constant

    International Nuclear Information System (INIS)

    Braun, R.T.; Tornow, W.; Gonzalez Trotter, D.E.; Howell, C.R.; Machleidt, R.; Roper, C.D.; Salinas, F.; Setze, H.R.; Walter, R.L.

    1995-01-01

    Recent reanalysis of scattering data by the Nijmegen group has led to new values for the πNN coupling constants, g 2 πdegree /4π and g 2 π± /4π, about 6% smaller than the previously accepted values. The impact of this finding is far reaching. Since the neutron-proton A y (θ) is dominated at low energies by the one-pion-exchange mechanism, accurate np data should provide unique information as to the magnitude of g 2 π± /4π. Using a new experimental setup consisting of a shielded neutron source, a five-pair neutron detector array, a n- 4 He polarimeter, and an intense polarized source with fast spin-flipping capability, we have measured a 15 point angular distribution of the neutron-proton A y (θ) at and incident neutron energy of 12 MeV to a statistical accuracy of 5x10 -4 . We will discuss the data taking procedures, the analysis, and the corrections applied to the data. Preliminary results will be presented

  14. Methodology of measurement of thermal neutron time decay constant in Canberra 35+ MCA system

    International Nuclear Information System (INIS)

    Drozdowicz, K.; Gabanska, B.; Igielski, A.; Krynicka, E.; Woznicka, U.

    1993-01-01

    A method of the thermal neutron time decay constant measurement in small bounded media is presented. A 14 MeV pulsed neutron generator is the neutron source. The system of recording of a die-away curve of thermal neutrons consists of a 3 He detector and of a multichannel time analyzer based on analyzer Canberra 35+ with multi scaler module MCS 7880 (microsecond range). Optimum parameters for the measuring system are considered. Experimental verification of a dead time of the instrumentation system is made and a count-loss correction is incorporated into the data treatment. An attention is paid to evaluate with a high accuracy the fundamental mode decay constant of the registered decaying curve. A new procedure of the determination of the decay constant by a multiple recording of the die-away curve is presented and results of test measurements are shown. (author). 11 refs, 12 figs, 4 tabs

  15. Methodology of measurement of thermal neutron time decay constant in Canberra 35+ MCA system

    Energy Technology Data Exchange (ETDEWEB)

    Drozdowicz, K; Gabanska, B; Igielski, A; Krynicka, E; Woznicka, U [Institute of Nuclear Physics, Cracow (Poland)

    1994-12-31

    A method of the thermal neutron time decay constant measurement in small bounded media is presented. A 14 MeV pulsed neutron generator is the neutron source. The system of recording of a die-away curve of thermal neutrons consists of a {sup 3}He detector and of a multichannel time analyzer based on analyzer Canberra 35+ with multi scaler module MCS 7880 (microsecond range). Optimum parameters for the measuring system are considered. Experimental verification of a dead time of the instrumentation system is made and a count-loss correction is incorporated into the data treatment. An attention is paid to evaluate with a high accuracy the fundamental mode decay constant of the registered decaying curve. A new procedure of the determination of the decay constant by a multiple recording of the die-away curve is presented and results of test measurements are shown. (author). 11 refs, 12 figs, 4 tabs.

  16. Methodology of measurement of thermal neutron time decay constant in Canberra 35+ MCA system

    Energy Technology Data Exchange (ETDEWEB)

    Drozdowicz, K.; Gabanska, B.; Igielski, A.; Krynicka, E.; Woznicka, U. [Institute of Nuclear Physics, Cracow (Poland)

    1993-12-31

    A method of the thermal neutron time decay constant measurement in small bounded media is presented. A 14 MeV pulsed neutron generator is the neutron source. The system of recording of a die-away curve of thermal neutrons consists of a {sup 3}He detector and of a multichannel time analyzer based on analyzer Canberra 35+ with multi scaler module MCS 7880 (microsecond range). Optimum parameters for the measuring system are considered. Experimental verification of a dead time of the instrumentation system is made and a count-loss correction is incorporated into the data treatment. An attention is paid to evaluate with a high accuracy the fundamental mode decay constant of the registered decaying curve. A new procedure of the determination of the decay constant by a multiple recording of the die-away curve is presented and results of test measurements are shown. (author). 11 refs, 12 figs, 4 tabs.

  17. 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)

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  19. COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    Energy Technology Data Exchange (ETDEWEB)

    Woo Y. Yoon; David W. Nigg

    2009-08-01

    COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete

  20. COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    International Nuclear Information System (INIS)

    Yoon, Woo Y.; Nigg, David W.

    2009-01-01

    COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete

  1. 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

  2. New ENDF/B-7.0 library

    International Nuclear Information System (INIS)

    Oblozinsky, P.

    2008-01-01

    We describe the new version of the Evaluated Nuclear Data File, Endf/B-7.0, of recommended nuclear data for advanced nuclear science and technology applications. The library, produced by the US Cross Section Evaluation Working Group, was released in December 2006. The library contains data in 14 sub-libraries, primarily for reactions with incident neutrons, protons and photons, based on the experimental data and nuclear reaction theory predictions. The neutron reaction sub-library contains data for 393 materials. The new library was extensively tested and shows considerable improvements over the earlier Endf/B-6.8 library. (author)

  3. Euratom Neutron Radiography Working Group

    DEFF Research Database (Denmark)

    Domanus, Joseph Czeslaw

    1986-01-01

    reactor fuel as well as establish standards for radiographic image quality of neutron radiographs. The NRWG meets once a year in each of the neutron radiography centers to review the progress made and draw plans for the future. Besides, ad-hoc sub-groups or. different topics within the field of neutron......In 1979 a Neutron Radiography Working Group (NRWG) was constituted within Buratom with the participation of all centers within the European Community at which neutron facilities were available. The main purpose of NRWG was to standardize methods and procedures used in neutron radiography of nuclear...... radiography are constituted. This paper reviews the activities and achievements of the NRWG and its sub-groups....

  4. Nuclear data libraries for Tripoli-3.5 code

    International Nuclear Information System (INIS)

    Vergnaud, Th.

    2001-01-01

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

  5. 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)

  6. Critical experiments analysis by ABBN-90 constant system

    Energy Technology Data Exchange (ETDEWEB)

    Tsiboulia, A.; Nikolaev, M.N.; Golubev, V. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)] [and others

    1997-06-01

    The ABBN-90 is a new version of the well-known Russian group-constant system ABBN. Included constants were calculated based on files of evaluated nuclear data from the BROND-2, ENDF/B-VI, and JENDL-3 libraries. The ABBN-90 is intended for the calculation of different types of nuclear reactors and radiation shielding. Calculations of criticality safety and reactivity accidents are also provided by using this constant set. Validation of the ABBN-90 set was made by using a computerized bank of evaluated critical experiments. This bank includes the results of experiments conducted in Russia and abroad of compact spherical assemblies with different reflectors, fast critical assemblies, and fuel/water-solution criticalities. This report presents the results of the calculational analysis of the whole collection of critical experiments. All calculations were produced with the ABBN-90 group-constant system. Revealed discrepancies between experimental and calculational results and their possible reasons are discussed. The codes and archives INDECS system is also described. This system includes three computerized banks: LEMEX, which consists of evaluated experiments and their calculational results; LSENS, which consists of sensitivity coefficients; and LUND, which consists of group-constant covariance matrices. The INDECS system permits us to estimate the accuracy of neutronics calculations. A discussion of the reliability of such estimations is finally presented. 16 figs.

  7. 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)

  8. Survey of group data libraries for use of the DYN3D program for WWER type reactors

    International Nuclear Information System (INIS)

    Mittag, S.

    1994-06-01

    So-called few-group neutron data have to be used as input data in core models (such as DYN3D) calculating the reactor behaviour. A survey is given of qualified data libraries for the reactor cores of Russian VVER. The information about primary data used in group data generation and the accuracy reached by the cell codes is compiled in tables. To assess the quality of the data, comparisons have been made between measured and calculated reactor parameters. The information available does not show significant differences concerning the quality of the data libraries. (orig.) [de

  9. Calculation of anisotropic few-group constants in asymptotic cells: the code ANICELL

    International Nuclear Information System (INIS)

    Devenyi, A.

    1985-10-01

    The theoretical background of the ANICELL computer program together with a user's manual is presented. ANICELL is a nuclear reactor neutron transport code which solves the traditional asymptotic and the so-called tilted flux transport problems in one-dimensional cylindrical geometry using linearly anisotropic scattering. The method of solution used is the first flight collision probability technique. Few-group constants including radial and axial diffusion coefficients for the cell are also prepared by the program. (author)

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

    International Nuclear Information System (INIS)

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

    1994-08-01

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

  11. Delayed neutron fraction and prompt decay constant measurement in the MINERVE reactor using the PSI instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Perret, Gregory [Paul Scherrer Institute, Villigen, 5232, (Switzerland)

    2015-07-01

    The critical decay constant (B/A), delayed neutron fraction (B) and generation time (A) of the Minerve reactor were measured by the Paul Scherrer Institut (PSI) and the Commissariat a l'Energie Atomique (CEA) in September 2014 using the Feynman-alpha and Power Spectral Density neutron noise measurement techniques. Three slightly subcritical configuration were measured using two 1-g {sup 235}U fission chambers. This paper reports on the results obtained by PSI in the near critical configuration (-2g). The most reliable and precise results were obtained with the Cross-Power Spectral Density technique: B = 708.4±9.2 pcm, B/A = 79.0±0.6 s{sup -1} and A 89.7±1.4 micros. Predictions of the same kinetic parameters were obtained with MCNP5-v1.6 and the JEFF-3.1 and ENDF/B-VII.1 nuclear data libraries. On average the predictions for B and B/A overestimate the experimental results by 5% and 11%, respectively. The discrepancy is suspected to come from either a corruption of the data or from the inadequacy of the point kinetic equations to interpret the measurements in the Minerve driven system. (authors)

  12. BROND-2.1. Russian Evaluated Neutron Reaction Data Library

    International Nuclear Information System (INIS)

    Manokhin, V.N.

    1993-01-01

    BROND-2, the computerized data library for evaluated neutron reaction data of the Russian Federation was released in 1992 and updated as BROND-2.1. Its content is summarized in this document. Upon request it is available on magnetic tape, cost free, from the IAEA Nuclear Data Section. It is also available within the IAEA online system NDIS. (author)

  13. Energy dependence of relative abundances and periods of separate groups of delayed neutrons at neutron induced fission of 239Pu in a range of neutrons energies 0.37 - 5 MeV

    International Nuclear Information System (INIS)

    Roschenko, V.A.; Piksaikin, V.M.; Kazakov, L.E.; Isaev, S.G.; Korolev, G.G.; Tarasko, M.Z.; Tertychnyi, R.G.

    2001-01-01

    The fundamental role of delayed neutrons in behavior, control and safety of reactors is well known today. Delayed neutron data are of great interest not only for reactor physics but also for nuclear fission physics and astrophysics. The purpose of the present work was the measurement of energy dependence of delayed neutrons (DN) group parameters at fission of nuclei 239 Pu in a range of energies of primary neutrons from 0.37 up to 5 MeV. The measurements were executed on installation designed on the basis of the electrostatic accelerator of KG - 2.5 SSC RF IPPE. The data are obtained in 6-group representation. It is shown, that there is a significant energy dependence of DN group parameters in a range of primary neutrons energies from thermal meanings up to 5 MeV, which is expressed in reduction of the average half-life of nuclei of the DN precursors on 10 %. The data, received in the present work, can be used at creation of a set of group constants for reactors with an intermediate spectrum of neutrons. (authors)

  14. Generation of Matxs-formated nuclear data libraries

    International Nuclear Information System (INIS)

    Vontobel, P.

    1989-01-01

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

  15. 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)

  16. 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...

  17. The method to set up file-6 in neutron data library of light nuclei below 20 MeV

    International Nuclear Information System (INIS)

    Zhang Jingshang; Han Yinlu

    2001-01-01

    So far there is no file-6 (double differential cross section data, DDX) of the light nuclei in the main evaluated neutron nuclear data libraries in the world. Therefore, locating a proper description on the double differential cross section of all kinds of outgoing particles from neutron induced light nucleus reaction below 20 MeV is necessary. The motivation for this work is to introduce a way to set up file-6 in the neutron data library

  18. A Monte Carlo Library Least Square approach in the Neutron Inelastic-scattering and Thermal-capture Analysis (NISTA) process in bulk coal samples

    Science.gov (United States)

    Reyhancan, Iskender Atilla; Ebrahimi, Alborz; Çolak, Üner; Erduran, M. Nizamettin; Angin, Nergis

    2017-01-01

    A new Monte-Carlo Library Least Square (MCLLS) approach for treating non-linear radiation analysis problem in Neutron Inelastic-scattering and Thermal-capture Analysis (NISTA) was developed. 14 MeV neutrons were produced by a neutron generator via the 3H (2H , n) 4He reaction. The prompt gamma ray spectra from bulk samples of seven different materials were measured by a Bismuth Germanate (BGO) gamma detection system. Polyethylene was used as neutron moderator along with iron and lead as neutron and gamma ray shielding, respectively. The gamma detection system was equipped with a list mode data acquisition system which streams spectroscopy data directly to the computer, event-by-event. A GEANT4 simulation toolkit was used for generating the single-element libraries of all the elements of interest. These libraries were then used in a Linear Library Least Square (LLLS) approach with an unknown experimental sample spectrum to fit it with the calculated elemental libraries. GEANT4 simulation results were also used for the selection of the neutron shielding material.

  19. Calculation of 14 MeV neutron transmission

    International Nuclear Information System (INIS)

    Vyrskij, M.Yu.; Dubinin, A.A.; Zhuravlev, V.I.; Isaev, N.V.; Klintsov, A.A.; Krivtsov, A.S.; Linge, I.I.; Panfilov, E.I.; Prit'mov, A.P.

    1979-01-01

    The possibility of using the 28 group constant system (28-GCS) for calculating the transport of neutrons with initial energy of 14 MeV in thermonuclear reactor blankets is studied. A blanket project suggested by the Oak Ridge National Laboratory is used as a test version to estimate applicability of the 28-GCS. Niobium is used in a blanket as a structural material. A mixture of lithium nuclides is used for tritium production. The results of blanket test calculation and the calculational results obtained using the 28-GCS from the UKNDL library are compared. The numerical 28-group calculation of blonket is carried out by means of the ROZ-6 and ROZ-9 codes but not by the Monte-Carlo method as compared with the test calculation. Time of the blanket calculation on the BESM-6 computer by means of the ROZ-9 code in 2P 5 approximation using the 28-GCS amounts to 10 min. It is noted that to create effective codes for the numerical blanket calculation different calculational grids are necessary for different energy grups. The calculations carried out have shown the possibility of using the 28-group library of cross sections for the numerical solution of the neutron transport equation in estimating analysis of blankets

  20. Prompt neutron decay constant estimation of RSG-GAS at high power noise experiment

    International Nuclear Information System (INIS)

    Jujuratisbela, U.; Kristedjo; Tukiran; Pinem, S.; Iman, J.; Puryono; Sanjaya, A.; Suwarno

    1998-01-01

    The determination of prompt neutron decay constant (α) of RGS-GAS by using low power noise experiment method at the equilibrium core indicated that the result is not good. The bad result was due to the small ratio of the noise signal to background which was caused by low detector efficiency or contaminated core after long time operation. To solve the problem is tried by using noise experiment technique at high power. The voltage output of neutron detectors at power of 5, 12, and 23 MW were connected to preamplifier and filter then to the Dynamic Signal Analyzer Version-2 and then the power spectral density of each channel of JKT04 and JKT03, the cut off frequency of each channel can be determined by using linear regression technique such that the prompt neutron decay constant can be estimated

  1. 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

  2. Prompt neutron decay constants and subcritical measurements for material control and accountability in SHEBA

    International Nuclear Information System (INIS)

    Sanchez, R.; Jaegers, P.

    1998-01-01

    Rossi-Alpha measurements were performed on the SHEBA assembly to determine the prompt neutron decay constants. These prompt neutron decay constants represent an eigenvalue characteristic of this particular assembly, which can be used to infer the amount of fissile material in the assembly. In addition, subcritical measurements using Rossi-Alpha and the source-jerk techniques were also performed on the SHEBA assembly. These measurements were compared against TWODANT calculations and agreed quite well. The subcritical measurements were also used to obtain a unique signature that represented the amount of material associated with the degree of subcriticality of the SHEBA assembly. Finally, the Feynman variance-to-mean technique in conjunction with TWODANT, were used to determine the effective delayed neutron fraction for the SHEBA assembly

  3. Neutron spectra measurement and calculations using data libraries CIELO, JEFF-3.2 and ENDF/B-VII.1 in iron benchmark assemblies

    Science.gov (United States)

    Jansky, Bohumil; Rejchrt, Jiri; Novak, Evzen; Losa, Evzen; Blokhin, Anatoly I.; Mitenkova, Elena

    2017-09-01

    The leakage neutron spectra measurements have been done on benchmark spherical assemblies - iron spheres with diameter of 20, 30, 50 and 100 cm. The Cf-252 neutron source was placed into the centre of iron sphere. The proton recoil method was used for neutron spectra measurement using spherical hydrogen proportional counters with diameter of 4 cm and with pressure of 400 and 1000 kPa. The neutron energy range of spectrometer is from 0.1 to 1.3 MeV. This energy interval represents about 85 % of all leakage neutrons from Fe sphere of diameter 50 cm and about of 74% for Fe sphere of diameter 100 cm. The adequate MCNP neutron spectra calculations based on data libraries CIELO, JEFF-3.2 and ENDF/B-VII.1 were done. Two calculations were done with CIELO library. The first one used data for all Fe-isotopes from CIELO and the second one (CIELO-56) used only Fe-56 data from CIELO and data for other Fe isotopes were from ENDF/B-VII.1. The energy structure used for calculations and measurements was 40 gpd (groups per decade) and 200 gpd. Structure 200 gpd represents lethargy step about of 1%. This relatively fine energy structure enables to analyze the Fe resonance neutron energy structure. The evaluated cross section data of Fe were validated on comparisons between the calculated and experimental spectra.

  4. 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)

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

    International Nuclear Information System (INIS)

    2005-01-01

    Description: FENDL: Fusion Evaluated Nuclear Data Library. Materials/nuclides: H 1 , H 2 , H 3 , He 3 , He 4 , Li 6 , Li 7 , Be 9 , B 10 , B 11 , C 12 , N 14 , N 15 , O 16 , F 19 , Na 23 , Mg nat , Al 27 , Si 28 , Si 29 , Si 30 , P 31 , S nat , Cl 35 , Cl 37 , K nat , Ca nat , Ti 46 , Ti 47 , Ti 48 , Ti 49 , Ti 50 , V nat , Cr 50 , Cr 52 , Cr 53 , Cr 54 , Mn 55 , Fe 54 , Fe 56 , Fe 57 , Fe 58 , Co 59 , Ni 58 , Ni 60 , Ni 61 , Ni 62 , Ni 64 , Cu 63 , Cu 65 , Ga nat , Zr nat , Nb 93 , Mo 92 , Mo 94 , Mo 95 , Mo 96 , Mo 97 , Mo 98 , Mo 100 , Sn nat , Ta 181 , W 182 , W 183 , W 184 , W 186 , Au 197 , Pb 206 , Pb 207 , Pb 208 , Bi 209 . Photo-atomic data. IAEA1364/02: FENDL version 2.0 consists of the following sub-libraries: - ACTIVATION (FENDL/A-2.0)- neutron activation cross sections for 13006 reactions on 739 targets ranging from 1-H up to 248-Cm at incident energies up to 20 MeV. Pointwise and processed data in different formats are included. Plots are available. - DECAY (FENDL/D-2.0) - decay properties (decay type, decay energy, and half life) for 1867 nuclides and isomers. FENDL/D-2.0 sub-library is complementary to the activation sub-library. Pointwise and processed data are included. - DOSIMETRY (FENDL/DS-2.0) - neutron cross sections to be used for reactor neutron dosimetry by foil activation, radiation damage cross-sections, and benchmark neutron spectra. This sub-library is identical to the International Reactor Dosimetry File (IRDF-90). Pointwise and processed data are included. - FUSION (FENDL/C-2.0) - charged-particle cross sections for the following fusion reactions: 1-H 2 (d,n)2-He 3 , 1-H 2 (d,p)1-H 3 , 2-He 3 (d,p)2-He 4 , 1-H-3(t,2n)2-He 4 , and 1-H 3 (d,n)2-He 4 . Pointwise and processed data are included. - TRANSPORT - validated basic nuclear data (neutron-nucleus interaction including photon production, and photon-atom interaction cross sections) for 57 nuclides relevant for fusion. In addition to the pointwise data (FENDL/E-2.0), the sub-library

  6. Prompt Neutron Decay Constant Determination Of Silicide Transition Core Using Noise Method

    International Nuclear Information System (INIS)

    Jujuratisbela, Uju; Yulianto, Yusi Eko; Cahyana

    2001-01-01

    Chairman of BATAN had decided to replace the Oxide fuel element type of RSG-GAS into silicide element type step by step. The replacement will create core transitions. Kinetic characteristic of the transition cores have to be monitored in order to know the deviation of core behavior. For that reason, the kinetic parameters have to be measured. Prompt neutron decay constant (alpha) is one of the kinetic parameters that has to be monitored continuously in the transition cores. In order not to disturb the normal operation of reactor, alpha parameter should be measured by using noise analysis method. The voltage of neutron flux at power of 15 MW is connected to preamplifier and filter then to the Dynamic Signal Analyzer Version-2 and then the auto power spectral density (APSD) was determined by using Fast Fourier transform. From the APSD curve of each channel of JKT03, the cut off frequency of each channel can be determined by using linear regression technique such that the prompt neutron decay constant can be estimated

  7. 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)

  8. Generation of a library of two-group diffusion parameters for SPPS-1,6 by HELIOS

    International Nuclear Information System (INIS)

    Petkov, P.T.; Haralampieva, C.V.; Simeonov, T.; Stojanova, I.; Kamenov, K.

    2000-01-01

    The two-group three-dimensional nodal diffusion code SPPS-1.6 has been used for many years for steady-state neutronics calculations of the WWER-440 reactors at Kozloduy NPP. The old library of two-group diffusion parameters for SPPS-1.6 has been generated by WIMSD4 with a nuclear data library compiled from three different libraries. The current paper presents our experience in generating a new library for SPPS-1.6 by the HELIOS lattice code. The accuracy of the current-coupling collision probability (CCCP) method in calculating a single WWER-440 assembly has been studied first. Among all possible angular discretization of the interface partial currents, called coupling orders, only coupling order 3 is suitable for hexagonal cells. Dividing each cell side into 3 segments an accuracy of 100 pcm has been achieved. The accuracy in calculating the absorber problem was estimated at 1%, which means about 10% error in the control assemblies efficiency. The accuracy for small core-reflector problems is 1% as well. The general conclusion is that HELIOS is accurate enough for assembly calculations, but inadequate for absorber and core-reflector problems. (Authors)

  9. Pulsed neutron determination of anisotropic diffusion constants in multi-layered slabs

    International Nuclear Information System (INIS)

    Sri Ram, K.

    1978-01-01

    Anisotropic neutron diffusion parameters for graphite and plexiglas slab assemblies were calculated using one-dimensional discrete ordinates code ANISN, and also Case's eigenfunction expansion technique as suggested by Leonard. These calculated values were checked with the pulsed neutron experimental results as well as simple diffusion theory calculations of Spinrad. Relatively little experimental work has been done with heterogeneous assemblies which do not contain voids. The present comparison shows that the experimental results agree well with transport theory calculations. It appears from the results and inter-comparison of this work in simple geometries, that the pulsed neutron method can yield accurate experimental anisotropic diffusion constants, and can therefore be applied to more complicated geometries which may be difficult to calculate. (author)

  10. An optimized ultra-fine energy group structure for neutron transport calculations

    International Nuclear Information System (INIS)

    Huria, Harish; Ouisloumen, Mohamed

    2008-01-01

    This paper describes an optimized energy group structure that was developed for neutron transport calculations in lattices using the Westinghouse lattice physics code PARAGON. The currently used 70-energy group structure results in significant discrepancies when the predictions are compared with those from the continuous energy Monte Carlo methods. The main source of the differences is the approximations employed in the resonance self-shielding methodology. This, in turn, leads to ambiguous adjustments in the resonance range cross-sections. The main goal of developing this group structure was to bypass the self-shielding methodology altogether thereby reducing the neutronic calculation errors. The proposed optimized energy mesh has 6064 points with 5877 points spanning the resonance range. The group boundaries in the resonance range were selected so that the micro group cross-sections matched reasonably well with those derived from reaction tallies of MCNP for a number of resonance absorbers of interest in reactor lattices. At the same time, however, the fast and thermal energy range boundaries were also adjusted to match the MCNP reaction rates in the relevant ranges. The resulting multi-group library was used to obtain eigenvalues for a wide variety of reactor lattice numerical benchmarks and also the Doppler reactivity defect benchmarks to establish its adequacy. (authors)

  11. 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)

  12. 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)

  13. Libraries of decay data and fission product yields in the ABBN-93 constant set

    International Nuclear Information System (INIS)

    Zabrodskaya, S.V.; Nikolaev, M.N.; Tsibulya, A.M.

    2001-01-01

    This paper describes three new libraries in the Abb. constant set which are essential for calculating radioactivity: basic decay data, radioactive decay photon spectra and fission product yields. (author)

  14. Resonance self-shielding effect analysis of neutron data libraries applied for the dual-cooled waste transmutation blanket of the fusion-driven subcritical system

    International Nuclear Information System (INIS)

    Liu Haibo; Wu Yican; Zheng Shanliang; Zhang Chunzao

    2004-01-01

    Based on the Fusion-Driven Subcritical System (FDS-I), the 25 groups, 175 groups and 620 groups neutron nuclear data libraries with/without resonance self-shielding correction are made with the Njoy and Transx codes, and the K eff and reaction rates are calculated with the Anisn code. The conclusion indicates that the resonance self-shielding effect affects the reaction rates strongly. (authors)

  15. 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

  16. 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.

  17. The management-retrieval code of the sub-library of atomic mass and characteristic constants for nuclear ground state

    International Nuclear Information System (INIS)

    Su Zongdi; Ma Lizhen

    1994-01-01

    The management code of the sub-library of atomic mass and characteristic constants for nuclear ground state (MCC) is used for displaying the basic information on the MCC sub-library on the screen, and retrieving the required data. The MCC data file contains the data of 4800 nuclides ranging from Z 0, A = 1 to Z = 122, A = 318. The MCC sub-library has been set up at Chinese Nuclear Data Center (CNDC), and has been used to provide the atomic masses and characteristic constants of nuclear ground states for the nuclear model calculation, nuclear data evaluations and other fields

  18. Group constant preparation for the estimate of neutron induced damage in structural materials

    International Nuclear Information System (INIS)

    Panini, G.C.

    1996-01-01

    Neutron heating (kerma), displacement per atom cross sections (DPA), gas and γ-ray production are important parameters for the estimate of the damage produced by neutron induced nuclear reactions in the structural materials. The NJOY System for Nuclear Data Processing has been extensively used in order to compute the above quantities; here the theory, the algorithms and the connected problems are described. (author). 6 refs, 3 tabs

  19. 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

  20. Sensitivity Analysis of Nuclide Importance to One-Group Neutron Cross Sections

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Nemoto, Atsushi; Yoshimura, Yoshikane

    2001-01-01

    The importance of nuclides is useful when investigating nuclide characteristics in a given neutron spectrum. However, it is derived using one-group microscopic cross sections, which may contain large errors or uncertainties. The sensitivity coefficient shows the effect of these errors or uncertainties on the importance.The equations for calculating sensitivity coefficients of importance to one-group nuclear constants are derived using the perturbation method. Numerical values are also evaluated for some important cases for fast and thermal reactor systems.Many characteristics of the sensitivity coefficients are derived from the derived equations and numerical results. The matrix of sensitivity coefficients seems diagonally dominant. However, it is not always satisfied in a detailed structure. The detailed structure of the matrix and the characteristics of coefficients are given.By using the obtained sensitivity coefficients, some demonstration calculations have been performed. The effects of error and uncertainty of nuclear data and of the change of one-group cross-section input caused by fuel design changes through the neutron spectrum are investigated. These calculations show that the sensitivity coefficient is useful when evaluating error or uncertainty of nuclide importance caused by the cross-section data error or uncertainty and when checking effectiveness of fuel cell or core design change for improving neutron economy

  1. INDL/V (85). IAEA Nuclear Data Library for various neutron data evaluations in ENDF-5 format

    International Nuclear Information System (INIS)

    Lemmel, H.D.; Goulo, V.; McLaughlin, K.; Pronyaev, V.; Schwerer, O.

    1985-06-01

    INDL/V is a computerized library for evaluated neutron reaction data from varying origin compiled in ENDF-5 format. The data are available costfree on magnetic tape from the IAEA Nuclear Data Section. This document summarizes the contents of the library in its version of March 1985. (author)

  2. 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)

  3. An efficient methodology of two groups spatial calculation for neutronic state and sensisivity coefficients in fast reactors

    International Nuclear Information System (INIS)

    Jachic, J.

    1985-01-01

    It is presented the ONEDM neutronic simulator for RZ spatial calculation, two energy groups, aiming at researching and optimization of a low power fast reactor design. The simulator's methodology is based in RZ calculation from radial and axial calculation iteractively coupled and in macroscopic cross sections corrected by power density and asymmetry of the spectrum in the feedback process with phase library for reference neutronic state. The transversal area which are determined by energy groups and material region in the iteration are introduced in the spatial calculation. The simulator efficiency is tested and compared with the CITATION and 2DB codes. The cross sections are generated by 1DX code. (M.C.K.) [pt

  4. 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

  5. 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.

  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

    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.

  7. INDL/A IAEA Nuclear Data Library for evaluated neutron reaction data of actinides

    International Nuclear Information System (INIS)

    Lemmel, H.D.

    1982-05-01

    This Library contains evaluations performed by participants of the IAEA Coordinated Research Project on the Intercomparison of Evaluations of Actinide Neutron Nuclear Data. The data are available on magnetic tape, free of charge, from the IAEA Nuclear Data Section. (author)

  8. 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

  9. 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

  10. INDL/V. IAEA Nuclear Data Library for various neutron data evaluations in ENDF/B-5 format

    International Nuclear Information System (INIS)

    Pronyaev, V.; Cullen, D.; Lemmel, H.D.; McLaughlin, K.; Schwerer, O.

    1982-05-01

    INDL/V is a computerized library for evaluated neutron reaction data of varying origin compiled in ENDF/B-5 format. The data are available costfree on magnetic tape from the IAEA Nuclear Data Section. This document summarizes the contents of the library, including graphical plots of all cross-section data. (author)

  11. 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

  12. Impact of nuclear library difference on neutronic characteristics of thorium-loaded light water reactor fuel

    International Nuclear Information System (INIS)

    Unesaki, H.; Isaka, S.; Nakagome, Y.

    2006-01-01

    Impact of nuclear library difference on neutronic characteristics of thorium-loaded light water reactor fuel is investigated through cell burnup calculations using SRAC code system. Comparison of k ∞ and nuclide composition was made between the results obtained by JENDL-3.3, ENDF/B-VI.8 and JEFF3.0 for (U, Th)O 2 fuels as well as UO 2 fuels, with special interest on the burnup dependence of the neutronic characteristics. The impact of nuclear data library difference on k ∞ of (U, Th)O 2 fuels was found to be significantly large compared to that of UO 2 fuels. Notable difference was also found in nuclide concentration of TRU nuclides. (authors)

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

    International Nuclear Information System (INIS)

    Tang Haibo; Chen Yixue; Wu Jun

    2013-01-01

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

  14. BROND-2.2. Russian evaluated neutron reaction data library. Summary documentation

    Energy Technology Data Exchange (ETDEWEB)

    Lemmel, H D; McLaughlin, P K

    1994-01-01

    BROND-2, the computerized data library for evaluated neutron reaction data of the Russian Federation was released in 1992 and updated in 1993 as BROND-2.2. Its content is summarized in this document. Upon request it is available on magnetic tape, costfree, from the IAEA Nuclear Data Section. It is also available within the IAEA online system NDIS. (author)

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  16. Standardization activities of the Euratom Neutron Radiography Working Group

    International Nuclear Information System (INIS)

    Domanus, J.

