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Sample records for capture gamma endf

  1. Reproducibility of (n,γ) gamma ray spectrum in Pb under different ENDF/B releases

    Energy Technology Data Exchange (ETDEWEB)

    Kebwaro, J.M., E-mail: jeremiahkebwaro@gmail.com [Department of Physical Sciences, Karatina University, P.O. Box 1957-10101, Karatina (Kenya); He, C.H.; Zhao, Y.L. [School of Nuclear Science and Technology, Xian Jiaotong University, Xian, Shaanxi 710049 (China)

    2016-04-15

    Radiative capture reactions are of interest in shielding design and other fundamental research. In this study the reproducibility of (n,γ) reactions in Pb when cross-section data from different ENDF/B releases are used in the Monte-Carlo code, MCNP, was investigated. Pb was selected for this study because it is widely used in shielding applications where capture reactions are likely to occur. Four different neutron spectra were declared as source in the MCNP model which consisted of a simple spherical geometry. The gamma ray spectra due to the capture reactions were recorded at 10 cm from the center of the sphere. The results reveal that the gamma ray spectrum produced by ENDF/B-V is in reasonable agreement with that produced when ENDF/B-VI.6 is used. However the spectrum produced by ENDF/B-VII does not reveal any primary gamma rays in the higher energy region (E > 3 MeV). It is further observed that the intensities of the capture gamma rays produced when various releases are used differ by a some margin showing that the results are not reproducible. The generated spectra also vary with the spectrum of the source neutrons. The discrepancies observed among various ENDF/B releases could raise concerns to end users and need to be addressed properly during benchmarking calculations before the next release. The evaluation from ENDF to ACE format that is supplied with MCNP should also be examined because errors might have arisen during the evaluation.

  2. Neutron Capture Gamma-Ray Libraries for Nuclear Applications

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  3. Neutron Capture Gamma-Ray Libraries for Nuclear Applications

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  4. Capture Gamma-Ray Libraries for Nuclear Applications

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  5. Evaluations of fission product capture cross sections for ENDF/B-V

    International Nuclear Information System (INIS)

    Schenter, R.E.; Johnson, D.L.; Mann, F.M.; Schmittroth, F.

    1979-01-01

    Capture cross section evaluations were made for the 36 most important fission product absorbers in a fast reactor system. These evaluations were obtained by use of a generalized least-squares approach with calculations being performed with the computer code FERRET. These results will provide the major revisions to the ENDF/B-IV Fission Product Cross Section File which will be released as part of ENDF/B-V. Input for the cross section adjustment calculations included 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, and 4000. Comparisons of these evaluations with recent capture measurements are presented. 14 figures

  6. Multi-parameter study of gammas capture

    International Nuclear Information System (INIS)

    Samama, R.; Nifenecker, H.; Carlos, P.; Delaitre, B.

    1966-06-01

    This equipment is intended for analyzing, recording, and reading simultaneous information from several 'gamma' detectors. It allows multiparameter study of γ-γ cascades emitted after thermal neutrons capture. (authors) [fr

  7. 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.

  8. 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

  9. 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)

  10. ZZ DLC-15 STORM-ISRAEL, Gamma Interaction Cross-Section Library in ENDF/B Format for Transport

    International Nuclear Information System (INIS)

    1972-01-01

    1 - Nature of physical problem solved: Format: Data in ENDF/B file 23 format. Number of groups: energies from 1 KeV to 100 MeV; Nuclides: elements Z=1 to 100. Origin: E. Storm and H.I. Israel compilation. For use in general purpose gamma-ray transport codes. 2 - Method of solution: A discussion of the evaluation and much of the data were published in ref.1. The data are given in barns/atom for energies 1 keV to 100 MeV and for elements Z=1 to 100. The material numbers (MAT) are equal to the atomic numbers (Z). D.J. Dudziak placed the data in ENDF/B BCD format. The reaction type numbers (MT) used are consistent with those recommended in ENDF publications where possible, although several had to be assigned for the purpose. In the newer nomenclature σa(tot) and σh(tot) may be regarded as kerma factors which should be applied to the spectral flux density in a fashion consistent with the transport calculation which determined the flux density. That is, if the transport model assumes bound-electron incoherent scattering and treats secondary photons from pair production and photoelectric reactions, σa(tot) should be used to calculate Kerma. If the model assumes free-electron incoherent scattering and treats pair production and photoelectric reactions as absorption, σh(tot) should be used

  11. Thermal neutron capture gamma-rays

    International Nuclear Information System (INIS)

    Tuli, J.K.

