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Sample records for multigroup cross-section library

  1. NDS multigroup cross section libraries

    International Nuclear Information System (INIS)

    DayDay, N.

    1981-12-01

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

  2. 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

  3. 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

  4. 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

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

    International Nuclear Information System (INIS)

    Gastaldi, B.

    1986-07-01

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

  6. 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

  7. 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)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  9. 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

  10. 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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

    A revised multigroup cross-section library based ON ENDF/B-VI Release 3 has been produced for light water reactor shielding and reactor pressure vessel dosimetry applications. This new broad-group library, which is designated BUGLE-96, represents an improvement over the BUGLE-93 library released in February 1994 and is expected to replace te BUGLE-93 data. The cross-section processing methodology is the same as that used for producing BUGLE-93 and is consistent with ANSI/ANS 6.1.2. As an added feature, cross-section sets having upscatter data for four thermal neutron groups are included in the BUGLE-96 package available from the Radiation Shielding Information Center. The upscattering data should improve the application of this library to the calculation of more accurate thermal fluences, although more computer time will be required. The incorporation of feedback from users has resulted in a data library that addresses a wider spectrum of user needs

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    1980-05-01

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

  14. 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)

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

    International Nuclear Information System (INIS)

    Ono, S.; Corcuera, R.P.

    1984-01-01

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

  16. 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

  17. 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

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

    International Nuclear Information System (INIS)

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

    1980-04-01

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

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

    International Nuclear Information System (INIS)

    Mori, Takamasa; Nakagawa, Masayuki; Ishiguro, Yukio

    2002-01-01

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

  20. 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)

  1. 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

  2. Review of multigroup nuclear cross-section processing

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-10-01

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

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

    International Nuclear Information System (INIS)

    Keinert, J.; Mattes, M.

    1975-01-01

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

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

    International Nuclear Information System (INIS)

    Kim, Jung-Do; Lee, Jong Tai

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  6. 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)

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

    International Nuclear Information System (INIS)

    Chalhoub, E.S.; Anaf, J.

    1988-02-01

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

  8. Evaluated cross section libraries

    International Nuclear Information System (INIS)

    Maqurno, B.A.

    1976-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-05-01

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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  12. 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).

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

    International Nuclear Information System (INIS)

    Renke, C.A.C.

    1982-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-11-08

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

  15. 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)

  16. RZ calculations for self shielded multigroup cross sections

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-07-01

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

  17. RZ calculations for self shielded multigroup cross sections

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  18. 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.

  19. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Alpan, A.; Haghighat, A.

    1999-07-01

    Multigroup cross-section libraries are needed in performing neutronics calculations. These libraries are referred to as broad-group libraries. The number of energy groups and group structure are highly dependent on the application and/or user's objectives. For example, for shielding calculations, broad-group libraries such as SAILOR and BUGLE with 47-neutron and 20-gamma energy groups are used. The common procedure to obtain a broad-group library is a three-step process: (1) processing pointwise ENDF (PENDF) format cross sections; (2) generating fine-group cross sections; and (3) collapsing fine-group cross sections to broad-group. The NJOY code is used to prepare fine-group cross sections by processing pointwise ENDF data. The code has several modules, each one performing a specific task. For instance, the module RECONR performs linearization and reconstruction of the cross sections, and the module GROUPR generates multigroup self-shielded cross sections. After fine-group, i.e., groupwise ENDF (GENDF), cross sections are produced, cross sections are self-shielded, and a one-dimensional transport calculation is performed to obtain flux spectra at specific regions in the model. These fluxes are then used as weighting functions to collapse the fine-group cross sections to obtain a broad-group cross-section library. The third step described is commonly performed by the AMPX code system. SMILER converts NJOY GENDF filed to AMPX master libraries, AJAX collects the master libraries. BONAMI performs self-shielding calculations, NITAWL converts the AMPX master library to a working library, XSDRNPM performs one-dimensional transport calculations, and MALOCS collapses fine-group cross sections to broad-group. Finally, ALPO is used to generate ANISN format libraries. In this three-step procedure, generally NJOY requires the largest amount of CPU time. This time varies depending on the user's specified parameters for each module, such as reconstruction tolerances

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

    International Nuclear Information System (INIS)

    Alpan, A.; Haghighat, A.

    1999-01-01

    Multigroup cross-section libraries are needed in performing neutronics calculations. These libraries are referred to as broad-group libraries. The number of energy groups and group structure are highly dependent on the application and/or user's objectives. For example, for shielding calculations, broad-group libraries such as SAILOR and BUGLE with 47-neutron and 20-gamma energy groups are used. The common procedure to obtain a broad-group library is a three-step process: (1) processing pointwise ENDF (PENDF) format cross sections; (2) generating fine-group cross sections; and (3) collapsing fine-group cross sections to broad-group. The NJOY code is used to prepare fine-group cross sections by processing pointwise ENDF data. The code has several modules, each one performing a specific task. For instance, the module RECONR performs linearization and reconstruction of the cross sections, and the module GROUPR generates multigroup self-shielded cross sections. After fine-group, i.e., groupwise ENDF (GENDF), cross sections are produced, cross sections are self-shielded, and a one-dimensional transport calculation is performed to obtain flux spectra at specific regions in the model. These fluxes are then used as weighting functions to collapse the fine-group cross sections to obtain a broad-group cross-section library. The third step described is commonly performed by the AMPX code system. SMILER converts NJOY GENDF filed to AMPX master libraries, AJAX collects the master libraries. BONAMI performs self-shielding calculations, NITAWL converts the AMPX master library to a working library, XSDRNPM performs one-dimensional transport calculations, and MALOCS collapses fine-group cross sections to broad-group. Finally, ALPO is used to generate ANISN format libraries. In this three-step procedure, generally NJOY requires the largest amount of CPU time. This time varies depending on the user's specified parameters for each module, such as reconstruction tolerances, temperatures

  2. 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)

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

    International Nuclear Information System (INIS)

    Rahman, Mafizur; Takano, Hideki

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Williams, M.L.

    1982-07-01

    The effect that interference between resolved resonances has on averaging multigroup cross sections is examined for thermal reactor-type problems. A simple and efficient numerical scheme is presented to correct a preprocessed multigroup library for interference effects. The procedure is implemented in a design oriented lattice physics computer code and compared with rigorous numerical calculations. The approximate method for computing resonance interference correction factors is applied to obtaining fine-group cross sections for a homogeneous uranium-plutonium mixture and a uranium oxide lattice. It was found that some fine group cross sections are changed by more than 40% due to resonance interference. The change in resonance interference correction factors due to burnup of a PWR fuel pin is examined and found to be small. The effect of resolved resonance interference on collapsed broad-group cross sections for thermal reactor calculations is discussed

  5. Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Clarno, Kevin T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gentry, Cole [Univ. of Tennessee, Knoxville, TN (United States); Wiarda, Dorothea [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Williams, Mark L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Kochunas, Brendan [Univ. of Michigan, Ann Arbor, MI (United States); Liu, Yuxuan [Univ. of Michigan, Ann Arbor, MI (United States); Palmtag, Scott [Core Physics, Inc., Wilmington, NC (United States); Godfrey, Andrew T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-03-01

    The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include (1) a subgroup method for resonance selfshielding and (2) a whole-core transport solver with a 2-D/1-D synthesis method. The MPACT code requires a cross section library to support all the MPACT core simulation capabilities which would be the most influencing component for simulation accuracy.

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  7. Nuclear data processing and multigroup cross section generation

    International Nuclear Information System (INIS)

    Kim, Jeong Do; Kil, Chung Sub

    1996-01-01

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

  8. 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)

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  12. 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)

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

    International Nuclear Information System (INIS)

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

    1982-10-01

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

  14. 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)

  15. 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.

  16. 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)

  17. 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)

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

    International Nuclear Information System (INIS)

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

    1995-05-01

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

  19. 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)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-10-01

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

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

    International Nuclear Information System (INIS)

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

    1992-10-01

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

  3. 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)

  4. 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.)

  5. 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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Ganesan, S.

    1994-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    Garg, S.B.; Kumar, Ashok

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    1988-11-01

    We have developed a group of computer codes to realize the accurate transport calculation by using the multi-group double-differential form cross section. This type of cross section can correctly take account of the energy-angle correlated reaction kinematics. Accordingly, the transport phenomena in materials with highly anisotropic scattering are accurately calculated by using this cross section. They include the following four codes or code systems: PROF-DD : a code system to generate the multi-group double-differential form cross section library by processing basic nuclear data file compiled in the ENDF / B-IV or -V format, ANISN-DD : a one-dimensional transport code based on the discrete ordinate method, DOT-DD : a two-dimensional transport code based on the discrete ordinate method, MORSE-DD : a three-dimensional transport code based on the Monte Carlo method. In addition to these codes, several auxiliary codes have been developed to process calculated results. This report describes the calculation algorithm employed in these codes and how to use them. (author)

  12. AFCI-2.0 Neutron Cross Section Covariance Library

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-03-01

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

  13. AFCI-2.0 Neutron Cross Section Covariance Library

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  14. 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

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

    International Nuclear Information System (INIS)

    Anaf, Jaime; Chalhoub, E.S.

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

    1976-06-01

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

  17. 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

  18. 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.

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

    International Nuclear Information System (INIS)

    Bastos, H.F.B.N.

    1979-01-01

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

  20. 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)

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

    International Nuclear Information System (INIS)

    Anton, V.

    1979-05-01

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

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

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1987-11-01

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

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

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1988-02-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  7. Obtaining incremental multigroup cross sections for CANDU super cells with reactivity devices

    International Nuclear Information System (INIS)

    Balaceanu, V.; Constantin, M.

    2001-01-01

    In the last 20 years a multigroup methodology WIMS - PIJXYZ (WP) was developed and validated at INR Pitesti for obtaining incremental cross sections for reactivity devices in CANDU reactors. This is an alternate methodology to the CANDU classic methodology (experimentally adjusted) based on the POWDERPUFS and MULTICELL computer codes. The 2D supercell calculation performed with the WIMS code, that is a NEA Data Bank transport code, and which produces multigroup cross sections (on 18 energy groups) for CANDU supercell material (standard and perturbed, with and without reactivity devices). To obtain an as correct as possible 3D modelling for the CANDU supercells containing reactivity devices, the WIMS cross sections are used as input data for the PIJXYZ code, thus obtaining homogenized cross sections for CANDU supercells. PIJXYZ is an integral transport code based on the formalism of the first collision probabilities. It is analogue to the SHETAN code and it was created for neutron analyzes at cell level for CANDU type reactors were the reactivity devices are perpendicular to the fuel channels. The coordinate system used in PIJXYZ is a mixed one, namely a rectangular-cylindrical system. The geometric model used in PIJXYZ is presented. The fuel beam is represented by a horizontal cylinder and the reactivity device by a vertical one both cylinders being immersed in the moderator. Two supercell types were considered: a perturbed supercell (containing a reactivity device) and the standard supercell were the place of reactivity device is occupied by the moderator. The incremental cross sections for reactivity device are obtained as differences between the homogenized over supercell cross sections (with reactivity device) and homogenized over standards supercell (without device) cross sections. The PIJXYZ computation may be done on an energy cutting with 2 up to 18 groups. The validation of VIMS - PIJXYZ was done on the basis of several benchmark and by comparison with

  8. 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

  9. Testing of cross section libraries for TRIGA criticality benchmark

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  12. 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)

  13. 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)

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

    International Nuclear Information System (INIS)

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

    1976-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  16. Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

    International Nuclear Information System (INIS)

    Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

    1986-02-01

    Multigroup cross sections (66 neutron groups and 22 photon groups) are described for neutron energies from thermal to 400 MeV. The elements considered are hydrogen, 10 B, 11 B, carbon, nitrogen, oxygen, sodium, magnesium, aluminum, silicon, sulfur, potassium, calcium, chromium, iron, nickel, tungsten, and lead. The cross section data presented are a revision of similar data presented previously. In the case of iron, transport calculations using the earlier and the revised cross sections are presented and compared, and significant differences are found. The revised cross sections are available from the Radiation Shielding information Center of the Oak Ridge National Laboratory. 32 refs., 5 figs., 3 tabs

  17. A new modelling of the multigroup scattering cross section in deterministic codes for neutron transport

    International Nuclear Information System (INIS)

    Calloo, A.A.