    1982-06-01

    In 1979 a working group on neutron radiography was formed at Euratom. The purpose of this group is the standardization of neutron radiographic methods in the field of nuclear fuel. Activities of this Neutron Radiography Working Group are revised. Classification of defects revealed by neutron radiography is illustrated in a special atlas. Beam purity and sensitivity indicators are tested together with a special calibration fuel pin. All the Euratom neutron radiography centers will perform comparative neutron radiography with those items. The measuring results obtained, using various measuring aparatus will form the basis to formulate conclusions about the best measuring methods and instruments to be used in that field. Besides the atlas of neutron radiographic findings in light water reactor fuel, the Euratom Neutron Radiogrphy Working Group has published a neutron radiography handbook in which the neutron radiography installations in the European Community are also described. (author)

  17. Generation of the V4.2m5 and AMPX and MPACT 51 and 252-Group Libraries with ENDF/B-VII.0 and VII.1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States). Consortium for Advanced Simulation of LWRs (CASL)

    2016-12-12

    The evaluated nuclear data file (ENDF)/B-7.0 v4.1m3 MPACT 47-group library has been used as a main library for the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronics simulator in simulating pressurized water reactor (PWR) problems. Recent analysis for the high void boiling water reactor (BWR) fuels and burnt fuels indicates that the 47-group library introduces relatively large reactivity bias. Since the 47- group structure does not match with the SCALE 6.2 252-group boundaries, the CASL Virtual Environment for Reactor Applications Core Simulator (VERA-CS) MPACT library must be maintained independently, which causes quality assurance concerns. In order to address this issue, a new 51-group structure has been proposed based on the MPACT 47- g and SCALE 252-g structures. In addition, the new CASL library will include a 19-group structure for gamma production and interaction cross section data based on the SCALE 19- group structure. New AMPX and MPACT 51-group libraries have been developed with the ENDF/B-7.0 and 7.1 evaluated nuclear data. The 19-group gamma data also have been generated for future use, but they are only available on the AMPX 51-g library. In addition, ENDF/B-7.0 and 7.1 MPACT 252-g libraries have been generated for verification purposes. Various benchmark calculations have been performed to verify and validate the newly developed libraries.

  18. Generation of the V4.2m5 and AMPX and MPACT 51 and 252-Group Libraries with ENDF/B-VII.0 and VII.1

    International Nuclear Information System (INIS)

    Kim, Kang Seog

    2016-01-01

    The evaluated nuclear data file (ENDF)/B-7.0 v4.1m3 MPACT 47-group library has been used as a main library for the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronics simulator in simulating pressurized water reactor (PWR) problems. Recent analysis for the high void boiling water reactor (BWR) fuels and burnt fuels indicates that the 47-group library introduces relatively large reactivity bias. Since the 47- group structure does not match with the SCALE 6.2 252-group boundaries, the CASL Virtual Environment for Reactor Applications Core Simulator (VERA-CS) MPACT library must be maintained independently, which causes quality assurance concerns. In order to address this issue, a new 51-group structure has been proposed based on the MPACT 47- g and SCALE 252-g structures. In addition, the new CASL library will include a 19-group structure for gamma production and interaction cross section data based on the SCALE 19- group structure. New AMPX and MPACT 51-group libraries have been developed with the ENDF/B-7.0 and 7.1 evaluated nuclear data. The 19-group gamma data also have been generated for future use, but they are only available on the AMPX 51-g library. In addition, ENDF/B-7.0 and 7.1 MPACT 252-g libraries have been generated for verification purposes. Various benchmark calculations have been performed to verify and validate the newly developed libraries.

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

    International Nuclear Information System (INIS)

    Alpan, F. Arzu; Haghighat, Alireza

    2008-01-01

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

  20. Reflector drums as control mechanism for craft thermionic reactors with constant emitter heating containing U-233 as fuel and beryllium as moderator

    International Nuclear Information System (INIS)

    Sahin, S.; Selvi, S.

    1980-01-01

    The suitability of borated reflector drums has been investigated and shown as a control mechanism for space craft thermionic reactors with constant emitter heating using U-233 as fuel and beryllium to be moderator, mainly due to their extremce compactness and their very soft neutron sepctrum. The achievable change in ksub(eff) allows long-term control operation with success. The use of reflector drums keeps the cone diameter and the mass of the radiation shield on minimum. The distortion of the emitter heating field remains under acceptable tolerances, mainly due to the enhanced neutron production at the outer core region and the remaining reflector part between the boron layer and the core. All neutron physics calculations have been carried out using the multigroup Ssub(N) methods. Three data groups for r-theta-calculations in S 4 -P 1 approximation (16 space angles) have been evaluated from a 123-energy-groups data library using transport theoretical methods. (orig.) [de

  1. BISERM version 2. Nuclear data library for evaluation of radiation effects in materials induced by neutrons of intermediate

    International Nuclear Information System (INIS)

    Korovin, Yu.A.; Stankovsky, A.Yu.; Konobeyev, A.Yu.; Pereslavtsev, P.E.

    1997-01-01

    This document describes the cross-section data library for studies of radiation effects induced by intermediate energy neutrons. The library contains hydrogen, helium-3 and helium-4 production cross-sections as well as nonelastic and total displacement cross-sections for neutrons in the energy range from threshold to 1 GeV. The cross-sections are given for 259 isotopes from 27 Al to 209 Bi. All this information is given in 18 compressed files plus one file with the catalogue of the data. The library was composed at the Institute of Nuclear Power Engineering in Obninsk, Russia. It is available on one PC diskette from the IAEA Nuclear Data Section costfree upon request. It requires 4 Mb of disk space after decompression. (author)

  2. The effects of nuclear data library processing on Geant4 and MCNP simulations of the thermal neutron scattering law

    Science.gov (United States)

    Hartling, K.; Ciungu, B.; Li, G.; Bentoumi, G.; Sur, B.

    2018-05-01

    Monte Carlo codes such as MCNP and Geant4 rely on a combination of physics models and evaluated nuclear data files (ENDF) to simulate the transport of neutrons through various materials and geometries. The grid representation used to represent the final-state scattering energies and angles associated with neutron scattering interactions can significantly affect the predictions of these codes. In particular, the default thermal scattering libraries used by MCNP6.1 and Geant4.10.3 do not accurately reproduce the ENDF/B-VII.1 model in simulations of the double-differential cross section for thermal neutrons interacting with hydrogen nuclei in a thin layer of water. However, agreement between model and simulation can be achieved within the statistical error by re-processing ENDF/B-VII.I thermal scattering libraries with the NJOY code. The structure of the thermal scattering libraries and sampling algorithms in MCNP and Geant4 are also reviewed.

  3. Generation of a Broad-Group HTGR Library for Use with SCALE

    International Nuclear Information System (INIS)

    Ellis, Ronald James; Lee, Deokjung; Wiarda, Dorothea; Williams, Mark L.; Mertyurek, Ugur

    2012-01-01

    With current and ongoing interest in high temperature gas reactors (HTGRs), the U.S. Nuclear Regulatory Commission (NRC) anticipates the need for nuclear data libraries appropriate for use in applications for modeling, assessing, and analyzing HTGR reactor physics and operating behavior. The objective of this work was to develop a broad-group library suitable for production analyses with SCALE for HTGR applications. Several interim libraries were generated from SCALE fine-group 238- and 999-group libraries, and the final broad-group library was created from Evaluated Nuclear Data File/B Version ENDF/B-VII Release 0 cross-section evaluations using new ORNL methodologies with AMPX, SCALE, and other codes. Furthermore, intermediate resonance (IR) methods were applied to the HTGR broadgroup library, and lambda factors and f-factors were incorporated into the library s nuclear data files. A new version of the SCALE BONAMI module named BONAMI-IR was developed to process the IR data in the new library and, thus, eliminate the need for the CENTRM/PMC modules for resonance selfshielding. This report documents the development of the HTGR broad-group nuclear data library and the results of test and benchmark calculations using the new library with SCALE. The 81-group library is shown to model HTGR cases with similar accuracy to the SCALE 238-group library but with significantly faster computational times due to the reduced number of energy groups and the use of BONAMI-IR instead of BONAMI/CENTRM/PMC for resonance self-shielding calculations.

  4. Constant-q data representation in Neutron Compton scattering on the VESUVIO spectrometer

    International Nuclear Information System (INIS)

    Senesi, R.; Pietropaolo, A.; Andreani, C.

    2008-01-01

    Standard data analysis on the VESUVIO spectrometer at ISIS is carried out within the Impulse Approximation framework, making use of the West scaling variable y. The experiments are performed using the time-of-flight technique with the detectors positioned at constant scattering angles. Line shape analysis is routinely performed in the y-scaling framework, using two different (and equivalent) approaches: (1) fitting the parameters of the recoil peaks directly to fixed-angle time-of-flight spectra; (2) transforming the time-of-flight spectra into fixed-angle y spectra, referred to as the Neutron Compton Profiles, and then fitting the line shape parameters. The present work shows that scattering signals from different fixed-angle detectors can be collected and rebinned to obtain Neutron Compton Profiles at constant wave vector transfer, q, allowing for a suitable interpretation of data in terms of the dynamical structure factor, S(q,ω). The current limits of applicability of such a procedure are discussed in terms of the available q-range and relative uncertainties for the VESUVIO experimental set up and of the main approximations involved

  5. Constant-q data representation in Neutron Compton scattering on the VESUVIO spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Senesi, R. [Dipartimento di Fisica, Universita degli Studi di Roma ' Tor Vergata' , Via della Ricerca Scientifica 1, 00133 Roma (Italy); Centro NAST, Nanoscienze and Nanotecnologie and Strumentazione, Universita degli Studi di Roma ' Tor Vergata' , Via della Ricerca Scientifica 1, 00133 Roma (Italy)], E-mail: roberto.senesi@roma2.infn.it; Pietropaolo, A.; Andreani, C. [Dipartimento di Fisica, Universita degli Studi di Roma ' Tor Vergata' , Via della Ricerca Scientifica 1, 00133 Roma (Italy); Centro NAST, Nanoscienze and Nanotecnologie and Strumentazione, Universita degli Studi di Roma ' Tor Vergata' , Via della Ricerca Scientifica 1, 00133 Roma (Italy)

    2008-09-01

    Standard data analysis on the VESUVIO spectrometer at ISIS is carried out within the Impulse Approximation framework, making use of the West scaling variable y. The experiments are performed using the time-of-flight technique with the detectors positioned at constant scattering angles. Line shape analysis is routinely performed in the y-scaling framework, using two different (and equivalent) approaches: (1) fitting the parameters of the recoil peaks directly to fixed-angle time-of-flight spectra; (2) transforming the time-of-flight spectra into fixed-angle y spectra, referred to as the Neutron Compton Profiles, and then fitting the line shape parameters. The present work shows that scattering signals from different fixed-angle detectors can be collected and rebinned to obtain Neutron Compton Profiles at constant wave vector transfer, q, allowing for a suitable interpretation of data in terms of the dynamical structure factor, S(q,{omega}). The current limits of applicability of such a procedure are discussed in terms of the available q-range and relative uncertainties for the VESUVIO experimental set up and of the main approximations involved.

  6. Constant- q data representation in Neutron Compton scattering on the VESUVIO spectrometer

    Science.gov (United States)

    Senesi, R.; Pietropaolo, A.; Andreani, C.

    2008-09-01

    Standard data analysis on the VESUVIO spectrometer at ISIS is carried out within the Impulse Approximation framework, making use of the West scaling variable y. The experiments are performed using the time-of-flight technique with the detectors positioned at constant scattering angles. Line shape analysis is routinely performed in the y-scaling framework, using two different (and equivalent) approaches: (1) fitting the parameters of the recoil peaks directly to fixed-angle time-of-flight spectra; (2) transforming the time-of-flight spectra into fixed-angle y spectra, referred to as the Neutron Compton Profiles, and then fitting the line shape parameters. The present work shows that scattering signals from different fixed-angle detectors can be collected and rebinned to obtain Neutron Compton Profiles at constant wave vector transfer, q, allowing for a suitable interpretation of data in terms of the dynamical structure factor, S(q,ω). The current limits of applicability of such a procedure are discussed in terms of the available q-range and relative uncertainties for the VESUVIO experimental set up and of the main approximations involved.

  7. 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)

  8. The JEFF-3.1 Nuclear Data Library - JEFF Report 21

    International Nuclear Information System (INIS)

    Koning, Arjan; Forrest, Robin; Kellett, Mark; Mills, Robert; Henriksson, Hans; Rugama, Yolanda; Bersillon, O.; Bouland, O.; Courcelle, A.; Duijvestijn, M.C.; Dupont, E.; Kopecky, J.; Leichtle, D.; Marie, F.; Mattes, M.; Menapace, E.; Morillon, B.; Mounier, C.; Noguerre, G.; Pereslavtsev, P.; Romain, P.; Serot, O.; Simakov, S.; Tagesen, S.; Vonach, H.; Batistoni, P.; Bem, P.; Gunsing, F.; Pillon, M.; Plompen, A.; Rullhusen, P.; Seidel, K.; Avrigeanu, M.; Avrigeanu, V.; Bauge, E.; Leeb, H.; Lopez Jimenez, M.J.; Bernard, D.; Bidaud, A.; Dagan, R.; Dean, C.; Dos-Santos-Uzarralde, P.; Fischer, U.; Hogenbirk, A.; Jacqmin, R.; Jouanne, C.; Kodeli, I.; Leppanen, J.; Marck, S.C. van der; Perel, R.; Perry, R.; Pescarini, M.; Santamarina, A.; Sublet, J.C.; Trkov, A.; Be, M.M.; Huynh, T.D.; Kellett, M.A.; Mills, R.; Nichols, A.; Henriksson, H.; Nordborg, C.; Nouri, A.; Rugama, Y.; Sartori, E.

    2006-01-01

    The Joint Evaluated Fission and Fusion (JEFF) Project is a collaborative effort among the member countries of the NEA Data Bank to develop a reference nuclear data library. The JEFF library contains sets of evaluated nuclear data, mainly for fission and fusion applications; it contains a number of different data types, including neutron and proton interaction data, radioactive decay data, fission yield data, thermal scattering law data and photo-atomic interaction data. The latest version of the JEFF library, JEFF-3.1, was released by the NEA in May 2005. JEFF-3.1 combines the efforts of the JEFF and EFF/EAF (European Fusion File/European Activation File) working groups who have contributed to this combined fission and fusion library. The neutron general purpose library contains incident neutron data for 381 materials from 1 H to 255 Fm. The activation library (based on the European Activation File, EAF-2003) contains 774 different targets from 1 H to 257 Fm. The radioactive decay data library contains data for 3 852 isotopes, of which 226 are stable. The proton special purpose library contains incident proton data for 26 materials from 40 Ca to 209 Bi. The thermal scattering law library covers 9 materials, and the fission yield library covers 19 isotopes of neutron induced fission yield from 232 Th to 245 Cm and 3 isotopes with spontaneous fission yields ( 242 Cm, 244 Cm and 252 Cf)

  9. Non-equilibrium thermodynamics and energy distribution function of neutron gas in constant power reactor under coupling of neutrons and medium

    International Nuclear Information System (INIS)

    Hayasaka, Hideo

    1983-01-01

    The thermodynamics and the energy distribution function of the neutron gas in a constant power reactor are considered, taking into account the burn-up of fuel. To separate the secular motion of neutrons owing to fuel burn-up and the microscopic fluctuations of neutrons around this motion, a long time of the order of several months is divided into m equal intervals, and the respective states corresponding to m small time intervals are treated as quasi-stationary states. The local energy distribution function of the neutron gas in the quasi-stationary state is given by a generalized Boltzmann distribution specified by the respective generalized activity coefficient for each subsystem. The effects of fuel burn-up on the respective distribution functions for successive small time intervals are taken into account through various quantities relating to reactor physics, depending upon the fuel burn-up, by successive approximation. (author)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    1976-07-01

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

  12. Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3

    International Nuclear Information System (INIS)

    Altaf, M.H.; Badrun, N.H.; Chowdhury, M.T.

    2015-01-01

    Highlights: • SRAC-PIJ code and SRAC-CITATION have been utilized to model the core. • Most of the simulated results show no significant differences with references. • Thermal peak flux varies a bit due to up condition of TRIGA. • ENDF/B-VII.0 and JENDL-3.3 libraries perform well for neutronics analysis of TRIGA. - Abstract: Important kinetic parameters such as effective multiplication factor, k eff , excess reactivity, neutron flux and power distribution, and power peaking factors of TRIGA Mark II research reactor in Bangladesh have been calculated using the comprehensive neutronics calculation code system SRAC 2006 with the evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3. In the code system, PIJ code was employed to obtain cross section of the core cells, followed by the integral calculation of neutronic parameters of the reactor conducted by CITATION code. All the analyses were performed using the 7-group macroscopic cross section library. Results were compared to the experimental data, the safety analysis report (SAR) of the reactor provided by General Atomic as well as to the simulated values by numerically benchmarked MCNP4C, WIMS-CITATION and SRAC-CITATION codes. The maximum power densities at the hot spot were found to be 169.7 W/cc and 170.1 W/cc for data libraries ENDF/B-VII.0 and JENDL-3.3, respectively. Similarly, the total peaking factors based on ENDF/B-VII.0 and JENDL-3.3 were calculated as 5.68 and 5.70, respectively, which were compared to the original SAR value of 5.63, as well as to MCNP4C, WIMS-CITATION and SRAC-CITATION results. It was found in most cases that the calculated results demonstrate a good agreement with our experiments and published works. Therefore, this analysis benchmarks the code system and will be helpful to enhance further neutronics and thermal hydraulics study of the reactor

  13. Effects of neutron data libraries and criticality codes on IAEA criticality benchmark problems

    International Nuclear Information System (INIS)

    Sarker, Md.M.; Takano, Makoto; Masukawa, Fumihiro; Naito, Yoshitaka

    1993-10-01

    In order to compare the effects of neutron data libraries and criticality codes to thermal reactors (LWR), the IAEA criticality benchmark calculations have been performed. The experiments selected in this study include TRX-1 and TRX-2 with a simple geometric configuration. Reactor lattice calculation codes WIMS-D/4, MCNP-4, JACS (MGCL, KENO), and SRAC were used in the present calculations. The TRX cores were analyzed by WIMS-D/4 using WIMS original library and also by MCNP-4, JACS (MGCL, KENO), and SRAC using the libraries generated from JENDL-3 and ENDF/B-IV nuclear data files. An intercomparison work for the above mentioned code systems and cross section libraries was performed by analyzing the LWR benchmark experiments TRX-1 and TRX-2. The TRX cores were also analyzed for supercritical and subcritical conditions and these results were compared. In the case of critical condition, the results were in good agreement. But for the supercritical and subcritical conditions, the difference of the results obtained by using the different cross section libraries become larger than for the critical condition. (author)

  14. Use of focus groups in a library's strategic planning process.

    Science.gov (United States)

    Higa-Moore, Mori Lou; Bunnett, Brian; Mayo, Helen G; Olney, Cynthia A

    2002-01-01

    The use of focus groups to determine patron satisfaction with library resources and services is extensive and well established. This article demonstrates how focus groups can also be used to help shape the future direction of a library as part of the strategic planning process. By responding to questions about their long-term library and information needs, focus group participants at the University of Texas Southwestern Medical Center at Dallas Library contributed an abundance of qualitative patron data that was previously lacking from this process. The selection and recruitment of these patrons is discussed along with the line of questioning used in the various focus group sessions. Of special interest is the way the authors utilized these sessions to mobilize and involve the staff in creating the library's strategic plan. This was accomplished not only by having staff members participate in one of the sessions but also by sharing the project's major findings with them and instructing them in how these findings related to the library's future. The authors' experience demonstrates that focus groups are an effective strategic planning tool for libraries and emphasizes the need to share information broadly, if active involvement of the staff is desired in both the development and implementation of the library's strategic plan.

  15. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Vergnaud, Th

    2001-07-01

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

  17. Validation of the 172 group ENDFB7GX library

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  18. MODICO, 1-D Time-Dependent 1 Group, 2 Group Neutron Diffusion with Delayed Neutron Precursors

    International Nuclear Information System (INIS)

    Camiciola, P.; Cundari, D.; Montagnini, B.

    1992-01-01

    1 - Description of program or function: The program solves the 1-D time-dependent one and two group coarse-mesh neutron diffusion equations, coupled with the equations for the delayed-neutron precursor, in plane geometry. 2 - Method of solution: The program is based on a simple coarse-mesh cubic approximation formula for the spatial behaviour of the flux inside each interval. An implicit scheme (the time-integrated method) is used for the advancement of the solution. The resulting (block three-diagonal) matrix is inverted at each time step by Thomas' method. 3 - Restrictions on the complexity of the problem: Number of coarse- mesh intervals LE 80; number of material regions LE 10; number of delayed-neutron precursor groups LE 10. Typical mesh sizes range from 5 cm to 20 cm; typical step length (non-prompt critical transients) ranges from 0.005 to 0.1 seconds

  19. Experiment and analysis of neutron spectra in a concrete assembly bombarded by 14 MeV neutrons

    International Nuclear Information System (INIS)

    Oishi, Koji; Tomioka, Kazuyuki; Ikeda, Yujiro; Nakamura, Tomoo.

    1988-01-01

    Neutron spectrum in concrete bombarded by 14 MeV neutrons was measured using a miniature NE213 spectrometer and multi-foil activation method. A good agreement between those two experimental methods was obtained within experimental errors. The measured spectrum was compared with calculated ones using two-dimensional transport code DOT3.5 with 125 group structure cross section libraries based on ENDF/B-IV, JENDL-2, and JENDL-3T (the testing version of JENDL-3.) In the D-T neutron peak region, measured and calculated neutron spectra agreed well with each other for those libraries. However, disagreements of about -10 % to +50 % and -30 % to +40 % were obtained in the MeV region and still lower neutron energy range, respectively. As a result, it was concluded that those discrepancies were caused by the overestimation of secondary neutrons emitted by inelastic scattering from O, Si, and/or Ca which were the main components of concrete. (author)

  20. 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

  1. Application of the decoupling scheme on complex neutron-gamma shielding problems

    Energy Technology Data Exchange (ETDEWEB)

    Feher, S. [Institute of Nuclear Technology, Technical University of Budapest, Budapest (Hungary); Leege, P.F.A. de; Hoogenboom, J.E.; Kloosterman, J.L. [Interfaculty Reactor Institute, Delft University of Technology, Delft (Netherlands)

    2000-03-01

    Coupled neutron-gamma shielding calculations using S{sub n} transport theory can be time consuming, especially for two- and three-dimensional geometries. In general, the CPU time of these calculations increases stronger than linear with increasing number of neutron and gamma energy groups, and depends on the order of Legendre expansion and number of S{sub n} directions used. This fact induced the idea of the decoupling method, which seems applicable to accelerate coupled neutron-gamma shielding calculations. The data included in a combined neutron-gamma library can be readily separated into a library containing neutron data only and another library containing gamma data only. Separate calculations for neutrons and gammas are performed on complex geometries using a different Legendre order expansion for neutrons and gammas. CPU savings of 60 to 85% can be achieved for the two-dimensional DORT and three-dimensional TORT calculations respectively. (author)

  2. Beam neutron energy optimization for boron neutron capture therapy using monte Carlo method

    International Nuclear Information System (INIS)

    Pazirandeh, A.; Shekarian, E.

    2006-01-01

    In last two decades the optimal neutron energy for the treatment of deep seated tumors in boron neutron capture therapy in view of neutron physics and chemical compounds of boron carrier has been under thorough study. Although neutron absorption cross section of boron is high (3836b), the treatment of deep seated tumors such as glioblastoma multiform requires beam of neutrons of higher energy that can penetrate deeply into the brain and thermalized in the proximity of the tumor. Dosage from recoil proton associated with fast neutrons however poses some constraints on maximum neutron energy that can be used in the treatment. For this reason neutrons in the epithermal energy range of 10eV-10keV are generally to be the most appropriate. The simulation carried out by Monte Carlo methods using MCBNCT and MCNP4C codes along with the cross section library in 290 groups extracted from ENDF/B6 main library. The ptimal neutron energy for deep seated tumors depends on the sue and depth of tumor. Our estimated optimized energy for the tumor of 5cm wide and 1-2cm thick stands at 5cm depth is in the range of 3-5keV

  3. The WIMS 69-group library tape 166259

    International Nuclear Information System (INIS)

    Taubman, C.J.

    1975-07-01

    This note describes the contents of the WIMS 69-group library, and includes a list of nuclides with details of data file or other source of data, resonance tabulations and thermal scattering models, and a list and details of resonance tabulations. Also included are condensation spectra used to obtain group cross-sections in fast energy range, group energy boundaries, and burn-up details, including fuel and fission product burn-up chains, fission product yields and energy release data. A fission spectrum for the 69-group library is given together with a lambda and sigma p values used in the calculation of resonance cross-sections, and 2200 m/sec absorption cross-sections and resonance absorption integrals. (U.K.)

  4. 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)

  5. Thermal reactor benchmark testing of 69 group library

    International Nuclear Information System (INIS)

    Liu Guisheng; Wang Yaoqing; Liu Ping; Zhang Baocheng

    1994-01-01

    Using a code system NSLINK, AMPX master library in WIMS 69 groups structure are made from nuclides relating to 4 newest evaluated nuclear data libraries. Some integrals of 10 thermal reactor benchmark assemblies recommended by the U.S. CSEWG are calculated using rectified PASC-1 code system and compared with foreign results, the authors results are in good agreement with others. 69 group libraries of evaluated data bases in TPFAP interface file are generated with NJOY code system. The k ∞ values of 6 cell lattice assemblies are calculated by the code CBM. The calculated results are analysed and compared

  6. Diffraction and single-crystal elastic constants of Inconel 625 at room and elevated temperatures determined by neutron diffraction

    International Nuclear Information System (INIS)

    Wang, Zhuqing; Stoica, Alexandru D.; Ma, Dong; Beese, Allison M.

    2016-01-01

    In this work, diffraction and single-crystal elastic constants of Inconel 625 have been determined by means of in situ loading at room and elevated temperatures using time-of-flight neutron diffraction. Theoretical models proposed by Voigt, Reuss, and Kroner were used to determine single-crystal elastic constants from measured diffraction elastic constants, with the Kroner model having the best ability to capture experimental data. The magnitude of single-crystal elastic moduli, computed from single-crystal elastic constants, decreases and the single crystal anisotropy increases as temperature increases, indicating the importance of texture in affecting macroscopic stress at elevated temperatures. The experimental data reported here are of great importance in understanding additive manufacturing of metallic components as: diffraction elastic constants are required for computing residual stresses from residual lattice strains measured using neutron diffraction, which can be used to validate thermomechanical models of additive manufacturing, while single-crystal elastic constants can be used in crystal plasticity modeling, for example, to understand mechanical deformation behavior of additively manufactured components.

  7. Diffraction and single-crystal elastic constants of Inconel 625 at room and elevated temperatures determined by neutron diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Zhuqing [Department of Materials Science and Engineering, Pennsylvania State University, University Park, PA 16802 (United States); Stoica, Alexandru D. [Chemical and Engineering Materials Division, Neutron Sciences Directorate, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Ma, Dong, E-mail: dongma@ornl.gov [Chemical and Engineering Materials Division, Neutron Sciences Directorate, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Beese, Allison M., E-mail: amb961@psu.edu [Department of Materials Science and Engineering, Pennsylvania State University, University Park, PA 16802 (United States)

    2016-09-30

    In this work, diffraction and single-crystal elastic constants of Inconel 625 have been determined by means of in situ loading at room and elevated temperatures using time-of-flight neutron diffraction. Theoretical models proposed by Voigt, Reuss, and Kroner were used to determine single-crystal elastic constants from measured diffraction elastic constants, with the Kroner model having the best ability to capture experimental data. The magnitude of single-crystal elastic moduli, computed from single-crystal elastic constants, decreases and the single crystal anisotropy increases as temperature increases, indicating the importance of texture in affecting macroscopic stress at elevated temperatures. The experimental data reported here are of great importance in understanding additive manufacturing of metallic components as: diffraction elastic constants are required for computing residual stresses from residual lattice strains measured using neutron diffraction, which can be used to validate thermomechanical models of additive manufacturing, while single-crystal elastic constants can be used in crystal plasticity modeling, for example, to understand mechanical deformation behavior of additively manufactured components.

  8. 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).

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

    International Nuclear Information System (INIS)

    Li Mancang; Wang Kan; Yao Dong

    2012-01-01

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

  10. Evaluation of WIMS-D/4 nuclear data library used on TRIGA reactor calculation

    International Nuclear Information System (INIS)

    Chen Wei; Xie Zhongsheng; Jiang Xinbiao; Chen Da

    1997-01-01

    The 69 groups constants of H in ZrH, 166 Er and 167 Er generated by NJOY and GASKET codes are inserted into WIMS nuclear data library WIMS-CNDC and WIMS-NINT libraries used on RTIGA reactor calculation are obtained. In order to check WIMS-CNDC and WIMS-NINT libraries, the scattering cross-section is compared with that in WIMS-IJS library. The group constant, K ∞ and temperature coefficient are calculated by using WIMS-CNDC, WIMS-NINT and WIMS-IJS. The results show the both libraries are suitable for calculation of TRIGA reactor

  11. Neutron cross-sections database for amino acids and proteins analysis

    Energy Technology Data Exchange (ETDEWEB)

    Voi, Dante L.; Ferreira, Francisco de O.; Nunes, Rogerio Chaffin, E-mail: dante@ien.gov.br, E-mail: fferreira@ien.gov.br, E-mail: Chaffin@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Rocha, Helio F. da, E-mail: hrocha@gbl.com.br [Universidade Federal do Rio de Janeiro (IPPMG/UFRJ), Rio de Janeiro, RJ (Brazil). Instituto de Pediatria

    2015-07-01

    Biological materials may be studied using neutrons as an unconventional tool of analysis. Dynamics and structures data can be obtained for amino acids, protein and others cellular components by neutron cross sections determinations especially for applications in nuclear purity and conformation analysis. The instrument used for this is the crystal spectrometer of the Instituto de Engenharia Nuclear (IEN-CNEN-RJ), the only one in Latin America that uses neutrons for this type of analyzes and it is installed in one of the reactor Argonauta irradiation channels. The experimentally values obtained are compared with calculated values using literature data with a rigorous analysis of the chemical composition, conformation and molecular structure analysis of the materials. A neutron cross-section database was constructed to assist in determining molecular dynamic, structure and formulae of biological materials. The database contains neutron cross-sections values of all amino acids, chemical elements, molecular groups, auxiliary radicals, as well as values of constants and parameters necessary for the analysis. An unprecedented analytical procedure was developed using the neutron cross section parceling and grouping method for data manipulation. This database is a result of measurements obtained from twenty amino acids that were provided by different manufactories and are used in oral administration in hospital individuals for nutritional applications. It was also constructed a small data file of compounds with different molecular groups including carbon, nitrogen, sulfur and oxygen, all linked to hydrogen atoms. A review of global and national scene in the acquisition of neutron cross sections data, the formation of libraries and the application of neutrons for analyzing biological materials is presented. This database has further application in protein analysis and the neutron cross-section from the insulin was estimated. (author)

  12. Neutron cross-sections database for amino acids and proteins analysis

    International Nuclear Information System (INIS)

    Voi, Dante L.; Ferreira, Francisco de O.; Nunes, Rogerio Chaffin; Rocha, Helio F. da

    2015-01-01

    Biological materials may be studied using neutrons as an unconventional tool of analysis. Dynamics and structures data can be obtained for amino acids, protein and others cellular components by neutron cross sections determinations especially for applications in nuclear purity and conformation analysis. The instrument used for this is the crystal spectrometer of the Instituto de Engenharia Nuclear (IEN-CNEN-RJ), the only one in Latin America that uses neutrons for this type of analyzes and it is installed in one of the reactor Argonauta irradiation channels. The experimentally values obtained are compared with calculated values using literature data with a rigorous analysis of the chemical composition, conformation and molecular structure analysis of the materials. A neutron cross-section database was constructed to assist in determining molecular dynamic, structure and formulae of biological materials. The database contains neutron cross-sections values of all amino acids, chemical elements, molecular groups, auxiliary radicals, as well as values of constants and parameters necessary for the analysis. An unprecedented analytical procedure was developed using the neutron cross section parceling and grouping method for data manipulation. This database is a result of measurements obtained from twenty amino acids that were provided by different manufactories and are used in oral administration in hospital individuals for nutritional applications. It was also constructed a small data file of compounds with different molecular groups including carbon, nitrogen, sulfur and oxygen, all linked to hydrogen atoms. A review of global and national scene in the acquisition of neutron cross sections data, the formation of libraries and the application of neutrons for analyzing biological materials is presented. This database has further application in protein analysis and the neutron cross-section from the insulin was estimated. (author)

  13. Prompt neutron decay constant for the Oak Ridge Research Reactor with 20 wt % 235U enriched fuel

    International Nuclear Information System (INIS)

    Ragan, G.E.; Mihalczo, J.T.

    1986-01-01

    This paper describes measurements of the prompt neutron decay constant at delayed criticality for the Oak Ridge Research Reactor (ORR) using 20 wt % 235 U enriched fuel and compares these measurements with similar measurements using 93.2 wt % 235 U enriched fuel. This reactor parameter is of interest because it affects the transient behavior of the reactor in prompt criticality accident situations. This experiment is part of a program to investigate the differences in the performance of research reactors fueled with highly enriched and low enriched uranium. The prompt neutron decay constants were obtained using noise analysis measurement techniques for a core with newly fabricated, unirradiated fuel elements

  14. 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)

  15. Multi-group neutron transport theory

    International Nuclear Information System (INIS)

    Zelazny, R.; Kuszell, A.

    1962-01-01

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

  16. MACKLIB-IV: a library of nuclear response functions generated with the MACK-IV computer program from ENDF/B-IV

    International Nuclear Information System (INIS)

    Gohar, Y.; Abdou, M.A.