    1983-01-01

    The energy and intensity of gamma rays as seen in thermal neutron capture are presented. Only those (n,α), E = thermal, reactions for which the residual nucleus mass number is greater than or equal to 45 are included. These correspond to evaluations published in Nuclear Data Sheets. The publication source data are contained in the Evaluated Nuclear Structure Data File (ENSDF). The data presented here do not involve any additional evaluation. Appendix I lists all the residual nuclides for which the data are included here. Appendix II gives a cumulated index to A-chain evaluations including the year of publication. The capture gamma ray data are given in two tables - the Table 1 is the list of all gamma rays seen in (n,#betta#) reaction given in the order of increasing energy; the Table II lists the gamma rays according to the nuclide

  12. 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

  13. 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

  14. New lithology compensated capture gamma ray system

    International Nuclear Information System (INIS)

    Peatross, R.F.

    1976-01-01

    The results of the HYDROCARBON* log after a series of field tests in which gamma rays resulting from thermal neutron capture were measured utilizing an energy analyzer and a scintillation counter of unique construction are reported. A brief discussion covers the nuclear physics required for an understanding of gamma spectral logging. Included in the explanation will be the effects of different atoms on neutrons and photons. The HYDROCARBON log utilizes these nuclear principles to record cased hole measurements and quantitatively distinguish possible productive zones from non-productive zones. Different field examples are illustrated showing the response to shaly sands, porosity and water salinity. Interpretation techniques are discussed both qualitatively and quantitatively. The HYDROCARBON log has proven to be a reliable device in the determination of water saturation in sands behind casing even when shale content and porosity are not well known. This technique is also valuable in the location of the present position of gas--oil contacts and water levels

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

    International Nuclear Information System (INIS)

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

    1977-05-01

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

  16. Evaluation of sodium-23 neutron capture cross section data for the ENDF/B V-III file

    International Nuclear Information System (INIS)

    Paik, N.C.; Pitterle, T.A.

    1975-01-01

    The evaluation of neutron cross sections of 23 Na, material number 1156, for the ENDF/B File is described. Cross sections were evaluated between 10 -5 eV and 15 MeV. Experimental data available up to March 1971 were included in the evaluation

  17. Analytical applications of neutron capture gamma-rays

    International Nuclear Information System (INIS)

    Lindstrom, R.M.; Paul, R.L.; Anderson, D.L.; Paul, R.L.

    1997-01-01

    Field and industrial applications of neutron capture gamma-ray spectrometry with isotopic sources or neutron generators are economically important. Geochemical exploration in boreholes is done routinely with neutron probes. Coal and ores are assayed with analyzers adjacent to a conveyor belt in dozens of industrial facilities. The use of capture gamma rays for explosives detection has been described in the literature, both for scanning airline baggage and for characterizing obsolete munitions; a packaged system for the latter is available commercially. Generalizations are drawn from the history of the field, and predictions are made about the future usefulness of capture gamma rays. (author)

  18. Virtual Gamma Ray Radiation Sources through Neutron Radiative Capture

    Energy Technology Data Exchange (ETDEWEB)

    Scott Wilde, Raymond Keegan

    2008-07-01

    The countrate response of a gamma spectrometry system from a neutron radiation source behind a plane of moderating material doped with a nuclide of a large radiative neutron capture cross-section exhibits a countrate response analogous to a gamma radiation source at the same position from the detector. Using a planar, surface area of the neutron moderating material exposed to the neutron radiation produces a larger area under the prompt gamma ray peak in the detector than a smaller area of dimensions relative to the active volume of the gamma detection system.

  19. Comparison of the BNAB-78 and ENDF/B-V evaluated 238U radiative capture data in the energy range from 0.5 to 15 MeV

    International Nuclear Information System (INIS)

    Tolstikov, V.A.