    2012-01-01

    In reactor physics, calculation schemes with deterministic codes are validated with respect to a reference Monte Carlo code. The remaining biases are attributed to the approximations and models induced by the multigroup theory (self-shielding models and expansion of the scattering law using Legendre polynomials) to represent physical phenomena (resonant absorption and scattering anisotropy respectively). This work focuses on the relevance of a polynomial expansion to model the scattering law. Since the outset of reactor physics, the latter has been expanded on a truncated Legendre polynomial basis. However, the transfer cross sections are highly anisotropic, with non-zero values for a very small range of the cosine of the scattering angle. Besides, the finer the energy mesh and the lighter the scattering nucleus, the more exacerbated is the peaked shape of this cross section. As such, the Legendre expansion is less suited to represent the scattering law. Furthermore, this model induces negative values which are non-physical. In this work, various scattering laws are briefly described and the limitations of the existing model are pointed out. Hence, piecewise-constant functions have been used to represent the multigroup scattering cross section. This representation requires a different model for the diffusion source. The discrete ordinates method which is widely employed to solve the transport equation has been adapted. Thus, the finite volume method for angular discretization has been developed and implemented in Paris environment which hosts the S n solver, Snatch. The angular finite volume method has been compared to the collocation method with Legendre moments to ensure its proper performance. Moreover, unlike the latter, this method is adapted for both the Legendre moments and the piecewise-constant functions representations of the scattering cross section. This hybrid-source method has been validated for different cases: fuel cell in infinite lattice

  18. AMPX: a modular system for multigroup cross-section generation and manipulation

    International Nuclear Information System (INIS)

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

    1978-01-01

    The AMPX system, developed at the Oak Ridge National Laboratory over the past seven years, is a collection of computer programs in a modular arrangement. Starting with ENDF-formatted nuclear data files, the system includes a full range of features needed to produce and use multigroup neutron, gamma-ray production, and gamma-ray interaction cross-section data. The balance between production and analysis is roughly even; thus, the system serves a wide variety of needs. The modularity is particularly attractive, since it allows the user to choose an arbitrary execution sequence from the approximately 40 to 50 modules available in the system. The modularity also allows selection from different treatments; e.g., the Nordheim method, a full-blown integral transport calculation, the Bondarenko method, or other alternative can be selected for resonance shielding. 2 figures

  19. 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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-05-16

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

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

    International Nuclear Information System (INIS)

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

    1980-05-01

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

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

    International Nuclear Information System (INIS)

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

    1980-06-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  5. Formats and processing of evaluated nuclear data into multigroup cross-sections

    International Nuclear Information System (INIS)

    Motta, M.

    1984-01-01

    The first part of these lectures concerns the data in nuclear files and their manipulation. The structure of the data files as divided into the resonance region (subdivided into the resolved and the unresolved regions) and the continuum region is presented. The reactions concerned are the elastic scattering; the radiative capture and the fission methods for averaging the cross sections are given. Then, the group averaging formulas and the self-shielding factors are presented in some detail. The second part concerns a presentation of nuclear data files handling and conversion. The main libraries are listed and several maintenance computer codes presented. The way the conversion among different files is handled is also presented. The listings of several BASIC programs for different cross section calculations are given. These codes are self-guided

  6. Library of neutron cross sections of the Thermos code

    International Nuclear Information System (INIS)

    Alonso V, G.; Hernandez L, H.

    1991-10-01

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

  7. 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)

  8. 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)

  9. 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.

  10. Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

    International Nuclear Information System (INIS)

    Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

    1986-01-01

    For a variety of applications, e.g., accelerator shielding design, neutrons in radiotherapy, radiation damage studies, etc., it is necessary to carry out transport calculations involving medium-energy (greater than or equal to20 MeV) neutrons. A previous paper described neutron-photon multigroup cross sections in the ANISN format for neutrons from thermal to 400 MeV. In the present paper the cross-section data presented previously have been revised to make them agree with available experimental data. 7 refs., 1 fig

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

    International Nuclear Information System (INIS)

    2001-01-01

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

  12. MOX Cross-Section Libraries for ORIGEN-ARP

    International Nuclear Information System (INIS)

    Gauld, I.C.

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    1977-08-01

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

  14. AXMIX program for cross section mixing and library arrangement

    International Nuclear Information System (INIS)

    Haynes, G.C.

    1976-03-01

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

  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. 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)

  17. 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)

  18. 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)

  19. 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)

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

    International Nuclear Information System (INIS)

    Gopalakrishnan, V.; Ganesan, S.

    1988-01-01

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

  1. TIMS-1, Multigroup Cross-Sections of Heavy Isotope Mixture with Resonance from ENDF/B

    International Nuclear Information System (INIS)

    Takano, Hideki; Ishiguro, Yukio; Matsui, Yasushi

    1984-01-01

    1 - Description of problem or function: TIMS-1 is a code for calculating the group constants of heavy resonant nuclei by using ENDF/ B-4 format data. This code calculates infinitely dilute cross sections and self-shielding factors as a function of composition sigma-0 temperature T and R-parameter, where R is the ratio of ato- mic number density of two different resonant nuclei. 2 - Method of solution: In the unresolved resonance region, a ladder of resonance parameters and levels is generated with Monte Carlo method. The temperature dependent cross sections are calculated with the Breit-Wigner single-level and multi-level formula. The neutron spectrum is accurately calculated by solving numerically the neutron slowing down equation using a recurrence formula for neutron slowing down source. 3 - Restrictions on the complexity of the problem: The maximum numbers of energy groups, temperatures and compositions are 60, 4 and 10 respectively

  2. TEMPEST-2, Thermalization Program for Neutron Spectra and Multigroup Cross-Sections

    International Nuclear Information System (INIS)

    Gowins, G.

    1984-01-01

    Description of problem or function: TEMPEST2 is a neutron thermalization program based upon the Wigner-Wilkins approximation for light moderators and the Wilkins approximation for heavy moderators. A Maxwellian distribution may also be used. The model used may be selected as a function of energy. The second-order differential equations are integrated directly rather than transformed to the Riccati equation. The program provides microscopic and macroscopic cross-section averages over the thermal neutron spectrum

  3. Application of Gauss quadratures to the calculation of resolved resonance contribution in multigroup cross sections

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1989-01-01

    A program (RESQ) based on quadratures that evaluates, from ENDF/B data, the resolved resonance contribution in group-averaged cross sections (capture, fission and scattering) was developed. Single and Multilevel Breit-Wigner parameters are accepted. Constant weighting function and zero Kelvin were considered. To assure convergence, different quadrature orders may be analysed. Results are compared with other codes' reconstruction and integration methods. (author) [pt

  4. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    Energy Technology Data Exchange (ETDEWEB)

    Díez, C.J., E-mail: cj.diez@upm.es [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Cabellos, O. [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Instituto de Fusión Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain); Martínez, J.S. [Dpto. de Ingeníera Nuclear, Universidad Politécnica de Madrid, 28006 Madrid (Spain)

    2015-01-15

    Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.

  5. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    International Nuclear Information System (INIS)

    Díez, C.J.; Cabellos, O.; Martínez, J.S.

    2015-01-01

    Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties

  6. Nuclear Data Uncertainty Propagation in Depletion Calculations Using Cross Section Uncertainties in One-group or Multi-group

    Science.gov (United States)

    Díez, C. J.; Cabellos, O.; Martínez, J. S.

    2015-01-01

    Several approaches have been developed in last decades to tackle nuclear data uncertainty propagation problems of burn-up calculations. One approach proposed was the Hybrid Method, where uncertainties in nuclear data are propagated only on the depletion part of a burn-up problem. Because only depletion is addressed, only one-group cross sections are necessary, and hence, their collapsed one-group uncertainties. This approach has been applied successfully in several advanced reactor systems like EFIT (ADS-like reactor) or ESFR (Sodium fast reactor) to assess uncertainties on the isotopic composition. However, a comparison with using multi-group energy structures was not carried out, and has to be performed in order to analyse the limitations of using one-group uncertainties.

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  8. 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)

  9. 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)

  10. 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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    Cox, L.J.

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Huang Shien; Wang Kan; Yu Ganglin

    2008-01-01

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

  14. 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

  15. 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)

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

    International Nuclear Information System (INIS)

    Westfall, R.M.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  18. 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

  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. 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)

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

    International Nuclear Information System (INIS)

    Pashchenko, A.B.; Wienke, H.

    1998-01-01

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

  2. 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

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

    DEFF Research Database (Denmark)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Attaya, H.

    1993-01-01

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

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

    CERN Document Server

    Zaidi, H

    1999-01-01

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

  6. 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

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

    International Nuclear Information System (INIS)

    Shoman, Nathan T.; Skutnik, Steven E.

    2016-01-01

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

  8. 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)

  9. 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)

  10. 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).

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

    International Nuclear Information System (INIS)

    Hernandez Lopez, H.

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    Jachic, J.

    1993-09-01

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

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

    International Nuclear Information System (INIS)

    Zhao Qiujuan; Wu Haicheng; Ge Zhigang

    2014-01-01

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

  15. 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

  16. 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)

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

    International Nuclear Information System (INIS)

    Cullen, D E

    2004-01-01

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

  18. 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)

  19. 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

  20. 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

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

    Science.gov (United States)

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

    2016-06-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Cullen, D E

    2011-04-07

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

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

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1978-01-01

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

  5. 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

  6. 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

  7. 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

  8. Re-evaluation of the neutron scattering dynamics in heavy water, generation of multigroup cross sections for THERM-126

    International Nuclear Information System (INIS)

    Keinert, J.

    1982-06-01

    In providing THERM-126 with cross section matrices for deuterium bound in heavy water the IKE phonon spectrum was reevaluated. The changes are modifications in the acoustic part and in the frequency of the second oscillator. Contrary to the phonon spectrum model for D in D 2 O in ENDF/B-IV the broad band of hindered rotations is assumed to be temperature dependent taking into account the diffusive motion of the molecule. With the new model scattering law data S (α, β) are generated in the temperature range 293.6 K-673.6 K. The THERM-126 scattering cross section matrices are calculated up to P 3 . As a validity check a lot of differential and integral cross sections are compared to experiments and benchmarks are recalculated. (orig.) [de

  9. 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)

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

    International Nuclear Information System (INIS)

    2002-01-01

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

  11. 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

  12. 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.

  13. 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)

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    Kim, Kang Seog; Lee, Chung Chan

    2007-03-01

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

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

    International Nuclear Information System (INIS)

    Nguyen Phuoc Lan.

    1982-11-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  18. Verification of the accuracy of Doppler broadened, self-shielded multigroup cross sections for fast power reactor applications

    International Nuclear Information System (INIS)

    Ganesan, S.; Gopalakrishnan, V.; Ramanadhan, M.M.; Cullen, D.E.

    1988-01-01

    Verification results for Doppler broadening and self-shielding are presented. One of the important results presented is that the original SIGMA1 method of numerical Doppler broadening has now been demonstrated to be inaccurate and not capable of producing results to within required accuracies. Fortunately, due to this study, the SIGMA1 method has been significantly improved and the new SIGMA1 is now capable of producing results to within required accuracies. Although this paper presents results based upon using only one code system, it is important to realize that the original SIGMA1 method is presently used in many cross-section processing code systems; the results of this paper indicate that unless these other code systems are updated to include the new SIGMA1 method, the results produced by these code systems could be very inaccurate. The objectives of the IAEA nuclear data processing code verification project are reviewed as well as the requirements for the accuracy of calculation of Doppler coefficients and the present status of these calculations. The initial results of Doppler broadening and self-shielding calculations are presented and the inconsistency of the results which led to the discovery of errors in the original SIGMA1 method of Doppler broadening are pointed out. Analysis of the errors found and improvements in the SIGMA1 method are presented. Improved results are presented in order to demonstrate that the new SIGMA1 method can produce results within required accuracies. Guidelines are presented to limit the uncertainty introduced due to cross-section processing in order to balance available computer resources to accuracy requirements. Finally cross-section processing code users are invited to participate in the IAEA processing code verification project in order to verify the accuracy of their calculated results. (author)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  20. 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)

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-06-01

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

  3. 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)

  4. 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

  5. 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.)

  6. Generation of SCALE 6 Input Data File for Cross Section Library of PWR Spent Fuel

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Cho, Dong Keun

    2010-11-01

    In order to obtain the cross section libraries of the Korean Pressurized water reactor (PWR) spent fuel (SF), SCALE 6 code input files have been generated. The PWR fuel data were obtained from the nuclear design report (NDR) of the current operating PWRs. The input file were prepared for 16 fuel types such as 4 types of Westinghouse 14x14, 3 types of OPR-1000 16x16, 4 types of Westinghouse 16x16, and 6 types of Westinghouse 17x17. For each fuel type, 5 kinds of fuel enrichments have been considered such as 1.5, 2.0 ,3.0, 4.0 and 5.0 wt%. In the SCALE 6 calculation, a ENDF-V 44 group was used. The 25 burnup step until 72000 MWD/T was used. A 1/4 symmetry model was used for 16x16 and 17x17 fuel assembly, and 1/2 symmetry model was used for 14x14 fuel assembly The generated cross section libraries will be used for the source-term analysis of the PWR SF

  7. 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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  9. 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

  10. 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)

  11. System THEMIS. Cross sections processing system from ENDF/B

    Energy Technology Data Exchange (ETDEWEB)

    Gonnord, J.

    1983-09-01

    The THEMIS system allowed to prepare a self punctual and multigroup library for codes solving the TRIPOLI-PROMETHEE transport equation, allowing comparisons with different methods and approximations. The contents of the THEMIS data base was fixed from its use by the PROMETHEE system (punctual Monte Carlo calculations, multigroup calculations, uncertainties analysis and sensitivity studies). The main characteristics of the THEMIS cross section processing system are briefly presented.