    1978-03-01

    MACKLIB-IV employs the CTR energy group structure of 171 neutron groups and 36 gamma groups. A retrieval computer program is included with the library to permit collapsing into any other energy group structure. The library is in the new format of the ''MACK-Activity Table'' which uses a fixed position for each specific response function. This permits the user when employing the library with present transport codes to obtain directly the nuclear responses (e.g. the total nuclear heating) summed for all isotopes and integrated over any geometrical volume. The response functions included in the library are neutron kerma factor, gamma kerma factor, gas production and tritium-breeding functions, and all important reaction cross sections. Pertinent information about the library and a graphical display of six response functions for all materials in the library are given

  17. Some aspects of preparation and testing of group constants group constant system ABBN-90

    International Nuclear Information System (INIS)

    Nikolaev, M.N.; Tsiboulia, A.M.; Manturov, G.N.

    1996-01-01

    This paper presents an overview of activities performed to prepare and test the group constants ABBN-90. The ABBN-90 set is designed for application calculations of fast, intermediate and thermal nuclear reactors. The calculations of subgroup parameters are discussed. The processing code system GRUCON is mentioned in comparison to the NJOY code system. Proposals are made for future activities. (author). Figs, tabs

  18. Investigating The Neutron Flux Distribution Of The Miniature Neutron Source Reactor MNSR Type

    International Nuclear Information System (INIS)

    Nguyen Hoang Hai; Do Quang Binh

    2011-01-01

    Neutron flux distribution is the important characteristic of nuclear reactor. In this article, four energy group neutron flux distributions of the miniature neutron source reactor MNSR type versus radial and axial directions are investigated in case the control rod is fully withdrawn. In addition, the effect of control rod positions on the thermal neutron flux distribution is also studied. The group constants for all reactor components are generated by the WIMSD code, and the neutron flux distributions are calculated by the CITATION code. The results show that the control rod positions only affect in the planning area for distribution in the region around the control rod. (author)

  19. A general formula considering one group delayed neutron under nonequilibrium condition

    International Nuclear Information System (INIS)

    Li Haofeng; Chen Wenzhen; Zhu Qian; Luo Lei

    2008-01-01

    A general neutron breeder formula is developed when the reactor does not reach the steady state and the reactivity changes in phase. This formula can be used to calculate the results of six groups delayed neutron model through a way of amending λ in one group delayed neutron model. The analysis shows that the solution of amended single group delayed neutron model is approximately equal to that of six-group delayed neutron model, and the amended model meets the engineering accuracy. (authors)

  20. Validation study of SRAC2006 code system based on evaluated nuclear data libraries for TRIGA calculations by benchmarking integral parameters of TRX and BAPL lattices of thermal reactors

    International Nuclear Information System (INIS)

    Khan, M.J.H.; Sarker, M.M.; Islam, S.M.A.

    2013-01-01

    Highlights: ► To validate the SRAC2006 code system for TRIGA neutronics calculations. ► TRX and BAPL lattices are treated as standard benchmarks for this purpose. ► To compare the calculated results with experiment as well as MCNP values in this study. ► The study demonstrates a good agreement with the experiment and the MCNP results. ► Thus, this analysis reflects the validation study of the SRAC2006 code system. - Abstract: The goal of this study is to present the validation study of the SRAC2006 code system based on evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3 for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. This study is achieved through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors. In integral measurements, the thermal reactor lattices TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 are treated as standard benchmarks for validating/testing the SRAC2006 code system as well as nuclear data libraries. The integral parameters of the said lattices are calculated using the collision probability transport code PIJ of the SRAC2006 code system at room temperature 20 °C based on the above libraries. The calculated integral parameters are compared to the measured values as well as the MCNP values based on the Chinese evaluated nuclear data library CENDL-3.0. It was found that in most cases, the values of integral parameters demonstrate a good agreement with the experiment and the MCNP results. In addition, the group constants in SRAC format for TRX and BAPL lattices in fast and thermal energy range respectively are compared between the above libraries and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation study of the SRAC2006 code system based on evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 and can also be essential to implement further neutronics calculations

  1. Optimization of a neutron detector design using adjoint transport simulation

    International Nuclear Information System (INIS)

    Yi, C.; Manalo, K.; Huang, M.; Chin, M.; Edgar, C.; Applegate, S.; Sjoden, G.

    2012-01-01

    A synthetic aperture approach has been developed and investigated for Special Nuclear Materials (SNM) detection in vehicles passing a checkpoint at highway speeds. SNM is postulated to be stored in a moving vehicle and detector assemblies are placed on the road-side or in chambers embedded below the road surface. Neutron and gamma spectral awareness is important for the detector assembly design besides high efficiencies, so that different SNMs can be detected and identified with various possible shielding settings. The detector assembly design is composed of a CsI gamma-ray detector block and five neutron detector blocks, with peak efficiencies targeting different energy ranges determined by adjoint simulations. In this study, formulations are derived using adjoint transport simulations to estimate detector efficiencies. The formulations is applied to investigate several neutron detector designs for Block IV, which has its peak efficiency in the thermal range, and Block V, designed to maximize the total neutron counts over the entire energy spectrum. Other Blocks detect different neutron energies. All five neutron detector blocks and the gamma-ray block are assembled in both MCNP and deterministic simulation models, with detector responses calculated to validate the fully assembled design using a 30-group library. The simulation results show that the 30-group library, collapsed from an 80-group library using an adjoint-weighting approach with the YGROUP code, significantly reduced the computational cost while maintaining accuracy. (authors)

  2. Generation of a library of two-group diffusion and kinetics parameters for DYN3D

    International Nuclear Information System (INIS)

    Petkov, P.T.; Christoskov, I.D.; Kamenov, K.; Antov, A.

    2002-01-01

    A library of two-group diffusion and kinetics parameters has been generated for the neutron kinetics code DYN3D for analysis of reactivity initiated accidents for the WWER-440 reactors, based on the MAGRU approximation methodology for the diffusion and kinetics parameters. The accuracy of this methodology has been tested and the conclusion is that it is not adequate. A new approximation methodology, based on interpolation for the most widely varying parameters, i.e. the moderator temperature and density, and on approximation for all other independent parameters, is presented. The methodology of calculation of the kinetics parameters using primary data from ENDF-B/VI is described in detail (Authors)

  3. Validation of The Deterministic Diffusion Method For The Neutronic Calculations of Thermal Research Reactors of TRIGA-Type Using The Wisdom-IAEA-69 Nuclear Data Library

    International Nuclear Information System (INIS)

    Hussein, H.M.; Sakr, A.M.; Amin, E.H.

    2011-01-01

    The objective of this paper is to assess the suitability and the accuracy of the deterministic diffusion method for the neutronic calculations of the TRIGA type research reactors in proposed condensed energy spectra of five and seven groups with one and three thermal groups respectively, using the calculational line: WIMSD-IAEA-69 nuclear data library/ WIMSD-5B lattice and cell calculations code/ CITVAP v3.1 core calculations code. Firstly, The assessment goes through analyzing the integral parameters - k e ff, ρ 238 , σ 235 , σ 238 , and C * - of the TRX and BAPL benchmark lattices and comparison with experimental and previous reference results using other ENDLs at the full energy spectra, which show good agreement with the references at both spectra. Secondly, evaluation of the 3D nuclear characteristics of three different cores of the TRR-1/M1 TRIGA Mark- III Thai research reactor, using the CITVAP v3.1 code and macroscopic cross-section libraries generated using the WIMSD-5B code at the proposed energy spectra separately. The results include the excess reactivities and the worth of control rods, which were compared with previous Monte Carlo results and experimental values, that show good agreement with the references at both energy spectra, albeit better accuracies are shown with the five groups spectrum. The results also includes neutron flux distributions which are settled for future comparisons with other calculational techniques, even, they are comparable to reactors and fuels of the same type. The study reflects the adequacy of using the pre-stated calculational line at the condensed energy spectra for evaluation of the neutronic parameters of the TRIGA type reactors, and future comparisons of the un-benchmarked results could assure this result for wider range of neutronics or safety-related parameters

  4. 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)

  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. Proceedings of the specialists' meeting on reactor group constants

    Energy Technology Data Exchange (ETDEWEB)

    Katakura, Jun-ichi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

    This report is the Proceedings of the Specialists' Meeting on Reactor Group Constants. The meeting was held on February 22-23, 2001 at Tokai Research Establishment of Japan Atomic Energy Research Institute with the participation of 59 specialists. The evaluation work for JENDL-3.3 is going on for the publication in a short time. The processing JENDL-3.3 file to make reactor group constants is needed when it is used in application fields. In the meeting, the present status of the reactor group constants was reviewed and the issues relating to them were discussed in such fields as thermal reactor, criticality safety, fast reactor, high energy region, burn-up calculation and radiation shielding. At the final session in the meeting, standardization of reactor group constants was discussed and the need of the reference group constants was confirmed by the participants. The 11 of the presented papers are indexed individually. (J.P.N.)

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

    International Nuclear Information System (INIS)

    Zhang Baocheng

    1998-01-01

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

  8. ERANOS 2.0, Modular code and data system for fast reactor neutronics analyses

    International Nuclear Information System (INIS)

    2008-01-01

    follow-up modules (connected in the PROJERIX procedures), a fine burn-up analysis subset named MECCYCO (mass balances, activities, decay heat, dose rates). Coupled neutron/gamma calculations are also possible using specific libraries. Nuclear data libraries: The ECCO/ERANOS 2.0 code package contains four neutron cross section libraries derived from the JEF-2.2 nuclear data evaluated files. They are: - a 1968-group library (41 main nuclides), - a 33-group library (246 nuclides, including pseudo fission products), - a 175-group library (VITAMIN-J energy group scheme), - a 172-group library (XMAS energy group scheme, 246 nuclides, including pseudo-FP). These libraries were obtained by processing the JEF-2.2 files with the NJOY and CALENDF codes. Probability tables are included for the main 37 resonant nuclides. The 172-group library (XMAS energy scheme) may be used for thermal spectrum calculations. The 175-group library (some cross-sections in P5, but no probability tables) is used for shielding calculations only. Other nuclear data (fission yields and energies, decay constants, gamma production and interaction libraries, etc.) are provided in separate files. Details about cell/lattice and burnup calculations are given. Besides the modules related to basic data preparation (creation of medium, geometry, and burn-up chain SETs, modelling of operating conditions, etc.), a variety of modules computes and/or extracts specific information from the code output (fluxes, concentrations, etc.). Here is a non-exhaustive sample of such modules: - Traverse extraction and processing, - Mass and atom balances by region, - Neutron balance by region, reaction and energy group, - Integrated reaction rate processing, - Equivalence coefficients and Breeding gain, - Beta effective, - Linear and bilinear integrals (with respect to the forward and possibly adjoint fluxes). Perturbation theory and sensitivity analysis: several modules of ERANOS are available for a modular processing of such

  9. Nuclear data evaluation and group constant generation for reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Lee, Jong Tae; Min, Byung Joo; Gil, Choong Sup [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)

    1991-01-01

    In nuclear or shielding design analysis for reactors or other facilities, nuclear data are one of the primary importances. Research project for nuclear data evaluation and their effective applications has been continuously performed. The objectives of this project are (1) to compile the latest evaluated nuclear data files, (2) to establish their processing code systems, and (3) to evaluate the multi- group constant library using the newly compiled data files and the code systems. As the results of this project, ENDF/B-VI Supplementary File including important nuclides, JENDL-3.1 and JEF-1 were compiled, and ENDF-6 international computer file format for evaluated nuclear data and its processing system NJOY89.31 were tested with ENDF/B-VI data. In order to test an applicability of the newly released data to thermal reactor problems, a number of benchmark calculations were performed, and the results were analyzed. Since preliminary benchmark testing of thermal reactor problems have been made the newly compiled data are expected to be positively used to develop advanced reactors. (Author).

  10. Nuclear data evaluation and group constant generation for reactor analysis

    International Nuclear Information System (INIS)

    Kim, Jung Do; Lee, Jong Tae; Min, Byung Joo; Gil, Choong Sup

    1991-01-01

    In nuclear or shielding design analysis for reactors or other facilities, nuclear data are one of the primary importances. Research project for nuclear data evaluation and their effective applications has been continuously performed. The objectives of this project are (1) to compile the latest evaluated nuclear data files, (2) to establish their processing code systems, and (3) to evaluate the multi- group constant library using the newly compiled data files and the code systems. As the results of this project, ENDF/B-VI Supplementary File including important nuclides, JENDL-3.1 and JEF-1 were compiled, and ENDF-6 international computer file format for evaluated nuclear data and its processing system NJOY89.31 were tested with ENDF/B-VI data. In order to test an applicability of the newly released data to thermal reactor problems, a number of benchmark calculations were performed, and the results were analyzed. Since preliminary benchmark testing of thermal reactor problems have been made the newly compiled data are expected to be positively used to develop advanced reactors. (Author)

  11. Chinese computerized nuclear data library

    International Nuclear Information System (INIS)

    Liang Qichang; Cai Dunjiu

    1996-01-01

    The Second Version of Chinese Evaluated Nuclear Data Library (CENDL-2) includes the complete neutron nuclear data sets of 54 important elements and isotopes used for nuclear science and engineering with the incident neutron energy from 10 -5 eV to 20 MeV, the international universal format ENDF/B-6 was adopted. Now, the Chinese Computerized nuclear data library has been developed and put into operation. That is, the users can make on-line use of the main data libraries for evaluated neutron reaction data in the world of EXFOR experimental nuclear data library on the terminal of computer via the perfect computer software system, carry out directly the nuclear engineering calculation or nuclear data evaluation, enjoy the use of the resource of our nuclear data libraries for their development of nuclear energy and nuclear technology applications

  12. Development of low-activation design method for reduction of radioactive waste (2). Precise neutron flux and activation estimation of nuclear power plants using MATXSLIB-J33T10

    International Nuclear Information System (INIS)

    Uematsu, Mikio; Hayashi, Katsumi; Nemezawa, Shigeki; Ogata, Tomohiro; Nakata, Mikihiro; Kinno, Masaharu; Yamaguchi, Katsuyoshi; Saito, Minoru; Hasegawa, Akira

    2008-01-01

    We have been developing low-activation concrete for biological shielding wall of nuclear power plants, for the purpose of reducing large amount of radioactive waste. Based on measurement of Eu and Co content in various aggregate candidates, limestone and electro-fused alumina were selected as the most feasible aggregate for low activation concrete. Induced activity in shielding wall was calculated for both low activation concrete and ordinary concrete using neutron flux obtained from DORT two-dimensional calculation made for typical ABWR and APWR models. We have prepared new cross section library named 'MATXSLIB-J33T10 that has multi-group structure in thermal energy. The library was processed from evaluated cross section library JENDL 3.3 by using NJOY 99.83. Activation cross section library for ORIGEN-79 code is prepared for each activation calculation case by collapsing JENDL-3.3 originated 183-group constants into 3-group activation cross section using 183-group neutron flux. One-group activation cross section was also prepared in the same manner for ORIGEN2 calculation. The ΣD/C value results for low-activation concrete was sufficiently low comparing to the ordinary concrete. By using the developed low-activation concrete, activation level of biological shielding wall concrete will be effectively decreased. The use of the developed low-activation concrete will contribute to economization of nuclear power plants decommissioning by reducing large amount of radioactive concrete waste. (author)

  13. Nuclear data evaluation and group constant generation for reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung Do; Gil, Choong Sup [Korea Atomic Energy Res. Inst., Taejon (Korea, Republic of)

    1993-12-01

    In nuclear or shielding design analysis for reactors including nuclear facilities, nuclear data are one of the primary importances. Research project for nuclear data evaluation and their effective applications has been continuously performed. The objectives of this project are (1) to compile the latest evaluated nuclear data files, (2) to establish their processing code systems, and (3) to evaluate the multigroup constant library using the newly compiled data files and the code systems. As the results of this project, JEF-2.2 which is latest version of Joint Evaluated File developed at OECD/NEA was compiled and COMPLOT and EVALPLOT utility codes were installed in personal computer, which are able to draw ENDF/B-formatted nuclear data for comparison and check. Computer system (NJOY/ACER) for generating continuous energy Monte Carlo code MCNP library was established and the system was validated by analyzing a number of experimental data. (Author).

  14. Tables and figures from JNDC Nuclear Data Library of fission products, version 2

    International Nuclear Information System (INIS)

    Ihara, Hitoshi

    1989-11-01

    The content of JNDC (Japanese Nuclear Data Committee) FP (Fission Product) Nuclear Data Library version 2 for 1227 fission products is presented in the form of tables and figures. The library is inclusive of evaluated decay data such as decay constant, Q-value, average energies of beta, gamma and internal conversion electron, spin-parity, branching ratio of each decay mode and fission yield. The neutron capture cross-sections are also contained for 166 nuclides. The mass number of the fission product nuclides ranges from A = 66 to A = 172. (author)

  15. Development of the fast reactor group constant set JFS-3-J3.2R based on the JENDL-3.2

    CERN Document Server

    Chiba, G

    2002-01-01

    It is reported that the fast reactor group constant set JFS-3-J3.2 based on the newest evaluated nuclear data library JENDL3.2 has a serious error in the process of applying the weighting function. As the error affects greatly nuclear characteristics, and a corrected version of the reactor constant set, JFS-3-J3.2R, was developed, as well as lumped FP cross sections. The use of JFS-3-J3.2R improves the results of analyses especially on sample Doppler reactivity and reaction rate in the blanket region in comparison with those obtained using the JFS-3-J3.2.

  16. Investigation of the neutron detection statistics in fast critical assembly BFS-24-1

    International Nuclear Information System (INIS)

    Avramov, A.M.; Tyutyunnikov, P.L.; Mikulski, A.T.; Rafalska, E.; Chwaszczewski, S.; Jablonski, K.

    1974-01-01

    The results of the neutron detection statistics investigation at the fast critical assembly BFS-24-1 are given. The Ross-α measurements were carried out using: digital flash-start unit and 256 channel time analyzer, 10 channel time analyzer, alphameter device. Parallely the measurements using the variable dead time method and zero probability method were performed. The prompt neutron decay constants, the effectiveness of neutron detector and the intensity of external neutron source are determined using the experimental data. The experimental values of prompt neutron decay constant are compared with the calculated ones. The codes used in the calculation are following: one dimensional, diffusion, 26-group code 26-M and EWA-1, one dimensional, multiregion, nonstationary diffusion 3-group code SPECTR, 26-group, diffusion code in buckling approximation, MIXSPECTR. In all codes the 26 group nuclear constants BNAB-26 and BNAB-70 are used. (author)

  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. 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. Importance of delayed neutron data in transmutation system

    International Nuclear Information System (INIS)

    Tsujimoto, Kazufumi

    1999-01-01

    The accelerator-driven transmutation system has been studied at the Japan Atomic Energy Research Institute. This system is a hybrid system which consists of a high intensity accelerator, a spallation target and a subcritical core region. The subcritical core is driven by neutrons generated by spallation reaction in the target region. There is no control rod in this system, so the power is controlled only by proton beam current. The beam current to keep constant power change with effective multiplication factor of subcritical core. So, the evaluation of delayed neutron fraction which is strongly connected to the measurement of subcritical level is important factor in operation of accelerator-driven system. In this paper, important nuclides for the delayed neutron fraction of ADS will be discussed, moreover, present state of delayed neutron data in evaluated nuclear data library is presented. (author)

  20. Nuclear energy and astrophysics applications of ENDF/B-VII.1 evaluated nuclear library

    International Nuclear Information System (INIS)

    Pritychenko, B.

    2012-01-01

    Recently released ENDF/B-VII.1 evaluated nuclear library contains the most up-to-date evaluated neutron cross section and covariance data. These data provide new opportunities for nuclear science and astrophysics application development. The improvements in neutron cross section evaluations and more extensive utilization of covariance files, by the Cross Section Evaluation Working Group (CSEWG) collaboration, allowed users to produce neutron thermal cross sections, Westcott factors, resonance integrals, Maxwellian-averaged cross sections and astrophysical reaction rates, and provide additional insights on the currently available neutron-induced reaction data. Nuclear reaction calculations using the ENDF/B-VII.1 library and current computer technologies will be discussed and new results will be presented

  1. Update of KASHIL-E6 library for shielding analysis and benchmark calculations

    International Nuclear Information System (INIS)

    Kim, D. H.; Kil, C. S.; Jang, J. H.

    2004-01-01

    For various shielding and reactor pressure vessel dosimetry applications, a pseudo-problem-independent neutron-photon coupled MATXS-format library based on the last release of ENDF/B-VI has been generated as a part of the update program for KASHIL-E6, which was based on ENDF/B-VI.5. It has VITAMIN-B6 neutron and photon energy group structures, i.e., 199 groups for neutron and 42 groups for photon. The neutron and photon weighting functions and the Legendre order of scattering are same as KASHIL-E6. The library has been validated through some benchmarks: the PCA-REPLICA and NESDIP-2 experiments for LWR pressure vessel facility benchmark, the Winfrith Iron88 experiment for validation of iron data, and the Winfrith Graphite experiment for validation of graphite data. These calculations were performed by the TRANSXlDANTSYS code system. In addition, the substitutions of the JENDL-3.3 and JEFF-3.0 data for Fe, Cr, Cu and Ni, which are very important nuclides for shielding analyses, were investigated to estimate the effects on the benchmark calculation results

  2. Neutron spectrum in small iron pile surrounded by lead reflector

    International Nuclear Information System (INIS)

    Kimura, Itsuro; Hayashi, S.A.; Kobayashi, Katsuhei; Matsumura, Tetsuo; Nishihara, Hiroshi.

    1978-01-01

    In order to save the quantity of sample material, a possibility to assess group constants of a reactor material through measurement and analysis of neutron spectrum in a small sample pile surrounded by a reflector of heavy moderator, was investigated. As the sample and the reflector, we chose iron and lead, respectively. Although the time dispersion in moderation of neutrons was considerably prolonged by the lead reflector, this hardly interferes with the assessment of group constants. Theoretical calculation revealed that both the neutron flux spectrum and the sensitivity coefficient of group constants in an iron sphere, 35 cm in diameter surrounded by the lead reflector, 25 cm thick, were close to those of the bare iron sphere, 108 cm in diameter. The neutron spectra in a small iron pile surrounded by a lead reflector were experimentally obtained by the time-of-flight method with an electron linear accelerator and the result was compared with the predicted values. It could be confirmed that a small sample pile surrounded by a reflector, such as lead, was as useful as a much larger bulk pile for the assessment of group constants of a reactor material. (auth.)

  3. Early fusion reactor neutronic calculations: A reevaluation

    International Nuclear Information System (INIS)

    Perry, R.T.

    1996-01-01

    Several fusion power plant design studies were made at a number of universities and laboratories in the late 1960s and early 1970s. These studies included such designs as the Princeton Plasma Physics Laboratory Fusion Power Plan and the University of Wisconsin UWMAK-I Reactor Neutronic analyses of the blankets and shields were part of the studies. During this time there were dissertations written on neutronic analysis systems and the results of neutronic analysis on several blanket and shield designs. The results were presented in the literature. Now in the fifth decade of fusion research, investigators often return to the earlier analyses for the neutronic results that are applicable to current blanket and shield designs, with the idea of using the older work as a basis for the new. However, the analyses of the past were made with cross-section data sets that have long been replaced with more modern versions. In addition, approximations were often made to the cross sections used because more exact data were not available. Because these results are used as guides, it is important to know if they are reproducible using more modern data. In this paper, several of the neutronic calculations made in the early studies are repeated using the MATXS-11 data library. This library is the ENDF/B-VI version of the MATXS-5 library. The library has 80 neutron groups. Tritium breeding ratios, heating rates, and fluxes are calculated and compared. This transport code used here is the one- dimensional S n code, ONEDANT. It is important to note that the calculations here are not to be considered as benchmarks because parameter and sensitivity studies were not made. They are used only to see if the results of older calculations are in reasonable agreement with a more modern library

  4. 8-group relative delayed neutron yields for monoenergetic neutron induced fission of 239Pu

    International Nuclear Information System (INIS)

    Piksaikin, V.M.; Kazakov, L.E.; Isaev, S.G.; Korolev, G.G.; Roshchenko, V.A.; Tertychnyj, R.G

    2002-01-01

    The energy dependence of the relative yield of delayed neutrons in an 8-group model representation was obtained for monoenergetic neutron induced fission of 239 Pu. A comparison of this data with the available experimental data by other authors was made in terms of the mean half-life of the delayed neutron precursors. (author)

  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. ZZ-CENPL, Chinese Evaluated Nuclear Parameter Library. ZZ CENPL-DLS, Discrete Level Schemes and Gamma Branching Ratios Library; ZZ CENPL-FBP, Fission Barrier Parameter Library; ZZ CENPL-GDRP, Giant Dipole Resonance Parameter Library; ZZ CENPL-NLD, Nuclear Level Density Parameter Library; ZZ CENPL-MCC, Nuclear Ground State Atomic Masses Library; ZZ CENPL-OMP, Optical Model Parameter Library

    International Nuclear Information System (INIS)

    Su Zongdi

    1995-01-01

    Description of program or function: CENPL - GDRP (Giant Dipole Resonance Parameters for Gamma-Ray): - Format: special format described in documentation; - Nuclides: V, Mn, Co, Ni, Cu, Zn, Ga, Ge, As, Se, Rb, Sr, Y, Zr, Nb, Mo, Rh, Pd, Ag, Cd, In, Sn, Sb, Te, I, Cs, Ba, La, Ce, Pr, Nd, Sm, Eu, Gd, Tb, Ho, Er, Lu, Ta, W, Re, Os, Ir, Pt, Au, Hg, Pb, Bi, Th, U, Np, Pu. - Origin: Experimental values offered by S.S. Dietrich and B.L. Berman. CENPL - FBP (Fission Barrier Parameter Sub-Library): - Format: special format described in documentation; - Nuclides: (1) 51 nuclei region from Th-230 to Cf-255, (2) 46 nuclei region from Th-229 to Cf-253, (3) 24 nuclei region from Pa-232 to Cf-253; - Origin: (1) Lynn, (2) Analysis of experimental data by Back et al., (3) Ohsawa. CENPL - DLS (Discrete level scheme and branch ratio of gamma decay: - Format: Special format described in documentation; - Origin: ENSDF - BNL. CENPL - NLD (Nuclear Level Density): - Format: Special format described in documentation; - Origin: Huang Zhongfu et al. CENPL - OMP (Optical model parameter sub-library): - Format: special format described in documentation ; - Origin: CENDL, ENDF/B-VI, JENDL-3. CENPL - MC (I) and (II) (Atomic masses and characteristic constants for nuclear ground states) : - Format: Brief table format; - Nuclides: 4760 nuclides ranging from Z=0 A=1 to Z=122 A=318. - Origin: Experimental data and systematic results evaluated by Wapstra, theoretical results calculated by Moller, ENSDF - BNL and Nuclear Wallet Cards. CENPL contains the following six sub-libraries: 1. Atomic Masses and Characteristic Constants for nuclear ground states (MCC). This data consists of calculated and in most cases also measured mass excesses, atomic masses, total binding energies, spins, parities, and half-lives of nuclear ground states, abundances, etc. for 4800 nuclides. 2. Discrete Level Schemes and branching ratios of gamma decay (DLS). The data on nuclear discrete levels are based on the Evaluated

  7. The WIMSLIB library - neutron data library for WIMS-D

    International Nuclear Information System (INIS)

    Liu Ping

    1998-05-01

    During a visit to the IAEA Nuclear Data Section from 13 June to 12 December 1997, the author processed the Chinese Evaluated Nuclear Data Library (CENDL), Version 2.1, using the NJOY Nuclear Data Processing System, Version 94.105, to generate the working library WIMSLIB for input to WIMS-D/4 and WIMS-D/5A. The WIMSLIB library was then used to perform benchmark testing of CENDL-2.1. (author)

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  10. Development of temperature related thermal neutron scattering database for MCNP

    International Nuclear Information System (INIS)

    Mei Longwei; Cai Xiangzhou; Jiang Dazhen; Chen Jingen; Guo Wei

    2013-01-01

    Based on ENDF/B-Ⅶ neutron library, the thermal neutron scattering library S(α, β) for molten salt reactor moderators was developed. The temperatures of this library were chose as the characteristic temperature of the molten salt reactor. The cross section of the thermal neutron scattering of ACE format was investigated, and this library was also validated by the benchmarks of ICSBEP. The uncertainties shown in the validation were in reasonable range when compared with the thermal neutron scattering library tmccs which included in the MCNP data library. It was proved that the thermal neutron scattering library processed in this study could be used in the molten salt reactor design. (authors)

  11. Renormalization group equations with multiple coupling constants

    International Nuclear Information System (INIS)

    Ghika, G.; Visinescu, M.

    1975-01-01

    The main purpose of this paper is to study the renormalization group equations of a renormalizable field theory with multiple coupling constants. A method for the investigation of the asymptotic stability is presented. This method is applied to a gauge theory with Yukawa and self-quartic couplings of scalar mesons in order to find the domains of asymptotic freedom. An asymptotic expansion for the solutions which tend to the origin of the coupling constants is given

  12. Results of neutron physics analyses of WWER-440 cores with modified reactor protection and control systems

    International Nuclear Information System (INIS)

    Lehmann, M.; Pecka, M.; Rocek, J.; Zalesky, K.

    1993-12-01

    Detailed results are given of neutron physics analyses performed to assess the efficiency and acceptability of modifications of the WWER-440 core protection and control system; the modifications have been proposed with a view to increasing the proportion of mechanical control in the compensation of reactivity effects during reactor unit operation in the variable load mode. The calculations were carried out using the modular MOBY-DICK macrocode system together with the SMV42G36 library of two-group parametrized diffusion constants, containing corrections which allow new-design WWER-440 fuel assemblies to be discriminated. (J.B). 37 tabs., 18 figs., 5 refs

  13. The fluence research of filter material for fast neutron fluence measurement

    International Nuclear Information System (INIS)

    Tang Xiding

    2010-01-01

    When the fast neutron fluence is measured by radioactivation techniques in the nuclear reactor the fast neutron is also filtered a little by the thermal neutron filter material, and if the filter material thickness increase the filtered fast neutron increases therewith. For fast neutron fluenc measurement, there are only cadmium, boron and gadolinium three elements filtering fluence can be calculated ordinarily. In order to calculate the filtered fast neutron fluence of the all elements in the filter material, the many total cross sections of nuclides had checked out from nuclear cross section data library, converted them into the same energy group structure, then element's total cross section, compound's total cross section and multilayer filters' total cross section had calculated from these total cross sections with same energy group structure, a new cross section data library can be obtained lastly through merging these cross sections into the old cross section data library used for neutron fluence measurement. The calculation analysis indicates that the results of the unit 2 surveillance capsule U of DAYA Bay NPP and the unit 1 surveillance capsule A of the Second Nuclear Power Plant of Qinshan by considering the all elements subtracting iron are smaller about 1.5% and 2.6% respectively than the ones only to consider cadmium, boron. The old measured results accord with the new values under the measurement uncertainty, are reliable. The new results are more accuracy. (authors)

  14. Fedgroup - a program system for producing group constants from evaluated nuclear data of files disseminated by IAEA

    International Nuclear Information System (INIS)

    Vertes, P.

    1978-03-01

    The program system FEDGROUP was originally released for general distribution in June 1976. It is widely applied for the generation of group constant libraries used by different spectral codes. In this revised version of report INDC/HUN/-13 errors, deficiencies and misprints in the original report have been removed and an extension is introduced and described. The basic computer of FEDGROUP is the CDC-3300. There exist, however, CYBER-72, BESM-6 and IBM-360 versions, too. The problems connected with the various computer versions are discussed. Results of test calculations are quoted and errata to the report INDC/HUN/-13 are given

  15. A study of group constant generation method in fast reactor analysis

    International Nuclear Information System (INIS)

    Takano, Hideki

    1983-05-01

    The methods of generating group constants have been studied to predict accurately the nuclear characteristics of fast reactors. In resonance energy region, the accuracy of the group constants was investigated, which were calculated by the approximate weighting spectrum used for a conventional standards group constant set such as ABBN. It was shown that the basic assumption of constant collision density for group constant calculation was not always satisfactory. Moreover, a multilevel formula was derived without losing the useful characteristics of the Breit-Wigner single level formula. Using this formula, the interference effect between resonances was examined. In addition, the mutual interference between different resonant nuclides was calculated. The cause of a systematic dependence of effective multiplication factors on U-238 concentration ratio was studied, and the cross section adjustment was performed. In the unresolved resonance region, the average resonance parameters were searched. As a result, the JFS-2 set was generated, and several studies were performed to advance the concept of the group constant set JFS-2. (Kako, I.)

  16. Neutronics codes

    International Nuclear Information System (INIS)

    Buckel, G.