    1991-01-01

    Evaluations of the 238 U capture cross-section as given in the BNAB-78 and ENDF/B-V evaluated data libraries are intercompared and their values compared to recently published data which had not been included in these evaluations. It is concluded that there is a need to re-assess the earlier experimental data, particularly those based on activation measurements, taking secondary neutron reactions and scattering effects into account. It is recommended that a precision measurement of the capture cross-section and its dependence on energy be done in the 1-7 MeV energy range. (author). 18 refs, 1 fig

  20. Gamma rays from fast neutron capture in silicon and sulphur

    International Nuclear Information System (INIS)

    Lindholm, A.; Nilsson, L.; Bergqvist, I.

    1975-01-01

    Gamma-ray spectra from neutron capture in natural samples of silicon and sulphur have been recorded at eight neutron energies between 4 and 15 MeV. Time-of-flight techniques were used to improve the signal-to-background ratio and the gamma radiation was detected by a large NaI(Tl) scintillator. Cross sections have been determined for transitions to individual (or groups of) levels in the final nucleus. Calculations based on the direct-semidirect model show that this model gives a reasonable description of the shapes of the gamma-ray spectra, but fails to account for observed excitation functions. The inclusion of the compound-nucleus capture process gives a conclusive improvement in the description of the excitation functions, in particular at low neutron energies. The ability of the compound-nucleus model to account for the shapes of the gamma-ray spectra is as good as that of the direct-semidirect model. At higher neutron energies, an improvement is obtained for transitions to the region of weakly bound levels, where the single-particle structure is poorly known. (Auth.)

  1. ENDF/B Format

    International Nuclear Information System (INIS)

    Khalil, M.A.

    1975-01-01

    This document is a brief user's description of the format of ENDF/B, the evaluated neutron nuclear data library of the US National Nuclear Data Center. This summary is an aid to customers of the IAEA Nuclear Data Section when receiving data retrievals in ENDF/B format. For more detailed information the report BNL-50274 (ENDF-102) should be consulted. (author)

  2. Neutron Capture Gamma-Ray Spectroscopy. Proceedings of the International Symposium on Neutron Capture Gamma-Ray Spectroscopy

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1969-11-15

    Experimental capabilities in the field of neutron capture gamma-ray spectroscopy have expanded greatly in the last few years; this has been due in large part to the advent of high-quality Ge(Li) detectors, improvements in electronic data processing, and improvements in bent-crystal spectrometers. Previously unsuspected phenomena, such as the '5. 5-MeV1 anomaly, have appeared and new research tools, such as neutron guide tubes, have been brought into use. Equally exciting developments have occurred in the theory of neutron capture. Complex spectra have yielded to analysis after account had been taken of such effects as vibration, rotation and Coriolis forces, and the theoretical prediction of capture spectra seems to be a future possibility. In view of the International Atomic Energy Agency's close interest in this subject and the need for an international exchange of ideas to analyse and study the latest developments, the organizers of the Symposium felt that work on neutron capture gamma-ray spectroscopy had achieved such valuable and significant results that the time had come for this information to be presented, examined and discussed internationally.

  3. Neutron Capture Gamma-Ray Spectroscopy. Proceedings of the International Symposium on Neutron Capture Gamma-Ray Spectroscopy

    International Nuclear Information System (INIS)

    1969-01-01

    Experimental capabilities in the field of neutron capture gamma-ray spectroscopy have expanded greatly in the last few years; this has been due in large part to the advent of high-quality Ge(Li) detectors, improvements in electronic data processing, and improvements in bent-crystal spectrometers. Previously unsuspected phenomena, such as the '5. 5-MeV1 anomaly, have appeared and new research tools, such as neutron guide tubes, have been brought into use. Equally exciting developments have occurred in the theory of neutron capture. Complex spectra have yielded to analysis after account had been taken of such effects as vibration, rotation and Coriolis forces, and the theoretical prediction of capture spectra seems to be a future possibility. In view of the International Atomic Energy Agency's close interest in this subject and the need for an international exchange of ideas to analyse and study the latest developments, the organizers of the Symposium felt that work on neutron capture gamma-ray spectroscopy had achieved such valuable and significant results that the time had come for this information to be presented, examined and discussed internationally