  12. System THEMIS. Cross sections processing system from ENDF/B

    International Nuclear Information System (INIS)

    Gonnord, J.

    1983-09-01

    The THEMIS system allowed to prepare a self punctual and multigroup library for codes solving the TRIPOLI-PROMETHEE transport equation, allowing comparisons with different methods and approximations. The contents of the THEMIS data base was fixed from its use by the PROMETHEE system (punctual Monte Carlo calculations, multigroup calculations, uncertainties analysis and sensitivity studies). The main characteristics of the THEMIS cross section processing system are briefly presented [fr

  13. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-10-08

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

  15. 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...

  16. 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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  18. 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)

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

    International Nuclear Information System (INIS)

    2002-01-01

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

  20. 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

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

    International Nuclear Information System (INIS)

    Pashchenko, A.B.

    1996-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-08-15

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

  4. 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

  5. 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

  6. 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

  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 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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  10. UFMGLIB: a multigroup library for the WIMS code

    International Nuclear Information System (INIS)

    Aboustta, Mohamed A.; Mello, Jair Carlos

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Kim Jung-Do; Gil Choong-Sup

    1996-01-01

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

  12. 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)

  13. 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.

  14. 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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

    Mosca, P.

    2009-12-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  19. 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)

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

    International Nuclear Information System (INIS)

    Forrest, R.A.

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    Cullen, D.E.

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Cullen, D E

    2012-02-26

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

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

    International Nuclear Information System (INIS)

    Haeck, W.; Verboomen, B.

    2006-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Haeck, W; Verboomen, B

    2006-10-15

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

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

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Ishikawa, Makoto; Numata, Kazuyuki

    1999-04-01

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

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

    International Nuclear Information System (INIS)

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

    2008-11-01

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

  7. 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.

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

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

  10. 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.

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  12. SCAMPI: A code package for cross-section processing

    International Nuclear Information System (INIS)

    Parks, C.V.; Petrie, L.M.; Bowman, S.M.; Broadhead, B.L.; Greene, N.M.; White, J.E.

    1996-01-01

    The SCAMPI code package consists of a set of SCALE and AMPX modules that have been assembled to facilitate user needs for preparation of problem-specific, multigroup cross-section libraries. The function of each module contained in the SCANTI code package is discussed, along with illustrations of their use in practical analyses. Ideas are presented for future work that can enable one-step processing from a fine-group, problem-independent library to a broad-group, problem-specific library ready for a shielding analysis

  13. SCAMPI: A code package for cross-section processing

    Energy Technology Data Exchange (ETDEWEB)

    Parks, C.V.; Petrie, L.M.; Bowman, S.M.; Broadhead, B.L.; Greene, N.M.; White, J.E.

    1996-04-01

    The SCAMPI code package consists of a set of SCALE and AMPX modules that have been assembled to facilitate user needs for preparation of problem-specific, multigroup cross-section libraries. The function of each module contained in the SCANTI code package is discussed, along with illustrations of their use in practical analyses. Ideas are presented for future work that can enable one-step processing from a fine-group, problem-independent library to a broad-group, problem-specific library ready for a shielding analysis.

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

    International Nuclear Information System (INIS)

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

    2014-03-01

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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    International Nuclear Information System (INIS)

    Zajac, R.; Necas, V.

    2006-01-01

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

  17. 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

  18. 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

  19. Multigroup discrete ordinates solution of Boltzmann-Fokker-Planck equations and cross section library development of ion transport

    International Nuclear Information System (INIS)

    Prinja, A.K.

    1995-08-01

    We have developed and successfully implemented a two-dimensional bilinear discontinuous in space and time, used in conjunction with the S N angular approximation, to numerically solve the time dependent, one-dimensional, one-speed, slab geometry, (ion) transport equation. Numerical results and comparison with analytical solutions have shown that the bilinear-discontinuous (BLD) scheme is third-order accurate in the space ad time dimensions independently. Comparison of the BLD results with diamond-difference methods indicate that the BLD method is both quantitavely and qualitatively superior to the DD scheme. We note that the form of the transport operator is such that these conclusions carry over to energy dependent problems that include the constant-slowing-down-approximation term, and to multiple space dimensions or combinations thereof. An optimized marching or inversion scheme or a parallel algorithm should be investigated to determine if the increased accuracy can compensate for the extra overhead required for a BLD solution, and then could be compared to other discretization methods such as nodal or characteristic schemes

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

    International Nuclear Information System (INIS)

    Iwasaki, Tomohiko; Unesaki, Hironobu; Kitada, Takanori

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-08-01

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

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

    International Nuclear Information System (INIS)

    Pelloni, S.; Stepanek, J.

    1987-01-01

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

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

    Science.gov (United States)

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

    2013-03-21

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

  4. Comparative analysis among several cross section sets

    International Nuclear Information System (INIS)

    Caldeira, A.D.

    1983-01-01

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

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

    International Nuclear Information System (INIS)

    Easter, M.E.

    1985-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  7. 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)

  8. 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

  9. 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)

  10. 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

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

    International Nuclear Information System (INIS)

    1984-01-01

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

  12. Nuclear data and multigroup methods in fast reactor calculations

    International Nuclear Information System (INIS)

    Gur, Y.

    1975-03-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  14. 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)

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

    International Nuclear Information System (INIS)

    Kodeli, Ivan-Alexander

    2005-01-01

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

  16. 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)

  17. 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.)

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

    Science.gov (United States)

    Frisoni, Manuela

    2017-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Goluoglu, S.

    2003-12-01

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

  20. Status of recent fast capture cross section evaluations for important fission product nuclides

    International Nuclear Information System (INIS)

    Gruppelaar, H.

    1982-01-01

    A comparison is made between recent evaluations of fission-product cross sections as given in the CNEN/CEA, ENDF/B-IV, ENDF/V-V, JENDL-1, RCN-2 and RCN-3 data libraries. The intercomparison is restricted to 24 important fission products in a fast power reactor. The evaluation methods used to obtain the various data files are reviewed and possible shortcomings are indicated. A survey is given of the experimental data based used in the various evaluations. Some graphs are included showing the new ENDF/B-V and RCN-3 fastcapture cross-section evaluations. Further intercomparisons are made by means of multi-group and one-group cross sections. It is shown that lumped fission-product cross sections calculated from the most recent versions of the data files are in quite good agreement with each other. This review concludes with a discussion on observed discrepancies and requests for new measurements. 78 references

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

    International Nuclear Information System (INIS)

    Lima Bezerra, J. de.

    1991-04-01

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

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

    Science.gov (United States)

    Pritychenko, B.; Mughabghab, S. F.

    2012-12-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  4. Preparation and benchmarking of ANSL-V cross sections for advanced neutron source reactor studies

    International Nuclear Information System (INIS)

    Arwood, J.W.; Ford, W.E. III; 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

    Research and development for the advanced neutron source (ANS) reactor is being funded by the US Dept. of Energy. This reactor is to provide the world's most intense steady-state source of low-energy neutrons for a national experimental user facility. Pseudo-problem-independent, multigroup cross-section libraries were generated to support ANS design work. The libraries, designated ANSL-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, and MORSE. Included in ANSL-V are 123-material P 3 neutron, 46-material P 0 or P 6 secondary gamma-ray production (SGRP), and 34-material P 6 gamma-ray interaction (GRI) libraries

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-11-12

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

  8. FEMA DFIRM Cross Sections

    Data.gov (United States)

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

  9. 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

  10. 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)

  11. Requests on domestic nuclear data library from BWR design

    International Nuclear Information System (INIS)

    Maruyama, Hiromi

    2003-01-01

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

  12. Status of neutron dosimetry cross sections

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  14. A subroutine for the calculation of resonance cross sections of U-238 in HTR fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Cuniberti, R; Marullo, G C

    1971-02-15

    In this paper, a survey of the codes used at Ispra for the calculations of resonance absorption in HTR fuel elements is presented and a subroutine for the calculation of resonance cross-sections, in a seven groups energy structure, for a HTR lattice of annular type is described. A library of homogeneous resonance integrals and a wide tabulation of lump and kernel Bell factors, and moderators efficiency is given. This paper deals mainly with the problem of taking into account the correct slowing down of neutrons in the graphite and with the derivation of Bell factors to be used in a multigroup calculation scheme.

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

    International Nuclear Information System (INIS)

    Knipe, A.D.

    1975-10-01

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

  16. FENDL/MC. Library of continuous energy cross sections in ACE format for neutron-photon transport calculations with the Monte Carlo N-particle Transport Code system MCNP 4A. Version 1.1 of March 1995. Summary documentation

    International Nuclear Information System (INIS)

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

    1996-01-01

    Selected neutron reaction nuclear data evaluations for elements of interest to the IAEA's program on Fusion Evaluated Nuclear Data Library (FENDL) have been processed into ACE format using the NJOY system by R.E. MacFarlane. This document summarizes the resulting continuous energy cross-section data library FENDL/MC version 1.1. The data are available cost free, upon request from the IAEA Nuclear Data Section, online or on magnetic tape. (author). 1 tab

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

    Energy Technology Data Exchange (ETDEWEB)

    Atchison, F.

    1998-09-01

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

  18. 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

  19. Jet inclusive cross sections

    International Nuclear Information System (INIS)

    Del Duca, V.

    1992-11-01

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

  20. Curves and tables of neutron cross sections

    International Nuclear Information System (INIS)

    Nakagawa, Tsuneo; Asami, Tetsuo; Yoshida, Tadashi

    1990-07-01

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

  1. 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)

  2. Energy meshing techniques for processing ENDF/B-VI cross sections using the AMPX code system

    International Nuclear Information System (INIS)

    Dunn, M.E.; Greene, N.M.; Leal, L.C.

    1999-01-01

    Modern techniques for the establishment of criticality safety for fissile systems invariably require the use of neutronic transport codes with applicable cross-section data. Accurate cross-section data are essential for solving the Boltzmann Transport Equation for fissile systems. In the absence of applicable critical experimental data, the use of independent calculational methods is crucial for the establishment of subcritical limits. Moreover, there are various independent modern transport codes available to the criticality safety analyst (e.g., KENO V.a., MCNP, and MONK). In contrast, there is currently only one complete software package that processes data from the Version 6 format of the Evaluated Nuclear Data File (ENDF) to a format useable by criticality safety codes. To facilitate independent cross-section processing, Oak Ridge National Laboratory (ORNL) is upgrading the AMPX code system to enable independent processing of Version 6 formats using state-of-the-art procedures. The AMPX code system has been in continuous use at ORNL since the early 1970s and is the premier processor for providing multigroup cross sections for criticality safety analysis codes. Within the AMPX system, the module POLIDENT is used to access the resonance parameters in File 2 of an ENDF/B library, generate point cross-section data, and combine the cross sections with File 3 point data. At the heart of any point cross-section processing code is the generation of a suitable energy mesh for representing the data. The purpose of this work is to facilitate the AMPX upgrade through the development of a new and innovative energy meshing technique for processing point cross-section data

  3. Process of cross section generation for radiation shielding calculations, using the NJOY code

    International Nuclear Information System (INIS)

    Ono, S.; Corcuera, R.P.

    1986-10-01

    The process of multigroup cross sections generation for radiation shielding calculations, using the NJOY code, is explained. Photon production cross sections, processed by the GROUPR module, and photon interaction cross sections processed by the GAMINR are given. These data are compared with the data produced by the AMPX system and published data. (author) [pt

  4. 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

  5. Floodplain Cross Section Lines

    Data.gov (United States)

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

  6. Multitrajectory eikonal cross sections

    International Nuclear Information System (INIS)

    Turner, R.E.

    1983-01-01

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

  7. 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.

  8. 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

  9. SENSIT: a cross-section and design sensitivity and uncertainty analysis code

    International Nuclear Information System (INIS)

    Gerstl, S.A.W.

    1980-01-01

    SENSIT computes the sensitivity and uncertainty of a calculated integral response (such as a dose rate) due to input cross sections and their uncertainties. Sensitivity profiles are computed for neutron and gamma-ray reaction cross sections of standard multigroup cross section sets and for secondary energy distributions (SEDs) of multigroup scattering matrices. In the design sensitivity mode, SENSIT computes changes in an integral response due to design changes and gives the appropriate sensitivity coefficients. Cross section uncertainty analyses are performed for three types of input data uncertainties: cross-section covariance matrices for pairs of multigroup reaction cross sections, spectral shape uncertainty parameters for secondary energy distributions (integral SED uncertainties), and covariance matrices for energy-dependent response functions. For all three types of data uncertainties SENSIT computes the resulting variance and estimated standard deviation in an integral response of interest, on the basis of generalized perturbation theory. SENSIT attempts to be more comprehensive than earlier sensitivity analysis codes, such as SWANLAKE

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-05

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

  11. 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

  12. 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

  13. 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

  14. 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

  15. Recommended evaluation procedure for photonuclear cross section

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-06-01

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

  18. 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

  19. Evaluation of cross sections of Th-232 and U-233

    International Nuclear Information System (INIS)

    Dias, A.M.