    1983-01-01

    The objectives are the development, testing and cultivation of reliable, efficient and user-optimized neutron-physical calculation methods and conformity with users' requirements concerning design of power reactors, planning and analysis of experiments necessary for their protection as well as research on physical key problems. A short outline of available computing programmes for the following objectives is given: - Provision of macroscopic group constants, - Calculation of neutron flux distribution in transport theory and diffusion approximation, - Evaluation of neutron flux-distribution, - Execution of disturbance calculations for the determination reactivity coefficients, and - graphical representation of results. (orig./RW) [de

  17. Criticality experiments: analysis, evaluation, and programs. 8. Prompt Neutron Decay Constants in Uranium Diluted with Matrix Material Systems

    International Nuclear Information System (INIS)

    Sanchez, Rene; Loaiza, David; Brunson, Glenn

    2001-01-01

    Rossi-Alpha measurements were performed on uranium diluted with matrix material systems to determine the prompt neutron decay constants. These constants represent an eigenvalue characteristic of these particular critical assemblies, which can be experimentally measured by the Rossi-Alpha or pulse neutron source techniques and calculated by a deterministic or Monte Carlo method. In the measurements presented in this paper, highly enriched foils diluted in various X/ 235 U ratios with polyethylene and SiO 2 , and polyethylene and aluminum were assembled to a high multiplication, and the prompt neutron decay constants were obtained by the Rossi-Alpha technique. The uranium diluted with matrix material experiments were fueled with highly enriched uranium foils. The average dimensions of the bare foils were 22.86 cm squared and 0.00762 cm thick. The foils were laminated with plastic sheets to reduce the amount of airborne contamination. Each foil weighed ∼70 g. The diluent material consisted of SiO 2 , or 6061 aluminum plates, which were embedded into polyethylene plates. The SiO 2 and aluminum plates were 22.86 cm square and 0.64 cm thick. The polyethylene plates were 39.12 cm square and 1.91 cm thick. Each polyethylene plate had a central recess whose dimensions were 22.86 cm by 22.86 cm by 0.64 cm deep and was used to accommodate the SiO 2 , or aluminum plates as well as the uranium foils. There were eight 39.12-cm-squared by 2.54-cm-thick high density polyethylene plates that form the top and bottom reflectors (four at the top and four at the bottom). Also, one of the polyethylene plates located in the center of the assembly had holes drilled in a radial direction to accommodate neutron detectors. Four 3 He detectors were placed in this plate. The 3 He detectors were 1.27 cm in diameter and ∼15 cm long. Rossi-Alpha measurements were performed at several subcritical separations for both experiments. The data were collected with a type I time analyzer (PATRM

  18. 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.)

  19. Measurements of fusion neutron yields by neutron activation technique: Uncertainty due to the uncertainty on activation cross-sections

    Energy Technology Data Exchange (ETDEWEB)

    Stankunas, Gediminas, E-mail: gediminas.stankunas@lei.lt [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos str. 3, LT-44403 Kaunas (Lithuania); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Batistoni, Paola [ENEA, Via E. Fermi, 45, 00044 Frascati, Rome (Italy); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Sjöstrand, Henrik; Conroy, Sean [Department of Physics and Astronomy, Uppsala University, PO Box 516, SE-75120 Uppsala (Sweden); EUROfusion Consortium, JET, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

    2015-07-11

    The neutron activation technique is routinely used in fusion experiments to measure the neutron yields. This paper investigates the uncertainty on these measurements as due to the uncertainties on dosimetry and activation reactions. For this purpose, activation cross-sections were taken from the International Reactor Dosimetry and Fusion File (IRDFF-v1.05) in 640 groups ENDF-6 format for several reactions of interest for both 2.5 and 14 MeV neutrons. Activation coefficients (reaction rates) have been calculated using the neutron flux spectra at JET vacuum vessel, both for DD and DT plasmas, calculated by MCNP in the required 640-energy group format. The related uncertainties for the JET neutron spectra are evaluated as well using the covariance data available in the library. These uncertainties are in general small, but not negligible when high accuracy is required in the determination of the fusion neutron yields.

  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. Creation and testing of an ENDF/B-VI neutron data library (ENDF60) for use with MCNP trademark

    International Nuclear Information System (INIS)

    Frankle, S.C.; MacFarlane, R.E.

    1995-01-01

    The continuous-energy neutron data library ENDF60, for use with the Monte Carlo N Particle radiation transport code MCNP4A, was released in the fall of 1994. The ENDF60 library is comprised of 124 nuclide data files based on the ENDF/B-VI evaluations through Release 2. Fifty-two percent of these ENDF/B-VI evaluations are translations from ENDF/B-V. The remaining forty-eight percent are new evaluations which have sometimes changed significantly. The new evaluations include important materials for criticality safety calculations, as well as significant enhancements such as isotopic evaluations, better resonance-range representations, and the new correlated energy-angle distributions for emitted particles. In particular, the upper energy limit for the resolved resonance region of 235 U, 238 U and 239 Pu has been extended from 0.082, 4.0, and 0.301 keV to 2.25, 10.0, and 2.5 keV respectively. As part of the overall quality assurance testing of the ENDF60 library, calculations for well known benchmark assemblies were performed. This benchmarking effort included revising the standard nine criticality benchmarks documented in previous Los Alamos National Laboratory Reports, LA-12212 and LA-12891, as well as the implementation of new Cross Section Evaluation Working Group (CSEWG) benchmarks. Comparisons of benchmark results for different data libraries can aid the user in understanding how well an evaluation performs for their application

  2. Neutronic design and analysis on dual-cooled waste transmutation blanket for the fusion driven sub-critical system

    International Nuclear Information System (INIS)

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

    2004-01-01

    Neutronics design and analysis of dual-cooled multi-functional waste transmutation blanket (DWTB) for the fusion driven sub-critical system (FDS) are performed to ensure the system be able to meet the requirements of fuel-sufficiency and more waste transmutation ratio with low initial loading fuel inventory, which is based on 1-D burn-up calculations with home-developed code Visual BUS and the multi-group (175 neutron groups-42 Gamma groups coupled) data library HENDL1.0/MG (Hybrid Evaluated Nuclear Data Library). (authors)

  3. Measurement of the neutron activation constants Q0 and k0 for the 27Al(n, γ)28Al reaction at the JSI TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Vladimir Radulovic; Andrej Trkov; Radojko Jacimovic; Robert Jeraj

    2013-01-01

    Measurements of the neutron activation constants Q 0 and k 0 for the 27 Al(n, γ) 28 Al reaction have been performed in two irradiation channels with different spectral characteristics at the JSI TRIGA Mark II reactor. In the determination of Q 0 the fission spectrum contribution to the reaction rates has been corrected for. The final experimental value of the Q 0 factor was found to differ significantly from the adopted value in the k 0 -database. The experimental value of the k 0 factor is in agreement with the recommended value in the k 0 -database. The thermal cross-section and resonance integral for the reaction were found to be in good agreement with the values calculated from the cross-sections from the ENDF/B-VII.1 library. (author)

  4. Report from the neutron diffraction work group

    International Nuclear Information System (INIS)

    1978-08-01

    This progress report of the neutron diffraction group at the Hahn Meitner Institute in Berlin comprises the following contributions: Three-dimensional critical properties of CsNiF 3 around the Neel point; Spin waves in CsNiF 3 with an applied magnetic field; Solitons in CsNiF 3 : Their experimental evidence and their thermodynamics; Neutron diffraction study of DAG at very low temperatures and in external magnetic field; Neutron diffraction investigation of tricritical behaviour in DyPO 4 ; Crystalline modifications and structural phase transitions of NaOH; Gitterdynamik von Cerhydrid; Investigation of the ferroelectric-ferroelastic phase transition in KH 2 PO 4 and RbH 2 PO 4 by means of γ-ray diffractometry; A γ-ray diffractometer for systematic measurements of absolute structure factors; Electron density in pyrite by combined γ-ray and neutron diffraction measurements: Thermal parameters from short wavelength neutron data; Accurate determination of temperature parameters from neutron diffraction data: Direct observation of the thermal diffuse scattering from silicon using perfect crystals; A Compton spectrometer for momentum density studies using 412 keV γ-radiation; Investigation of the electronic structure of Niobiumhydrides by means of gamma-ray Compton scattering; Interpretation of Compton profile data in position space; High resolution neutron scattering measurements on single crystals using a horizontally bent monochromator and a multidetecter; Statistical analysis of neutron diffraction studies of proteins. (orig.) [de

  5. Neutron radiography working group test programme

    International Nuclear Information System (INIS)

    Domanus, J.C.

    1989-03-01

    Scope and results of the Euratom Neutron Radiography Working Group Test Program are described. Seven NR centers from six European Community countries have performed this investigation using eleven NR facilities. Four test items were neutron radiographed using 30 different film/converter combinations. From film density measurements neutron beam components were determined. Radiographic sensitivity was assessed from visual examinations of the radiographs. About 25,000 dimensional measurements were made and were used for the assessment of accuracies of dimensional measurements from neutron radiographs. The report gives a description of the test items used for the Test Program, the film density and dimensional measurements, and concentrates on the assessment of the measuring results. The usefulness of the beam purity and sensitivity indicators was assessed with the conclusion that they are not suitable for neutron radiography of nuclear reactor fuel. Ample information is included in the report about measuring accuracies which can be reached in dimensional measurements of fuel pins. After a general comparison of measuring accuracies is discussed. Results from different NR facilities are treated separately as are the different kinds of dimensions of the fuel pins. Finally human and instrument factors are discussed. After presenting final conclusions (which take into account the above-mentioned factors) results of other investigations about dimensional measurements are shortly reviewed

  6. Application of the IEAF-2001 activation data library to activation analyses of the IFMIF high flux test module

    International Nuclear Information System (INIS)

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

    2002-01-01

    A complete activation data library IEAF-2001 (intermediate energy activation file) has been developed in standard ENDF-6 format with neutron-induced activation cross sections for 679 target nuclides from Z=1 (hydrogen) to Z=84 (polonium) and incident neutron energies up to 150 MeV. Using the NJOY processing code, an IEAF-2001 working library has been prepared in a 256 energy group structure for enabling activation analyses of the International Fusion Material Irradiation Facility (IFMIF) D-Li neutron source. This library was applied to the activation analysis of the IFMIF high flux test module using the recent Analytical and Laplacian Adaptive Radioactivity Analysis activation code which is capable of handling the variety of reaction channels open in the energy domain above 20 MeV. The IEAF-2001 activation library was thus shown to be suitable for activation analyses in fusion technology and intermediate energy applications such as the IFMIF D-Li neutron source

  7. Reasons for 2011 Release of the Evaluated Nuclear Data Library (ENDL2011.0)

    Energy Technology Data Exchange (ETDEWEB)

    Brown, D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Escher, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hoffman, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Luu, T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ormand, W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Summers, N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Thompson, I. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-09-22

    LLNL's Computational Nuclear Physics Group and Nuclear Theory and Modeling Group have collaborated to create the 2011 release of the Evaluated Nuclear Data Library (ENDL2011). ENDL2011 is designed to sup- port LLNL's current and future nuclear data needs. This database is currently the most complete nuclear database for Monte Carlo and deterministic transport of neutrons and charged particles, surpassing ENDL2009.0 [1]. The ENDL2011 release [2] contains 918 transport-ready eval- uations in the neutron sub-library alone. ENDL2011 was assembled with strong support from the ASC program, leveraged with support from NNSA science campaigns and the DOE/Offce of Science US Nuclear Data Pro- gram.

  8. 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

  9. Deterministic calculation of the effective delayed neutron fraction without using the adjoint neutron flux - 299

    International Nuclear Information System (INIS)

    Talamo, A.; Gohar, Y.; Aliberti, G.; Zhong, Z.; Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C.; Serafimovich, I.

    2010-01-01

    In 1997, Bretscher calculated the effective delayed neutron fraction by the k-ratio method. The Bretscher's approach is based on calculating the multiplication factor of a nuclear reactor core with and without the contribution of delayed neutrons. The multiplication factor set by the delayed neutrons (the delayed multiplication factor) is obtained as the difference between the total and the prompt multiplication factors. Bretscher evaluated the effective delayed neutron fraction as the ratio between the delayed and total multiplication factors (therefore the method is often referred to as k-ratio method). In the present work, the k-ratio method is applied by deterministic nuclear codes. The ENDF/B nuclear data library of the fuel isotopes ( 238 U and 238 U) have been processed by the NJOY code with and without the delayed neutron data to prepare multigroup WIMSD nuclear data libraries for the DRAGON code. The DRAGON code has been used for preparing the PARTISN macroscopic cross sections. This calculation methodology has been applied to the YALINA-Thermal assembly of Belarus. The assembly has been modeled and analyzed using PARTISN code with 69 energy groups and 60 different material zones. The deterministic and Monte Carlo results for the effective delayed neutron fraction obtained by the k-ratio method agree very well. The results also agree with the values obtained by using the adjoint flux. (authors)

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

    International Nuclear Information System (INIS)

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

    1988-03-01

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

  11. FENDL multigroup libraries

    International Nuclear Information System (INIS)

    Ganesan, S.; Muir, D.W.

    1992-01-01

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

  12. Diffraction plane dependency of elastic constants in ferritic steel in neutron stress measurement

    International Nuclear Information System (INIS)

    Hayashi, M.; Ishiwata, M.; Minakawa, N.; Funahashi, S.

    1993-01-01

    Neutron diffraction measurements have been made to investigate the elastic properties of the ferritic steel obtained from socket weld. The Kroner elastic model is found to account for the [hkl]-dependence of Young's modulus and Poisson's ratio in the material. Maps of residual stress are later to be made by measuring lattice strain from shifts in the (112) diffraction peak, for which the diffraction elastic constants the herein found to be E=243±5GPa and ν=0.28±0.01. (author)

  13. ANITA-2000 activation code package - updating of the decay data libraries and validation on the experimental data of the 14 MeV Frascati Neutron Generator

    Directory of Open Access Journals (Sweden)

    Frisoni Manuela

    2016-01-01

    Full Text Available ANITA-2000 is a code package for the activation characterization of materials exposed to neutron irradiation released by ENEA to OECD-NEADB and ORNL-RSICC. The main component of the package is the activation code ANITA-4M that computes the radioactive inventory of a material exposed to neutron irradiation. The code requires the decay data library (file fl1 containing the quantities describing the decay properties of the unstable nuclides and the library (file fl2 containing the gamma ray spectra emitted by the radioactive nuclei. The fl1 and fl2 files of the ANITA-2000 code package, originally based on the evaluated nuclear data library FENDL/D-2.0, were recently updated on the basis of the JEFF-3.1.1 Radioactive Decay Data Library. This paper presents the results of the validation of the new fl1 decay data library through the comparison of the ANITA-4M calculated values with the measured electron and photon decay heats and activities of fusion material samples irradiated at the 14 MeV Frascati Neutron Generator (FNG of the NEA-Frascati Research Centre. Twelve material samples were considered, namely: Mo, Cu, Hf, Mg, Ni, Cd, Sn, Re, Ti, W, Ag and Al. The ratios between calculated and experimental values (C/E are shown and discussed in this paper.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  15. Neutron data testing for plutonium isotopes in experiments at fast critical assemblies

    International Nuclear Information System (INIS)

    Bednyakov, S.M.; Dulin, V.A.; Manturov, G.N.; Mozhaev, V.K.; Semenov, M.Yu.; Tsibulya, A.M.

    1996-01-01

    Experimental results on checking neutron data, obtained at the fast critical assemblies, are presented. They constitute sufficiently large collection of data making it possible to test nuclear neutron constants of plutonium isotopes for the new system of group constants BNAB-93. The work contains comparison of the measurement results on average fission cross section ratios and reactivity coefficients ratios for 239,240,241 Pu (to 235 U) with calculational data, obtained on the basis of the new testing system of the BNAB-93 group constants system. 14 refs., 6 figs

  16. Means and method for controlling the neutron output of a neutron generator tube

    International Nuclear Information System (INIS)

    Langford, O.M.; Peelman, H.E.

    1980-01-01

    A gas filled neutron tube in a nuclear well logging tool has a target an ion source voltage and a replenisher connected to ground. A negative high voltage is applied to the target by a power supply also providing a target current corresponding to the neutron output of the neutron generator tube. A constant current source provides a constant current. A network receiving the target current and the constant current provides a portion of the constant current as a replenisher current which is applied to the replenisher in a neutron generating tube. The network controls the magnitude of the replenisher current in accordance with the target current so as to control the neutron output of the neutron generating tube. (auth)

  17. Identification of force constants in β-brass

    DEFF Research Database (Denmark)

    Norvell, J. C.; Als-Nielsen, Jens Aage

    1969-01-01

    The phonon dispersion curves of β-brass have previously been measured by Gilat and Dolling and a fit was obtained to a Born-von Kármán model with forces extending to the fourth nearest neighbours. Although a factor of 10 was found between the second-nearest-neighbour Cu-Cu and Zn-Zn force constants......, the data did not allow an identification of these constants. By comparisons of neutron group intensities from two β-brass crystals, one with normal Cu and the other isotopically enriched with 65Cu, we are able to identify conclusively these force constants: αZn-Zn2nd similar, equals 10αCu-Cu2nd....

  18. A one-dimensional, one-group absorption-production nodal method for neutron flux and power distributions calculations

    International Nuclear Information System (INIS)

    Ferreira, C.R.

    1984-01-01

    It is presented the absorption-production nodal method for steady and dynamical calculations in one-dimension and one group energy. It was elaborated the NOD1D computer code (in FORTRAN-IV language). Calculations of neutron flux and power distributions, burnup, effective multiplication factors and critical boron concentration were made with the NOD1D code and compared with results obtained through the CITATION code, which uses the finite difference method. The nuclear constants were produced by the LEOPARD code. (M.C.K.) [pt

  19. 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

  20. Uppsala neutron-proton scattering measurements and the πNN coupling constant

    International Nuclear Information System (INIS)

    Olsson, N.; Blomgren, J.; Conde, H.; Dangtip, S.; Elmgren, K.; Rahm, J.; Roennqvist, T.; Zorro, R.; Loiseau, B.

    2000-01-01

    The differential np scattering cross section has been measured at 96 MeV and 162 MeV at backward angles at the neutron beam facility of the The Svedberg Laboratory in Uppsala. The angular distributions have been normalized to the experimental total np cross section. Between 150 and 180 , the angular distributions are steeper than for most previous measurements and nucleon-nucleon potential predictions, but for all the angular range covered, the data agree very well in shape with the recent PSI data. At 180 , the difference versus older data amounts to about 10%, implying serious consequences because of the fundamental importance of this cross section. Values of the charged πNN coupling constant have been extracted from the data. (orig.)

  1. McStas 1.1. A freeware package for neutron Monte Carlo ray-tracing simulations

    International Nuclear Information System (INIS)

    Lefmann, K.; Nielsen, K.

    1999-01-01

    Neutron simulation is becoming an indispensable tool for neutron instrument design. At Risoe National Laboratory, a user-friendly, versatile, and fast simulation package, McStas has been developed, which may be freely downloaded from our website. An instrument is described in the McStas meta-language and is composed of elements from the McStas component library, which is under constant development and debugging by both the users and us. The McStas front- and back-ends take care of performing the simulations and displaying their results, respectively. McStas 1.1 facilities detailed simulations of complicated triple-axis instruments like the Riso RITA spectrometer, and it is equally well equipped for time-of flight spectrometers. At ECNS'99, a brief tutorial of McStas including a few on-line demonstrations is presented. Further, results from the latest simulation work in the growing McStas user group are presented and the future of this project is discussed. (author)

  2. Critical parameters and decay constants for one-speed neutrons in slabs and spheres with anisotropic scattering

    International Nuclear Information System (INIS)

    Yildiz, C.

    2001-01-01

    Time-dependent, one-speed neutron transport equations with strong forward and backward scattering together with isotropic scattering are studied in homogeneous slabs and spheres. First, a simple formal equivalence between the transport equations for a critical and for a time-decaying system is established. Then, the transport equation is converted into a more conventional one. The F N method of solving the resulting transport equation is applied to the calculation of the critical parameters and decay constants for the fundamental mode of the flux distribution and one-speed neutrons in spheres and infinite slabs. Numerical results are given for a number of significant figures and compared with those already available in the literature. (orig.) [de

  3. Inter-atomic force constants of BaF{sub 2} by diffuse neutron scattering measurement

    Energy Technology Data Exchange (ETDEWEB)

    Sakuma, Takashi, E-mail: sakuma@mx.ibaraki.ac.jp; Makhsun,; Sakai, Ryutaro [Institute of Applied Beam Science, Ibaraki University, Mito 310-8512 (Japan); Xianglian [College of Physics and Electronics Information, Inner Mongolia University for the Nationalities, Tongliao 028043 (China); Takahashi, Haruyuki [Institute of Applied Beam Science, Ibaraki University, Hitachi 316-8511 (Japan); Basar, Khairul [Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Bandung 40132 (Indonesia); Igawa, Naoki [Quantum Beam Science Directorate, Japan Atomic Energy Agency, Tokai 319-1195 (Japan); Danilkin, Sergey A. [Bragg Institute, Australian Nuclear Science and Technology Organisation, Kirrawee DC NSW 2232 (Australia)

    2015-04-16

    Diffuse neutron scattering measurement on BaF{sub 2} crystals was performed at 10 K and 295 K. Oscillatory form in the diffuse scattering intensity of BaF{sub 2} was observed at 295 K. The correlation effects among thermal displacements of F-F atoms were obtained from the analysis of oscillatory diffuse scattering intensity. The force constants among neighboring atoms in BaF{sub 2} were determined and compared to those in ionic crystals and semiconductors.

  4. 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

  5. Analytical approximation of neutron physics data

    International Nuclear Information System (INIS)

    Badikov, S.A.; Vinogradov, V.A.; Gaj, E.V.; Rabotnov, N.S.

    1984-01-01

    The method for experimental neutron-physical data analytical approximation by rational functions based on the Pade approximation is suggested. It is shown that the existence of the Pade approximation specific properties in polar zones is an extremely favourable analytical property essentially extending the convergence range and increasing its rate as compared with polynomial approximation. The Pade approximation is the particularly natural instrument for resonance curve processing as the resonances conform to the complex poles of the approximant. But even in a general case analytical representation of the data in this form is convenient and compact. Thus representation of the data on the neutron threshold reaction cross sections (BOSPOR constant library) in the form of rational functions lead to approximately twenty fold reduction of the storaged numerical information as compared with the by-point calculation at the same accWracy

  6. 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)

  7. Analytical calculations of neutron slowing down and transport in the constant-cross-section problem

    International Nuclear Information System (INIS)

    Cacuci, D.G.

    1978-01-01

    Some aspects of the problem of neutron slowing down and transport in an infinite medium consisting of a single nuclide that scatters elastically and isotropically and has energy-independent cross sections were investigated. The method of singular eigenfunctions was applied to the Boltzmann equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. A new sufficient condition for the convergence of the coefficients of the expansion of the scattering kernel in Legendre polynomials was rigorously derived for this energy-dependent problem. Formulas were obtained for the lethargy-dependent spatial moments of the scalar flux that are valid for medium to large lethargies. In deriving these formulas, use was made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations were greatly aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use was also made of the methods of combinatorial analysis and of computer evaluation, via FORMAC, of complicated sequences of manipulations. For the case of no absorption it was possible to obtain for materials of any atomic weight explicit corrections to the age-theory formulas for the spatial moments M/sub 2n/(u) of the scalar flux that are valid through terms of the order of u -5 . The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, is one of the end products of this investigation. In addition, an exact expression for the second spatial moment, M 2 (u), valid for arbitrary (constant) absorption, was derived. It is now possible to calculate analytically and rigorously the ''age'' for the constant-cross-section problem for arbitrary (constant) absorption and nuclear mass. 5 figures, 1 table

  8. Analytical calculations of neutron slowing down and transport in the constant-cross-section problem

    International Nuclear Information System (INIS)

    Cacuci, D.G.

    1978-04-01

    Aspects of the problem of neutron slowing down and transport in an infinite medium consisting of a single nuclide that scatters elastically and isotropically and has energy-independent cross sections were investigated. The method of singular eigenfunctions was applied to the Boltzmann Equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. A new sufficient condition for the convergence of the coefficients of the expansion of the scattering kernel in Legendre polynomials was rigorously derived for this energy-dependent problem. Formulas were obtained for the lethargy-dependent spatial moments of the scalar flux that are valid for medium to large lethargies. Use was made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations were aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use was also made of the methods of combinatorial analysis and of computer evaluation of complicated sequences of manipulations. For the case of no absorption it was possible to obtain for materials of any atomic weight explicit corrections to the age-theory formulas for the spatial moments M/sub 2n/(u) of the scalar flux that are valid through terms of the order of u -5 . The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, represent one of the end products of this investigation. In addition, an exact expression for the second spatial moment, M 2 (u), valid for arbitrary (constant) absorption, was derived. It is now possible to calculate analytically and rigorously the ''age'' for the constant-cross-section problem for arbitrary (constant) absorption and nuclear mass. 5 figures, 1 table

  9. Analytical calculations of neutron slowing down and transport in the constant-cross-section problem

    Energy Technology Data Exchange (ETDEWEB)

    Cacuci, D.G.

    1978-04-01

    Aspects of the problem of neutron slowing down and transport in an infinite medium consisting of a single nuclide that scatters elastically and isotropically and has energy-independent cross sections were investigated. The method of singular eigenfunctions was applied to the Boltzmann Equation governing the Laplace transform (with respect to the lethargy variable) of the neutron flux. A new sufficient condition for the convergence of the coefficients of the expansion of the scattering kernel in Legendre polynomials was rigorously derived for this energy-dependent problem. Formulas were obtained for the lethargy-dependent spatial moments of the scalar flux that are valid for medium to large lethargies. Use was made of the well-known connection between the spatial moments of the Laplace-transformed scalar flux and the moments of the flux in the ''eigenvalue space.'' The calculations were aided by the construction of a closed general expression for these ''eigenvalue space'' moments. Extensive use was also made of the methods of combinatorial analysis and of computer evaluation of complicated sequences of manipulations. For the case of no absorption it was possible to obtain for materials of any atomic weight explicit corrections to the age-theory formulas for the spatial moments M/sub 2n/(u) of the scalar flux that are valid through terms of the order of u/sup -5/. The evaluation of the coefficients of the powers of n, as explicit functions of the nuclear mass, represent one of the end products of this investigation. In addition, an exact expression for the second spatial moment, M/sub 2/(u), valid for arbitrary (constant) absorption, was derived. It is now possible to calculate analytically and rigorously the ''age'' for the constant-cross-section problem for arbitrary (constant) absorption and nuclear mass. 5 figures, 1 table.

  10. 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)

  11. 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

  12. Procedure to Generate the MPACT Multigroup Library

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-17

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

  13. 8-group relative delayed neutron yields for epithermal neutron induced fission of 235U and 239Pu

    International Nuclear Information System (INIS)

    Piksaikin, V.M.; Kazakov, L.E.; Isaev, S.G.; Korolev, G.G.; Roshchenko, V.A.; Tertychnyj, R.G

    2002-01-01

    An 8-group representation of relative delayed neutron yields was obtained for epithermal neutron induced fission of 235 U and 239 Pu. These data were compared with ENDF/B-VI data in terms of the average half- life of the delayed neutron precursors and on the basis of the dependence of reactivity on the asymptotic period. (author)

  14. Benchmarking of the 99-group ANSL-V library

    International Nuclear Information System (INIS)

    Wright, R.Q.; Ford, W.E. III; Greene, N.M.; Petrie, L.M.; Primm, R.T. III; Westfall, R.M.

    1987-01-01

    The purpose of this paper is to present thermal benchmark data testing results for the BAPL-1, TRX-1, and SEEP-1 lattices, using selected processed cross-sections from the ANSL-V 99-group library. 7 refs., 1 tab

  15. 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)

  16. Integrated system for production of neutronics and photonics calculational constants. Supplemental neutron-induced interactions (Z less than or equal to 35): graphical, experimental data

    International Nuclear Information System (INIS)

    Cullen, D.E.; Howerton, R.J.; MacGregor, M.H.; Perkins, S.T.

    1976-01-01

    This report (Vol. 8) presents graphs of supplemental neutron-induced cross sections in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976. It consists of interactions where more than one data set is needed to show cross-section behavior. In contrast, Vol. 7 of this UCRL-50400 series consists primarily of interactions where a single data set contains enough points to show cross-section behavior. In Vol. 7 can be found the total, elastic, capture, and fission cross sections (along with the parameters anti ν, α, and eta). Volume 8 contains all other reactions. Data are plotted with associated cross-section error bars (when given) and compared with the Evaluated Nuclear Data Library (ENDL) as of July 4, 1976. The plots are arranged in ascending order of atomic number (Z) and atomic weight (A). Part A contains the plots for Z = 1 to 35; Part B contains the plots for Z greater than 35

  17. Integrated system for production of neutronics and photonics calculational constants. Major neutron-induced interactions (Z less than or equal to 55): graphical, experimental data

    International Nuclear Information System (INIS)

    Cullen, D.E.; Howerton, R.J.; MacGregor, M.H.; Perkins, S.T.

    1976-01-01

    This report (vol. 7) presents graphs of major neutron-induced interaction cross sections in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976. It consists primarily of interactions where a single data set contains enough points to show cross-section behavior. In contrast, vol. 8 of this UCRL-50400 series consists of interactions where more than one data set is needed to show cross section behavior. Thus, you can find the total, elastic, capture, and fission cross sections (along with the parameters anti ν, α, and eta) in vol. 7 and all other reactions in vol. 8. Data are plotted with associated cross section error bars (when given) and compared with the Evaluated Nuclear Data Library (ENDL) as of July 4, 1976. The plots are arranged in ascending order of atomic number (Z) and atomic weight (A). Part A contains the plots for Z = 1 to 55; Part B contains the plots for Z greater than 55

  18. Integrated system for production of neutronics and photonics calculational constants. Major neutron-induced interactions (Z less than or equal to 55): graphical, experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Cullen, D.E.; Howerton, R.J.; MacGregor, M.H.; Perkins, S.T.

    1976-07-04

    This report (vol. 7) presents graphs of major neutron-induced interaction cross sections in the Experimental Cross Section Information Library (ECSIL) as of July 4, 1976. It consists primarily of interactions where a single data set contains enough points to show cross-section behavior. In contrast, vol. 8 of this UCRL-50400 series consists of interactions where more than one data set is needed to show cross section behavior. Thus, you can find the total, elastic, capture, and fission cross sections (along with the parameters anti ..nu.., ..cap alpha.., and eta) in vol. 7 and all other reactions in vol. 8. Data are plotted with associated cross section error bars (when given) and compared with the Evaluated Nuclear Data Library (ENDL) as of July 4, 1976. The plots are arranged in ascending order of atomic number (Z) and atomic weight (A). Part A contains the plots for Z = 1 to 55; Part B contains the plots for Z greater than 55.

  19. Calculation of neutron and gamma-ray emission spectra produced by p +2''2'Al reactions

    International Nuclear Information System (INIS)

    Arthur, E.D.

    1985-01-01

    As a contribution to the US/Japan cooperative program in fusion neutronics, we have prepared a library of multigroup neutron cross sections, scattering matrices, and covariances (uncertainties and their correlations). This 74-group library, called COVFILS-2, is being used at Los Alamos and at the University of California at Los Angeles in the sensitivity and uncertainty analysis of the Li 2 O integral experiment recently performed at the Fast Neutron Source (FNS) in Japan. Another intended use of this library is in the estimation of the uncertainty in key performance parameters (such as breeding ratio) of conceptual fusion reactors. The 14 materials included in the first version of COVFILS-2 are H, 6 Li, 7 Li, Be, C, N, O, Na, Al, Si, Cr, Fe, Ni, and Pb

  20. 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)

  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. Information about AER WG A on improvement, extension and validation of parametrized few-group libraries for VVER 440 and VVER 1000

    International Nuclear Information System (INIS)

    Mikolas, P.

    2010-01-01

    Joint AER Working Group A on 'Improvement, extension and validation of parameterized few-group libraries for WWER-440 and WWER-1000' and AER Working Group B on 'Core design' nineteenth meeting was hosted by VUJE a. s. in Modra - Harmonia (Slovakia) during the period of 20. to 22. April 2010. There were present altogether 12 participants from 8 member organizations and 9 papers were presented (8 of them in written form). Objectives of the meeting of WG A are: Issues connected with spectral calculations and few-groups libraries preparation, their accuracy and validation. Presentations were devoted to some aspects of transport and diffusion calculations and to the benchmark dealing with WWER-1000 core periphery power tilt. Tamas Parko (co-authors Istvan Pos and Sandor Patai Szabo) described in his presentation 'Application of Discontinuity factors in C-PORCA 7 code', Radoslav Zajac (co-authors Petr Darilek and Vladimir Necas) spoke about 'Fast Reactor Nodalisation in HELIOS Code', Gabriel Farkas presented 'Calculation of Spatial Weighting Functions of Ex-Core Neutron Detectors for WWER-440 Using Monte Carlo Approach' and Daniel Sprinzl (co-authors Vaclav Krysl, Pavel Mikolas and Jiri Svarny) provided a definition of a benchmark in ' 'MIDICORE' WWER-1000 core periphery power tilt benchmark proposal'. (Author)

  3. A group of neutronics calculations in the MNSR using the MCNP-4C code

    International Nuclear Information System (INIS)

    Khattab, K.; Sulieman, I.