  4. Improving Capture-gamma Libraries for Nonproliferation Applications

    Energy Technology Data Exchange (ETDEWEB)

    Sleaford, Brad W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hurst, Aaron M. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Univ. of California, Berkeley, CA (United States)

    2016-11-01

    This report describes the measurement, evaluation and incorporation of new -ray spectroscopic data into the Evaluated Nuclear Data File (ENDF) for nonproliferation applications. Analysis and processing techniques are described along with key deliverables that have been met over the course of this project. A total of nine new ENDF libraries have been submitted to the National Nuclear Data Center at the Brookhaven National Laboratory and are now available in the ENDF/B-VIII.beta2 release. Furthermore, this project has led to more than ten peer-reviewed publications and provided theses for ve graduate students. This project is a component of the NA-22 venture collaboration on \\Correlated Nuclear Data in Fission Events" (LA14-V-CorrData-PD2Jb).

  5. ENDF/B format

    International Nuclear Information System (INIS)

    Khalil, M.A.; Lemmel, H.D.

    1986-09-01

    This document is a brief user's description of the format of ENDF/B. This format, originally designed for the US Evaluated Nuclear Data File, is recommended for international use. This summary is an aid to customers of the IAEA Nuclear Data Section when receiving data retrievals in ENDF/B format. For more detailed information the report BNL-NCS-50496 (ENDF 102) should be consulted. An Appendix to the present document gives a summary of the format differences between ENDF/B-4 and ENDF/B-5. (author)

  6. Numerical study on determining formation porosity using a boron capture gamma ray technique and MCNP.

    Science.gov (United States)

    Liu, Juntao; Zhang, Feng; Wang, Xinguang; Han, Fei; Yuan, Zhelong

    2014-12-01

    Formation porosity can be determined using the boron capture gamma ray counting ratio with a near to far detector in a pulsed neutron-gamma element logging tool. The thermal neutron distribution, boron capture gamma spectroscopy and porosity response for formations with different water salinity and wellbore diameter characteristics were simulated using the Monte Carlo method. We found that a boron lining improves the signal-to-noise ratio and that the boron capture gamma ray counting ratio has a higher sensitivity for determining porosity than total capture gamma. Copyright © 2014 Elsevier Ltd. All rights reserved.

  7. Gamma spectrum following neutron capture in {sup 167}Er

    Energy Technology Data Exchange (ETDEWEB)

    Visser, D.; Khoo, T.L.; Lister, C.J. [and others

    1995-08-01

    Statistical decay from a highly excited state samples all the lower-lying states and, hence, provides a sensitive measure of the level density. Pairing has a major impact on the level density, e.g. creating a pair gap between the 0- and 2-quasiparticle configurations. Hence the shape of the statistical spectrum contains information on pairing, and can be used to provide information on the reduction of pairing with thermal excitation energy. For this reason, we measured the complete spectrum of {gamma}rays following thermal neutron capture in {sup 167}Er. The experiment was performed at the Brookhaven reactor using Compton-suppressed Ge detectors from TESSA. The spectrum, which was corrected for detector response and efficiency, reveals primary (first-step, high-energy) transitions up to nearly 8 MeV, secondary (last-step, lower-energy) transitions, as we as a continuous statistical component. Effort was expanded to identify all lines from contaminant sources and an upper limit of 5% was tentatively set for their contributions. The spectral shape of the statistical spectrum will be compared with theoretical spectra obtained from a calculation of pairing which accounts for a stepwise reduction of the pair correlations as the number of quasiparticles increases. The primary lines which decay directly to the near-yrast states will also be used to deduce the level densities.

  8. Development of advanced sensing system for antipersonnel mines with neutron capture gamma-ray analysis

    International Nuclear Information System (INIS)

    Iguchi, Tetsuo

    2006-01-01

    Neutron induced prompt gamma-ray analysis (NPGA) for survey of antipersonnel landmines is developed. A concept of sensor system with compact strong accelerator neutron source, simulation of detection and simulation results by trial examinations are stated. The measurement principles, objects, system construction, development of compact accelerator neutron source and high performance neutron capture gamma-ray detector, simulation of detection of landmine are reported. It can detect 10.8 MeV gamma-rays and estimate the incident angle of gamma-ray. Schematic layouts of the compact accelerator neutron resource, the compact Compton gamma camera and sensor unit, the estimation principle of incident angle of gamma-ray, experiments and comparison between the experimental results and the estimation results, a preliminary trial experiment system for sensing antipersonnel mines with neutron capture gamma-ray analysis are illustrated. (S.Y.)