    1978-01-01

    The cross sections in multigroups of Th-232 and U-233 are evaluated by comparison of theoretical results and experimental data obtained through experiments on the fast reactors IBR-I, EBR-II, BR-I and AETR. The deviation between calculated values and experimental results is about 10%. They are therefore satisfatory for neutronic calculations [pt

  20. FENDL2/A-MCNP, FENDL2/A-VITJE and FENDL2/A-VITJFLAT. The processed FENDL-2 neutron activation cross-section data files. Summary documentation

    International Nuclear Information System (INIS)

    Pashchenko, A.B.; Wienke, H.

    1997-01-01

    This document summarizes the libraries of neutron activation cross-section data processed into the following three formats: continuous energy format as used by the Monte Carlo neutron/photon transport code MCNP4A; VITAMIN-J 175 multigroup format weighted with the VITAMIN-E weighting spectrum as used by the transmutation codes REAC*2/3 and FOUR ACES; VITAMIN-J 175 multigroup ENDF-6 format, with a flat weighting spectrum. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape. (author)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-01-01

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

  4. 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

  5. Radar cross section

    CERN Document Server

    Knott, Gene; Tuley, Michael

    2004-01-01

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

  6. 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 ...

  7. 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)

  8. 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)

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

    International Nuclear Information System (INIS)

    Krishnani, P.D.

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    Petrizzi, L.

    1989-01-01

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

  11. Calculation of atom displacement cross section for structure material

    International Nuclear Information System (INIS)

    Liu Ping; Xu Yiping

    2015-01-01

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

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

    Science.gov (United States)

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

    2018-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  16. Standard cross-section data

    International Nuclear Information System (INIS)

    Carlson, A.D.

    1984-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, W.C.

    1993-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, W.C.

    1993-02-01

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

  19. Effects of space-dependent cross sections on core physics parameters for compact fast spectrum space power reactors

    International Nuclear Information System (INIS)

    Lell, R.M.; Hanan, N.A.

    1987-01-01

    Effects of multigroup neutron cross section generation procedures on core physics parameters for compact fast spectrum reactors have been examined. Homogeneous and space-dependent multigroup cross section sets were generated in 11 and 27 groups for a representative fast reactor core. These cross sections were used to compute various reactor physics parameters for the reference core. Coarse group structure and neglect of space-dependence in the generation procedure resulted in inaccurate computations of reactor flux and power distributions and in significant errors regarding estimates of core reactivity and control system worth. Delayed neutron fraction was insensitive to cross section treatment, and computed reactivity coefficients were only slightly sensitive. However, neutron lifetime was found to be very sensitive to cross section treatment. Deficiencies in multigroup cross sections are reflected in core nuclear design and, consequently, in system mechanical design

  20. SENSIT: a cross-section and design sensitivity and uncertainty analysis code. [In FORTRAN for CDC-7600, IBM 360

    Energy Technology Data Exchange (ETDEWEB)

    Gerstl, S.A.W.

    1980-01-01

    SENSIT computes the sensitivity and uncertainty of a calculated integral response (such as a dose rate) due to input cross sections and their uncertainties. Sensitivity profiles are computed for neutron and gamma-ray reaction cross sections of standard multigroup cross section sets and for secondary energy distributions (SEDs) of multigroup scattering matrices. In the design sensitivity mode, SENSIT computes changes in an integral response due to design changes and gives the appropriate sensitivity coefficients. Cross section uncertainty analyses are performed for three types of input data uncertainties: cross-section covariance matrices for pairs of multigroup reaction cross sections, spectral shape uncertainty parameters for secondary energy distributions (integral SED uncertainties), and covariance matrices for energy-dependent response functions. For all three types of data uncertainties SENSIT computes the resulting variance and estimated standard deviation in an integral response of interest, on the basis of generalized perturbation theory. SENSIT attempts to be more comprehensive than earlier sensitivity analysis codes, such as SWANLAKE.

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

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    2002-03-01

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

  2. Cross sections for atmospheric corrections

    International Nuclear Information System (INIS)

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

    1975-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-09-01

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

  4. Status of pseudo fission product cross sections for fast reactors. Results of the SWG 17, International working party on evaluation coordination of the nuclear science committee, NEA- OECD

    International Nuclear Information System (INIS)

    Gruppelaar, H.; Kloosterman, J.L.; Pijlgroms, B.J.; Rimpault, G.; Smith, P.; Ignatyuk, A.; Koshcheev, V.; Nikolaev, M.; Thsiboulia, A.; Kawai, M.; Nakagawa, T.; Watanabe, T.; Zukeran, A.; Nakajima, Y.; Matsunobu, H.

    1998-08-01

    Within the framework of the SWG17 benchmark organized by a Working Party of the Nuclear Science Committee of the Nuclear Energy Agency (NEA), a comparison of lumped or pseudo fission product cross sections for fast reactors has been made. Four institutions participated with data libraries based on the JEF2.2, EAF-4.2, BROND-2, FONDL-2.1, ADL-3 and JENDL-3.2 evaluated nuclear data files. Several parameters have been compared with each other: the one-group cross sections and reactivity worths of the lumped nuclide for several partial absorption and scattering cross sections, and the one-group cross sections of the individual fission products. Also graphs of the multi-group cross sections of the lumped nuclide have been compared, as well as graphs of capture cross sections for 27 nuclides. From two contributions based on JEF2.2, it can be concluded that the data processing influences the capture cross section by about 1% and the inelastic scattering cross section by 2%. The differences between the lumped cross sections of the different data libraries are surprisingly small: maximum 6% for capture and 9% for the inelastic scattering. Similar results are obtained for the reactivity effects. Since the reactivity worth of the lumped nuclide is dominated by the capture reaction, the maximum spread in the total reactivity worth is still only 5.3%. There is a systematic difference between total, elastic and capture cross sections of JENDL-3.2 and JEF2.2 of the same order of magnitude. Possible reasons for this discrepancy have been indicated. The one-group capture and inelastic scattering cross sections of most of the important individual fission products differ by less than 10% (root mean square values). Larger differences are observed for unstable nuclides where there is a lack of experimental data. For the (n,2n) group cross sections, which are rather sensitive to the weighting spectrum in the fast energy range, these differences are several tens of percents. The final

  5. Calculation of self-shielding factors for cross-sections in the unresolved resonance region using the GRUCON applied program package

    International Nuclear Information System (INIS)

    Sinitsa, V.V.

    1984-11-01

    The author gives a scheme for the calculation of the self-shielding factors in the unresolved resonance region using the GRUCON applied program package. This package is especially created to be used in the conversion of evaluated neutron cross-section data, as available in existing data libraries, into multigroup microscopic constants. A detailed description of the formulae and algorithms used in the programs is given. Some typical examples of calculation are considered and the results are compared with those of other authors. The calculation accuracy is better than 2%

  6. 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)

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Pashchenko, A.B.

    1995-02-01

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

  10. ZZ MATXSLIBJ33, JENDL-3.3 based, 175 N-42 photon groups (VITAMIN-J) MATXS library for discrete ordinates multi-group

    International Nuclear Information System (INIS)

    Kosako, K.; Yamano, N.; Fukahori, T.; Shibata, K.; Hasegawa, A.

    2006-01-01

    1 - Description of program or function: JENDL-3.3 based, 175 neutron-42 photon groups (VITAMIN-J) MATXS library for discrete ordinates multi-group transport codes. Format: MATXS. Number of groups: 175 neutron, 42 gamma-ray. Nuclides: 337 nuclides contained in JENDL-3.3: H-1, H-2, He-3, He-4, Li-6, Li-7, Be-9, B-10, B-11, C-Nat, N-14, N-15, O-16, F-19, Na-23, Mg-24, Mg-25, Mg-26, Al-27, Si-28, Si-29, Si-30, P-31, S-32, S-33, S-34, S-36, Cl-35, Cl-37, Ar-40, K-39, K-40, K-41, Ca-40, Ca-42, Ca-43, Ca-44, Ca-46, Ca-48, Sc-45, Ti-46, Ti-47, Ti-48, Ti-49, Ti-50, V-Nat, Cr-50, Cr-52, Cr-53, Cr-54, Mn-55, Fe-54, Fe-56, Fe-57, Fe-58, Co-59, Ni-58, Ni-60, Ni-61, Ni-62, Ni-64, Cu-63, Cu-65, Ga-69, Ga-71, Ge-70, Ge-72, Ge-73, Ge-74, Ge-76, As-75, Se-74, Se-76, Se-77, Se-78, Se-79, Se-80, Se-82, Br-79, Br-81, Kr-78, Kr-80, Kr-82, Kr-83, Kr-84, Kr-85, Kr-86, Rb-85, Rb-87, Sr-86, Sr-87, Sr-88, Sr-89, Sr-90, Y-89, Y-91, Zr-90, Zr-91, Zr-92, Zr-93, Zr-94, Zr-95, Zr-96, Nb-93, Nb-94, Nb-95, Mo-92, Mo-94, Mo-95, Mo-96, Mo-97, Mo-98, Mo-99, Mo-100, Tc-99, Ru-96, Ru-98, Ru-99, Ru-100, Ru-101, Ru-102, Ru-103, Ru-104, Ru-106, Rh-103, Rh-105, Pd-102, Pd-104, Pd-105, Pd-106, Pd-107, Pd-108, Pd-110, Ag-107, Ag-109, Ag-110m, Cd-106, Cd-108, Cd-110, Cd-111, Cd-112, Cd-113, Cd-114, Cd-116, In-113, In-115, Sn-112, Sn-114, Sn-115, Sn-116, Sn-117, Sn-118, Sn-119, Sn-120, Sn-122, Sn-123, Sn-124, Sn-126, Sb-121, Sb-123, Sb-124, Sb-125, Te-120, Te-122, Te-123, Te-124, Te-125, Te-126, Te-127m, Te-128, Te-129m, Te-130, I-127, I-129, I-131, Xe-124, Xe-126, Xe-128, Xe-129, Xe-130, Xe-131, Xe-132, Xe-133, Xe-134, Xe-135, Xe-136, Cs-133, Cs-134, Cs-135, Cs-136, Cs-137, Ba-130, Ba-132, Ba-134, Ba-135, Ba-136, Ba-137, Ba-138, Ba-140, La-138, La-139, Ce-140, Ce-141, Ce-142, Ce-144, Pr-141, Pr-143, Nd-142, Nd-143, Nd-144, Nd-145, Nd-146, Nd-147, Nd-148, Nd-150, Pm-147, Pm-148, Pm-148m, Pm-149, Sm-144, Sm-147, Sm-148, Sm-149, Sm-150, Sm-151, Sm-152, Sm-153, Sm-154, Eu-151, Eu-152, Eu-153, Eu-154, Eu-155, Eu

  11. Simplified polynomial representation of cross sections for reactor calculation

    International Nuclear Information System (INIS)

    Dias, A.M.; Sakai, M.

    1985-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  13. Mechanized evaluation of neutron cross-sections

    International Nuclear Information System (INIS)

    Horsley, A.; Parker, J.B.

    1967-01-01

    The evaluation work to provide accurate and consistent neutron cross-section data for multigroup neutronics calculations is not fully exploiting the available theoretical and experimental results; this has been so particularly since the introduction of on-line data handling techniques enabled experimenters to turn out vast quantities of numbers. This situation can be radically improved only by mechanizing the evaluation processes. Systems such as the SC1SRS tape will not only largely overcome the task of collecting data but will provide speedy access to it; by using computers and graph-plotting machines to tabulate and display this data, the labour of evaluation can be very greatly reduced. With some types of cross-section there is hope that by using modern curve-fitting techniques the actual evaluation and statistical accounting of the data can be performed automatically. Some areas where automatic evaluation would seem likely to succeed are specified and a discussion of the mathematical difficulties incurred, such as the elimination of anomalous data, is given. Particularly promising is the use of splines in the mechanized evaluation of data. Splines are the mathematical analogues of the draughtsman's spline used in drawing smooth curves. Their principal properties are the excellent approximations they give to the derivatives of a function; in contrast to conventional polynomial fitting, this feature ensures good interpolation and, when required, stable extrapolation. Various methods of using splines in data graduation and the problem of marrying these methods to standard statistical procedures are examined. The results of work done at AWRE with cubic splines on the mechanized evaluation of neutron scattering total cross-section and angular distribution data are presented. (author)

  14. Development of automatic cross section compilation system for MCNP

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Sakurai, Kiyoshi

    1999-01-01

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

  15. 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).

  16. XCOM: Photon Cross Sections Database

    Science.gov (United States)

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

  17. Doppler broadening of cross sections

    International Nuclear Information System (INIS)

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

    1962-12-01

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

  18. Positive Scattering Cross Sections using Constrained Least Squares

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Brown, Forrest B.

    2016-01-01

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

  20. ZZ SNLRML, Dosimetry Cross-Section Recommendations

    International Nuclear Information System (INIS)

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  2. 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

  3. Criticality benchmark comparisons leading to cross-section upgrades

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Lee, Dong Hyuk; Shim, Hyung Jin

    2015-01-01

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

  5. Calculated neutron-activation cross sections for E/sub n/ /le/ 100 MeV for a range of accelerator materials

    International Nuclear Information System (INIS)

    Bozoian, M.; Arthur, E.D.; Perry, R.T.; Wilson, W.B.; Young, P.G.