    2009-11-01

    The MCNP-4C code was used to model the 3-D core configuration for the Syrian Miniature Neutron Source Reactor (MNSR). The continuous energy neutron cross sections were evaluated from ENDF/B-VI library to calculate the thermal and fast neutron fluxes in the MNSR inner and outer irradiation sites. The thermal fluxes in the MNSR inner irradiation sites were measured for the first time using the multiple foil activation method. Good agreements were noticed between the calculated and measured results. This model is used as well to calculate neutron flux spectrum in the reactor inner and outer irradiation sites and the reactor thermal power. Three 3-D neutronic models for the Syrian MNSR reactor using the MCNP-4C code were developed also to assess the possibility of fuel conversion from 89.87 % HEU fuel (UAl 4 -Al) to 19.75 % LEU fuel (UO 2 ). This model is used in this paper to calculate the following reactor core physics parameters: clean cold core excess reactivity, calibration of the control rod worth and calculation its shut down margin, calibration of the top beryllium shim plate reflector, axial neutron flux distributions in the inner and outer irradiation sites and the kinetics parameters ( ι p l and β e ff). (authors)

  4. Capture Gamma-Ray Libraries for Nuclear Applications

    International Nuclear Information System (INIS)

    Sleaford, B.W.; Firestone, Richard B.; Summers, N.; Escher, J.; Hurst, A.; Krticka, M.; Basunia, S.; Molnar, G.; Belgya, T.; Revay, Z.; Choi, H.D.

    2010-01-01

    The neutron capture reaction is useful in identifying and analyzing the gamma-ray spectrum from an unknown assembly as it gives unambiguous information on its composition. This can be done passively or actively where an external neutron source is used to probe an unknown assembly. There are known capture gamma-ray data gaps in the ENDF libraries used by transport codes for various nuclear applications. The Evaluated Gamma-ray Activation file (EGAF) is a new thermal neutron capture database of discrete line spectra and cross sections for over 260 isotopes that was developed as part of an IAEA Coordinated Research Project. EGAF has been used to improve the capture gamma production in ENDF libraries. For medium to heavy nuclei the quasi continuum contribution to the gamma cascades is not experimentally resolved. The continuum contains up to 90% of all the decay energy an is modeled here with the statistical nuclear structure code DICEBOX. This code also provides a consistency check of the level scheme nuclear structure evaluation. The calculated continuum is of sufficient accuracy to include in the ENDF libraries. This analysis also determines new total thermal capture cross sections and provides an improved RIPL database. For higher energy neutron capture there is less experimental data available making benchmarking of the modeling codes more difficult. We use CASINO, a version of DICEBOX that is modified for this purpose. This can be used to simulate the neutron capture at incident neutron energies up to 20 MeV to improve the gamma-ray spectrum in neutron data libraries used for transport modelling of unknown assemblies.

  5. JEF-2.2. The evaluated neutron nuclear data library of the NEA Data Bank. Summary of contents

    International Nuclear Information System (INIS)

    Lemmel, H.D.; Schwerer, O.

    1996-01-01

    This document summarizes the contents of JEF-2.2, the Joint Evaluated File of neutron nuclear data compiled by the OECD Nuclear Energy Agency Data Bank, finalized in 1992 and released in 1993. The entire library or retrievals of selected materials are available from the IAEA Nuclear Data Section free of charge, either on magnetic tape, or online from NDIS, the interactive Nuclear Data Information System. (author)

  6. 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)

  7. KAERI photonuclear library

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jong Hwa; Lee, Young Ouk; Han, Yin Iu

    2000-03-01

    This report contains summary information and figures depicting the KAERI photonuclear data library that extends up to 140 MeV of incident photon. The library consists of 143 isotopes from C-12 to Bi-209, providing the photoabsorption cross section and the emission spectra for neutron, proton, deuteron, triton, alpha particles, and all residual nuclides in ENDF6 format. The contents of this report and ENDF-6 format data library are available at http://atom.kaeri.re.kr/.

  8. EASY-2005.1, European Neutron Activation System

    International Nuclear Information System (INIS)

    Forrest, R.A.; Sublet, Jean-Christophe

    2008-01-01

    that can use the deuteron library. The EAF-2005 data library covers: - Cross section data for neutron-induced reactions; - Uncertainty data for neutron-induced reactions; - Decay data; - Fission yield data; - Biological hazard data; - Legal transport data; - Clearance data; - Charged particle ranges in materials; - Emitted particle spectral data (from neutron-induced reactions); - Charged particle cross section data; - Gamma absorption data. The EAF-2005 contains the following data libraries: EAF-XS-2005 is the point-wise cross section library. EAF-GXS-2005 is a set of multi-group libraries; the available group structures are: WIMS (69), GAM-II (100), XMAS (172), VITAMIN-J (175), VITAMIN-J+ (211), TRIPOLI (315) and TRIPOLI+ (351). EAF-UN-2005 contains uncertainty data for all cross sections. EAF-DEC-2005 contains decay data information for 2192 nuclides, it is based primarily on the JEF-2.2 radioactive decay data library [9], with additional data from recent UK evaluations. EAF-FIS-2005 is taken completely from the JEF-2.2 fission yield library. It is a library of independent fission yields in ENDF/B-6 format. EAF-HAZ-2005 is a library of values for each radionuclide describing the potential biological impact of that nuclide on human beings. EAF-A2-2005 is a library of values relevant for the transport of radionuclides in shielded flasks. EAF-STOP-2005 contains new data compared to the previous versions. EAF-STOP contains the differential ranges for protons (p), deuterons (d), hellions (h), tritons (t) and alpha particles in all the elements from H to U. EAF-SPEC-2005: EAF-XS contains the cross section data for (n,x) reactions, while EAF-SPEC describes the energy distribution of the charged particles emitted in these reactions. EAF-XN-2005 describes the cross sections of p, d, h, t and alpha particles on 775 targets. EAF-ABS contains the photon mass energy attenuation coefficient for all the elements Z= 1 - 100 in increasing Z order. 2 - Methods: FISPACT is an

  9. Developments in capture-γ libraries for nonproliferation applications

    Science.gov (United States)

    Hurst, A. M.; Firestone, R. B.; Sleaford, B. W.; Bleuel, D. L.; Basunia, M. S.; Bečvář, F.; Belgya, T.; Bernstein, L. A.; Carroll, J. J.; Detwiler, B.; Escher, J. E.; Genreith, C.; Goldblum, B. L.; Krtička, M.; Lerch, A. G.; Matters, D. A.; McClory, J. W.; McHale, S. R.; Révay, Zs.; Szentmiklosi, L.; Turkoglu, D.; Ureche, A.; Vujic, J.

    2017-09-01

    The neutron-capture reaction is fundamental for identifying and analyzing the γ-ray spectrum from an unknown assembly because it provides unambiguous information on the neutron-absorbing isotopes. Nondestructive-assay applications may exploit this phenomenon passively, for example, in the presence of spontaneous-fission neutrons, or actively where an external neutron source is used as a probe. There are known gaps in the Evaluated Nuclear Data File libraries corresponding to neutron-capture γ-ray data that otherwise limit transport-modeling applications. In this work, we describe how new thermal neutron-capture data are being used to improve information in the neutron-data libraries for isotopes relevant to nonproliferation applications. We address this problem by providing new experimentally-deduced partial and total neutron-capture reaction cross sections and then evaluate these data by comparison with statistical-model calculations.

  10. Semi-empirical neutron tool calibration (one and two-group approximation)

    International Nuclear Information System (INIS)

    Czubek, J.A.

    1988-01-01

    The physical principles of the new method of calibration of neutron tools for the rock porosity determination are given. A short description of the physics of neutron transport in the matter is presented together with some remarks on the elementary interactions of neutrons with nuclei (cross sections, group cross sections etc.). The definitions of the main integral parameters characterizing the neutron transport in the rock media are given. The three main approaches to the calibration problem: empirical, theoretical and semi-empirical are presented with some more detailed description of the latter one. The new semi-empirical approach is described. The method is based on the definition of the apparent slowing down or migration length for neutrons sensed by the neutron tool situated in the real borehole-rock conditions. To calculate this apparent slowing down or migration lengths the ratio of the proper space moments of the neutron distribution along the borehole axis is used. Theoretical results are given for one- and two-group diffusion approximations in the rock-borehole geometrical conditions when the tool is in the sidewall position. The physical and chemical parameters are given for the calibration blocks of the Logging Company in Zielona Gora. Using these data the neutron parameters of the calibration blocks have been calculated. An example, how to determine the calibration curve for the dual detector tool applying this new method and using the neutron parameters mentioned above together with the measurements performed in the calibration blocks, is given. The most important advantage of the new semi-empirical method of calibration is the possibility of setting on the unique calibration curve all experimental calibration data obtained for a given neutron tool for different porosities, lithologies and borehole diameters. 52 refs., 21 figs., 21 tabs. (author)

  11. Revision of fast reactor group constant set JFS-3-J2

    International Nuclear Information System (INIS)

    Takano, Hideki; Kaneko, Kunio.

    1989-10-01

    To improve the fast reactor group constant set JFS-3-J2 to be applicable for high burnup reactor calculations, group constants for 155 fission product nuclides and the lumped group cross sections for four mother fission isotopes of U-235, U-238, Pu-239 and Pu-241 have been generated. Furthermore, the group constants for higher actinides such as Am and Cm have been produced on the basis of the JENDL-2 nuclear data, so as to be able to use for TRU-transmutation calculations. Benchmark test of this revised set has been performed by analysing the 21 fast critical experimental assemblies. Benchmark calculation system based on one-dimensional Sn-method has been developed to investigate the accuracy of one-dimensional diffusion calculations. Significant difference between the results obtained with the diffusion and transport calculations was observed for small cores and the assemblies with iron or nickel reflector. (author)

  12. Neutron density decay constant in a non-multiplying lattice of finite size

    International Nuclear Information System (INIS)

    Deniz, V.C.

    1965-01-01

    This report presents a general theory, using the integral transport method, for obtaining the neutron density decay constant in a finite non-multiplying lattice. The theory is applied to obtain the expression for the diffusion coefficient. The case of a homogeneous medium with 1/v absorption and of finite size in all directions is treated in detail, assuming an isotropic scattering law. The decay constant is obtained up to the B 6 term. The expressions for the diffusion coefficient and for the diffusion cooling coefficient are the same as those obtained for a slab geometry by Nelkin, using the expansion in spherical harmonics of the Fourier transform in the spatial variable. Furthermore, explicit forms are obtained for the flux and the current. It is shown that the deviation of the actual flux from a Maxwellian is the flux generated in the medium, extended to infinity and deprived of its absorbing power, by various sources, each of which has a zero integral over all velocities. The study of the current permits the generalization of Fick's law. An independent integral method, valid for homogeneous media, is also presented. (author) [fr

  13. CASMO5 JENDL-4.0 and ENDF/B-VII.1beta4 libraries

    International Nuclear Information System (INIS)

    Rhodes, J.; Gheorghiu, N.; Ferrer, R.

    2012-01-01

    This paper details the generation of neutron data libraries for the CASMO5 lattice physics code based on the recently released JENDL-4.0 and ENDF/B-VII.1beta4 nuclear data evaluations. This data represents state-of-the-art nuclear data for late-2011. The key features of the new evaluations are briefly described along with the procedure for processing of this data into CASMO5, 586-energy group neutron data libraries. Finally some CASMO5 results for standard UO 2 and MOX critical experiments for the two new libraries and the current ENDF/B-VII.0 CASMO5 library are presented including the B and W 1810 series, DIMPLE S06A, S06B, TCA reflector criticals with iron plates and the PNL-30-35 MOX criticals. The results show that CASMO5 with the new libraries is performing well for these criticals with a very slight edge in results to the JENDL-4.0 nuclear data evaluation over the ENDF/B-VII.1beta4 evaluation. Work is currently underway to generate a CASMO5 library based on the final ENDF/B-VII.R1 evaluation released Dec. 22, 2011. (authors)

  14. 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)

  15. A multi-group neutron noise simulator for fast reactors

    International Nuclear Information System (INIS)

    Tran, Hoai Nam; Zylbersztejn, Florian; Demazière, Christophe; Jammes, Christian; Filliatre, Philippe

    2013-01-01

    Highlights: • The development of a neutron noise simulator for fast reactors. • The noise equation is solved fully in a frequency-domain. • A good agreement with ERANOS on the static calculations. • Noise calculations induced by a localized perturbation of absorption cross section. - Abstract: A neutron noise simulator has been developed for fast reactors based on diffusion theory with multi-energy groups and several groups of delayed neutron precursors. The tool is expected to be applicable for core monitoring of fast reactors and also for other reactor types with hexagonal fuel assemblies. The noise sources are modeled through small stationary fluctuations of macroscopic cross sections, and the induced first order noise is solved fully in the frequency domain. Numerical algorithms are implemented for solving both the static and noise equations using finite differences for spatial discretization, where a hexagonal assembly is radially divided into finer triangular meshes. A coarse mesh finite difference (CMFD) acceleration has been used for accelerating the convergence of both the static and noise calculations. Numerical calculations have been performed for the ESFR core with 33 energy groups and 8 groups of delayed neutron precursors using the cross section data generated by the ERANOS code. The results of the static state have been compared with those obtained using ERANOS. The results show an adequate agreement between the two calculations. Noise calculations for the ESFR core have also been performed and demonstrated with an assumption of the perturbation of the absorption cross section located at the central fuel ring

  16. 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

  17. 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)

  18. ZZ AIRFEWG, Gamma, Neutron Transport Calculation in Air Using FEWG1 Cross-Section

    International Nuclear Information System (INIS)

    1985-01-01

    1 - Description of program or function: Format: ANISN; Number of groups: 37 neutron / 21 gamma-ray; Nuclides: air (79% N and 21% O); Origin: DLC-0031/FEWG1 cross sections (ENDF/B-IV). Weighting spectrum: 1/E. The AIRFEWG library has been generated by an ANISN multigroup calculation of gamma-ray, neutron, and secondary gamma-ray transport in infinite homogeneous air using DLC-0031/FEWG1 cross sections. 2 - Method of solution: The results were generated with a P3, ANISN run with a source in a single energy group. Thus, 58 such runs were required. For sources in the 37 neutron groups, both neutron and secondary gamma-ray fluence results were calculated. For gamma-ray sources only gamma-ray fluences were calculated

  19. Derivation of Inter-Atomic Force Constants of Cu2O from Diffuse Neutron Scattering Measurement

    Directory of Open Access Journals (Sweden)

    T. Makhsun

    2013-04-01

    Full Text Available Neutron scattering intensity from Cu2O compound has been measured at 10 K and 295 K with High Resolution Powder Diffractometer at JRR-3 JAEA. The oscillatory diffuse scattering related to correlations among thermal displacements of atoms was observed at 295 K. The correlation parameters were determined from the observed diffuse scattering intensity at 10 and 295 K. The force constants between the neighboring atoms in Cu2O were estimated from the correlation parameters and compared to those of Ag2O

  20. Charge dependence of the pion-nucleon coupling constant

    Directory of Open Access Journals (Sweden)

    V. A. Babenko

    2015-07-01

    Full Text Available On the basis of the Yukawa potential we study the pion-nucleon coupling constants for the neutral and charged pions assuming that nuclear forces at low energies are mainly determined by the exchange of virtual pions. We obtain the charged pseudovector pion-nucleon coupling constant f2π± = 0.0804(7 by making the use of experimental low-energy scattering parameters for the singlet pp- and np-scattering, and also by use of the neutral pseudovector pion-nucleon coupling constant f2π0 = 0.0749(7. Corresponding value of the charged pseudoscalar pion-nucleon coupling constant g2π0 / 4π = 14.55(13 is also determined. This calculated value of the charged pseudoscalar pion-nucleon coupling constant is in fully agreement with the experimental constant g2π0 / 4π = 14.52(26 obtained by the Uppsala Neutron Research Group. Our results show considerable charge splitting of the pion-nucleon coupling constant.

  1. Neutronics model of the bulk shielding reactor (BSR): validation by comparison of calculations with the experimental measurements

    International Nuclear Information System (INIS)

    Johnson, J.O.; Miller, L.F.; Kam, F.B.K.

    1981-05-01

    A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with 58 Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR

  2. Neutronics model of the bulk shielding reactor (BSR): validation by comparison of calculations with the experimental measurements

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.O.; Miller, L.F.; Kam, F.B.K.

    1981-05-01

    A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with /sup 58/Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR.

  3. Progress on Nuclear Data Covariances: AFCI-1.2 Covariance Library

    International Nuclear Information System (INIS)

    Oblozinsky, P.; Oblozinsky, P.; Mattoon, C.M.; Herman, M.; Mughabghab, S.F.; Pigni, M.T.; Talou, P.; Hale, G.M.; Kahler, A.C.; Kawano, T.; Little, R.C.; Young, P.G

    2009-01-01

    Improved neutron cross section covariances were produced for 110 materials including 12 light nuclei (coolants and moderators), 78 structural materials and fission products, and 20 actinides. Improved covariances were organized into AFCI-1.2 covariance library in 33-energy groups, from 10 -5 eV to 19.6 MeV. BNL contributed improved covariance data for the following materials: 23 Na and 55 Mn where more detailed evaluation was done; improvements in major structural materials 52 Cr, 56 Fe and 58 Ni; improved estimates for remaining structural materials and fission products; improved covariances for 14 minor actinides, and estimates of mubar covariances for 23 Na and 56 Fe. LANL contributed improved covariance data for 235 U and 239 Pu including prompt neutron fission spectra and completely new evaluation for 240 Pu. New R-matrix evaluation for 16 O including mubar covariances is under completion. BNL assembled the library and performed basic testing using improved procedures including inspection of uncertainty and correlation plots for each material. The AFCI-1.2 library was released to ANL and INL in August 2009.

  4. The LIPAR-5 resonance parameter library

    International Nuclear Information System (INIS)

    Abagyan, L.P.

    1997-08-01

    The LIPAR-5 neutron resolved resonance parameter library has been elaborated. It contains data for 94 isotopes. The author's evaluations are included in LIPAR. Other authors' results are also included after re-evaluation. The codes used for the evaluation are described briefly. Tables of results are included for every isotope: the boundaries of the resolved resonance region, the numbers of s- and p-resonances, the thermal neutron partial cross-sections and the resonance integrals. The parameters are presented in ENDF/B-6 format. LIPAR is part of the nuclear data library of the MCU Monte Carlo code for neutron transport calculations. LIPAR was verified by comparing the benchmark experiment and Monte Carlo calculation results. (author). 44 refs, 6 tabs

  5. Neutronic study of spherical cold-neutron sources composed of liquid hydrogen and liquid deuterium

    CERN Document Server

    Matsuo, Y; Nagaya, Y

    2003-01-01

    Using the cross-section model for neutron scattering in liquid H sub 2 and D sub 2 , a neutron transport analysis is performed for spherical cold-neutron sources composed of either para H sub 2 , normal H sub 2 or normal D sub 2. A special effort is made to generate a set of energy-averaged cross-sections (80 group constants between 0.1 mu eV and 10 eV) for liquid H sub 2 and D sub 2 at melting and boiling points. A number of conclusions on the spherical cold-neutron source configurations are drawn. It is especially shown that the highest cold-neutron flux is obtainable from the normal D sub 2 source with a radius of about 50 cm, while the normal- and para-H sub 2 sources with radii around 3-4 cm produce maximum cold-neutron fluxes at the center.

  6. AUS, Neutron Transport and Gamma Transport System for Fission Reactors and Fusion Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    1 - Description of program or function: AUS is a neutronics code system which may be used for calculations of a wide range of fission reactors, fusion blankets and other neutron applications. The present version, AUS98, has a nuclear cross section library based on ENDF/B-VI and includes modules which provide for reactor lattice calculations, one-dimensional transport calculations, multi-dimensional diffusion calculations, cell and whole reactor burnup calculations, and flexible editing of results. Calculations of multi-region resonance shielding, coupled neutron and photon transport, energy deposition, fission product inventory and neutron diffusion are combined within the one code system. The major changes from the previous release, AUS87, are the inclusion of a cross-section library based on ENDF/B-VI, the addition of the POW3D multi-dimensional diffusion module, the addition of the MICBURN module for controlling whole reactor burnup calculations, and changes to the system as a consequence of moving from IBM mainframe computers to UNIX workstations. 2 - Method of solution: AUS98 is a modular system in which the modules are complete programs linked by a path given in the input stream. A simple path is simply a sequence of modules, but the path is actually pre-processed and compiled using the Fortran 77 compiler. This provides for complex module linking if required. Some of the modules included in AUS98 are: MIRANDA Cross-section generation in a multi-region resonance subgroup calculation and preliminary group condensation. ANAUSN One-dimensional discrete ordinates calculation. ICPP Isotropic collision probability calculation in one dimension and for rod clusters. POW3D Multi-dimensional neutron diffusion calculation including feedback-free kinetics. AUSIDD One-dimensional diffusion calculation. EDITAR Reaction-rate editing and group collapsing following a transport calculation. CHAR Lattice and global burnup calculation. MICBURN Control of global burnup

  7. New ENDF/B-V nuclear data library for WIMS-D4M

    International Nuclear Information System (INIS)

    Deen, J.R.; Woodruff, W.L.; Costescu, C.I.

    1994-01-01

    A new 69-group 96-material library has been created for use in WIMS-D4M. The latest SUN version of NJOY (91.27) was used to generate the ENDF/B-V-based cross-section library. The library also includes ENDF/B-V based fission yields, energy fission and energy per capture data. The upper energy boundary has been extended from 10 to 20 MeV in order to model high energy neutron reactions. Additional fuel and moderator temperatures have been included to better predict temperature coefficients. More excess potential scattering points have been added to increase the accuracy of self-shielded resonance cross-sections. Several benchmark comparisons have been made to validate the new library. (author)

  8. Up-date of the BCG code library

    International Nuclear Information System (INIS)

    Caldeira, A.D.; Garcia, R.D.M.

    1990-01-01

    Procedures for generating an up-date material library for the BCG code were established. A new library was generated by processing ENDF/B-IV data with the 89-1 version of the LINEAR, RECENT and SIGMA1 programs. The effect of library change in the neutron spectrum and effective multiplication factor of a fast reactor cell was analized. During the course of this study, an error was detected in the BCG code. Although localized in a narrow energy range, the discrepancies in neutron spectrum caused by the error were large enough to yield a difference of about 1% in the effective multiplication factor of the test cell. (author)

  9. ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Chadwick, M B; Oblozinsky, P; Herman, M; Greene, N M; McKnight, R D; Smith, D L; Young, P G; MacFarlane, R E; Hale, G M; Haight, R C; Frankle, S; Kahler, A C; Kawano, T; Little, R C; Madland, D G; Moller, P; Mosteller, R; Page, P; Talou, P; Trellue, H; White, M; Wilson, W B; Arcilla, R; Dunford, C L; Mughabghab, S F; Pritychenko, B; Rochman, D; Sonzogni, A A; Lubitz, C; Trumbull, T H; Weinman, J; Brown, D; Cullen, D E; Heinrichs, D; McNabb, D; Derrien, H; Dunn, M; Larson, N M; Leal, L C; Carlson, A D; Block, R C; Briggs, B; Cheng, E; Huria, H; Kozier, K; Courcelle, A; Pronyaev, V; der Marck, S

    2006-10-02

    We describe the next generation general purpose Evaluated Nuclear Data File, ENDF/B-VII.0, of recommended nuclear data for advanced nuclear science and technology applications. The library, released by the U.S. Cross Section Evaluation Working Group (CSEWG) in December 2006, contains data primarily for reactions with incident neutrons, protons, and photons on almost 400 isotopes. The new evaluations are based on both experimental data and nuclear reaction theory predictions. The principal advances over the previous ENDF/B-VI library are the following: (1) New cross sections for U, Pu, Th, Np and Am actinide isotopes, with improved performance in integral validation criticality and neutron transmission benchmark tests; (2) More precise standard cross sections for neutron reactions on H, {sup 6}Li, {sup 10}B, Au and for {sup 235,238}U fission, developed by a collaboration with the IAEA and the OECD/NEA Working Party on Evaluation Cooperation (WPEC); (3) Improved thermal neutron scattering; (4) An extensive set of neutron cross sections on fission products developed through a WPEC collaboration; (5) A large suite of photonuclear reactions; (6) Extension of many neutron- and proton-induced reactions up to an energy of 150 MeV; (7) Many new light nucleus neutron and proton reactions; (8) Post-fission beta-delayed photon decay spectra; (9) New radioactive decay data; and (10) New methods developed to provide uncertainties and covariances, together with covariance evaluations for some sample cases. The paper provides an overview of this library, consisting of 14 sublibraries in the same, ENDF-6 format, as the earlier ENDF/B-VI library. We describe each of the 14 sublibraries, focusing on neutron reactions. Extensive validation, using radiation transport codes to simulate measured critical assemblies, show major improvements: (a) The long-standing underprediction of low enriched U thermal assemblies is removed; (b) The {sup 238}U, {sup 208}Pb, and {sup 9}Be reflector

  10. New Beta-delayed Neutron Measurements in the Light-mass Fission Group

    Energy Technology Data Exchange (ETDEWEB)

    Agramunt, J. [Instituto de Física Corpuscular, CSIC-Univ. Valencia, Apdo. Correos 22085, E-46071 Valencia (Spain); García, A.R. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, E-28040 Madrid (Spain); Algora, A. [Instituto de Física Corpuscular, CSIC-Univ. Valencia, Apdo. Correos 22085, E-46071 Valencia (Spain); Äystö, J. [University of Jyväskylä, FI-40014 Jyväskyä (Finland); Caballero-Folch, R.; Calviño, F. [Secció d' Enginyeria Nuclear, Universitat Politécnica de Catalunya, E-08028 Barcelona (Spain); Cano-Ott, D. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, E-28040 Madrid (Spain); Cortés, G. [Secció d' Enginyeria Nuclear, Universitat Politécnica de Catalunya, E-08028 Barcelona (Spain); Domingo-Pardo, C. [Instituto de Física Corpuscular, CSIC-Univ. Valencia, Apdo. Correos 22085, E-46071 Valencia (Spain); Eronen, T. [University of Jyväskylä, FI-40014 Jyväskyä (Finland); Gelletly, W. [Department of Physics, University of Surrey, Guildford GU2 7XH (United Kingdom); Gómez-Hornillos, M.B. [Secció d' Enginyeria Nuclear, Universitat Politécnica de Catalunya, E-08028 Barcelona (Spain); and others

    2014-06-15

    A new accurate determination of beta-delayed neutron emission probabilities from nuclei in the low mass region of the light fission group has been performed. The measurements were carried out using the BELEN 4π neutron counter at the IGISOL-JYFL mass separator in combination with a Penning trap. The new results significantly improve the uncertainties of neutron emission probabilities for {sup 91}Br, {sup 86}As, {sup 85}As, and {sup 85}Ge nuclei.

  11. Testing of the ENDF/B-VI neutron data library ENDF60 for use with MCNP trademark

    International Nuclear Information System (INIS)

    Frankle, S.C.; MacFarlane, R.E.

    1995-01-01

    The continuous-energy neutron data library ENDF60, for use with the Monte Carlo N-Particle radiation transport code MCNP4A, was released in the fall of 1994. It is comprised of 124 nuclide data files based on the ENDF/B-Vi evaluations through Release 2. Forty-eight percent of these materials are new or modified evaluations, while the balance are translations from ENDF/B-V. The new evaluations include most of the important materials for criticality safety calculations, and include significant enhancements such as more isotopic evaluations, better resonance-range representations, and the new correlated energy-angle distributions for emitted particles. As part of the overall quality assurance testing of the ENDF60 library, calculations for well known benchmark assemblies were performed. The results of these calculations help the user to know how the combination of ENDF60 and MCNP4A will perform for real problems

  12. Development of the tool for generating ORIGEN2 library based on JENDL-3.2 for FBR

    International Nuclear Information System (INIS)

    Ohkawachi, Yasushi; Fukushima, Manabu

    1999-05-01

    ORIGEN2 is one of the most widely-used burnup analysis code in the world. This code has one-grouped cross section libraries compiled for various types of reactors. However, these libraries have some problems. One is that these libraries were developed from old nuclear data libraries (ENDF/B-IV,V) and the other is that core and fuel designs from which these libraries are generated do not match the current analysis. In order to solve the problems, analysis tool is developed for generating ORIGEN2 library from JENDL-3.2 considering multi-energy neutron spectrum. And eight new libraries are prepared using this tool for analysis of sodium-cooled FBR. These new libraries are prepared for eight kinds of cores in total. Seven of them are made by changing core size (small core - large core), fuel type (oxide, nitride, metal) and Pu vector as a parameter. The eighth one is a Pu burner core. Burnup calculation using both new and original libraries, shows large difference in buildup or depletion numbers of nuclides among the libraries. It is estimated that the analysis result is greatly influenced by the neutron spectrum which is used in collapse of cross section. By using this tool or new libraries, it seems to improve evaluation accuracy of buildup or depletion numbers of nuclides in transmutation research on FBR fuel cycle. (author)

  13. Evaluation of the criticality constant from Pulsed Neutron Source measurements in the Yalina-Booster subcritical assembly

    International Nuclear Information System (INIS)

    Bécares, V.; Villamarín, D.; Fernández-Ordóñez, M.; González-Romero, E.M.; Berglöf, C.; Bournos, V.; Fokov, Y.; Mazanik, S.; Serafimovich, I.

    2013-01-01

    Highlights: ► New methodology proposed to determine the reactivity of subcritical systems. ► Methodology tested in PNS experiments at the Yalina-Booster subcritical assembly. ► The area-ratio and the prompt decay constant methods have been used for validation. ► The absolute reactivity of the system is determined in spite of large spatial effects. - Abstract: The prompt decay constant method and the area-ratio (Sjöstrand) method constitute the reference techniques for measuring the reactivity of a subcritical system using Pulsed Neutron Source experiments (PNS). However, different experiments have shown that in many cases it is necessary to apply corrections to the experimental results in order to take into account spectral and spatial effects. In these cases, the approach usually followed is to develop different specific correction procedures for each method. In this work we discuss the validity of prompt decay constant method and the area-ratio method in the Yalina-Booster subcritical assembly and propose a general correction procedure based on Monte Carlo simulations

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

    International Nuclear Information System (INIS)

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

    1981-03-01

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

  15. 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

  16. Energy dependence of relative abundances and periods of delayed neutron separate groups from neutron induced fission of 239Pu in the virgin neutron energy range 0.37-4.97 MeV

    International Nuclear Information System (INIS)

    Piksajkin, V.M.; Kazakov, L.E.; Isaev, S.T.; Korolev, G.G.; Roshchenko, V.A.; Tertychnyj, R.G.

    2002-01-01

    Relative yield and group period of delayed neutrons induced by the 239 Pu fission in the 0.37-4.97 MeV range were measured. Comparative analysis of experimental data was conducted in terms of middle period of half-life of delayed neutron nuclei-precursors. Character and scale of changing values of delayed neutron group parameters as changing excitation energy of fission compound-nucleus have been demonstrated for the first time. Considerable energy dependence of group parameters under the neutron induced 239 Pu fission that was expressed by the decreasing middle period of half-life of nuclei-precursors by 10 % in the 2.85 eV - 5 MeV range of virgin neutrons was detected [ru

  17. Knowledge Management in the Neutronics Group of CAREM Project

    International Nuclear Information System (INIS)

    Torres, L.; Lopasso, E.

    2016-01-01

    Full text: An analysis of the Neutronics Group of CAREM25 project was performed in order to plan for the gradual implementation of knowledge management. The group structure, performed tasks and the way these tasks are linked together were studied. Staff functions within the group, profiles of each position and the training and education of human resources were also analyzed. (author

  18. Characteristic analysis on moderating material for obtaining epithermal neutron beam

    International Nuclear Information System (INIS)

    Jiang Xinbiao; Chen Da; Zhang Ying

    2000-01-01

    The one dimension discrete coordinates transport code ANISN was used to calculate three-group constants of 11 elements which could be used to consist moderating epithermal neutron material of beam. Moderating character of simple substances, compounds and mixtures consisted of the optimized elements analyzed three kinds of moderating materials were optimized for epithermal neutron beam

  19. 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)

  20. 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

  1. Evaluation of the Neutron Data Standards

    Science.gov (United States)

    Carlson, A. D.; Pronyaev, V. G.; Capote, R.; Hale, G. M.; Chen, Z.-P.; Duran, I.; Hambsch, F.-J.; Kunieda, S.; Mannhart, W.; Marcinkevicius, B.; Nelson, R. O.; Neudecker, D.; Noguere, G.; Paris, M.; Simakov, S. P.; Schillebeeckx, P.; Smith, D. L.; Tao, X.; Trkov, A.; Wallner, A.; Wang, W.

    2018-02-01

    With the need for improving existing nuclear data evaluations, (e.g., ENDF/B-VIII.0 and JEFF-3.3 releases) the first step was to evaluate the standards for use in such a library. This new standards evaluation made use of improved experimental data and some developments in the methodology of analysis and evaluation. In addition to the work on the traditional standards, this work produced the extension of some energy ranges and includes new reactions that are called reference cross sections. Since the effort extends beyond the traditional standards, it is called the neutron data standards evaluation. This international effort has produced new evaluations of the following cross section standards: the H(n,n), 6Li(n,t), 10B(n,α), 10B(n,α1 γ), natC(n,n), Au(n,γ), 235U(n,f) and 238U(n,f). Also in the evaluation process the 238U(n,γ) and 239Pu(n,f) cross sections that are not standards were evaluated. Evaluations were also obtained for data that are not traditional standards: the Maxwellian spectrum averaged cross section for the Au(n,γ) cross section at 30 keV; reference cross sections for prompt γ-ray production in fast neutron-induced reactions; reference cross sections for very high energy fission cross sections; the 252Cf spontaneous fission neutron spectrum and the 235U prompt fission neutron spectrum induced by thermal incident neutrons; and the thermal neutron constants. The data and covariance matrices of the uncertainties were obtained directly from the evaluation procedure.