  9. Studies of weak capture-gamma-ray resonances via coincidence techniques

    CERN Document Server

    Rowland, C; Champagne, A E; Dummer, A K; Fitzgerald, R; Harley, E C T; Mosher, J; Runkle, R

    2002-01-01

    A method for measuring weak capture-gamma-ray resonances via gamma gamma-coincidence counting techniques is described. The coincidence apparatus consisted of a large-volume germanium detector and an annular NaI(Tl) crystal. The setup was tested by measuring the weak E sub R =227 keV resonance in sup 2 sup 6 Mg(p,gamma) sup 2 sup 7 Al. Absolute germanium and NaI(Tl) counting efficiencies for a range of gamma-ray energies and for different detector-target geometries are presented. Studies of the gamma-ray background in our spectra are described. Compared to previous work, our method improves the detection sensitivity for weak capture-gamma-ray resonances by a factor of approx 100. The usefulness of the present technique for investigations of interest to nuclear astrophysics is discussed.

  10. Measurements of keV-neutron capture {gamma} rays of fission products. 3

    Energy Technology Data Exchange (ETDEWEB)

    Igashira, Masayuki [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors

    1997-03-01

    {gamma} rays from the keV-neutron capture reactions by {sup 143,145}Nd and {sup 153}Eu have been measured in a neutron energy region of 10 to 80 keV, using a large anti-Compton NaI(Tl) {gamma}-ray spectrometer and the {sup 7}Li(p,n){sup 7}Be pulsed neutron source with a 3-MV Pelletron accelerator. The preliminary results for the capture cross sections and {gamma}-ray spectra of those nuclei are presented and discussed. (author)

  11. Summary of ENDF/B-V evaluations for carbon, calcium, iron, copper, and lead and ENDF/B-V Revision 2 for calcium and iron

    Energy Technology Data Exchange (ETDEWEB)

    Fu, C Y

    1982-09-01

    This report, together with documents already published, describes the ENDF/B-V evaluations of the neutron and gamma-ray-production cross sections for carbon, calcium, iron, copper, and lead and the ENDF/B-V Revision 2 evaluations for calcium and iron.

  12. Tangential channel for nuclear gamma-resonance spectroscopy in thermal neutron capture

    International Nuclear Information System (INIS)

    Belogurov, V.N.; Bondars, H.Ya.; Lapenas, A.A.; Reznikov, R.S.; Senkov, P.E.

    1979-01-01

    Design of a tangential reactor channel which has been made to replace the radial one in the pulsed research reactor IRT-2000 is described. It allows to use the same hole in biological reactor schielding. Characteristics of neutron and gamma-background spectra at the excit of the channel are given and compared with analogous characteristics of the radial one. The gamma background in the tangential channel is lower than in the radial channel. The gamma spectra in the Gd 155 (n, γ)Gd 156 , Gd 157 (n, γ)Gd 158 , Er 167 (n, γ)Er 168 and Hf 177 (n, γ)Hf 178 reactions show that the application of X-ray detection units BDR with the tangential channel allows to carry out the gamma spectrometry of gamma quanta emitted in the thermal neutron capture by both high and low neutron capture cross section nuclei (e.g., Gdsup(157, 155) and Er 167 , Hf 177 , respectively)

  13. Experimental arrangement for production and use of gamma radiation from neutron capture

    International Nuclear Information System (INIS)

    Mafra, Olga Yajgunovitch

    1969-01-01

    This dissertation presents the main characteristics and construction details of collimator system for gamma radiation emitted by atomic nuclei after capturing thermal neutrons. This construction was made in one of the cross channels of IEAR-1 swimming pool reactor of the Atomic Energy Institute of Sao Paulo, Brazil. The energies of gamma radiation available vary range from about 4 MeV and 11 MeV, discreetly. With this experimental arrangement is obtained: high intensity, good collimation and monochrome gamma radiation, important for conducting experiments with gamma radiation. It is also present in this dissertation the description of the techniques employed in determining the intensity of gamma radiation and the extent of contamination in the neutron beam as well as the program list GAMAU that adjusts the gamma spectrum photopeak taken as a Gaussian curve. We intend to use this experimental arrangement for the measurement of cross sections of photonuclear reactions

  14. Neutron-capture gamma-ray analysis of coal for sulfur, iron, silicon and moisture

    International Nuclear Information System (INIS)

    Fay, D.A.