    1988-01-01

    Activation problems associated with particle accelerators are commonly dominated by reactions of secondary neutrons produced in reactions of beam particles with accelerator or beam stop materials. Measured values of neutron-activation cross sections above a few MeV are sparse. Calculations with the GNASH code have been made for neutrons incident on all stable nuclides of a range of elements common to accelerator materials. These elements include B, C, N, O, Ne, Mg, Al, Si, P, S, Ar, K, Ca, Cr, Mn, Fe, Co, Ni, Cu, Zn, Zr, Mo, Nd, and Sm. Calculations were made for a grid of incident neutron energies extending to 100 MeV. Cross sections leading to the direct production of as many as 87 activation products for each of 84 target nuclide were tabulated on this grid of neutron energies, each beginning with the threshold for the product nuclide's formation. Multigrouped values of these cross sections have been calculated and are being integrated into the cross-section library of the REAC-2 neutron activation code. Illustrative cross sections are presented. 20 refs., 6 figs., 1 tab

  6. WIMSD5, Deterministic Multigroup Reactor Lattice Calculations

    International Nuclear Information System (INIS)

    2004-01-01

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

  7. COMBINE/PC - a portable neutron spectrum and cross-section generation program

    International Nuclear Information System (INIS)

    Nigg, D.W.; Grimesey, R.A.; Curtis, R.L.

    1990-01-01

    Use of personal computers and engineering workstations for complex scientific computations has expanded rapidly in the past few years. This trend is expected to continue in the future with the introduction of increasingly sophisticated microprocessors and microcomputer systems. In response to this, an integrated system of neutronics and radiation transport software suitable for operation in an IBM personal computer (PC)-class environment has been under development at the Idaho National Engineering Laboratory (INEL) for the past 3 years. A key component of this system will be module to produce application-specific multigroup cross-section libraries that can be used in various neutron transport and diffusion theory code modules. This software module, referred to as COMBINE/PC, was recently completed at INEL and is the subject of this paper. COMBINE/PC was developed to provide an ENDF/B-based neutron cross-section generation capability of sufficient sophistication to handle a wide variety of practical fission and fusion-related applications while maintaining a compact machine-independent structure

  8. Photon-splitting cross sections

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  9. Cross section recondensation method via generalized energy condensation theory

    International Nuclear Information System (INIS)

    Douglass, Steven; Rahnema, Farzad

    2011-01-01

    Highlights: → A new method is presented which corrects for core environment error from specular boundaries at the lattice cell level. → Solution obtained with generalized energy condensation provides improved approximation to the core level fine-group flux. → Iterative recondensation of the cross sections and unfolding of the flux provides on-the-fly updating of the core cross sections. → Precomputation of energy integrals and fine-group cross sections allows for easy implementation and efficient solution. → Method has been implemented in 1D and shown to correct the environment error, particularly in strongly heterogeneous cores. - Abstract: The standard multigroup method used in whole-core reactor analysis relies on energy condensed (coarse-group) cross sections generated from single lattice cell calculations, typically with specular reflective boundary conditions. Because these boundary conditions are an approximation and not representative of the core environment for that lattice, an error is introduced in the core solution (both eigenvalue and flux). As current and next generation reactors trend toward increasing assembly and core heterogeneity, this error becomes more significant. The method presented here corrects for this error by generating updated coarse-group cross sections on-the-fly within whole-core reactor calculations without resorting to additional cell calculations. In this paper, the fine-group core flux is unfolded by making use of the recently published Generalized Condensation Theory and the cross sections are recondensed at the whole-core level. By iteratively performing this recondensation, an improved core solution is found in which the core-environment has been fully taken into account. This recondensation method is both easy to implement and computationally very efficient because it requires precomputation and storage of only the energy integrals and fine-group cross sections. In this work, the theoretical basis and development

  10. Neutron Cross Sections for Aluminium

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, Leif

    1963-08-15

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

  11. Accurate Cross Sections for Microanalysis

    OpenAIRE

    Rez, Peter

    2002-01-01

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

  12. 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

  13. Mixed Legendre moments and discrete scattering cross sections for anisotropy representation

    International Nuclear Information System (INIS)

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

    2012-01-01

    This paper deals with the resolution of the integro-differential form of the Boltzmann transport equation for neutron transport in nuclear reactors. In multigroup theory, deterministic codes use transfer cross sections which are expanded on Legendre polynomials. This modelling leads to negative values of the transfer cross section for certain scattering angles, and hence, the multigroup scattering source term is wrongly computed. The first part compares the convergence of 'Legendre-expanded' cross sections with respect to the order used with the method of characteristics (MOC) for Pressurised Water Reactor (PWR) type cells. Furthermore, the cross section is developed using piecewise-constant functions, which better models the multigroup transfer cross section and prevents the occurrence of any negative value for it. The second part focuses on the method of solving the transport equation with the above-mentioned piecewise-constant cross sections for lattice calculations for PWR cells. This expansion thereby constitutes a 'reference' method to compare the conventional Legendre expansion to, and to determine its pertinence when applied to reactor physics calculations. (authors)

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

    International Nuclear Information System (INIS)

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

    1979-12-01

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

  15. Low Energy Neutrino Cross Sections

    International Nuclear Information System (INIS)

    Zeller, G.P.

    2004-01-01

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

  16. Annotated bibliography covering generation and use of evaluated cross section uncertainty files

    International Nuclear Information System (INIS)

    Peelle, R.W.; Burrows, T.W.

    1983-03-01

    Literature references related to definition, generation, and use of evaluated cross section uncertainty (variance-covariance) files are listed with comments intended primarily to guide the reader toward materials of immediate interest. Papers are also cited that cover covariance information for individual experiments and that relate to production and use of multigroup covariance matrices. Titles are divided among several major categories

  17. Inelastic neutron spectra and cross sections for 238 U

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-03-15

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

  19. Terahertz radar cross section measurements

    DEFF Research Database (Denmark)

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

    2010-01-01

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

  20. Pion-nucleus cross sections

    International Nuclear Information System (INIS)

    Barashenkov, V.S.

    1990-01-01

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

  1. Parametric equations for calculation of macroscopic cross sections

    International Nuclear Information System (INIS)

    Botelho, Mario Hugo; Carvalho, Fernando

    2015-01-01

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

  2. FENDL/A-MCNP and FENDL/A-VITJE. The processed neutron activation cross-section data files of the FENDL project. Version 1.1 of March 1995. Summary documentation

    International Nuclear Information System (INIS)

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

    1996-01-01

    This document summarizes a neutron activation cross-section database processed in two formats as generated by F.M. Mann within the project of the Fusion Evaluated Nuclear Data Library (FENDL): in continuous energy format as used by the Monte Carlo neutron/photon transport code MCNP; and in 175 group multigroup format with VIT-E weighting spectrum, as used by the transmutation code REAC*2/3. The data are available from the IAEA Nuclear Data Section online via INTERNET by FTP command, or on magnetic tape. (author). 2 refs, 1 tab

  3. Shielding Factor Method for producing effective cross sections: MINX/SPHINX and the CCCC interface system

    International Nuclear Information System (INIS)

    MacFarlane, R.E.; Weisbin, C.R.; Paik, N.C.

    1978-01-01

    The Shielding Factor Method (SFM) is an economical designer-oriented method for producing the coarse-group space and energy self-shielded cross sections needed for reactor-core analysis. Extensive experience with the ETOX/1DX and ENDRUN/TDOWN systems has made the SFM the method of choice for most US fast-reactor design activities. The MINX/SPHINX system was designed to expand upon the capabilities of the older SFM codes and to incorporate the new standard interfaces for fast-reactor cross sections specified by the Committee for Computer Code Coordination (CCCC). MINX is the cross-section processor. It generates multigroup cross sections, shielding factors, and group-to-group transfer matriccs from ENDF/B-IV and writes them out as CCCC ISOTXS and BRKOXS files. It features detailed pointwise resonance reconstruction, accurate Doppler broadening, and an efficient treatment of anisotropic scattering. SPHINX is the space-and-energy shielding code. It uses specific mixture and geometry information together with equivalence principles to construct shielded macroscopic multigroup cross sections in as many as 240 groups. It then makes a flux calculation by diffusion or transport methods and collapses to an appropriate set of cell-averaged coarse-group effective cross sections. The integration of MINX and SPHINX with the CCCC interface system provides an efficient, accurate, and convenient system for producing effective cross sections for use in fast-reactor problems. The system has also proved useful in shielding and CTR applications. 3 figures, 4 tables

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

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1979-01-01

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

  5. Neutron cross sections for fusion

    International Nuclear Information System (INIS)

    Haight, R.C.

    1979-10-01

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

  6. Negative ion detachment cross sections

    International Nuclear Information System (INIS)

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

    1992-10-01

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  8. Microscopic cross sections: An utopia?

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  9. Microscopic cross sections: An utopia?

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    2007-01-01

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

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

    International Nuclear Information System (INIS)

    Carlson, A.D.

    1990-01-01

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

  13. Sensitivity of LWR fuel cycle costs to uncertainties in detailed thermal cross sections

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Becker, M.; Harris, D.R.

    1979-01-01

    Cross sections averaged over the thermal energy (< 1 or 2 eV) group have been shown to have an important economic role for light-water reactors. Cost implications of thermal cross section uncertainties at the few-group level were reported earlier. When it has been determined that costs are sensitive to a specific thermal-group cross section, it becomes desirable to determine how specific energy-dependent cross sections influence fuel cycle costs. Multigroup cross-section sensitivity coefficients vary with fuel exposure. By changing the shape of a cross section displayed on a view-tube through an interactive graphics system, one can compute the change in few-group cross section using the exposure dependent sensitivity coefficients. With the changed exposure dependent few-group cross section, a new fuel cycle cost is computed by a sequence of batch depletion, core analysis, and fuel batch cost code modules. Fuel cycle costs are generally most sensitive to cross section uncertainties near the peak of the hardened Maxwellian flux

  14. Wind Turbine Radar Cross Section

    Directory of Open Access Journals (Sweden)

    David Jenn

    2012-01-01

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

  15. Measurements and analysis of the 127I and 129I neutron capture and total cross sections

    International Nuclear Information System (INIS)

    Noguere, G.

    2005-01-01

    Most of the experimental work on the interaction of neutrons with matter has focused on materials important to reactor physics and reactor structures. By comparison, the corresponding data for minor actinides or long-lived fission products are poor. A significant demand has developed for improved neutron cross-section data of these little-studied nuclides due to the surge of interest in the transmutation of nuclear waste. With 400 kg of 129 I produced yearly in the reactors of the EU countries and a very long β - half-life of 1.57 x 10 7 years, iodine requires disposal strategies that will isolate this isotope from the environment for long periods of time. Therefore, 129 I is potentially a key long-lived fission product for transmutation applications, since 129 I transmutes in 130 I after a single neutron capture and decays to 130 Xe with a 12.36 h half-life. Accurate capture cross sections would help to reduce uncertainties in waste management concepts. For that purpose, Time-Of-Flight measurements covering the [0.5 eV-100 keV] energy range have been carried out at the 150 MeV pulsed neutron source GELINA of the Institute for Reference Materials and Measurements (IRMM). Two types of experiments have been performed at the IRMM, namely capture and transmission experiments. They are respectively related to the neutron capture and total cross sections. Since the PbI 2 samples used in this work contain natural and radioactive iodine, extensive measurements of 129 I have been carried out under the same experimental conditions as for the 129 I. The data reduction process was performed with the AGS system, and the resonance parameters were extracted with the SAMMY and REFIT shape analysis codes. In a last step, the parameters have been converted into ENDF-6 format and processed with the NJOY code to produce point-wise and multigroup cross sections, as well as MCNP and ERANOS libraries. (author)

  16. 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

  17. 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

  18. ACTIV87 Fast neutron activation cross section file 1987

    International Nuclear Information System (INIS)

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

    1989-10-01

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

  19. Generation of neutron scattering cross sections for silicon dioxide

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  20. K+ nucleus total cross sections

    International Nuclear Information System (INIS)

    Sawafta, R.

    1990-01-01

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

  1. Terahertz radar cross section measurements.

    Science.gov (United States)

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

    2010-12-06

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

  2. [Fast neutron cross section measurements

    International Nuclear Information System (INIS)

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  4. Neutron cross sections: Book of curves

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  5. Nuclear Forensics and Radiochemistry: Cross Sections

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-11-08

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

  6. JENDL gas-production cross section file

    International Nuclear Information System (INIS)

    Nakagawa, Tsuneo; Narita, Tsutomu

    1992-05-01

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

  7. Integral nucleus-nucleus cross sections

    International Nuclear Information System (INIS)

    Barashenkov, V.S.; Kumawat, H.