  2. FENDL: International reference nuclear data library for fusion applications

    International Nuclear Information System (INIS)

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

    1996-01-01

    The IAEA nuclear data section, in co-operation with several national nuclear data centres and research groups, has created the first version of an internationally available fusion evaluated nuclear data library (FENDL-1). The FENDL library has been selected to serve as a comprehensive source of processed and tested nuclear data tailored to the requirements of the engineering design activity (EDA) of the ITER project and other fusion-related development projects. The present version of FENDL consists of the following sublibraries covering the necessary nuclear input for all physics and engineering aspects of the material development, design, operation and safety of the ITER project in its current EDA phase: FENDL/A-1.1: neutron activation cross-sections, selected from different available sources, for 636 nuclides, FENDL/D-1.0: nuclear decay data for 2900 nuclides in ENDF-6 format, FENDL/DS-1.0: neutron activation data for dosimetry by foil activation, FENDL/C-1.0: data for the fusion reactions D(d,n), D(d,p), T(d,n), T(t,2n), He-3(d,p) extracted from ENDF/B-6 and processed, FENDL/E-1.0:data for coupled neutron-photon transport calculations, including a data library for neutron interaction and photon production for 63 elements or isotopes, selected from ENDF/B-6, JENDL-3, or BROND-2, and a photon-atom interaction data library for 34 elements. The benchmark validation of FENDL-1 as required by the customer, i.e. the ITER team, is considered to be a task of high priority in the coming months. The well tested and validated nuclear data libraries in processed form of the FENDL-2 are expected to be ready by mid 1996 for use by the ITER team in the final phase of ITER EDA after extensive benchmarking and integral validation studies in the 1995-1996 period. The FENDL data files can be electronically transferred to users from the IAEA nuclear data section online system through INTERNET. A grand total of 54 (sub)directories with 845 files with total size of about 2 million

  3. Measurement of anisotropy constant in US with polarized neutrons

    DEFF Research Database (Denmark)

    Lander, G.H.; Brooks, M.S.S.; Lebech, B.

    1991-01-01

    than found in TbFe2 at 0 K. The method we have used is with polarized neutrons. Because the neutron interaction with the magnetic moment is vectorial in nature we can determine individually the magnitude and direction of the moment in an applied field. In many cases this method has advantages over...

  4. Means and method for controlling the neutron output of a neutron generator tube

    International Nuclear Information System (INIS)

    1980-01-01

    A specification is given for an energizing and regulating circuit for a gas filled neutron generator tube consisting of a target, an ion source and a replenisher, the circuit consisting of a power supply to provide a negative high voltage to the target and a target current corresponding to the neutron output of the tube, a constant current source, and control means connected to the power supply and to the constant current source, the control means being responsive to the target current to provide a portion of the constant current to the replenisher substantially to regulate the neutron output of the tube. (author)

  5. SPOTS4. Group data library and computer code, preparing ENDF/B-4 data for input to LEOPARD

    International Nuclear Information System (INIS)

    Kim, J.D.; Lee, J.T.

    1981-09-01

    The magnetic tape SPOTS4 contains in file 1 a data library to be used as input to the SPOTS4 program which is contained in file 2. The data library is based on ENDF/B-4 and consists of two parts in TEMPEST format (246 groups) and MUFT format (54 groups) respectively. From this library the SPOTS4 program produces a 172 + 54 group library for LEOPARD input. A copy of the magnetic tape is available from the IAEA Nuclear Data Section. (author)

  6. MCNP4c JEFF-3.1 Based Libraries. Eccolib-Jeff-3.1 libraries

    International Nuclear Information System (INIS)

    Sublet, J.Ch.

    2006-01-01

    Continuous-energy and multi-temperatures MCNP Ace types libraries, derived from the Joint European Fusion-Fission JEFF-3.1 evaluations, have been generated using the NJOY-99.111 processing code system. They include the continuous-energy neutron JEFF-3.1/General Purpose, JEFF-3.1/Activation-Dosimetry and thermal S(α,β) JEFF-3.1/Thermal libraries and data tables. The processing steps and features are explained together with the Quality Assurance processes and records linked to the generation of such multipurpose libraries. (author)

  7. 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)

  8. 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.

  9. Information about the new 8-group delayed neutron set preparation

    International Nuclear Information System (INIS)

    Svarny, J.

    1998-01-01

    Some comments to the present state concerning delayed neutron data preparation is given and preliminary analysis of the new 8-group delayed data (relative abundances) is presented. Comparisons of the 8-group to 6-group set is given for rod drop experiment (Unit 1, Cycle 14, NPP Dukovany).(Author)

  10. AER working group A on improvement extension and validation of parametrized few-group libraries for VVER-440 and VVER-1000

    International Nuclear Information System (INIS)

    Svarny, J.

    1998-01-01

    The AER Working Groups A and B held its sixth meeting at SKODA JS, Plzen in April 28 and 29, 1998. There were altogether 13 participants from 6 member organizations. The list of participants and the list of papers are attached. Main topics of the meeting were: A few-group cross-section library preparation methodology (standard few-group libraries, kinetics parameters, SPND signal interpretation parametrization) and its validation; Participation on intercomparisons of spectral codes (spectral codes benchmark); of kinetics parameters calculations (kinetics parameters benchmark). (author)

  11. Theory of neutron resonance cross sections for safety applications

    International Nuclear Information System (INIS)

    Froehner, F.H.

    1992-09-01

    Neutron resonances exert a strong influence on the behaviour of nuclear reactors, especially on their response to the temperature changes accompanying power excursions, and also on the efficiency of shielding materials. The relevant theory of neutron resonance cross sections including the practically important approximations is reviewed, both for the resolved and the unresolved resonance region. Numerical techniques for Doppler broadening of resonances are presented, and the construction of group constants and especially of self-shielding factors for neutronics calculations is outlined. (orig.) [de

  12. The measurement of thermal neutron constants of the soil; application to the calibration of neutron moisture gauges and to the pedological study of soil

    International Nuclear Information System (INIS)

    Couchat, P.; Marcesse, J.; Carre, C.; Le Ho, J.

    1975-01-01

    The neutronic method for measuring the water content of soils is more and more used by agronomists, hydrogeologists and pedologists. On the other hand the studies on the phenomena of slowing down and diffusion process have shown a narrow relation between the thermal absorption (Σ(a)) and diffusion (Σ(d)) constants and the thermal flux developed in the soil around a fast neutron source like Am-Be. Two original applications of the direct measurement of Σ(a) and Σ(d) are then presented. The method described consists in the measurement, in a cube of graphite with Am-Be source in the middle, on one side of the perturbation of the thermal flux, obtained by the introduction of 300g of soil, and on the other side of the transmitted thermal flux measured through the same sample of soil, on a side of the cube. After calibrating the device, these two parameters give Σ(a) and Σ(d) which are easily introduced in the calibration equation of neutron moisture gauge. Also these two values are useful for the pedologists because Σ(d) is connected to clay content in the soil and Σ(a) is connected to the type of clay by the way of rare earth contents [fr

  13. Neutron cross-sections of deuterium in the energy range 0.0001eV-15MeV

    International Nuclear Information System (INIS)

    Bazazyants, N.O.; Zabrodskaya, A.S.; Larina, A.F.; Nikolaev, M.N.

    1978-08-01

    The paper describes the evaluation of deuterium neutron cross-sections, the spectra of neutrons from the reaction D(n,2n)P and the angular distributions of neutrons from this reaction and of neutrons elastically scattered on deuterium. The evaluation results are presented in the SOCRATOR format. The 26-group system of constants for deuterium is also presented. (author)

  14. JEFF 3.1.2 - Joint evaluated nuclear data library for fission and fusion applications - February 2012 (DVD)

    International Nuclear Information System (INIS)

    2012-02-01

    The Joint Evaluated Fission and Fusion File (JEFF) project is a collaboration between NEA Data Bank member countries. The JEFF library combines the efforts of the JEFF and EFF/EAF Working Groups to produce a common sets of evaluated nuclear data, mainly for fission and fusion applications. The JEFF-3.1.2 version, released in February 2012, contains a number of different data types, including neutron and proton interaction data, radioactive decay data, fission yields, and thermal scattering law data. Currently, JEFF-3.1.2 data are available in ENDF-6 format (neutron library) from the Web. This new release is an update from JEFF-3.1.1 which concerns 115 material files from the general purpose incident neutron library which have been modified since JEFF-3.1.1. Modifications include: Hf isotopes: 6 new Hf evaluations have replaced previous ones; Gamma production data from neutron capture (MF=6 MT=102) has been added to 89 fission products (FP) evaluations; 47 of these FP have been replaced by ENDF-B/VII.0 evaluations, with gamma data added in this release. Corrections from JEFF-Beta feedback have been incorporated for 15 materials. Corrections that solve NJOY covariance processing problems and JANIS warnings have been made to 6 files. This DVD contains: - General purpose incident neutron data in ENDF-6 and ACE formats; - Activation data; - Thermal scattering data; - Incident proton data; - Radioactive decay data; - Neutron-induced fission yields data; - Spontaneous fission yields data

  15. ENDL-1978. LLL evaluated Nuclear Data Library 1978

    International Nuclear Information System (INIS)

    1979-07-01

    The contents and documentation of the 1978 version of the Evaluated Nuclear Data Library of the Lawrence Livermore Laboratory, USA are summarized. The Library contains numerical neutron reaction data for 88 isotopes or elements

  16. Creating new library services through collaboration with resident groups : Aimimg at human resource development and information literacy education in ways only libraries can do : Study on activities of an NPO called Ueda Library Club

    Science.gov (United States)

    Morita, Utako

    Creating new library services through collaboration with resident groups : Aimimg at human resource development and information literacy education in ways only libraries can do : Study on activities of an NPO called Ueda Library Club

  17. Wines: water inelastic neutron scattering experimental study

    International Nuclear Information System (INIS)

    Risch, P.; Ait Abderrahim, H.; D'hondt, P.; Malabu, E.

    1997-01-01

    An intercomparison of calculated fast neutron flux (E > 1 MeV) traverse through a very thick water zone obtained using both S N , (DORT) and Monte-Carlo (TRIPOLI and MCBEND) codes in combination with different cross-sections libraries (based on ENDF/B-III, IV, V and VI), showed small discrepancies either between S N , and Monte-Carlo results or even between S N , or Monte-Carlo results when we consider different cross-sections libraries except for S N , calculation when using P 0 , cross-sections. In order to validate our calculations we looked for experimental data. Unfortunately no experiment, dedicated for the fast neutron transport in large thickness of water, was found in the literature. Therefore SCK-CEN and EDF decided to launch the WINES experiment which is dedicated to study this phenomenon. WINES sands for Water Inelastic Neutron scattering Experimental Study. The aim of this experiment is to provide-experimental data for validation of neutron transport codes and nuclear cross-sections libraries used for LWR surveillance dosimetry analysis. The experimental device is made of 1 m 3 cubic plexiglass container filled with demineralized water. At one face of this cube, a 235 U neutron fission source system is screwed. The source device is made of a 235 U (93 % weight enriched) 18.55 x 16 cm 2 plate cladded with aluminium which is inserted in neutron beam emerging from the graphite gas-cooled BR1 reactor. Fission chambers ( 238 U(n,f), 232 Th(n,f), 237 Np(n,f) and 235 U(n,f)) are used to measure the flux traverses on the central axis of the water cube perpendicular to the fission sources. In this paper we will compare the experimental data to the calculated results using the S N , transport code DORT with the P 3 , ELXSIR library, based on ENDF/B-V, and the P 7 -BUGLE-93 library, based on ENDF/B-VI as well as the Monte-Carlo transport code TRIPOLI with a cross-section library based on ENDF/B IV and ENDF/B-VI. (authors)

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

    International Nuclear Information System (INIS)

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

    2017-03-01

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

  19. ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology

    International Nuclear Information System (INIS)

    Chadwick, M.B.; Oblozinsky, P.; Herman, M.

    2006-01-01

    We describe the next generation general purpose Evaluated Nuclear Data File, ENDF/B-VII.0, of recommended nuclear data for advanced nuclear science and technology applications. The library, released by the U.S. Cross Section Evaluation Working Group (CSEWG) in December 2006, contains data primarily for reactions with incident neutrons, protons, and photons on almost 400 isotopes, based on experimental data and theory predictions. The principal advances over the previous ENDF/B-VI library are the following: (1) New cross sections for U, Pu, Th, Np and Am actinide isotopes, with improved performance in integral validation criticality and neutron transmission benchmark tests; (2) More precise standard cross sections for neutron reactions on H, 6 Li, 10 B, Au and for 235,238 U fission, developed by a collaboration with the IAEA and the OECD/NEA Working Party on Evaluation Cooperation (WPEC); (3) Improved thermal neutron scattering; (4) An extensive set of neutron cross sections on fission products developed through a WPEC collaboration; (5) A large suite of photonuclear reactions; (6) Extension of many neutron- and proton-induced evaluations up to 150 MeV; (7) Many new light nucleus neutron and proton reactions; (8) Post-fission beta-delayed photon decay spectra; (9) New radioactive decay data; (10) New methods for uncertainties and covariances, together with covariance evaluations for some sample cases; and (11) New actinide fission energy deposition. The paper provides an overview of this library, consisting of 14 sublibraries in the same ENDF-6 format as the earlier ENDF/B-VI library. We describe each of the 14 sublibraries, focusing on neutron reactions. Extensive validation, using radiation transport codes to simulate measured critical assemblies, show major improvements: (a) The long-standing underprediction of low enriched uranium thermal assemblies is removed; (b) The 238 U and 208 Pb reflector biases in fast systems are largely removed; (c) ENDF/B-VI.8 good

  20. Measurements of DT and DD neutron yields by neutron activation on TFTR

    International Nuclear Information System (INIS)

    Barnes, C.W.; Larson, A.R.; LeMunyan, G.

    1994-01-01

    A variety of elemental foils have been activated by neutron fluence from TFTR under conditions with the DT neutron yield per shot ranging from 10 12 to over 10 18 , and with the DT/(DD+DT) neutron ratio varying from 0.5% (from triton burnup) to unity. Linear response over this large dynamic range is obtained by reducing the mass of the foils and increasing the cooling time, all while accepting greatly improved counting statistics. Effects on background gamma-ray lines from foil-capsule-material contaminants. and the resulting lower limits on activation foil mass, have been determined. DT neutron yields from dosimetry standard reactions on aluminum, chromium, iron, nickel, zirconium, and indium are in agreement within the ±9% (one-sigma,) accuracy of the measurements: also agreeing are yields from silicon foils using the ACTL library cross-section. While the ENDF/B-V library has too low a cross-section. Preliminary results from a variety of other threshold reactions are presented. Use of the 115 In(n,n) 115m In reaction (0.42 times as sensitive to DT neutrons as DD neutrons) in conjunction with pure-DT reactions allows a determination of the DT/(DD+DT) ratio in trace tritium or low-power tritium beam experiments

  1. Measurements of DT and DD neutron yields by neutron activation on TFTR

    International Nuclear Information System (INIS)

    Barnes, C.W.; Larson, A.R.; LeMunyan, G.

    1995-03-01

    A variety of elemental foils have been activated by neutron fluence from TFTR under conditions with the DT neutron yield per shot ranging from 10 12 to over 10 18 , and with the DT/(DD+DT) neutron ratio varying from 0.5% (from triton burnup) to unity. Linear response over this large dynamic range is obtained by reducing the mass of the foils and increasing the cooling time, all while accepting greatly improved counting statistics. Effects on background gamma-ray lines from foil-capsule-material contaminants, and the resulting lower limits on activation foil mass, have been determined. DT neutron yields from dosimetry standard reactions on aluminum, chromium, iron, nickel, zirconium, and indium are in agreement within the ±9% (one-sigma) accuracy of the measurements; also agreeing are yields from silicon foils using the ACTL library cross-section, while the ENDF/B-V library has too low a cross-section. Preliminary results from a variety of other threshold reactions are presented. Use of the 115 In(n.n') 115m In reaction (0.42 times as sensitive to DT neutrons as DD neutrons) in conjunction with pure-DT reactions allows a determination of the DT/(DD+DT) ratio in trace tritium or low-power tritium beam experiments

  2. JENDL-4.0 benchmarking for effective delayed neutron fraction of fast neutron systems

    International Nuclear Information System (INIS)

    Chiba, Go; Tsuji, Masashi; Sugiyama, Ken-ichiro; Narabayashi, Tadashi

    2011-01-01

    The performance of the latest Japanese evaluated nuclear data library JENDL-4.0 for the prediction of effective delayed neutron fraction β eff is assessed using experimental data of a wide range of fast neutron systems. Covariance data of JENDL-4.0 are used to quantify nuclear-data-induced uncertainties. Calculations with other libraries. JENDL-3.3, ENDF/B-VII.0, and JEFF-3.1, are also carried out for a quantitative comparison. JENDL-4.0 results in good agreement between calculation and experimental values within total uncertainties, and consistency between the differential nuclear data and integral experimental data is confirmed. While the other libraries also show good performance for β eff prediction, there are small differences in the predicted values of β eff among different libraries and ENDF/B-VII.0 gives the most stable results. Furthermore, a simple and convenient procedure to calculate sensitivity profiles of β eff to nuclear data is proposed. (author)

  3. Measurement of leakage neutron spectra from silicon carbide cylinders with D–T neutrons and validation of evaluated nuclear data

    Energy Technology Data Exchange (ETDEWEB)

    Luo, F. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Han, R. [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Key Laboratory of Nuclear Data, China Institute of Atomic Energy, Beijing 102413 (China); Nie, Y. [Key Laboratory of Nuclear Data, China Institute of Atomic Energy, Beijing 102413 (China); Chen, Z., E-mail: zqchen@impcas.ac.cn [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Zhang, S. [College of Physics Electronic Information, Inner Mongolia University for the Nationalities, Tongliao 028000 (China); Shi, F.; Lin, W.; Ren, P.; Tian, G.; Sun, Q.; Gou, B. [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Ruan, X.; Ren, J. [Key Laboratory of Nuclear Data, China Institute of Atomic Energy, Beijing 102413 (China); Ye, M. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China)

    2016-11-15

    Highlights: • Evaluated data for SiC are validated by a high precision benchmark experiment. • Leakage neutron spectra from SiC cylinders are measured at 60° and 120° using time-of-flight method. • The experimental results are compared with the MCNP-4C calculations with ENDF-BVII.1, JENDL-4.0 and CENDL-3.1 libraries. • The SiC evaluated nuclear data from CENDL-3.1 library was checked for the first time and proved to be reliable. - Abstract: Benchmarking of evaluated nuclear data libraries was performed for 14 MeV neutrons on silicon carbide samples. The experiments were carried out by using the benchmark experimental facility at China Institute of Atomic Energy (CIAE). The leakage neutron spectra from SiC (Φ13 cm × 20 cm) at 60° and 120° and SiC (Φ13 cm × 2 cm) at 60° were measured by the TOF method. The measured spectra are well reproduced by MCNP-4C calculations with the CENDL-3.1, ENDF/B-VII.1 and JENDL-4.0 evaluated nuclear data libraries, except 5–8 MeV range for 20 cm thickness. The discrepancies are mostly considered as caused by the improper evaluation of the angular distribution and secondary neutron energy distribution of the elastic scattering and inelastic scattering in evaluated nuclear data libraries.

  4. Measurement of leakage neutron spectra from silicon carbide cylinders with D–T neutrons and validation of evaluated nuclear data

    International Nuclear Information System (INIS)

    Luo, F.; Han, R.; Nie, Y.; Chen, Z.; Zhang, S.; Shi, F.; Lin, W.; Ren, P.; Tian, G.; Sun, Q.; Gou, B.; Ruan, X.; Ren, J.; Ye, M.

    2016-01-01

    Highlights: • Evaluated data for SiC are validated by a high precision benchmark experiment. • Leakage neutron spectra from SiC cylinders are measured at 60° and 120° using time-of-flight method. • The experimental results are compared with the MCNP-4C calculations with ENDF-BVII.1, JENDL-4.0 and CENDL-3.1 libraries. • The SiC evaluated nuclear data from CENDL-3.1 library was checked for the first time and proved to be reliable. - Abstract: Benchmarking of evaluated nuclear data libraries was performed for 14 MeV neutrons on silicon carbide samples. The experiments were carried out by using the benchmark experimental facility at China Institute of Atomic Energy (CIAE). The leakage neutron spectra from SiC (Φ13 cm × 20 cm) at 60° and 120° and SiC (Φ13 cm × 2 cm) at 60° were measured by the TOF method. The measured spectra are well reproduced by MCNP-4C calculations with the CENDL-3.1, ENDF/B-VII.1 and JENDL-4.0 evaluated nuclear data libraries, except 5–8 MeV range for 20 cm thickness. The discrepancies are mostly considered as caused by the improper evaluation of the angular distribution and secondary neutron energy distribution of the elastic scattering and inelastic scattering in evaluated nuclear data libraries.

  5. Constant round group key agreement protocols: A comparative study

    NARCIS (Netherlands)

    Makri, E.; Konstantinou, Elisavet

    2011-01-01

    The scope of this paper is to review and evaluate all constant round Group Key Agreement (GKA) protocols proposed so far in the literature. We have gathered all GKA protocols that require 1,2,3,4 and 5 rounds and examined their efficiency. In particular, we calculated each protocol’s computation and

  6. Determination of the Axial-Vector Weak Coupling Constant with Ultracold Neutrons

    International Nuclear Information System (INIS)

    Liu, J.; Mendenhall, M. P.; Carr, R.; Filippone, B. W.; Hickerson, K. P.; Perez Galvan, A.; Russell, R.; Holley, A. T.; Hoagland, J.; VornDick, B.; Back, H. O.; Pattie, R. W. Jr.; Young, A. R.; Bowles, T. J.; Clayton, S.; Currie, S.; Hogan, G. E.; Ito, T. M.; Makela, M.; Morris, C. L.

    2010-01-01

    A precise measurement of the neutron decay β asymmetry A 0 has been carried out using polarized ultracold neutrons from the pulsed spallation ultracold neutron source at the Los Alamos Neutron Science Center. Combining data obtained in 2008 and 2009, we report A 0 =-0.119 66±0.000 89 -0.00140 +0.00123 , from which we determine the ratio of the axial-vector to vector weak coupling of the nucleon g A /g V =-1.275 90 -0.00445 +0.00409 .

  7. Beginning-of-life neutronic analysis of a 3000-MW(t) HTGR

    International Nuclear Information System (INIS)

    Vigil, J.C.

    1975-12-01

    The results of a study of safety-related neutronic characteristics for the beginning-of-life core of a 3000-MW(t) High-Temperature Gas-Cooled Reactor are presented. Emphasis was placed on the temperature-dependent reactivity effects of fuel, moderator, control poisons, and fission products. Other neutronic characteristics studied were gross and local power distributions, neutron kinetics parameters, control rod and other material worths and worth distributions, and the reactivity worth of a selected hypothetical perturbation in the core configuration. The study was performed for the most part using discrete-ordinates transport theory codes and neutron cross sections that were interpolated from a four-parameter nine-group library supplied by the HTGR vendor. A few comparison calculations were also performed using nine-group data generated with an independent cross-section processing code system. Results from the study generally agree well with results reported by the HTGR vendor

  8. Particle physics with cold neutrons

    International Nuclear Information System (INIS)

    Dubbers, D.

    1991-01-01

    Slow neutrons are used in a large number of experiments to study the physics of particles and their fundamental interactions. Some of these experiments search for manifestations of ''new physics'' like baryon- or lepton-number nonconservation, time reversal nonconservation, new particles, right-handed currents, nonzero neutron charge, nonlinear terms in the Schrodinger equation, exotic e + e - states, and others. Other slow neutron experiments test the present Standard Model. The parity nonconserving weak neutron-nucleon interaction is studied in a variety of experiments. Free neutron beta decay gives precise values for the weak vector and axialvector coupling constants, which allow precise tests of basic symmetries like the conservation of the weak vector current, the unitarity of the weak quark mixing matrix, SU(3) flavour symmetry, and right-handed currents. Neutron beta decay data are further needed to calculate weak cross-sections, for applications, in big bang cosmology, in astrophysics, in solar physics and the solar neutrino problem, and in such mundane things as neutrino detection efficiencies in neutrino oscillation or proton decay experiments. Neutron-nucleon, neutron-nucleus and neutron-electron scattering lengths are determined in high precision experiments, which use methods like neutron interferometry or neutron gravity spectrometry. The experiments give information on quantities like the neutron charge radius or the neutron electric polarizability. Precision measurements of other fundamental constants lead to a better, model-independent value of the fine structure constant. Finally, the fundamental experiments on quantum mechanics, like spinor 4π -rotation, Berry's phase, dressed neutrons, Aharanov - Casher effect, or gravitational effects on the neutron's phase will be briefly discussed. (author)

  9. New nuclear data set ABBN-90 and its testing on macroscopic experiments

    International Nuclear Information System (INIS)

    Kosh'cheev, V.N.; Manturov, G.N.; Nikolaev, M.N.; Rineyskiy, A.A.; Sinitsa, V.V.; Tsyboolya, A.M.; Zabrodskaya, S.V.

    1993-01-01

    The new group constant set ABBN-90 is developed now. It based on the FOND-2 evaluated neutron data library processed with the code GRUCON. Some results of the testing ABBN-90 set in different macroscopic experiments are presented. (author)

  10. The effect of temperature and the control rod position on the spatial neutron flux distribution in the Syrian Miniature Neutron Source Reactor

    International Nuclear Information System (INIS)

    Khattab, K.; Omar, H.; Ghazi, N.

    2007-01-01

    The effect of water and fuel temperature increase and changes in the control rod positions on the spatial neutron flux distribution in the Syrian Miniature Neutron Source Reactor (MNSR) is discussed. The cross sections of all the reactor components at different temperatures are generated using the WIMSD4 code. These group constants are used then in the CITATION code to calculate the special neutron flux distribution using four energy groups. This work shows that water and fuel temperature increase in the reactor during the reactor daily operating time does not affect the spatial neutron flux distribution in the reactor. Changing the control rod position does not affect as well the spatial neutron flux distribution except in the region around the control rod position. This stability in the spatial neutron flux distribution, especially in the inner and outer irradiation sites, makes MNSR as a good tool for the neutron activation analysis (NAA) technique and production of radioisotopes with medium or short half lives during the reactor daily operating time. (author)

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

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1985-12-01

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

  12. Evaluated Nuclear Data Library for Transport Calculations at Energies up to 150 MeV

    International Nuclear Information System (INIS)

    Korovin, Yu.A.; Konobeyev, A.Yu.; Pilnov, G.B.; Stankovskiy, A.Yu.

    2005-01-01

    A new evaluated nuclear data library has been created. The library consists of two sub-libraries for neutron and proton incident particles. The first version of neutron sub-library has been completed and described in the present paper. The library contains nuclear data for transport, heating, and shielding applications for 242 nuclides ranging in atomic number from 8 to 82 in the energy region of primary neutrons from 10-5 eV to 150 MeV. Data below 20 MeV are taken mainly from ENDF/B-VI (Revision 8) and for some nuclides, from the JENDL-3.3 and JEFF-3.0 libraries. The evaluation of emitted particle energy and angular distributions at the energies above 20 MeV was performed with the help of the ALICE/ASH code and the analysis of available experimental data. The total cross sections, elastic cross sections, and elastic scattering angular distributions were calculated with the help of the coupled channel model. The results of the calculation were adjusted to the data from ENDF/B-VI, JENDL-3.3m or JEFF-3.0 at the neutron energy equal to 20 MeV. The library is written in ENDF/B-VI format using the MF=3/MT=5 and MF=6/MT=5 representations

  13. Neutron slowing down and transport in monoisotopic media with constant cross sections or with a square-well minimum

    International Nuclear Information System (INIS)

    Peng, W.H.

    1977-01-01

    A specialized moments-method computer code was constructed for the calculation of the even spatial moments of the scalar flux, phi/sub 2n/, through 2n = 80. Neutron slowing-down and transport in a medium with constant cross sections was examined and the effect of a superimposed square-well cross section minimum on the penetrating flux was studied. In the constant cross section case, for nuclei that are not too light, the scalar flux is essentially independent of the nuclide mass. The numerical results obtained were used to test the validity of existing analytic approximations to the flux at both small and large lethargies relative to the source energy. As a result it was possible to define the regions in the lethargy--distance plane where these analytic solutions apply with reasonable accuracy. A parametric study was made of the effect of a square-well cross section minimum on neutron fluxes at energies below the minimum. It was shown that the flux at energies well below the minimum is essentially independent of the position of the minimum in lethargy. The results can be described by a convolution-of-sources model involving only the lethargy separation between detector and source, the width and the relative depth of the minimum. On the basis of the computations and the corresponding model, it is possible to predict, e.g., the conditions under which transport in the region of minimum completely determines the penetrating flux. At the other extreme, the model describes when the transport in the minimum can be treated in the same manner as in any comparable lethargy interval. With the aid of these criteria it is possible to understand the apparent paradoxical effects of certain minima in neutron penetration through such media as iron and sodium

  14. Nuclear libraries for SCALE5.1 system

    International Nuclear Information System (INIS)

    Vertes, P.

    2009-01-01

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

  15. Nuclear libraries for SCALE5.1 system

    International Nuclear Information System (INIS)

    Vertes, P.

    2009-01-01

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

  16. Generation of the problem-dependent data libraries for IFIN-HH WWR-S spent fuel storage criticality and dose calculation

    International Nuclear Information System (INIS)

    Ene, Daniela; Tigau, F.

    1998-01-01

    The methods used for the radioactivity inventory calculation and dose evaluation of the fuel elements irradiated in the WWR-S IFIN-HH reactor are discussed in this work. A particular attention is paid to the processed problem-dependent nuclear libraries. SAS2H, a complex sequence of the SCALE-4.3 code system containing the modules BONAMI - NITAWL - XSDRNPM - COUPLE - ORIGEN-S - XSDOSE, has been assimilated on the IFIN-HH computer and applied to update the ORIGEN-S libraries by producing problem-dependent processed data libraries needed to perform the depletion and shielding analysis. This sequence uses one of the eight associated data libraries of the SCALE-4.3 system according to the choice of the user. The method consists in the following analysis processes: i) lattice cell neutron analysis to produce the flux weighting spectrum for activation library updating; ii) update of the nuclear data constants of the ORIGEN-S libraries; iii) depletion and decay analysis for a specified fuel assembly and irradiation history in order to generate gamma and neutron source strength and spectra. iv) one-dimensional radial shielding calculation for the evaluation of the angular neutron and gamma flux at the surface of a spent fuel shipping cask and further calculation of the dose rates at various points outside the cask. An efficient alternative of the calculation sequence mentioned above is the ARP (Automatic Rapid Processing) method conceived in order to generate independently ORIGEN-S libraries and to reduce substantially the running time. The substance of this method is the generation of the problem-dependent libraries from basis libraries a priori created by SAS2H for specific fuel assembly type and further interpolation of two independent variables, enrichment and burnup. Specific applications concerning WWR-S spent fuel were performed: i) generation of three problem-dependent libraries for the S-36 fuel assembly taking into account the maximum value of the burnup of this

  17. Formulae for thermal feedback of group constants in digital reactor simulation

    International Nuclear Information System (INIS)

    Perneczky, L.; Toth, I.; Vigassy, J.

    1976-01-01

    The problem, how the feedback of the thermohydraulic field to the neutron density in a reactor can be calculated is analysed. After a brief survey of the digital models in reactor simulation the applied model based on the time-dependent two-group diffusion equations is described. Using the reactor physical code system THERESA numerical results for the VVER-440 reactor are presented. (Sz.Z.)

  18. NSDUAZ unfolding package for neutron spectrometry and dosimetry with Bonner spheres

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R.; Martinez B, M. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Calle Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Ortiz R, J. M., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Av. Ramon Lopez Velarde 801, Col. Centro, 98000 Zacatecas (Mexico)

    2011-10-15

    NSDUAZ (Neutron Spectrometry and Dosimetry for the Universidad Autonoma de Zacatecas) is a user friendly neutron unfolding package for Bonner sphere spectrometer with {sup 6}Lil(Eu) developed under Lab View environment. Unfolding is carried out using a recursive iterative procedure with the SPUNIT algorithm, where the starting spectrum is obtained from a library initial guess spectrum to start the iterations, the package include a statistical procedure based on the count rates relative to the count rate in the 8 inches-diameter sphere to select the initial spectrum. Neutron spectrum is unfolded in 32 energy groups ranging from 10{sup -8} up to 231.2 MeV. (Author)

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

    International Nuclear Information System (INIS)

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

    2017-03-01

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

  20. Reports from the working group on neutron scattering

    International Nuclear Information System (INIS)

    1979-06-01

    The present report contains papers dating from July 1978 until May 1979. During this period the experimental facilities have been expanded; a new four-circuit neutron spectrometer was installed and, together with the Fritz Hafer Institute, a measuring point was set up for investigations of ideal crystals, the Compton scattering equipment has been essentially improved. The report contains a contribution on the mechanics and the control of the neutron diffractometers existing at BER II. The main subjects of the scientific research work were magnetic structures and phase transitions, electron densities and chemical bonds, structure and dynamics of molecular crystals. At the BER II reactor measuring opportunities could be offered to a number of guest groups. Their research activities are reported, too. In addition to those made at the Berlin reactor BER II measurements could be made at the accelerator VICKSI of the Hahn-Meitner Institute and at the reactors of the Institute Laue-Langevin at Grenoble and of the Research Establishment at Riso by the working groups. (orig.) [de

  1. The JFS libraries and the effect of beryllium on the breeding

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Gouveia, A.S. de.