    1979-05-01

    Samples of coal weighing approximately 200 grams placed in a collimated beam of neutrons from the thermal column of the Ames Laboratory Research Reactor produced capture gamma-rays which could be used for the simultaneous determination of sulfur and iron. Spectra from NaI(Tl) and Ge(Li) detectors were used and interferences were located by examining spectra of the major elemental components of coal. In determining sulfur, iron is a potential source of interference when gamma-ray spectra are collected with a NaI(Tl) detector. Corrections for iron interference were made by use of a higher energy iron peak. The possibility of determining silicon in coal was investigated but this element determination was unsuccessful since capture gamma-ray spectrometry lacked the necessary sensitivity for silicon. A linear relation was found between the area of the hydrogen capture peak at 2.23 MeV and the amount of water added to coal

  15. Assembly and calibration of a new experimental apparatus for production and utilization of capture gamma rays

    International Nuclear Information System (INIS)

    Semmler, R.

    1993-01-01

    A new experimental apparatus has been mounted at the tangential beam tube B H 4/12 of the IPEN IEA-R1 (2 MW) reactor, for production and utilization of capture gamma rays. In this type of experiment, monochromatic gamma radiation, with energy resolution of about 10 eV, is produced by thermal neutron capture in several materials placed near the reactor core. By changing the target material it was possible to obtain up to 30 gamma lines in the 5 to 11 MeV energy range and so, the present experimental arrangement may be considered as an excellent gamma ray source for photonuclear reactions studies in low excitation energies. (author)

  16. EVALPLOT2007, ENDF Plots Cross Section, Angular Distribution and Energy Distribution

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of program or function: EVALPLOT is designed to plot evaluated cross sections in the ENDF/B format. The program plots cross sections, angular distributions, energy distributions and other parameters. IAEA1322/16: This version include the updates up to January 30, 2007. Changes in ENDF/B-VII Format and procedures, as well as the evaluations themselves, make it impossible for versions of the ENDF/B pre-processing codes earlier than PREPRO 2007 (2007 Version) to accurately process current ENDF/B-VII evaluations. The present code can handle all existing ENDF/B-VI evaluations through release 8, which will be the last release of ENDF/B-VI. Modifications from previous versions: Evalplot Vers. 2007-1 (Jan. 2007): - checked against all ENDF/B-VII; - increased page size from 600,000 to 2,400,000; - increased the number of energies vs. legendre coefficients from 20,000 to 80,000 (must be 1/30 page size); - added (n,remainder) to first plot. 2 - Method of solution: In the case of processing neutron and photon cross sections (MF=3 or 23) and parameters (MF=1 or 27), all data in a file (MF) is read, grouped together by type, and plotted. All reactions of a data type appear on the same plot. The data types for MF=1 and 3 (neutrons) are: (1) total, elastic, capture, fission and total inelastic; (2) (n,2n), (n,3n) and (n,n' charged particle); (3) (n,charged particle); (4) particle production (proton, deuteron, etc.) and damage; (5) total, first, second, etc. chance fission; (6) total inelastic, inelastic discrete levels and continuum; (7) (n,p) total and levels (only if levels are given); (8) (n,d) total and levels (only if levels are given); (9) (n,t) total and levels (only if levels are given); (10) (n, 3 He) total and levels (only if levels are given); (11) (n, 4 He) total and levels (only if levels are given); (12) parameters mu-bar, xi and gamma; (13) nu-bar - total, prompt an delayed. The data types for MF=23 and 27 (photons) are: (14) total, coherent

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-09-30

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

  18. Determination of protein content in grains by radioactive thermal neutron capture prompt gamma rays analysis

    International Nuclear Information System (INIS)

    Carbonari, A.W.