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-21

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  11. Code implementation of partial-range angular scattering cross sections: GAMMER and MORSE

    International Nuclear Information System (INIS)

    Ward, J.T. Jr.

    1978-01-01

    A partial-range (finite-element) method has been previously developed for representing multigroup angular scattering in Monte Carlo photon transport. Computer application of the method, with preliminary quantitative results is discussed here. A multigroup photon cross section processing code, GAMMER, was written which utilized ENDF File 23 point data and the Klein--Nishina formula for Compton scattering. The cross section module of MORSE, along with several execution routines, were rewritten to permit use of the method with photon transport. Both conventional and partial-range techniques were applied for comparison to calculating angular and spectral penetration of 6-MeV photons through a six-inch iron slab. GAMMER was found to run 90% faster than SMUG, with further improvement evident for multiple-media situations; MORSE cross section storage was reduced by one-third; cross section processing, greatly simplified; and execution time, reduced by 15%. Particle penetration was clearly more forward peaked, as moment accuracy is retained to extremly high order. This method of cross section treatment offers potential savings in both storage and handling, as well as improved accuracy and running time in the actual execution phase. 3 figures, 4 tables

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

    International Nuclear Information System (INIS)

    Prati, A.; Claro, L.H.

    1990-01-01

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

  13. [Fast neutron cross section measurements

    International Nuclear Information System (INIS)

    Knoll, G.F.

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    1976-09-01

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

  15. Relativistic photon-Maxwellian electron cross sections

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  16. Evaluation of fission product neutron cross sections for JENDL

    International Nuclear Information System (INIS)

    1984-01-01

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

  17. Reaction cross section calculation of some alkaline earth elements

    Science.gov (United States)

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

    2017-09-01

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

  18. Reaction cross section calculation of some alkaline earth elements

    Directory of Open Access Journals (Sweden)

    Tel Eyyup

    2017-01-01

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

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

    International Nuclear Information System (INIS)

    López Aldama, Daniel

    2014-11-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  2. CHARTB multigroup transport package

    International Nuclear Information System (INIS)

    Baker, L.

    1979-03-01

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

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

    Directory of Open Access Journals (Sweden)

    Kabach Ouadie

    2017-12-01

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

  4. Background-cross-section-dependent subgroup parameters

    International Nuclear Information System (INIS)

    Yamamoto, Toshihisa

    2003-01-01

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

  5. Scattering cross section for various potential systems

    Directory of Open Access Journals (Sweden)

    Myagmarjav Odsuren

    2017-08-01

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

  6. Scattering cross section for various potential systems

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-15

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

  7. Cross-section methodology in SIMMER

    International Nuclear Information System (INIS)

    Soran, P.D.

    1975-11-01

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

  8. Cross-section methodology in SIMMER

    International Nuclear Information System (INIS)

    Soran, P.D.

    1976-05-01

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

  9. High ET jet cross sections at CDF

    International Nuclear Information System (INIS)

    Flaugher, B.

    1996-08-01

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

  10. Measurements of neutron capture cross sections

    International Nuclear Information System (INIS)

    Nakajima, Yutaka

    1984-01-01

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

  11. Total neutron cross section of lead

    International Nuclear Information System (INIS)

    Kanda, K.; Aizawa, O.

    1976-01-01

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

  12. 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

  13. Electron collision cross sections of mercury

    International Nuclear Information System (INIS)

    Suzuki, Susumu; Kuzuma, Kiyotaka; Itoh, Haruo

    2006-01-01

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

  14. First measurement of the Rayleigh cross section

    NARCIS (Netherlands)

    Naus, H.; Ubachs, W.

    2000-01-01

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

  15. Total cross section of highly excited strings

    International Nuclear Information System (INIS)

    Lizzi, F.; Senda, I.

    1990-01-01

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

  16. Compilation of cross-sections. Pt. 1

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  17. Total cross section results for deuterium electrodisintegration

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  18. Model cross section calculations using LAHET

    International Nuclear Information System (INIS)

    Prael, R.E.

    1992-01-01

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

  19. Vibrational enhancement of total breakup cross sections

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  20. Interference analysis of fission cross section

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  1. Compilation of cross-sections. Pt. 4

    International Nuclear Information System (INIS)

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

    1987-01-01

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

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

    International Nuclear Information System (INIS)

    2005-01-01

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

  3. Activation cross section data file, (1)

    International Nuclear Information System (INIS)

    Yamamuro, Nobuhiro; Iijima, Shungo.

    1989-09-01

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

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  5. Integral-capture measurements and cross-section adjustments for Nd, Sm, and Eu

    International Nuclear Information System (INIS)

    Anderl, R.A.; Schmittroth, F.; Harker, Y.D.

    1981-07-01

    Integral-capture reaction rates are reported for 143 Nd, 144 Nd, 145 Nd, 147 Sm, 151 Eu, 152 Eu, 153 Eu, and 154 Eu irradiated in different neutron spectra in EBR-II. These reaction rates are based primarily on mass-spectrometric measurements of the isotopic atom ratios of the capture product to the target nuclide. The neutron spectra are characterized using passive neutron dosimetry and spectrum-unfolding with the FERRET least-squares data analysis code. Reaction rates for the neutron spectrum monitors were determined by the radiometric technique using Ge(Li) spectrometers. These rates are also reported here. The integral data for the rare-earth samples and for the spectrum monitors were used in multigroup flux/cross-section adtustment analyses with FERRET to generate adjustments to 47 group representations of the ENDF/B-IV capture cross sections for the rare-earth isotopes. These adjusted cross sections are in good agreement with recent differential data and with adjusted cross sections based on STEK integral data. Examples are given of the use of the adjusted cross sections and covariance matrices for cross-section evaluation

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

    International Nuclear Information System (INIS)

    Silbar, R.R.

    1979-01-01

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

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

    International Nuclear Information System (INIS)

    Larsen, Edward

    2013-01-01

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

  8. Measurement of MA fission cross sections at YAYOI

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    International Nuclear Information System (INIS)

    Aggery, A.

    1999-12-01

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

  10. Electron-impact cross sections of Ne

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    Science.gov (United States)

    Hudson, Nathanael Harrison

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

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

    International Nuclear Information System (INIS)

    Matausek, M.V.; Kunc, M.

    1974-01-01

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

  13. Cross-sectional anatomy for computed tomography

    International Nuclear Information System (INIS)

    Farkas, M.L.

    1988-01-01

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

  14. NNLO jet cross sections by subtraction

    International Nuclear Information System (INIS)

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

    2010-06-01

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

  15. Differential Top Cross-section Measurements

    CERN Document Server

    Fenton, Michael James; The ATLAS collaboration

    2017-01-01

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

  16. Hardon cross sections at ultra high energies

    International Nuclear Information System (INIS)

    Yodh, G.B.

    1987-01-01

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

  17. NNLO jet cross sections by subtraction

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-06-15

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

  18. NNLO jet cross sections by subtraction

    CERN Document Server

    Somogyi, Gabor; Trocsanyi, Zoltan

    2010-01-01

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

  19. Calculation of the resonance cross section functions

    International Nuclear Information System (INIS)

    Slipicevic, K.F.

    1967-11-01

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

  20. Fission cross section measurements at intermediate energies

    International Nuclear Information System (INIS)

    Laptev, Alexander

    2005-01-01

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

  1. Measurement of multinucleon transfer cross-sections

    Indian Academy of Sciences (India)

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

  2. Capture cross sections on unstable nuclei

    Science.gov (United States)

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

    2017-09-01

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

  3. Calculation of the resonance cross section functions

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-11-15

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

  4. Pion-nucleus cross sections approximation

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  5. Tachyonic ionization cross sections of hydrogenic systems

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-03-11

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

  6. a cross-sectional analytic study 2014

    African Journals Online (AJOL)

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

  7. Capture cross sections on unstable nuclei

    Directory of Open Access Journals (Sweden)

    Tonchev A.P.

    2017-01-01

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

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

    International Nuclear Information System (INIS)

    Wagschal, J.; Yeivin, Y.

    1985-01-01

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

  9. Methods for calculating anisotropic transfer cross sections

    International Nuclear Information System (INIS)

    Cai, Shaohui; Zhang, Yixin.

    1985-01-01

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

  10. Heisenberg rise of total cross sections

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  11. Evaluation methods for neutron cross section standards

    International Nuclear Information System (INIS)

    Bhat, M.R.

    1980-01-01

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

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

    International Nuclear Information System (INIS)

    Henrikson, D.J.

    1995-01-01

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

  13. Evaluation of ETOG-3Q, ETOG-3, FLANGE-II, XLACS, NJOY and LINEAR/RECENT/GROUPIE computer codes concerning to the resonance contribution and background cross sections

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1988-12-01

    The NJOY and LINEAR/RECENT/GROUPIE calculational procedures for the resolved and unresolved resonance contributions and background cross sections are evaluated. Elastic scattering, fission and capture multigroup cross sections generated by these codes and the previously validated ETOG-3Q, ETOG-3, FLANGE-II and XLACS are compared. Constant weighting function and zero Kelvin temperature are considered. Discrepancies are presented and analysed. (author) [pt

  14. Evaluation of ETOG-3Q/ETOG-3, FLANGE-II, XLACS, NJOY and linear/recent/groupie codes for calculations of resonance and reference cross sections

    International Nuclear Information System (INIS)

    Anaf, J.; Chalhoub, E.S.

    1991-01-01

    The NJOY and LINEAR/RECENT/GROUPIE calculational procedures for the resolved and unresolved resonance contributions and background cross sections are evaluated. Elastic scattering, fission and capture multigroup cross sections generated by these codes and the previously validated ETOG-3Q, ETOG-3, FLANGE-II and XLACS are compared. Constant weighting function and zero Kelvin temperature are considered. Discrepancies are presented and analyzed. (author)

  15. A Pebble Bed Reactor cross section methodology

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  16. Top quark production cross-section measurements

    CERN Document Server

    Chen, Ye; The ATLAS collaboration

    2017-01-01

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

  17. Homogenized group cross sections by Monte Carlo

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  18. Partial cross sections near the higher resonances

    International Nuclear Information System (INIS)

    Falk-Vairant, P.; Valladas, G.

    1961-07-01

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

  19. Photoproton cross section for /sup 19/F

    Energy Technology Data Exchange (ETDEWEB)

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

    1975-02-01

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

  20. Prospects for Precision Neutrino Cross Section Measurements

    Energy Technology Data Exchange (ETDEWEB)

    Harris, Deborah A. [Fermilab

    2016-01-28

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

  1. NNLO jet cross sections by subtraction

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-08-15

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

  2. NNLO jet cross sections by subtraction

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  3. Average cross sections calculated in various neutron fields

    International Nuclear Information System (INIS)

    Shibata, Keiichi

    2002-01-01

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

  4. Optical Model and Cross Section Uncertainties

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-10-05

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

  5. Cross sections for charm production by neutrinos

    Energy Technology Data Exchange (ETDEWEB)

    Ushida, N [Aichi Univ. of Education, Kariya (Japan). Dept. of Physics; Kondo, T [Fermi National Accelerator Lab., Batavia, IL (USA); Fujioka, G; Fukushima, J; Takahashi, Y; Tatsumi, S; Yokoyama, C [Kobe Univ. (Japan). Dept. of Physics; Homma, Y; Tsuzuki, Y [Kobe Univ. (Japan). Coll. of Liberal Arts; Bahk, S

    1983-02-03

    The production of charmed particles has been measured using a hybrid emulsion spectrometer in the Fermilab wide-band neutrino beam. The relative cross section for charged current charmed particle production is sigma(v -> ..mu../sup -/c)/sigma(v -> ..mu../sup -/) = 6.5 +- 1.9/1.8%, and the energy dependence of the cross section is presented. One event with charm pair production was observed. A limit of sigma(v -> ..mu..canti c)/sigma(v -> ..mu..c) < 6% (90% CL) is found for the ratio of charged current pair and single charm production.

  6. Covariance Evaluation Methodology for Neutron Cross Sections

    Energy Technology Data Exchange (ETDEWEB)

    Herman,M.; Arcilla, R.; Mattoon, C.M.; Mughabghab, S.F.; Oblozinsky, P.; Pigni, M.; Pritychenko, b.; Songzoni, A.A.

    2008-09-01

    We present the NNDC-BNL methodology for estimating neutron cross section covariances in thermal, resolved resonance, unresolved resonance and fast neutron regions. The three key elements of the methodology are Atlas of Neutron Resonances, nuclear reaction code EMPIRE, and the Bayesian code implementing Kalman filter concept. The covariance data processing, visualization and distribution capabilities are integral components of the NNDC methodology. We illustrate its application on examples including relatively detailed evaluation of covariances for two individual nuclei and massive production of simple covariance estimates for 307 materials. Certain peculiarities regarding evaluation of covariances for resolved resonances and the consistency between resonance parameter uncertainties and thermal cross section uncertainties are also discussed.