    1981-02-01

    The Japanese group constants libraries for fast reactor analysis JFS-25 and JFS-70 are compared. The effect of beryllium on the breeding characteristics of thorium cycle breeders is examined. The results show that the 25-group set predicts shorter reactor doubling times than the 70-group set and that the effect of Be(n,2n) reactions is negligible. (Author) [pt

  2. Evaluated nuclear data file libraries use in nuclear-physical calculations

    International Nuclear Information System (INIS)

    Gritsaj, O.O.; Kalach, N.Yi.; Kal'chenko, O.Yi.; Kolotij, V.V.; Vlasov, M.F.

    1994-01-01

    The necessity of nuclear updated usage is founded for neutron experiment modeling calculations, for preparation of suitable data for reactor calculations and for other applications that account of detail energetic structure of cross section is required. The scheme of system to coordinate the work to collect and to prepare evaluated nuclear data on an international scale is presented. Main updated and recommended nuclear data libraries and associated computer programs are reviewed. Total neutron cross sections for 28 energetic groups calculated on the base of natural mixture iron isotopes evaluated nuclear data file (BROND-2, 1991) have been compared with BNAB-78 data. (author). 7 refs., 1 tab., 4 figs

  3. Neutron transmission study of the rotacional freedom of methyl groups in polydimethylsiloxane

    International Nuclear Information System (INIS)

    Amaral, L.Q.; Vinhas, L.A.; Herdade, S.B.

    1973-01-01

    The total neutron cross section of polydimethylsiloxane has been measured as a function of neutron wavelenght in the range of 4A to 10A, at room temperature, using a slow-neutron chopper and time-of-flight spectrometer. Scattering cross sections per hydrogen atom were obtained and the slope (12.2 +- 0.2) barns/A has been derived. Comparison with calculated neutron cross sections using the Krieger-Nelkin formalism for different dynamical situations as well as comparison with calibration curves relating the slope to the barrier hindering internal rotation indicates the existence of pratically free rotation of CH 3 groups about their C 3 axis

  4. Means and method for controlling the neutron output of a neutron generator tube

    International Nuclear Information System (INIS)

    Langford, O.M.; Peelman, H.E.

    1978-01-01

    Means and method are described for energizing and regulating a neutron generator tube having a target, an ion source and a replenisher. It providing a negative high voltage to the target and monitoring the target current. A constant current from a constant current source is divided into a shunt current and a replenisher current in accordence with the target current. The replenisher current is applied to the replenisher in a neutron generator tube so as to control the neutron output in accordance with the target current

  5. 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)

  6. Effect of updated WIMSD libraries on neutron energy spectrum at irradiation site of Pakistan Research Reactor-1 using 3D modeling

    International Nuclear Information System (INIS)

    Ahmad, Siraj-ul-Islam; Ahmad, Nasir

    2005-01-01

    International Atomic Energy Agency (IAEA) has recently released new WIMSD libraries based on current cross-section evaluations. Using these libraries the effect of different evaluated data sets on effective multiplication factor and neutron energy spectrum was studied with the help of 3D reactor simulation code CITATION. Simulation methodology adopted in this work was validated by analyzing IAEA 10 MW benchmark reactor. The k eff values obtained using all newly released libraries are within 0.45% to the experimental value, while the old library released in 1981 resulted in calculated value 1.05% larger than experimental. The flux spectrum obtained for standard fuel element using 3D modeling is smaller in fast energy range and higher in thermal energy range than is calculated using the 1D model for the standard cell. In the flux trap, differences of about -4% to 13% were found in thermal flux using the newly released libraries as compared to that obtained using 1981 WIMSD library. The major differences in the flux spectra between newly available libraries and the 1981 WIMSD library in thermal energy range are due to the differences in cross-sections of hydrogen bound-in-water. The use of only newly available cross-sections of hydrogen bound-in-water with 1981 WIMSD library resulted in significant improvement in value of k eff as well as in the flux spectrum. Moreover the differences among new libraries in the thermal energy range are also due to these cross-sections. Difference in fission spectra from different libraries is responsible for differences of flux spectra in the fast energy range. These differences in flux are reduced significantly in the fast energy range by only replacement of fission spectra

  7. Three-group albedo method applied to the diffusion phenomenon with up-scattering of neutrons

    International Nuclear Information System (INIS)

    Terra, Andre M. Barge Pontes Torres; Silva, Jorge A. Valle da; Cabral, Ronaldo G.

    2007-01-01

    The main objective of this research is to develop a three-group neutron Albedo algorithm considering the up-scattering of neutrons in order to analyse the diffusion phenomenon in nonmultiplying media. The neutron Albedo method is an analytical method that does not try to solve describing explicit equations for the neutron fluxes. Thus the neutron Albedo methodology is very different from the conventional methodology, as the neutron diffusion theory model. Graphite is analyzed as a model case. One major application is in the determination of the nonleakage probabilities with more understandable results in physical terms than conventional radiation transport method calculations. (author)

  8. Effects of neutron spectrum and external neutron source on neutron multiplication parameters in accelerator-driven system

    International Nuclear Information System (INIS)

    Shahbunder, Hesham; Pyeon, Cheol Ho; Misawa, Tsuyoshi; Lim, Jae-Yong; Shiroya, Seiji

    2010-01-01

    The neutron multiplication parameters: neutron multiplication M, subcritical multiplication factor k s , external source efficiency φ*, play an important role for numerical assessment and reactor power evaluation of an accelerator-driven system (ADS). Those parameters can be evaluated by using the measured reaction rate distribution in the subcritical system. In this study, the experimental verification of this methodology is performed in various ADS cores; with high-energy (100 MeV) proton-tungsten source in hard and soft neutron spectra cores and 14 MeV D-T neutron source in soft spectrum core. The comparison between measured and calculated multiplication parameters reveals a maximum relative difference in the range of 6.6-13.7% that is attributed to the calculation nuclear libraries uncertainty and accuracy for energies higher than 20 MeV and also dependent on the reaction rate distribution position and count rates. The effects of different core neutron spectra and external neutron sources on the neutron multiplication parameters are discussed.

  9. MURALB - a programme for calculating neutron fluxes in many groups

    International Nuclear Information System (INIS)

    MacDougall, J.

    1977-09-01

    The program MURALB solves the multi-group transport equation (with no upscatter) in many equal lethargy groups to produce neutron fluxes in these groups. The code has been made very flexible by confining the spatial flux solution to a single subroutine which takes as input the cross section data and source for a single group and calculates the flux for that group. In this way by supplying different versions of this routine different geometries and methods of solution of the transport equation may be treated. At present plane, cylindrical and spherical diffusion theory and collision probability solutions are available, together with a two region collision probability solution for a rod in a square cell. There is no basic restriction to one dimension but the practical size of problem tends to be limited to about 30 spatial regions by core storage requirements. In addition to the flux solution, the code calculates neutron balance, reaction rates and few groups cross sections for each mesh region, together with the values averaged over the system (cell or reactor). The program is available both as a stand-alone code and integrated into the COSMOS system. (author)

  10. Validation of a new library of nuclear constants of the WIMS code

    International Nuclear Information System (INIS)

    Aguilar H, F.

    1991-10-01

    The objective of the present work is to reproduce the experimental results of the thermal reference problems (benchmarks) TRX-1, TRX-2 and BAPL-1 to BAPL-3 with the WIMS code. It was proceeded in two stages, the first one consisted on using the original library of the code, while in the second one, a library that only contains the present elements in the benchmarks: H 1 , O 16 , Al 27 , U 235 and U 238 was generated. To generate the present nuclear data in the WIMS library, it was used the ENDF/B-IV database and the Data processing system of Nuclear Data NJOY, the library was generated using the FIXER code. (Author)

  11. 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.)

  12. Means and method for controlling the neutron output of a neutron generator tube

    International Nuclear Information System (INIS)

    1977-01-01

    A means and method for energizing and regulating a neutron generator tube is described. It has a target, an ion source and a replenisher. A negative high voltage is applied to the target and the target current monitored. A constant current from a constant current source is divided into a shunt current and a replenisher current in accordance with the target current. The replenisher current is applied to the replenisher in a neutron generator tube so as to control the neutron output in accordance with the target current. (C.F.)

  13. Comparative evaluation of group constants from UKNDL and the BNAB-70 system

    International Nuclear Information System (INIS)

    Bobkov, Yu.G.; Kolesov, V.E.; Krivtsov, A.S.; Manokhin, V.N.; Solov'ev, N.A.; Usachev, L.N.

    1976-01-01

    The comparison is made between the 26-group constants BNAB-70 with similar constants obtained from the evaluated UNKDL data. The data are compared by the capture and fission cross-section of Pu-239, U-235, U-238, the capture cross-section of Fe-56 and absorption of B-10 within an energy range from 100 eV to 10 MeV

  14. 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)

  15. CERN Library - Scientific journal cancellations

    CERN Multimedia

    2004-01-01

    Due to the constant increase of the subscription costs of scientific journals and the current budget restrictions, the Scientific Information Policy Board has mandated the Working Group for Acquisitions (WGA) together with the Library to propose a list of titles to be cancelled at the end of 2004. As a first step, the WGA has identified the scientific journals listed at the web site below as candidates for cancellation. The choice has been guided by the personal experience of the WGA members, consultation of other expert CERN staff for highly specialized titles, and by criteria such as subscription price, impact factor, and - where available - access statistics for electronic journals. The list also accounts for the fact that many titles are subscribed to in 'packages' such that a cancellation of individual titles would not lead to any cost savings. We invite users to carefully check the list on the Library homepage (http://library.cern.ch/). If you find any title that you consider critically important for y...

  16. 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)

  17. America's Star Libraries, 2010: Top-Rated Libraries

    Science.gov (United States)

    Lyons, Ray; Lance, Keith Curry

    2010-01-01

    The "LJ" Index of Public Library Service 2010, "Library Journal"'s national rating of public libraries, identifies 258 "star" libraries. Created by Ray Lyons and Keith Curry Lance, and based on 2008 data from the IMLS, it rates 7,407 public libraries. The top libraries in each group get five, four, or three stars. All included libraries, stars or…

  18. 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

  19. Spectrum of neutrons leaking from an iron sphere with a central 14 MeV neutron source

    International Nuclear Information System (INIS)

    Borisov, A.A.; Zagryadskij, V.A.; Chuvilin, D.Yu.; Kralik, M.; Pulpan, J.; Tichy, M.

    1991-01-01

    Following a review of the present state of nuclear data requisite for the calculation of the transport of 14 MeV neutrons through iron of natural isotopic composition, the results are given of the calculation of the energy spectrum of such neutrons after their passage through an iron sphere 240 mm o.d. and 90 mm i.d., the neutron source being accommodated in the centre of the sphere. The calculations were made using the one-dimensional code BLANK working with the nuclear data libraries ENDL-75, ENDL-83, ENDL/B-IV, JENDL-2 and BROND, and using the three-dimensional code BRAND with the library ENDL-78. The calculated spectra were compared with the experimental spectrum measured at a distance of 3 m from the sphere by means of an NE-213 scintillator, which records reflected protons. The reflected proton spectrum was processed by the matrix method (program FORIST), and the result was normalized to one neutron emitted by the source, as were the calculated spectra. The comparison demonstrates that the experiment is best fitted by the spectrum calculated by using the library JENDL-2, where the integrals of the observed and calculated spectra over the 1-15 MeV range differ as little as approximately 10%. (author). 3 figs., 5 tabs., 16 refs

  20. Measurement and Analysis of Neutron Leakage Spectra from Pb and LBE Cylinders with D-T Neutrons

    Science.gov (United States)

    Chen, Size; Gan, Leting; Li, Taosheng; Han, Yuncheng; Liu, Chao; Jiang, Jieqiong; Wu, Yican

    2017-09-01

    For validating the current evaluated neutron data libraries, neutron leakage spectra from lead and lead bismuth eutectic (LBE) cylinders have been measured using an intense D-T pulsed neutron source with time-of-flight (TOF) method by Institute of Nuclear Energy Safety Technology (INEST), Chinese Academy of Sciences (CAS). The measured leakage spectra have been compared with the calculated ones using Super Monte Carlo Simulation Program for Nuclear and Radiation Process (SuperMC) with the evaluated pointwise data of lead and bismuth processed from ENDF/B-VII.1, JEFF-3.1 and JENDL-4.0 libraries. This work shows that calculations of the three libraries are all generally consistent with the lead experimental result. For LBE experiment, the JEFF-3.1 and JENDL-4.0 calculations both agree well with the measurement. However, the result of ENDF/B-VII.1 fails to fit with the measured data, especially in the energy range of 5.5 and 7 MeV with difference more than 80%. Through sensitivity analysis with partial cross sections of 209Bi in ENDF/B-VII.1 and JEFF, the difference between the measurement and the ENDF/B-VII.1 calculation in LBE experiment is found due to the neutron data of 209Bi.

  1. Radiation transport and shielding information, computer codes, and nuclear data for use in CTR neutronics research and development

    International Nuclear Information System (INIS)

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

    1976-01-01

    The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components

  2. Development of 3D multi-group neutron diffusion code for hexagonal geometry

    International Nuclear Information System (INIS)

    Sun Wei; Wang Kan; Ni Dongyang; Li Qing

    2013-01-01

    Based on the theory of new flux expansion nodal method to solve the neutron diffusion equations, the intra-nodal fluence rate distribution was expanded in a series of analytic basic functions for each group. In order to improve the accuracy of calculation result, continuities of neutron fluence rate and current were utilized across the nodal surfaces. According to the boundary conditions, the iteration method was adopted to solve the diffusion equation, where inner iteration speedup method is Gauss-Seidel method and outer is Lyusternik-Wagner. A new speedup method (one-outer-iteration and multi-inner-iteration method) was proposed according to the characteristic that the convergence speed of multiplication factor is faster than that of neutron fluence rate and the update of inner iteration matrix is slow. Based on the proposed model, the code HANDF-D was developed and tested by 3D two-group vver440 benchmark, experiment 2 of HFETR, 3D four-group thermal reactor benchmark, and 3D seven-group fast reactor benchmark. The numerical results show that HANDF-D can predict accurately the multiplication factor and nodal powers. (authors)

  3. Parallel computational in nuclear group constant calculation

    International Nuclear Information System (INIS)

    Su'ud, Zaki; Rustandi, Yaddi K.; Kurniadi, Rizal

    2002-01-01

    In this paper parallel computational method in nuclear group constant calculation using collision probability method will be discuss. The main focus is on the calculation of collision matrix which need large amount of computational time. The geometry treated here is concentric cylinder. The calculation of collision probability matrix is carried out using semi analytic method using Beckley Naylor Function. To accelerate computation speed some computer parallel used to solve the problem. We used LINUX based parallelization using PVM software with C or fortran language. While in windows based we used socket programming using DELPHI or C builder. The calculation results shows the important of optimal weight for each processor in case there area many type of processor speed

  4. Comparison of serpent and triton generated FEW group constants for APR1400 nuclear reactor core

    International Nuclear Information System (INIS)

    Elsawi, Mohamed A.; Alnoamani, Zainab

    2015-01-01

    The accuracy of full-core reactor power calculations using diffusion codes is strongly dependent on the quality of the homogenized cross sections and other few-group constants generated by lattice codes. For many years, deterministic lattice codes have been used to generate these constants using different techniques: the discrete ordinates, collision probability or the method of characteristics, just to name a few. These codes, however, show some limitations, for example, on complex geometries or near heavy absorbers as in modern pressurized water reactor (PWR) designs like the Korean Advanced Power Reactor 1400 (APR1400) core. The use of continuous-energy Monte Carlo (MC) codes to produce nuclear constants can be seen as an attractive option when dealing with fuel or reactor types that lie beyond the capabilities of conventional deterministic lattice transport codes. In this paper, the few-group constants generated by two of the state-of-the-art reactor physics codes, SERPENT and SCALE/TRITON, will be critically studied and their reliability for being used in subsequent diffusion calculations will be evaluated. SERPENT is a 3D, continuous-energy, Monte Carlo reactor physics code which has a built-in burn-up calculation capability. It has been developed at the Technical Research Center of Finland (VTT) since 2004. SCALE/TRITON, on the other hand, is a control module developed within the framework of SCALE package that enables performing deterministic 2-D transport calculations on nuclear reactor core lattices. The approach followed in this paper is as follows. First, the few-group nuclear constants for the APR1400 reactor core were generated using SERPENT (version 2.1.22) and NEWT (in SCALE version 6.1.2) codes. For both codes, the critical spectrum, calculated using the B1 method, was used as a weighting function. Second, 2-D diffusion calculations were performed using the US NRC core simulator PARCS employing the two few-group constant sets generated in the first

  5. JENDL-3. The Japanese evaluated nuclear data library

    International Nuclear Information System (INIS)

    Lemmel, H.D.

    1992-01-01

    This document summarizes the contents of JENDL-3.1, the Japanese evaluated data library for neutron nuclear data, released in 1989 and revised in Dec. 1990. It also summarizes the JENDL-3 fission-products cross-section data library released in 1990. The entire library or retrievals of selected materials are available on magnetic tape from the IAEA Nuclear Data Section free of charge. (author)

  6. Research into the Impact of Facebook as a Library Marketing Tool is Inconclusive. A Review of: Xia, D. Z. (2009. Marketing library services through Facebook groups. Library Management 30(6/7, 469-477.

    Directory of Open Access Journals (Sweden)

    David Herron

    2010-09-01

    Full Text Available Objective – To investigate whether Facebook Groups are useful for library marketing.Design – Content analysis of membership and activity of university library-related Facebook Groups.Setting – Two global Facebook Groups, and the Facebook Groups of two academic libraries in the US (Rutgers University and Indiana University, both with populations in excess of 30 000 students.Subjects – A total of 28 Facebook Groups were analyzed.Methods – Facebook global Groups are open to all users, while Groups based in a network (e.g., a university only allow access for those in the network. Therefore, to collect data, theauthor used personal connections to log on to members’ profiles within university networks.The 26 university Groups were selected by searching Facebook for Groups belonging to the two university networks, using the word ‚library.‛ Groups unrelated to library business were discarded. A total of 11 Groups within the Rutgers network were analyzed. Of these, only one was organized by a librarian; the rest were organized by students. From Indiana, 15 Groups were identified, three of which were organized by librarians.In Table 1 (p. 474, all Groups are listed: 2 global Groups and 26 Groups within the two university networks. The author then visited all Groups, read all posts, and recorded the total number of members; status of each member, divided into faculty, staff and students; dates of first and last post; and discussion activity. The author analyzed group activity by keeping a tally of how often each member participated in discussions, as there was no way to see the number of times a member returned. The author also paid special attention to Groups with a large number of staff and faculty members, to gain information about the efforts of librarians to support or start new Groups.Main Results – There were a total of 652 members in the 26 university Groups (mean number of members was 25, ranging from 2 - 176. The two global

  7. Subcriticality calculation in nuclear reactors with external neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE). Programa de Engenharia Nuclear]. E-mails: asilva@con.ufrj.br; aquilino@lmp.ufrj.br; fernando@con.ufrj.br

    2007-07-01

    The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)

  8. Subcriticality calculation in nuclear reactors with external neutron sources

    International Nuclear Information System (INIS)

    Silva, Adilson Costa da; Martinez, Aquilino Senra; Silva, Fernando Carvalho da

    2007-01-01

    The main objective of this paper consists on the development of a methodology to monitor subcriticality. We used the inverse point kinetic equation with 6 precursor groups and external neutron sources for the calculation of reactivity. The input data for the inverse point kinetic equation was adjusted, in order to use the neutron counting rates obtained from the subcritical multiplication (1/M) in a nuclear reactor. In this paper, we assumed that the external neutron sources strength is constant and we define it in terms of a known initial condition. The results obtained from inverse point kinetic equation with external neutron sources were compared with the results obtained with a benchmark calculation, and showed good accuracy (author)

  9. Random pulsing of neutron source for inelastic neutron scattering gamma ray spectroscopy

    International Nuclear Information System (INIS)

    Hertzog, R.C.

    1981-01-01

    Method and apparatus are described for use in the detection of inelastic neutron scattering gamma ray spectroscopy. Data acquisition efficiency is enhanced by operating a neutron generator such that a resulting output burst of fast neutrons is maintained for as long as practicably possible until a gamma ray is detected. Upon the detection of a gamma ray the generator burst output is terminated. Pulsing of the generator may be accomplished either by controlling the burst period relative to the burst interval to achieve a constant duty cycle for the operation of the generator or by maintaining the burst period constant and controlling the burst interval such that the resulting mean burst interval corresponds to a burst time interval which reduces contributions to the detected radiation of radiation occasioned by other than the fast neutrons

  10. Leadership in academic and public libraries a time of change

    CERN Document Server

    Düren, Petra

    2013-01-01

    In a time when libraries have to face constant change, this book provides examples and advises on how to lead when change is needed (for example, when quality management is implemented or when libraries have to merge or to relocate). Engaging with how constant change affects leadership in libraries and how leaders in libraries act in times of change, this book is aimed at practitioners and students of Library and Information Science (LIS) alike, and is based on both theory and expert interviews from leaders in academic and public libraries that are in the midst, or are now coming out of a proc

  11. Improvement of Modeling HTGR Neutron Physics by Uncertainty Analysis with the Use of Cross-Section Covariance Information

    Science.gov (United States)

    Boyarinov, V. F.; Grol, A. V.; Fomichenko, P. A.; Ternovykh, M. Yu

    2017-01-01

    This work is aimed at improvement of HTGR neutron physics design calculations by application of uncertainty analysis with the use of cross-section covariance information. Methodology and codes for preparation of multigroup libraries of covariance information for individual isotopes from the basic 44-group library of SCALE-6 code system were developed. A 69-group library of covariance information in a special format for main isotopes and elements typical for high temperature gas cooled reactors (HTGR) was generated. This library can be used for estimation of uncertainties, associated with nuclear data, in analysis of HTGR neutron physics with design codes. As an example, calculations of one-group cross-section uncertainties for fission and capture reactions for main isotopes of the MHTGR-350 benchmark, as well as uncertainties of the multiplication factor (k∞) for the MHTGR-350 fuel compact cell model and fuel block model were performed. These uncertainties were estimated by the developed technology with the use of WIMS-D code and modules of SCALE-6 code system, namely, by TSUNAMI, KENO-VI and SAMS. Eight most important reactions on isotopes for MHTGR-350 benchmark were identified, namely: 10B(capt), 238U(n,γ), ν5, 235U(n,γ), 238U(el), natC(el), 235U(fiss)-235U(n,γ), 235U(fiss).

  12. Influence of the number of energy groups on the accuracy of neutron fluence calculations

    International Nuclear Information System (INIS)

    Barz, H.U.; Konheiser, J.

    1999-01-01

    The question how many groups are necessary to obtain all needed integral quantities for the neutron load of pressure vessels and detector positions outside the vessel with sufficient accuracy is of general interest. Until now, there are no systematic investigations on this question. In principle 3-dimensional consideration is required for such neutron load calculations. Therefore, an estimation of the needed number of groups can be of interest to minimize calculation time. One general problem is the P L -approximation of the angular distributions for the transfers between different groups. For elastic scattering this P L -approximation becomes poorer with increasing number of groups. As deterministic methods generally use the P L -approximation they cannot be used for investigations of the errors caused by the group approximation. We have investigated this problem applying group Monte-Carlo but nearly exact representation of this elastic slowing down without P L -approximation. The calculations were directed to assess the neutron fluence of a Russian WWER-1000 reactor. For that a simplified geometrical model of this reactor type has been used. (orig.)

  13. Thermal neutron scattering kernels for sapphire and silicon single crystals

    International Nuclear Information System (INIS)

    Cantargi, F.; Granada, J.R.; Mayer, R.E.

    2015-01-01

    Highlights: • Thermal cross section libraries for sapphire and silicon single crystals were generated. • Debye model was used to represent the vibrational frequency spectra to feed the NJOY code. • Sapphire total cross section was measured at Centro Atómico Bariloche. • Cross section libraries were validated with experimental data available. - Abstract: Sapphire and silicon are materials usually employed as filters in facilities with thermal neutron beams. Due to the lack of the corresponding thermal cross section libraries for those materials, necessary in calculations performed in order to optimize beams for specific applications, here we present the generation of new thermal neutron scattering kernels for those materials. The Debye model was used in both cases to represent the vibrational frequency spectra required to feed the NJOY nuclear data processing system in order to produce the corresponding libraries in ENDF and ACE format. These libraries were validated with available experimental data, some from the literature and others obtained at the pulsed neutron source at Centro Atómico Bariloche

  14. Fission product data library

    International Nuclear Information System (INIS)

    Hep, J.; Valenta, V.

    1975-01-01

    Reprints of values from BIBFP for 39 isotopes for which either a mistake in the BIBFP Library or updated values in the literature have been found, are given. Most corrections concern the branching ratios for isotopes which are the precursors of delayed neutron emitters

  15. Delayed neutron yield from fast neutron induced fission of 238U

    International Nuclear Information System (INIS)

    Piksaikin, V.M.; Kazakov, L.E.; Isaev, S.G.; Roshchenko, V.A.; Goverdovski, A.A.; Tertytchnyi, R.G.

    2002-01-01

    The measurements of the total delayed neutron yield from fast neutron induced fission of 238 U were made. The experimental method based on the periodic irradiation of the fissionable sample by neutrons from a suitable nuclear reaction had been employed. The preliminary results on the energy dependence of the total delayed neutron yield from fission of 238 U are obtained. According to the comparison of experimental data with our prediction based on correlation properties of delayed neutron characteristics, it is concluded that the value of the total delayed neutron yield near the threshold of (n,f) reaction is not a constant. (author)

  16. Verification and validation of ACE-format library created from ENDF/B-VII.0

    International Nuclear Information System (INIS)

    Chen Chaobin; Hu Zehua; Zhang Benai; Chen Yixue; Wu Jun

    2009-01-01

    ENDF/B-VII.0, released by the USA Cross Section Evaluation Working Group(CSEWG) in December 2006, was developed in five years since the previous release of ENDF/B-VI.8 and was demonstrated to contain much better physical representations of the data and to perform much better than previus ENDF evaluations over a broad range of applications. We generated ACE-format pointwise cross section library from the ENDF/B-VII.0 neutron reaction sublibrary with the processing code NJOY. The paper provides an overview of ENDF/B-VII.0, a summary of the ACE-format files producing process and a detail description of the validation of the ACE-format library. The conclusion is that the ACE-format library produced is correct. (authors)

  17. Determination of the decay constants and relative abundances of delayed neutrons by noise analysis in zero-power reactors

    International Nuclear Information System (INIS)

    Diniz, Ricardo

    2005-01-01

    A reactor noise approach has been employed at the IPEN/MB-01 research reactor facility in order to determine experimentally the effective delayed neutron parameters β i and λ i in a six group model and assuming the point reactor. The method can be considered a novice one because exploits the very low frequency domain of the spectral densities. The proposed method has some advantages to other in-pile methods since it does not disturb the reactor system and consequently does not 'excite' any sort of harmonic modes. As a byproduct and a consistency check, the β eff parameter was obtained without the need of the Diven factor and the power normalization and it is in excellent agreement with independent measurements. The theory/experiment comparison shows that for the abundances the JENDL 3.3 presents the best performance while for the decay constants the revised version of ENDF/B-VI.8 shows the best agreement. The best performance for the β eff determination is obtained with JENDL3.3. In contrast, ENDF/B-VI.8 and its revised version performed at LANL overestimate β eff by as much as 4%. The β eff results of this work support totally the proposal of reducing the thermal delayed neutron number for 235 U fission as made by Sakurai and Okajima. A new observed effect related to the correlation between the fluctuations of both measurement channels is also presented and discussed. This effect can be considered as an indirect evidence for the use of the point reactor model in this work as well as a possible useful tool in the understanding of reactor dynamics. (author)

  18. BIPAL - a data library for computing the burnup of fissionable isotopes and products of their decay

    International Nuclear Information System (INIS)

    Kralovcova, E.; Hep, J.; Valenta, V.

    1978-01-01

    The BIPAL databank contains data on 100 heavy metal isotopes starting with 206 Tl and finishing with 253 Es. Four are stable, the others are unstable. The following data are currently stored in the databank: the serial number and name of isotopes, decay modes and, for stable isotopes, the isotopic abundance (%), numbers of P decays and Q captures, numbers of corresponding final products, branching ratios, half-lives and their units, decay constants, thermal neutron captures, and fission cross sections, and other data (mainly alpha, beta and gamma intensities). The description of data and a printout of the BIPAL library are presented. (J.B.)

  19. Measurement of leakage neutron spectra for Tungsten with D-T neutrons and validation of evaluated nuclear data

    International Nuclear Information System (INIS)

    Zhang, S.; Chen, Z.; Nie, Y.; Wada, R.; Ruan, X.; Han, R.; Liu, X.; Lin, W.; Liu, J.; Shi, F.; Ren, P.; Tian, G.; Luo, F.; Ren, J.; Bao, J.

    2015-01-01

    Highlights: • Evaluated data for Tungsten are validated by integral experiment. • Leakage neutron spectra from the irradiation of D-T neutrons on Tungsten are measured at 60° and 120° by using a time-of-flight method. • The measured results are compared to the MCNP-4C calculated ones with evaluated data of the different libraries. - Abstract: Integral neutronics experiments have been investigated at Institute of Modern Physics, Chinese Academy of Sciences (IMP, CAS) in order to validate evaluated nuclear data related to the design of Chinese Initiative Accelerator Driven Systems (CIADS). In the present paper, the accuracy of evaluated nuclear data for Tungsten has been examined by comparing measured leakage neutron spectra with calculated ones. Leakage neutron spectra from the irradiation of D-T neutrons on Tungsten slab sample were experimentally measured at 60° and 120° by using a time-of-flight method. Theoretical calculations are carried out by Monte Carlo neutron transport code MCNP-4C with evaluated nuclear data of the ADS-2.0, ENDF/B-VII.0, ENDF/B-VII.1, JENDL-4.0 and CENDL-3.1 libraries. From the comparisons, it is found that the calculations with ADS-2.0 and ENDF/B-VII.1 give good agreements with the experiments in the whole energy regions at 60°, while a large discrepancy is observed at 120° in the elastic scattering peak, caused by a slight difference in the oscillation pattern of the elastic angular distribution at angles larger than 20°. However, the calculated spectra using data from ENDF/B-VII.0, JENDL-4.0 and CENDL-3.1 libraries showed larger discrepancies with the measured ones, especially around 8.5–13.5 MeV. Further studies are presented for these disagreements

  20. 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

  1. ZZ SAIL, Albedo Scattering Data Library for 3-D Monte-Carlo Radiation Transport in LWR Pressure Vessel

    International Nuclear Information System (INIS)

    1982-01-01

    1 - Description of problem or function: Format: SAIL format; Number of groups: 23 neutron / 17 gamma-ray; Nuclides: Type 04 Concrete and Low Carbon Steel (A533B). Origin: Science Applications, Inc (SAI); Weighting spectrum: yes. SAIL is a library of albedo scattering data to be used in three-dimensional Monte Carlo codes to solve radiation transport problems specific to the reactor pressure vessel cavity region of a LWR. The library contains data for Type 04 Concrete and Low Carbon Steel (A533B). 2 - Method of solution: The calculation of the albedo data was perform- ed with a version of the discrete ordinates transport code DOT which treats the transport of neutrons, secondary gamma-rays and gamma- rays in one dimension, while maintaining the complete two-dimension- al treatment of the angular dependence

  2. Simultaneous measurement of neutron-induced fission and capture cross sections for {sup 241}Am at neutron energies below fission threshold

    Energy Technology Data Exchange (ETDEWEB)

    Hirose, K., E-mail: hirose.kentaro@jaea.go.jp [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Nishio, K.; Makii, H.; Nishinaka, I.; Ota, S. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Nagayama, T. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Graduate School of Science and Engineering, Ibaraki University, Mito 310-0056 (Japan); Tamura, N. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Graduate School of Science and Technology, Niigata University, Niigata 950-2181 (Japan); Goto, S. [Graduate School of Science and Technology, Niigata University, Niigata 950-2181 (Japan); Andreyev, A.N. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Vermeulen, M.J. [Advanced Science Research Center, Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki 319-1195 (Japan); Gillespie, S.; Barton, C. [Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Kimura, A.; Harada, H. [Nuclear Science and Engineering Center, JAEA, Tokai, Ibaraki 319-1195 (Japan); Meigo, S. [J-PARC Center, JAEA, Tokai, Ibaraki 319-1195 (Japan); Chiba, S. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo 152-8550 (Japan); Ohtsuki, T. [Research Reactor Institute, Kyoto University, Kumatori-cho S' ennangun,Osaka 590-0494 (Japan)

    2017-06-01

    Fission and capture reactions were simultaneously measured in the neutron-induced reactions of {sup 241}Am at the spallation neutron facility of the Japan Proton Accelerator Research Complex (J-PARC). Data for the neutron energy range of E{sub n}=0.1–20 eV were taken with the TOF method. The fission events were observed by detecting prompt neutrons accompanied by fission using liquid organic scintillators. The capture reaction was measured by detecting γ rays emitted in the deexcitation of the compound nuclei using the same detectors, where the prompt fission neutrons and capture γ rays were separated by a pulse shape analysis. The cross sections were obtained by normalizing the relative yields at the first resonance to evaluations or other experimental data. The ratio of the fission to capture cross sections at each resonance is compared with those from an evaluated nuclear data library and other experimental data. Some differences were found between the present values and the library/literature values at several resonances.