  7. New remarks on KERMA factors and DPA cross section data in ACE files

    International Nuclear Information System (INIS)

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

    2016-01-01

    KERMA factors and DPA cross section data are essential for nuclear heating and material damage estimation in fusion reactor designs. Recently we compared KERMA factors and DPA cross section data in the latest official ACE files of JENDL-4.0, ENDF/B-VII.1, JEFF-3.2 and FENDL-3.0 and it was found out that the KERMA factors and DPA cross section data of a lot of nuclei did not always agree among the nuclear data libraries. We investigated the nuclear data libraries and the nuclear data processing code NJOY and specified new reasons for the discrepancies; (1) incorrect nuclear data and NJOY bugs, (2) huge helium production cross section data, (3) gamma production data format in the nuclear data, (4) no detailed secondary particle data (energy–angular distribution data). These problems should be resolved based on this study.

  8. New remarks on KERMA factors and DPA cross section data in ACE files

    Energy Technology Data Exchange (ETDEWEB)

    Konno, Chikara, E-mail: konno.chikara@jaea.go.jp; Sato, Satoshi; Ohta, Masayuki; Kwon, Saerom; Ochiai, Kentaro

    2016-11-01

    KERMA factors and DPA cross section data are essential for nuclear heating and material damage estimation in fusion reactor designs. Recently we compared KERMA factors and DPA cross section data in the latest official ACE files of JENDL-4.0, ENDF/B-VII.1, JEFF-3.2 and FENDL-3.0 and it was found out that the KERMA factors and DPA cross section data of a lot of nuclei did not always agree among the nuclear data libraries. We investigated the nuclear data libraries and the nuclear data processing code NJOY and specified new reasons for the discrepancies; (1) incorrect nuclear data and NJOY bugs, (2) huge helium production cross section data, (3) gamma production data format in the nuclear data, (4) no detailed secondary particle data (energy–angular distribution data). These problems should be resolved based on this study.

  9. Cross section sensitivity study for fusion blankets incorporating lead neutron multiplier

    International Nuclear Information System (INIS)

    Pelloni, S.; Cheng, E.T.

    1983-01-01

    In the recent European INTOR design, lead has been considered for incorporation in the blanket as either an explicit or implicit neutron multiplier. The blanket employs either Li 2 SiO 3 or Li 17 Pb 83 as tritium breeding material. Nucleonic analysis was performed for this blanket using the DLC37 and DLC41 cross section libraries. The reaction rates were estimated using the reaction cross sections provided with both libraries. In addition to that, they were estimated using the MACKLIB-IV response library. The calculated tritium breeding ratio was found to be 5% less and 15% more in the calculations with DLC41 and DLC41 plus MACKLIB-IV libraries, respectively, than in the calculation with the DLC37 library. The Fe, Pb, and Li cross sections given by the ENDF/B-IV and V were reviewed. A sensitivity study of these cross section uncertainties shows that the tritium breeding ratio is relatively insensitive to the above mentioned partial cross sections. The calculated tritium breeding ratio can be known within +-2%. (Auth.)

  10. Approximation of the cross-sections for charged-particle emission reactions near the threshold

    International Nuclear Information System (INIS)

    Badikov, S.A.; Pashchenko, A.B.

    1990-01-01

    We perform an analytical approximation of the energy dependence of the cross-sections for the reactions (n,p) and (n,γ) from the BOSPOR library, correct them for the latest differential and integral experimental data using the common features, characteristic of the energy dependence of the threshold reaction cross-section and making some physical assumptions. 19 refs, 1 fig., 1 tab

  11. Graphs of neutron cross sections in JSD1000 for radiation shielding safety analysis

    International Nuclear Information System (INIS)

    Yamano, Naoki

    1984-03-01

    Graphs of neutron cross sections and self-shielding factors in the JSD1000 library are presented for radiation shielding safety analysis. The compilation contains various reaction cross sections for 42 nuclides from 1 H to 241 Am in the energy range from 3.51 x 10 -4 eV to 16.5 MeV. The Bondarenko-type self-shielding factors of each reaction are given by the background cross sections from σ 0 = 0 to σ 0 = 10000. (author)

  12. Ecological Panel Inference from Repeated Cross Sections

    NARCIS (Netherlands)

    Pelzer, Ben; Eisinga, Rob; Franses, Philip Hans

    2004-01-01

    This chapter presents a Markov chain model for the estimation of individual-level binary transitions from a time series of independent repeated cross-sectional (RCS) samples. Although RCS samples lack direct information on individual turnover, it is demonstrated here that it is possible with these

  13. Stability of tokamaks with elongated cross section

    International Nuclear Information System (INIS)

    An, C.H.; Bateman, G.

    1978-08-01

    Fixed boundary n = 1 MHD instabilities are studied computationally as a function of diamagnetism (β/sub pol/) and current profile in elongated toroidal equilibria (1 2) or a diamagnetic plasma (β/sub pol/ > 1) with only a mildly elongated cross section

  14. Modelisation of the fission cross section

    International Nuclear Information System (INIS)

    Morariu, Claudia

    2013-03-01

    The neutron cross sections of four nuclear systems (n+ 235 U, n+ 233 U, n+ 241 Am and n+ 237 Np) are studied in the present document. The target nuclei of the first case, like 235 U and 239 Pu, have a large fission cross section after the absorption of thermal neutrons. These nuclei are called 'fissile' nuclei. The other type of nuclei, like 237 Np and 241 Am, fission mostly with fast neutrons, which exceed the fission threshold energy. These types of nuclei are called 'fertile'. The compound nuclei of the fertile nuclei have a binding energy higher than the fission barrier, while for the fissile nuclei the binding energy is lower than the fission barrier. In this work, the neutron induced cross sections for both types of nuclei are evaluated in the fast energy range. The total, reaction and shape-elastic cross sections are calculated by the coupled channel method of the optical model code ECIS, while the compound nucleus mechanism are treated by the statistical models implemented in the codes STATIS, GNASH and TALYS. The STATIS code includes a refined model of the fission process. Results from the theoretical calculations are compared with data retrieved from the experimental data base EXFOR. (author) [fr

  15. Photoelectric absorption cross sections with variable abundances

    Science.gov (United States)

    Balucinska-Church, Monika; Mccammon, Dan

    1992-01-01

    Polynomial fit coefficients have been obtained for the energy dependences of the photoelectric absorption cross sections of 17 astrophysically important elements. These results allow the calculation of X-ray absorption in the energy range 0.03-10 keV in material with noncosmic abundances.

  16. (, 3) Differential cross section of He

    Indian Academy of Sciences (India)

    The angular distribution of the five-fold differential cross section for the electron impact double ionization of He (21 ) and He (23 ) has been studied. The kinematical conditions for maxima/minima in the angular distribution for the two cases have been compared. The two-step process for the double ionization is found to ...

  17. Precise relative cross sections for np scattering

    International Nuclear Information System (INIS)

    Goetz, J.; Brogli-Gysin, C.; Hammans, M.; Haffter, P.; Henneck, R.; Jourdan, J.; Masson, G.; Qin, L.M.; Robinson, S.; Sick, I.; Tuccillo, M.

    1994-01-01

    We present data on the differential cross section for neutron-proton scattering for an incident neutron energy of 67 MeV. These data allow a precise determination of the 1 P 1 phase which, in phase-shift analyses, is strongly correlated with the S-D amplitude which we are measuring via different observables. ((orig.))

  18. Symmetric charge transfer cross section of uranium

    International Nuclear Information System (INIS)

    Shibata, Takemasa; Ogura, Koichi

    1995-03-01

    Symmetric charge transfer cross section of uranium was calculated under consideration of reaction paths. In the charge transfer reaction a d 3/2 electron in the U atom transfers into the d-electron site of U + ( 4 I 9/2 ) ion. The J value of the U atom produced after the reaction is 6, 5, 4 or 3, at impact energy below several tens eV, only resonant charge transfer in which the product atom is ground state (J=6) takes place. Therefore, the cross section is very small (4-5 x 10 -15 cm 2 ) compared with that considered so far. In the energy range of 100-1000eV the cross section increases with the impact energy because near resonant charge transfer in which an s-electron in the U atom transfers into the d-electron site of U + ion. Charge transfer cross section between U + in the first excited state (289 cm -1 ) and U in the ground state was also obtained. (author)

  19. LAMBDA p total cross-section measurement

    CERN Multimedia

    CERN PhotoLab

    1970-01-01

    A view of the apparatus used for the LAMBDA p total cross-section measurement at the time of its installation. The hyperons decaying into a proton and a pion in the conical tank in front were detected in the magnet spectrometer in the upper half of the picture. A novel detection technique using exclusively multiwire proportional chambers was employed.

  20. Measurement cross sections for radioisotopes production

    International Nuclear Information System (INIS)

    Garrido, E.

    2011-01-01

    New radioactive isotopes for nuclear medicine can be produced using particle accelerators. This is one goal of Arronax, a high energy - 70 MeV - high intensity - 2*350 μA - cyclotron set up in Nantes. A priority list was established containing β - - 47 Sc, 67 Cu - β + - 44 Sc, 64 Cu, 82 Sr/ 82 Rb, 68 Ge/ 68 Ga - and α emitters - 211 At. Among these radioisotopes, the Scandium 47 and the Copper 67 have a strong interest in targeted therapy. The optimization of their productions required a good knowledge of their cross-sections but also of all the contaminants created during irradiation. We launched on Arronax a program to measure these production cross-sections using the Stacked-Foils' technique. It consists in irradiating several groups of foils - target, monitor and degrader foils - and in measuring the produced isotopes by γ-spectrometry. The monitor - nat Cu or nat Ni - is used to correct beam loss whereas degrader foils are used to lower beam energy. We chose to study the nat Ti(p,X) 47 Sc and 68 Zn(p,2p) 67 Cu reactions. Targets are respectively natural Titanium foil - bought from Goodfellow - and enriched Zinc 68 deposited on Silver. In the latter case, Zn targets were prepared in-house - electroplating of 68 Zn - and a chemical separation between Copper and Gallium isotopes has to be made before γ counting. Cross-section values for more than 40 different reactions cross-sections have been obtained from 18 MeV to 68 MeV. A comparison with the Talys code is systematically done. Several parameters of theoretical models have been studied and we found that is not possible to reproduce faithfully all the cross-sections with a given set of parameters. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Aggery, A

    1999-12-01

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

  2. Evaluation of cross-section uncertainties using physical constraints for 238U, 239Pu

    International Nuclear Information System (INIS)

    De Saint Jean, Cyrille; Privas, Edwin; Archier, Pascal; Noguere, Gilles; Litaize, Olivier; Leconte, Pierre; Bernard, David

    2014-01-01

    Neutron-induced reactions between 0 eV and 20 MeV are based on various physical properties such as nuclear reaction models, microscopic and integral measurements. Most of the time, the evaluation work is done independently between the resolved resonance range and the continuum, giving rise to mismatches for the cross-sections, larger uncertainties on boundary and no cross-correlation between high-energy domain and resonance range. In addition the use of integral experiment is sometimes only related to central values (evaluation is 'working fine' on a dedicated set of benchmarks) and reductions of uncertainties are not straightforward on cross-sections themselves: working fine could be mathematically reflected by a reduced uncertainty. As the CIELO initiative is to bring experts in each field to propose/discuss these matters, after having presented the status of 238 U and 239 Pu cross-sections covariances evaluation (for JEFF-3.2 as well as the WPEC SG34 subgroup), this paper will present several methodologies that may be used to avoid such effects on covariances. A first idea based on the use of experiments overlapping two energy domains appeared in the near past. It was reviewed and extended to the use of systematic uncertainties (normalisation for example) and for integral experiments as well. In addition, we propose a methodology taking into account physical constraints on an overlapping energy domain where both nuclear reaction models are used (continuity of both cross-sections and derivatives for example). The use of Lagrange multiplier (related to these constraints) in a classical generalised least square procedure will be exposed. Some academic examples will then be presented for both point-wise and multi-group cross-sections to present the methodologies. In addition, new results for 239 Pu will be presented on resonance range and higher energies to reduce capture and fission cross-section uncertainties by using integral experiments (JEZEBEL experiment as

  3. POLIDENT: A Module for Generating Continuous-Energy Cross Sections from ENDF Resonance Data

    Energy Technology Data Exchange (ETDEWEB)

    Dunn, M.E.; Greene, N.M.

    2000-12-01

    POLIDENT (Point Libraries of Data from ENDF/B Tapes) is an AMPX module that accesses the resonance parameters from File 2 of an ENDF/B library and constructs the continuous-energy cross sections in the resonance energy region. The cross sections in the resonance range are subsequently combined with the File 3 background data to construct the cross-section representation over the complete energy range. POLIDENT has the capability to process all resonance reactions that are identified in File 2 of the ENDF/B library. In addition, the code has the capability to process the single- and multi-level Breit-Wigner, Reich-Moore and Adler-Adler resonance formalisms that are identified in File 2. POLIDENT uses a robust energy-mesh-generation scheme that determines the minimum, maximum and points of inflection in the cross-section function in the resolved-resonance region. Furthermore, POLIDENT processes all continuous-energy cross-section reactions that are identified in File 3 of the ENDF/B library and outputs all reactions in an ENDF/B TAB1 format that can be accessed by other AMPX modules.