  3. Description of WIMS Library Update Project (WLUP)

    International Nuclear Information System (INIS)

    Leszczynski, Francisco

    2002-01-01

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

  4. 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)

  5. The use of multi-energy-group neutron diffusion theory to numerically evaluate the relative utility of three dial-detector neutron porosity well logging tools

    International Nuclear Information System (INIS)

    Zalan, T.A.

    1988-01-01

    Multi-energy-group neutron diffusion theory is used to numerically evaluate the utility of two different dual-detector neutron porosity logging devices, a 14 MeV (accelerator) neutron source - epithermal neutron detector device and a 4 MeV neutron source - capture gamma-ray detector device, relative to the traditional 4 MeV neutron source - thermal neutron detector device. Fast and epithermal neutron diffusion parameters are calculated using Monte Carlo - derived neutron flux distributions. Thermal parameters are calculated from tabulated cross sections. An existing analytical method to describe the transport of gamma-rays through common earth materials is modified in order to accommodate the modeling of the 4 MeV neutron - capture gamma-ray device. The 14 MeV neutron - epithermal neutron device is found to be less sensitive to porosity than the 4 MeV neutron - capture gamma-ray device, which in turn is found to be less sensitive to porosity than the traditional 4 MeV neutron - thermal neutron device. Salinity effects are found to be comparable for the 4 MeV neutron - capture gamma-ray and 4 MeV neutron - thermal neutron devices. The 4 MeV neutron capture gamma-ray measurement is found to be deepest investigating

  6. Critical experiments analyses by using 70 energy group library based on ENDF/B-VI

    Energy Technology Data Exchange (ETDEWEB)

    Tahara, Yoshihisa; Matsumoto, Hideki [Mitsubishi Heavy Industries Ltd., Yokohama (Japan). Nuclear Energy Systems Engineering Center; Huria, H.C.; Ouisloumen, M.

    1998-03-01

    The newly developed 70-group library has been validated by comparing kinf from a continuous energy Monte-Carlo code MCNP and two dimensional spectrum calculation code PHOENIX-CP. The code employs Discrete Angular Flux Method based on Collision Probability. The library has been also validated against a large number of critical experiments and numerical benchmarks for assemblies with MOX and Gd fuels. (author)

  7. Henry's Constants of Persistent Organic Pollutants by a Group-Contribution Method Based on Scaled-Particle Theory.

    Science.gov (United States)

    Razdan, Neil K; Koshy, David M; Prausnitz, John M

    2017-11-07

    A group-contribution method based on scaled-particle theory was developed to predict Henry's constants for six families of persistent organic pollutants: polychlorinated benzenes, polychlorinated biphenyls, polychlorinated dibenzodioxins, polychlorinated dibenzofurans, polychlorinated naphthalenes, and polybrominated diphenyl ethers. The group-contribution model uses limited experimental data to obtain group-interaction parameters for an easy-to-use method to predict Henry's constants for systems where reliable experimental data are scarce. By using group-interaction parameters obtained from data reduction, scaled-particle theory gives the partial molar Gibbs energy of dissolution, Δg̅ 2 , allowing calculation of Henry's constant, H 2 , for more than 700 organic pollutants. The average deviation between predicted values of log H 2 and experiment is 4%. Application of an approximate van't Hoff equation gives the temperature dependence of Henry's constants for polychlorinated biphenyls, polychlorinated naphthalenes, and polybrominated diphenyl ethers in the environmentally relevant range 0-40 °C.

  8. Monte Carlo simulation of neutron scattering instruments

    International Nuclear Information System (INIS)

    Seeger, P.A.

    1995-01-01

    A library of Monte Carlo subroutines has been developed for the purpose of design of neutron scattering instruments. Using small-angle scattering as an example, the philosophy and structure of the library are described and the programs are used to compare instruments at continuous wave (CW) and long-pulse spallation source (LPSS) neutron facilities. The Monte Carlo results give a count-rate gain of a factor between 2 and 4 using time-of-flight analysis. This is comparable to scaling arguments based on the ratio of wavelength bandwidth to resolution width

  9. Amino acids analysis using grouping and parceling of neutrons cross sections techniques

    International Nuclear Information System (INIS)

    Voi, Dante Luiz Voi; Rocha, Helio Fenandes da

    2002-01-01

    Amino acids used in parenteral administration in hospital patients with special importance in nutritional applications were analyzed to compare with the manufactory data. Individual amino acid samples of phenylalanine, cysteine, methionine, tyrosine and threonine were measured with the neutron crystal spectrometer installed at the J-9 irradiation channel of the 1 kW Argonaut Reactor of the Instituto de Engenharia Nuclear (IEN). Gold and D 2 O high purity samples were used for the experimental system calibration. Neutron cross section values were calculated from chemical composition, conformation and molecular structure analysis of the materials. Literature data were manipulated by parceling and grouping neutron cross sections. (author)

  10. Verification of 3-D generation code package for neutronic calculations of WWERs

    International Nuclear Information System (INIS)

    Sidorenko, V.D.; Aleshin, S.S.; Bolobov, P.A.; Bolshagin, S.N.; Lazarenko, A.P.; Markov, A.V.; Morozov, V.V.; Syslov, A.A.; Tsvetkov, V.M.

    2000-01-01

    Materials on verification of the 3 -d generation code package for WWERs neutronic calculations are presented. The package includes: - spectral code TVS-M; - 2-D fine mesh diffusion code PERMAK-A for 4- or 6-group calculation of WWER core burnup; - 3-D coarse mesh diffusion code BIPR-7A for 2-group calculations of quasi-stationary WWERs regimes. The materials include both TVS-M verification data and verification data on PERMAK-A and BIPR-7A codes using constant libraries generated with TVS-M. All materials are related to the fuel without Gd. TVS-M verification materials include results of comparison both with benchmark calculations obtained by other codes and with experiments carried out at ZR-6 critical facility. PERMAK-A verification materials contain results of comparison with TVS-M calculations and with ZR-6 experiments. BIPR-7A materials include comparison with operation data for Dukovany-2 and Loviisa-1 NPPs (WWER-440) and for Balakovo NPP Unit 4 (WWER-1000). The verification materials demonstrate rather good accuracy of calculations obtained with the use of code package of the 3 -d generation. (Authors)

  11. Neutron dosimetry system SAIPS: Manual for users and programmers (Version 87-02)

    International Nuclear Information System (INIS)

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

    1988-07-01

    SAIPS is a system used for neutron dosimetry by foil activation, containing a package of programs and a data base of neutron activation cross-sections. A description is given of the SAIPS indexed procedures and users language, which are designed for producing input data for programs unfolding neutron spectra from reaction rate measurements, for carrying out calculations and processing and comparing the results obtained, for utilizing the additional capabilities of the system, and for setting up a working version of the system from the magnetic tapes used for distribution. A description is given of the logical structure of the data sets containing the libraries of neutron cross-section and a priori spectra and also the libraries of calculated spectra. The annexes give examples of SAIPS in use, of the contents of the a priori spectra and neutron cross-section libraries, and of the contents of the SAIPS distribution tapes. SAIPS contains programs in PL/1 (opt), FORTRAN IV(H) and ASSEMBLER. 25 refs

  12. ZZ CAD, 51 Neutron-Group, 25 Gamma-Group Albedo Data for 4 Materials from DOT Flux

    International Nuclear Information System (INIS)

    1992-01-01

    A - Description of problem or function: Format: BREESE tape-writing program, MORSE; Number of groups: 51 neutron, 25 gamma-ray group albedo data. Nuclides: 1) 12 inches of water. 2) 12 inches of ordinary concrete. 3) 9 inches of carbon steel (SA508). 4) 1/2 inches of steel over 12 inches of concrete. (O, Ca, Al, C, Si, H, K, Mg, Fe, Na, Mn); Origin: DOT angular flux tape. CAD is a set of 51 neutron, 25 gamma-ray group albedo data for the following four materials: 1) 12 inches of water. 2) 12 inches of ordinary concrete. 3) 9 inches of carbon steel (SA508). 4) 1/2 inches of steel over 12 inches of concrete. The differential angular albedos are a function of the five incident polar directions and 30 reflected directions. B - Method of solution: The data has been generated from a DOT angular flux tape using the code CARP (abstract PSR-0131). C - Restrictions on the complexity of the problem: Since the amount of data is so large, it is necessary to run CARP, using the group reduction option, in order to run a problem on most computers

  13. /sup 13/C-/sup 13/C spin-spin coupling constants in structural investigations. V. The direct carbon-carbon coupling constants in the vinyl group

    Energy Technology Data Exchange (ETDEWEB)

    Krivdin, L.B.; Shcherbakov, V.V.; Kalabin, G.A.

    1988-03-10

    The direct spin-spin coupling constants in the vinyl group were measured in 100 mono-substituted ethylene derivatives. The inductive effect of the substituent was found to be the major factor in the variation of this constant and, in some cases, the stereospecific effect of the unshared electron pairs of heteratoms makes a significant contribution to the /sup 13/C-/sup 13/C coupling constants.

  14. Cold neutron production in liquid para- and normal-H sub 2 moderators

    CERN Document Server

    Morishima, N

    2002-01-01

    A neutron transport analysis is performed for liquid H sub 2 moderators with 100% para and normal (ortho:para=0.75:0.25) fractions. Four sets of energy-averaged cross-sections (group constants) for liquid ortho- and para-H sub 2 at melting and boiling points are generated and neutron energy range between 0.1 mu eV and 10 eV is broken into 80 groups. Basic moderating characteristics are studied of a model cold-neutron source in a one-dimensional bare-slab geometry. It is shown that liquid para-H sub 2 is superior in cold neutron production to liquid normal H sub 2 on account of a para-to-ortho transition (molecular rotational excitation) and a good transmission property with a mean free path of about 10 cm. In the case of neutron extraction from the inside of the source, high intensity of cold neutrons is possible with liquid normal H sub 2 at higher temperatures up to the boiling point.

  15. The KINA neutronic module of the LEGO code for steady-state and transient PWR plant simulations

    International Nuclear Information System (INIS)

    Nicolopoulos, D.; Pollacchini, L.; Vimercati, G.; Spelta, S.

    1989-01-01

    The Automation Research Center (CRA) of ENEl has implemented some models for analyzing both incidental and operational transients in PWR power plants. For such models an axial neutron kinetics module characterized by high computational efficency with adequate results accuracy was called for. CISE has been entrusted with the task of implementing such a module named KINA and based on IQS (Improved Quasi Static) method, to be included in the library of LEGO modular code used by CRA to set up PWR power models. Moreover, The KINA module has been adapted to the neutron constants computing model developed by the EdF-SEPTEN, which has been using and improving the LEGO code for a long time in cooperation with ENEL-CRA. In this paper, after some remarks on the LEGO code, a general description of KINA neutronic module is given. The resylts of a preliminary validation activity of KINA for an EdF 1300 MWe PWR plant are also presented

  16. Proceedings of the specialists' meeting on delayed neutron nuclear data

    International Nuclear Information System (INIS)

    Katakura, Jun-ichi

    1999-07-01

    This report is the Proceedings of the Specialists' Meeting on Delayed Neutron Nuclear Data. The meeting was held on January 28-29, 1999, at the Tokai Research Establishment of Japan Atomic Energy Research Institute with the participation of thirty specialists, who are evaluators, theorist, experimentalists. Although the fraction of the delayed neutron is no more than 1% in the total neutrons emitted in the fission process, it plays an important roll in the control of fission reactor. In the meeting, the following topics were reported: the present status of delayed neutron data in the major evaluated data libraries, measurements of effective delayed neutron fraction using FCA (Fast Critical Assembly) and TCA (Tank-type Critical Assembly) and their analyses, sensitivity analysis for fast reactor, measurements of delayed neutron emission from actinides and so on. As another topics, delayed neutron in transmutation system and fission yield data were also presented. Free discussion was held on the future activity of delayed neutron data evaluation. The discussion was helpful for the future activity of the delayed neutron working group of JNDC aiming to the evaluation of delayed neutron data for JENDL-3.3. The 15 of the presented papers are indexed individually. (J.P.N.)

  17. Parallel processing of neutron transport in fuel assembly calculation

    International Nuclear Information System (INIS)

    Song, Jae Seung

    1992-02-01

    Group constants, which are used for reactor analyses by nodal method, are generated by fuel assembly calculations based on the neutron transport theory, since one or a quarter of the fuel assembly corresponds to a unit mesh in the current nodal calculation. The group constant calculation for a fuel assembly is performed through spectrum calculations, a two-dimensional fuel assembly calculation, and depletion calculations. The purpose of this study is to develop a parallel algorithm to be used in a parallel processor for the fuel assembly calculation and the depletion calculations of the group constant generation. A serial program, which solves the neutron integral transport equation using the transmission probability method and the linear depletion equation, was prepared and verified by a benchmark calculation. Small changes from the serial program was enough to parallelize the depletion calculation which has inherent parallel characteristics. In the fuel assembly calculation, however, efficient parallelization is not simple and easy because of the many coupling parameters in the calculation and data communications among CPU's. In this study, the group distribution method is introduced for the parallel processing of the fuel assembly calculation to minimize the data communications. The parallel processing was performed on Quadputer with 4 CPU's operating in NURAD Lab. at KAIST. Efficiencies of 54.3 % and 78.0 % were obtained in the fuel assembly calculation and depletion calculation, respectively, which lead to the overall speedup of about 2.5. As a result, it is concluded that the computing time consumed for the group constant generation can be easily reduced by parallel processing on the parallel computer with small size CPU's

  18. LabVIEW Library to EPICS Channel Access

    CERN Document Server

    Liyu, Andrei; Thompson, Dave H

    2005-01-01

    The Spallation Neutron Source (SNS) accelerator systems will deliver a 1.0 GeV, 1.4 MW proton beam to a liquid mercury target for neutron scattering research. The accelerator complex consists of a 1 GeV linear accelerator, an accumulator ring and associated transport lines. The SNS diagnostics platform is PC-based and will run Windows for its OS and LabVIEW as its programming language. Data acquisition hardware will be based on PCI cards. There will be about 300 rack-mounted computers. The Channel Access (CA) protocol of the Experimental Physics and Industrial Control System (EPICS) is the SNS control system communication standard. This paper describes the approaches, implementation, and features of LabVIEW library to CA for Windows, Linux, and Mac OS X. We also discuss how the library implements the asynchronous CA monitor routine using LabVIEW's occurrence mechanism instead of a callback function (which is not available in LabVIEW). The library is used to acquire accelerator data and applications have been ...

  19. Monitoring of the Irradiated Neutron Fluence in the Neutron Transmutation Doping Process of Hanaro

    Science.gov (United States)

    Kim, Myong-Seop; Park, Sang-Jun

    2009-08-01

    Neutron transmutation doping (NTD) for silicon is a process of the creation of phosphorus impurities in intrinsic or extrinsic silicon by neutron irradiation to obtain silicon semiconductors with extremely uniform dopant distribution. HANARO has two vertical holes for the NTD, and the irradiation for 5 and 6 inch silicon ingots has been going on at one hole. In order to achieve the accurate neutron fluence corresponding to the target resistivity, the real time neutron flux is monitored by self-powered neutron detectors. After irradiation, the total irradiation fluence is confirmed by measuring the absolute activity of activation detectors. In this work, a neutron fluence monitoring method using zirconium foils with the mass of 10 ~ 50 mg was applied to the NTD process of HANARO. We determined the proportional constant of the relationship between the resistivity of the irradiated silicon and the neutron fluence determined by using zirconium foils. The determined constant for the initially n-type silicon was 3.126 × 1019 n·Ω/cm. It was confirmed that the difference between this empirical value and the theoretical one was only 0.5%. Conclusively, the practical methodology to perform the neutron transmutation doping of silicon was established.

  20. 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)

  1. Direct 13C-1H coupling constants in the vinyl group of 1-vinylpyrazoles

    International Nuclear Information System (INIS)

    Afonin, A.V.; Voronov, V.K.; Es'kova, L.A.; Domnina, E.S.; Petrova, E.V.; Zasyad'ko, O.V.

    1987-01-01

    In a continuation of a study of the rotational isomerism of 1-vinylpyrazoles, they studied the direct 13 C- 1 H coupling constants in the vinyl group of 1-vinylpyrazole, 1-vinyl-4-bromopyrazole, 1-vinyl-3-methylpyrazole, 1-vinyl-5-methylpyrazole, 1-vinyl-3,5-dimethylpyrazole, and 1-vinyl-4-nitro-3,5-dimethylpyrazole. The 13 C- 1 H direct coupling constants in the vinyl group of 1-vinylpyrazoles are stereo-specific and vary with change in the conformer ratio

  2. A simple neutron-gamma discriminating system

    International Nuclear Information System (INIS)

    Liu Zhongming; Xing Shilin; Wang Zhongmin

    1986-01-01

    A simple neutron-gamma discriminating system is described. A detector and a pulse shape discriminator are suitable for the neutron-gamma discriminating system. The influence of the constant fraction discriminator threshold energy on the neutron-gamma resolution properties is shown. The neutron-gamma timing distributions from an 241 Am-Be source, 2.5 MeV neutron beam and 14 MeV neutron beam are presented

  3. Evaluation of the total gamma-ray production cross-sections for nonelastic interaction of fast neutrons with iron nuclei

    International Nuclear Information System (INIS)

    Savin, M.V.; Nefedov, Yu.Ya; Livke, A.V.; Zvenigorodskij, A.G.

    2001-01-01

    Experimental data on the total gamma-ray production cross-sections for inelastic interaction of fast neutrons with iron nuclei were analysed. The total gamma-ray production cross-sections, grouped according to E γ , were evaluated in the neutron energy range 0.5-19 MeV. The statistical spline approximation method was used to evaluate the experimental data. Evaluated data stored in the ENDF, JENDL, BROND, and other libraries on gamma-ray production spectra and cross-sections for inelastic interaction of fast neutrons with iron nuclei, were analysed. (author)

  4. ADL-3. Nuclear data library for activation and transmutation calculations

    International Nuclear Information System (INIS)

    Grudzevich, O.T.; Zelenetskij, A.V.; Ignatyuk, A.V.; Pashchenko, A.B.

    1995-07-01

    It is shown that the use of simplified approaches to calculate threshold neutron reaction cross-sections is not acceptable for the generation of cross-section libraries. Although rigorous models are complex and involve laborious calculations, they provide the only reliable means for evaluating cross-sections when no experimental data are available. A brief description is given of the new version of the library ADL-3 generated by the authors. It contains 18,200 excitation functions of reactions induced by neutrons of up to 20 MeV. The threshold reaction cross-sections have been calculated in the Hauser-Feshbach-Moldauer formalism with allowance for the contribution of non-equilibrium processes. The cross-sections obtained have been tested by comparison with experimental data and evaluations from other libraries. (author)

  5. MCNP4c JEFF-3.1 Based Libraries. Eccolib-Jeff-3.1 libraries; Les bibliotheques Eccolib-Jeff-3.1

    Energy Technology Data Exchange (ETDEWEB)

    Sublet, J.Ch

    2006-07-01

    Continuous-energy and multi-temperatures MCNP Ace types libraries, derived from the Joint European Fusion-Fission JEFF-3.1 evaluations, have been generated using the NJOY-99.111 processing code system. They include the continuous-energy neutron JEFF-3.1/General Purpose, JEFF-3.1/Activation-Dosimetry and thermal S({alpha},{beta}) JEFF-3.1/Thermal libraries and data tables. The processing steps and features are explained together with the Quality Assurance processes and records linked to the generation of such multipurpose libraries. (author)

  6. Burnup calculation with estimated neutron spectrum of JMTR irradiation field. Development of the burnup calculation method for fuel pre-irradiated in the JMTR

    International Nuclear Information System (INIS)

    Okonogi, Kazunari; Nakamura, Takehiko; Yoshinaga, Makio; Hosoyamada, Ryuji

    1999-03-01

    As a series of the pulse irradiation tests with the irradiated fuel, the high-enriched fuel rods pre-irradiated in the JMTR as well as the fuels irradiated in commercial reactors have been irradiated in the NSRR. In the pre-irradiation at the JMTR, the test fuels were placed at the irradiation holes in the reflector region far from the driver core to keep the linear heat generation rate of the test fuel low. Accordingly, neutron energy spectra of the irradiation holes for the test fuels are softened due to the higher moderator ratio than in those of the ordinary LWR core, which causes quite different burnup characteristics. JMTR post irradiation condition corresponds to the pre-test condition in the NSRR. Therefore, proper understanding of the condition is quite important for the precise evaluating the energy deposition and FP generation in the test. Then, neutron spectra at the JMTR irradiation field were evaluated and its effects on the burnup calculation were quantified. Basing on the configuration of the JMTR core in the operation cycle No.85, neutron diffusion calculations of 107 groups were executed in 2-D slab (X-Y) geometry of CITATION of SRAC95 code system, and neutron energy spectra of the irradiation hole for the test fuels were evaluated. Burnup calculations of Test JMN-1 fuel with the estimated neutron energy spectra were performed and the results were compared to both the measurements and calculation results with the PWR and BWR libraries in ORIGEN2 code. SWAT code was used to collapse the 107 groups spectra into 1 group libraries for the ORIGEN2 use. The calculation results for both the generation and depletion of U, Pu and Nd with the JMTR libraries obtained in the present study were in the reasonably good agreement with the measurements, while in the case of calculation with the PWR and BWR libraries in ORIGEN2, the generation of fission products having mass numbers from 105 to 130 and some actinides were overestimated by about 1.5 to 3.5 times

  7. Measurement of neutron spectra through composed material block bombarded with D-T neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Zhu, T.H. [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, P.O. BOX 919-213, Mian yang 621900 (China)], E-mail: zhutonghua@yahoo.com.cn; Liu, R.; Lu, X.X.; Jiang, L.; Wen, Z.W.; Wang, M.; Lin, J.F. [Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, P.O. BOX 919-213, Mian yang 621900 (China)

    2009-12-15

    A 2-dimensional composed material assembly made of the iron and hydric block has been established. The neutron spectra from the assembly bombarded with 14-MeV neutrons at neutron generator have been obtained using the proton recoil technique with a stillbene detector. The detector positions were selected at the 60 deg., 120 deg., 180 deg. on the surface of the iron spherical shell. The background neutron spectra consisted of background and room return radiation were subtracted with combination of methods of experimental shielding and MCNP calculation. The uncertainty of results was 6.3-7.4%. The experiment results were analyzed and simulated by MCNP code and two data library. The difference is integral neutron flux (background neutron subtracted) of measured results greater than calculations with maximum of 21.2% in the range of 1-16 MeV.

  8. Comparison of reactor RA-4 kinetics with simulations with Matlab-Simulink for one group and six groups of delayed neutrons

    International Nuclear Information System (INIS)

    Orso, J A

    2012-01-01

    The critical state of a nuclear reactor is an unstable equilibrium. The nuclear reactor can go from critical to subcritical state or can go from critical to hypercritical state. Although the evolution of the system in these cases is slow, it requires the intervention of an operator to correct deviations. For this reason an automatic control technique was designed, based on the kinetic point to a group of delayed neutrons, which corrects deviations automatically. In this paper we study the point kinetics models in a group and six groups of delayed neutrons for different values of reactivity using the simulations software MATLAB, Simulink. A comparison of two models with the reactor kinetic behavior is made (author)

  9. Neutron metrology file NMF-90. An integrated database for performing neutron spectrum adjustment calculations

    International Nuclear Information System (INIS)

    Kocherov, N.P.

    1996-01-01

    The Neutron Metrology File NMF-90 is an integrated database for performing neutron spectrum adjustment (unfolding) calculations. It contains 4 different adjustment codes, the dosimetry reaction cross-section library IRDF-90/NMF-G with covariances files, 6 input data sets for reactor benchmark neutron fields and a number of utility codes for processing and plotting the input and output data. The package consists of 9 PC HD diskettes and manuals for the codes. It is distributed by the Nuclear Data Section of the IAEA on request free of charge. About 10 MB of diskspace is needed to install and run a typical reactor neutron dosimetry unfolding problem. (author). 8 refs

  10. Assessment of nuclear data needs for broad-group SCALE library related to WWER spent fuel applications

    International Nuclear Information System (INIS)

    Zalesky, K.; Markova, L.

    1999-12-01

    A preliminary study aimed at the issue of feasibility to generate a broad-group SCALE library related to WWER spent fuel applications was made. The SCALE code system has been installed and is being used in many countries operating WWER-type reactors for criticality and shielding analyses as well as spent fuel isotopic inventory calculations but still without an extensive validation and verification for the WWER environment. This study should be a contribution to QA connected with the SCALE code system application for the WWER calculations as a basis on which the generation of the specific WWER SCALE library can be prepared. Possible ways of the broad-group library development are described. (author)

  11. Translation of selected reports on neutron spectrum unfolding

    International Nuclear Information System (INIS)

    Berzonis, M.; Bondars, Kh.Ya.; Taimina, D.

    1982-05-01

    The paper provides the information needed by users of the SAIPS information system on the neutron cross-section libraries accessible and on the principles upon which they are based. Neutron cross-section integrals in fission and fusion spectra are given. (author)

  12. User Perceptions of the Library's Web Pages: A Focus Group Study at Texas A&M University.

    Science.gov (United States)

    Crowley, Gwyneth H.; Leffel, Rob; Ramirez, Diana; Hart, Judith L.; Armstrong, Tommy S., II

    2002-01-01

    This focus group study explored library patrons' opinions about Texas A&M library's Web pages. Discusses information seeking behavior which indicated that patrons are confused when trying to navigate the Public Access Menu and suggests the need for a more intuitive interface. (Author/LRW)

  13. Report of the first United States conference on utility experience with neutron noise analysis

    International Nuclear Information System (INIS)

    Fry, D.N.; Horne, G.P.; Mayo, C.W.

    1984-01-01

    An informal meeting was held in Washington, D.C. on April 3 and 4, 1984, to discuss the current state of the art and experiences with neutron noise analysis in US pressurized water reactors (PWRs). The meeting was attended by 33 persons representing 11 utilities and 3 PWR reactor vendors as well as consultants, universities, and research laboratories. Presentations at the meeting covered several applications of neutron noise for diagnosing such things as vibrations induced by baffle jetting, detection of mechanical degradation of thermal shield supports, and electrical degradation of nuclear instrumentation channels. Twenty-one responses were obtained from a questionnaire circulated to all participants requesting their viewpoints and experiences regarding neutron noise analysis. The meeting participants concluded that a working group on neutron noise analysis should be formed to (1) establish a baseline library of neutron noise data, (2) provide a forum for communicating experiences with neutron noise surveillance, and (3) develop good practices and quality assurance procedures for neutron noise measurement and interpretation

  14. Monte Carlo simulation of neutron scattering instruments

    International Nuclear Information System (INIS)

    Seeger, P.A.; Daemen, L.L.; Hjelm, R.P. Jr.

    1998-01-01

    A code package consisting of the Monte Carlo Library MCLIB, the executing code MC RUN, the web application MC Web, and various ancillary codes is proposed as an open standard for simulation of neutron scattering instruments. The architecture of the package includes structures to define surfaces, regions, and optical elements contained in regions. A particle is defined by its vector position and velocity, its time of flight, its mass and charge, and a polarization vector. The MC RUN code handles neutron transport and bookkeeping, while the action on the neutron within any region is computed using algorithms that may be deterministic, probabilistic, or a combination. Complete versatility is possible because the existing library may be supplemented by any procedures a user is able to code. Some examples are shown

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

    International Nuclear Information System (INIS)

    Lopez Aldama, D.; Noy, R. Capote

    2011-12-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  17. n+235U resonance parameters and neutron multiplicities in the energy region below 100 eV

    Directory of Open Access Journals (Sweden)

    Pigni Marco T.

    2017-01-01

    Full Text Available In August 2016, following the recent effort within the Collaborative International Evaluated Library Organization (CIELO pilot project to improve the neutron cross sections of 235U, Oak Ridge National Laboratory (ORNL collaborated with the International Atomic Energy Agency (IAEA to release a resonance parameter evaluation. This evaluation restores the performance of the evaluated cross sections for the thermal- and above-thermal-solution benchmarks on the basis of newly evaluated thermal neutron constants (TNCs and thermal prompt fission neutron spectra (PFNS. Performed with support from the US Nuclear Criticality Safety Program (NCSP in an effort to provide the highest fidelity general purpose nuclear database for nuclear criticality applications, the resonance parameter evaluation was submitted as an ENDF-compatible file to be part of the next release of the ENDF/B-VIII.0 nuclear data library. The resonance parameter evaluation methodology used the Reich-Moore approximation of the R-matrix formalism implemented in the code SAMMY to fit the available time-of-flight (TOF measured data for the thermal induced cross section of n+235U up to 100 eV. While maintaining reasonably good agreement with the experimental data, the validation analysis focused on restoring the benchmark performance for 235U solutions by combining changes to the resonance parameters and to the prompt resonance v̅ below 100 eV.

  18. Parallel solutions of the two-group neutron diffusion equations

    International Nuclear Information System (INIS)

    Zee, K.S.; Turinsky, P.J.

    1987-01-01

    Recent efforts to adapt various numerical solution algorithms to parallel computer architectures have addressed the possibility of substantially reducing the running time of few-group neutron diffusion calculations. The authors have developed an efficient iterative parallel algorithm and an associated computer code for the rapid solution of the finite difference method representation of the two-group neutron diffusion equations on the CRAY X/MP-48 supercomputer having multi-CPUs and vector pipelines. For realistic simulation of light water reactor cores, the code employees a macroscopic depletion model with trace capability for selected fission product transients and critical boron. In addition to this, moderator and fuel temperature feedback models are also incorporated into the code. The validity of the physics models used in the code were benchmarked against qualified codes and proved accurate. This work is an extension of previous work in that various feedback effects are accounted for in the system; the entire code is structured to accommodate extensive vectorization; and an additional parallelism by multitasking is achieved not only for the solution of the matrix equations associated with the inner iterations but also for the other segments of the code, e.g., outer iterations

  19. Two-dimensional shielding benchmarks for iron at YAYOI, (1)

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; An, Shigehiro; Kasai, Shigeru; Miyasaka, Shun-ichi; Koyama, Kinji.

    The aim of this work is to assess the collapsed neutron and gamma multigroup cross sections for two dimensional discrete ordinate transport code. Two dimensional distributions of neutron flux and gamma ray dose through a 70cm thick and 94cm square iron shield were measured at the fast neutron source reactor ''YAYOI''. The iron shield was placed over the lead reflector in the vertical experimental column surrounded by heavy concrete wall. The detectors used in this experiment were threshold detectors In, Ni, Al, Mg, Fe and Zn, sandwitch resonance detectors Au, W and Co, activation foils Au for neutrons and thermoluminescence detectors for gamma ray dose. The experimental results were compared with the calculated ones by the discrete ordinate transport code ANISN and TWOTRAN. The region-wise, coupled neutron-gamma multigroup cross-sections (100n+20gamma, EURLIB structure) were generated from ENDF/B-IV library for neutrons and POPOP4 library for gamma-ray production cross-sections by using the code system RADHEAT. The effective microscopic neutron cross sections were obtained from the infinite dilution values applying ABBN type self-shielding factors. The gamma ray production multigroup cross-sections were calculated from these effective microscopic neutron cross-sections. For two-dimensional calculations the group constants were collapsed into 10 neutron groups and 3 gamma groups by using ANISN. (auth.)

  20. Off-line correction for excessive constant-fraction-discriminator walk in neutron time-of-flight experiments

    International Nuclear Information System (INIS)

    Heilbronn, Lawrence; Iwata, Yoshiyuki; Iwase, H.

    2003-01-01

    A method for reducing excessive constant-fraction-discriminator walk that utilizes experimental data in the off-line analysis stage is introduced. Excessive walk is defined here as any walk that leads to an overall timing resolution that is much greater than the intrinsic timing resolution of the detection system. The method is able to reduce the contribution to the overall timing resolution from the walk that is equal to or less than the intrinsic timing resolution of the detectors. Although the method is explained in the context of a neutron time-of-flight experiment, it is applicable to any data set that satisfies two conditions. (1) A measure of the signal amplitude for each event must be recorded on an event-by-event basis; and (2) There must be a distinguishable class of events present where the timing information is known a priori