  4. (n,{alpha}) cross section measurement of gaseous sample using gridded ionization chamber. Cross section determination

    Energy Technology Data Exchange (ETDEWEB)

    Sanami, Toshiya; Baba, Mamoru; Saito, Keiichiro; Ibara, Yasutaka; Hirakawa, Naohiro [Tohoku Univ., Sendai (Japan). Faculty of Engineering

    1997-03-01

    We are developing a method of (n,{alpha}) cross section measurement using gaseous samples in a gridded ionization chamber (GIC). This method enables cross section measurements in large solid angle without the distortion by the energy loss in a sample, but requires a method to estimate the detection efficiency. We solve this problem by using GIC signals and a tight neutron collimation. The validity of this method was confirmed through the {sup 12}C(n,{alpha}{sub 0}){sup 9}Be measurement. We applied this method to the {sup 16}O(n,{alpha}){sup 13}C cross section around 14.1 MeV. (author)

  5. Rotational averaging of multiphoton absorption cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Friese, Daniel H., E-mail: daniel.h.friese@uit.no; Beerepoot, Maarten T. P.; Ruud, Kenneth [Centre for Theoretical and Computational Chemistry, University of Tromsø — The Arctic University of Norway, N-9037 Tromsø (Norway)

    2014-11-28

    Rotational averaging of tensors is a crucial step in the calculation of molecular properties in isotropic media. We present a scheme for the rotational averaging of multiphoton absorption cross sections. We extend existing literature on rotational averaging to even-rank tensors of arbitrary order and derive equations that require only the number of photons as input. In particular, we derive the first explicit expressions for the rotational average of five-, six-, and seven-photon absorption cross sections. This work is one of the required steps in making the calculation of these higher-order absorption properties possible. The results can be applied to any even-rank tensor provided linearly polarized light is used.

  6. Measurements of neutron spallation cross section. 2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, E.; Nakamura, T. [Tohoku Univ., Sendai (Japan). Cyclotron and Radioisotope Center; Imamura, M.; Nakao, N.; Shibata, S.; Uwamino, Y.; Nakanishi, N.; Tanaka, Su.

    1997-03-01

    Neutron spallation cross section of {sup 59}Co(n,xn){sup 60-x}Co, {sup nat}Cu(n,sp){sup 56}Mn, {sup nat}Cu(n,sp){sup 58}Co, {sup nat}Cu(n,xn){sup 60}Cu, {sup nat}Cu(n,xn){sup 61}Cu and {sup nat}Cu(n,sp){sup 65}Ni was measured in the quasi-monoenergetic p-Li neutron fields in the energy range above 40 MeV which have been established at three AVF cyclotron facilities of (1) INS of Univ. of Tokyo, (2) TIARA of JAERI and (3) RIKEN. Our experimental data were compared with the ENDF/B-VI high energy file data by Fukahori and the calculated cross section data by Odano. (author)

  7. Structured ion impact: Doubly differential cross sections

    International Nuclear Information System (INIS)

    DuBois, R.D.

    1987-01-01

    The electron emission in coincidence with a projectile that has been ionized has been measured, thus making it possible to separate and identify electrons resulting from these various mechanisms. In 1985, coincidence doubly differential cross sections were measured for 400 to 750 keV/atomic mass unit (amu) He + impact on He, Ne, Ar, Kr, and H 2 O. Cross sections were measured for selected angles and for electron energies ranging from 10 to 1000 eV. Because of the coincidence mode of measurement, the total electron emission was subdivided into its target emission and its projectile emission components. The most interesting findings were that target ionization does not account for the electron emission spectrum at lower electron energies. A sizable percentage of these low-energy electrons were shown to originate as a result of simultaneous projectile/target ionizations. Similar features were observed for all targets and impact energies that were studied

  8. Electron-collision cross sections for iodine

    International Nuclear Information System (INIS)

    Zatsarinny, O.; Bartschat, K.; Garcia, G.; Blanco, F.; Hargreaves, L.R.; Jones, D.B.; Murrie, R.; Brunton, J.R.; Brunger, M.J.; Hoshino, M.; Buckman, S.J.

    2011-01-01

    We present results from a joint experimental and theoretical study of elastic electron scattering from atomic iodine. The experimental results were obtained by subtracting known cross sections from the measured data obtained with a pyrolyzed mixed beam containing a variety of atomic and molecular species. The calculations were performed using both a fully relativistic Dirac B-spline R-matrix (close-coupling) method and an optical model potential approach. Given the difficulty of the problem, the agreement between the two sets of theoretical predictions and the experimental data for the angle-differential and the angle-integrated elastic cross sections at 40 eV and 50 eV is satisfactory.

  9. Absolute partial photoionization cross sections of ethylene

    Science.gov (United States)

    Grimm, F. A.; Whitley, T. A.; Keller, P. R.; Taylor, J. W.

    1991-07-01

    Absolute partial photoionization cross sections for ionization out of the first four valence orbitals to the X 2B 3u, A 2B 3g, B 2A g and C 2B 2u states of the C 2H 4+ ion are presented as a function of photon energy over the energy range from 12 to 26 eV. The experimental results have been compared to previously published relative partial cross sections for the first two bands at 18, 21 and 24 eV. Comparison of the experimental data with continuum multiple scattering Xα calculations provides evidence for extensive autoionization to the X 2B 3u state and confirms the predicted shape resonances in ionization to the A 2B 3g and B 2A g states. Identification of possible transitions for the autoionizing resonances have been made using multiple scattering transition state calculations on Rydberg excited states.

  10. Radar cross section measurements using terahertz waves

    DEFF Research Database (Denmark)

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

    2010-01-01

    Radar cross sections at terahertz frequencies are measured on scale models of aircrafts. A time domain broadband THz system generates freely propagating THz pulses measured with sub-picosecond time resolution. The THz radiation is generated using fs laser pulses by optical rectification...... in order to measure realistic radar cross sections. RCS polar and azimuthal angle plots of F-16 and F-35 are presented....... in a lithium niobate crystal with application of the tilted wave front method, resulting in high electric field THz pulses with a broad band spectrum from 100 GHz up to 4 THz. The corresponding wave lengths are two orders of magnitude smaller than normal radars and we therefore use scale models of size 5-10 cm...

  11. Electron Capture Cross Sections for Stellar Nucleosynthesis

    Directory of Open Access Journals (Sweden)

    P. G. Giannaka

    2015-01-01

    Full Text Available In the first stage of this work, we perform detailed calculations for the cross sections of the electron capture on nuclei under laboratory conditions. Towards this aim we exploit the advantages of a refined version of the proton-neutron quasiparticle random-phase approximation (pn-QRPA and carry out state-by-state evaluations of the rates of exclusive processes that lead to any of the accessible transitions within the chosen model space. In the second stage of our present study, we translate the abovementioned e--capture cross sections to the stellar environment ones by inserting the temperature dependence through a Maxwell-Boltzmann distribution describing the stellar electron gas. As a concrete nuclear target we use the 66Zn isotope, which belongs to the iron group nuclei and plays prominent role in stellar nucleosynthesis at core collapse supernovae environment.

  12. Test of RIPL-2 cross section calculations

    International Nuclear Information System (INIS)

    Herman, M.

    2002-01-01

    The new levels and optical segments and microscopic HF-BCS level densities (part of the density segment) were tested in practical calculations of cross sections for neutron induced reactions on 22 targets (40-Ca, 47-Ti, 52-Cr, 55-Mn, 58-Ni, 63-Cu, 71-Ga, 80-Se, 92-Mo, 93-Nb, 100-Mo, 109-Ag, 114-Cd, 124-Sn, 127-I, 133-Cs, 140-Ce, 153-Eu, 169-Tm, 186-W, 197-Au, 208-Pb). For each target all reactions involving up to 3 neutron, 1 proton and 1 α-particle emissions (subject to actual reaction thresholds) were considered in the incident energy range from 1 keV up to 20 MeV (in some cases up to 27 MeV). In addition, total, elastic, and neutron capture cross sections were calculated

  13. Double differential cross sections of ethane molecule

    Science.gov (United States)

    Kumar, Rajeev

    2018-05-01

    Partial and total double differential cross sections corresponding to various cations C2H6+, C2H4+, C2H5+, C2H3+, C2H2+, CH3+, H+, CH2+, C2H+, H2+, CH+, H3+, C2+ and C+ produced during the direct and dissociative electron ionization of Ethane (C2H6) molecule have been calculated at fixed impinging electron energies 200 and 500eV by using modified Jain-Khare semi empirical approach. The calculation for double differential cross sections is made as a function of energy loss suffered by primary electron and angle of incident. To the best of my knowledge no other data is available for the comparison.

  14. Cross sections required for FMIT dosimetry

    International Nuclear Information System (INIS)

    Gold, R.; McElroy, W.N.; Lippincott, E.P.; Mann, F.M.; Oberg, D.L.; Roberts, J.H.; Ruddy, F.H.

    1980-01-01

    The Fusion Materials Irradiation Test (FMIT) facility, currently under construction, is designed to produce a high flux of high energy neutrons for irradiation effects experiments on fusion reactor materials. Characterization of the flux-fluence-spectrum in this rapidly varying neutron field requires adaptation and extension of currently available dosimetry techniques. This characterization will be carried out by a combination of active, passive, and calculational dosimetry. The goal is to provide the experimenter with accurate neutron flux-fluence-spectra at all positions in the test cell. Plans have been completed for a number of experimental dosimetry stations and provision for these facilities has been incorporated into the FMIT design. Overall needs of the FMIT irradiation damage program delineate goal accuracies for dosimetry that, in turn, create new requirements for high energy neutron cross section data. Recommendations based on these needs have been derived for required cross section data and accuracies

  15. Elliptical cross section fuel rod study II

    International Nuclear Information System (INIS)

    Taboada, H.; Marajofsky, A.

    1996-01-01

    In this paper it is continued the behavior analysis and comparison between cylindrical fuel rods of circular and elliptical cross sections. Taking into account the accepted models in the literature, the fission gas swelling and release were studied. An analytical comparison between both kinds of rod reveals a sensible gas release reduction in the elliptical case, a 50% swelling reduction due to intragranular bubble coalescence mechanism and an important swelling increase due to migration bubble mechanism. From the safety operation point of view, for the same linear power, an elliptical cross section rod is favored by lower central temperatures, lower gas release rates, greater gas store in ceramic matrix and lower stored energy rates. (author). 6 refs., 8 figs., 1 tab

  16. Electron collision cross sections and radiation chemistry

    International Nuclear Information System (INIS)

    Hatano, Y.

    1983-01-01

    A survey is given of the cross section data needs in radiation chemistry, and of the recent progress in electron impact studies on dissociative excitation of molecules. In the former some of the important target species, processes, and collision energies are presented, while in the latter it is demonstrated that radiation chemistry is a source of new ideas and information in atomic collision research. 37 references, 4 figures

  17. Absolute photoionization cross sections of atomic oxygen

    Science.gov (United States)

    Samson, J. A. R.; Pareek, P. N.

    1985-01-01

    The absolute values of photoionization cross sections of atomic oxygen were measured from the ionization threshold to 120 A. An auto-ionizing resonance belonging to the 2S2P4(4P)3P(3Do, 3So) transition was observed at 479.43 A and another line at 389.97 A. The experimental data is in excellent agreement with rigorous close-coupling calculations that include electron correlations in both the initial and final states.

  18. Total dissociation cross section of halo nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Formanek, J. [Karlova Univ., Prague (Czech Republic). Fakulta Matematicko-Fyzikalni; Lombard, R.J. [Paris-11 Univ., 91 - Orsay (France). Inst. de Physique Nucleaire

    1996-10-01

    Calculations of the total dissociation cross section is performed in the impact parameter representation. The case of {sup 11}Be and {sup 11}Li loosing one and two neutron(s), respectively, by collision on a {sup 12}C target, which remains in its ground state are discussed. The results are found to depend essentially on the rms radius of the halo wave function. (author). 12 refs.

  19. Measurements of Fission Cross Sections of Actinides

    CERN Multimedia

    Wiescher, M; Cox, J; Dahlfors, M

    2002-01-01

    A measurement of the neutron induced fission cross sections of $^{237}$Np, $^{241},{243}$Am and of $^{245}$Cm is proposed for the n_TOF neutron beam. Two sets of fission detectors will be used: one based on PPAC counters and another based on a fast ionization chamber (FIC). A total of 5x10$^{18}$ protons are requested for the entire fission measurement campaign.

  20. Cross section of the CMS solenoid

    CERN Multimedia

    Tejinder S. Virdee, CERN

    2005-01-01

    The pictures show a cross section of the CMS solenoid. One can see four layers of the superconducting coil, each of which contains the superconductor (central part, copper coloured - niobium-titanium strands in a copper coating, made into a "Rutherford cable"), surrounded by an ultra-pure aluminium as a magnetic stabilizer, then an aluminium alloy as a mechanical stabilizer. Besides the four layers there is an aluminium mechanical piece that includes pipes that transport the liquid helium.