    1983-01-01

    The radioactive thermal neutron capture prompt gamma rays technique can be used to determinate the nitrogen content in grains without chemical destruction, with good precision and relative rapidity. This determination is based on the detection of prompt gamma rays emitted by the 14 N(n,γ) 15 N reaction product. The samples has been irradiated the tanGencial tube of the IEA-R1 research reator and a pair spectrometer has been used for the detection of the prompt gamma rays. The nitrogen content is determinated in several samples of soybean, commonbean, peas and rice, and the results is compared with typical nitrogen content for each grain. (Autor) [pt

  19. ENDF-201, ENDF/B-VI summary documentation supplement 1, ENDF/HE-VI summary documentation

    International Nuclear Information System (INIS)

    McLane, V.

    1996-12-01

    The National Nuclear Data Center (NNDC) provides coordination for and serves as the secretariat to the Cross Section Evaluation Working Group (CSWEG). CSEWG is responsible for the oversight of the ENDF/B Evaluated Nuclear Data File. All data are checked and reviewed by CSEWG, and the file is maintained at the NNDC. For a description of the ENDF/B-VI file, see the ENDF-102 Data Formats and Procedures for the Evaluated Nuclear Data File ENDF-6. The purpose of this addendum to the ENDF/B-VI Summary Documentation is to provide documentation of Releases 1, 2, 3, and 4 for the ENDF/B-VI and ENDF/HE-VI evaluated nuclear data libraries. These releases contain many new and revised evaluations for the neutron, photo-atomic interaction, radioactive decay data, spontaneous fission product yield, neutron-induced fission product yield, thermal neutron scattering, proton, deuteron, and triton sublibraries. The summaries have been extracted mainly from the ENDF/B-VI File 1 comments (MT = 451), which have been checked, edited, and may also include supplementary information. Some summaries have been provided by the evaluators in electronic format, while others are extracted from reports on the evaluations. All references have been checked and corrected, or updated where appropriate. A list of the laboratories which have contributed evaluations used in ENDF/B-VI is given

  20. ENDF-201, ENDF/B-VI summary documentation supplement 1, ENDF/HE-VI summary documentation

    Energy Technology Data Exchange (ETDEWEB)

    McLane, V.

    1996-12-01

    The National Nuclear Data Center (NNDC) provides coordination for and serves as the secretariat to the Cross Section Evaluation Working Group (CSWEG). CSEWG is responsible for the oversight of the ENDF/B Evaluated Nuclear Data File. All data are checked and reviewed by CSEWG, and the file is maintained at the NNDC. For a description of the ENDF/B-VI file, see the ENDF-102 Data Formats and Procedures for the Evaluated Nuclear Data File ENDF-6. The purpose of this addendum to the ENDF/B-VI Summary Documentation is to provide documentation of Releases 1, 2, 3, and 4 for the ENDF/B-VI and ENDF/HE-VI evaluated nuclear data libraries. These releases contain many new and revised evaluations for the neutron, photo-atomic interaction, radioactive decay data, spontaneous fission product yield, neutron-induced fission product yield, thermal neutron scattering, proton, deuteron, and triton sublibraries. The summaries have been extracted mainly from the ENDF/B-VI File 1 comments (MT = 451), which have been checked, edited, and may also include supplementary information. Some summaries have been provided by the evaluators in electronic format, while others are extracted from reports on the evaluations. All references have been checked and corrected, or updated where appropriate. A list of the laboratories which have contributed evaluations used in ENDF/B-VI is given.

  1. Use of gamma ray strength functions for predicting the neutron capture cross section of 88Y

    International Nuclear Information System (INIS)

    Gardner, D.G.; Gardner, M.A.

    1977-01-01

    The present study indicates that the estimation of the gamma-ray strength function is the approach least subject to error when unmeasured capture cross sections are to be computed. An estimate is given for the 88 γ(n,γ) cross section

  2. Analysis of selected critical experiments using ENDF/B-IV and ENDF/B-V data

    International Nuclear Information System (INIS)

    Crump, M.W.; Durston, C.; Jonsson, A.; Singh, U.N.

    1983-01-01

    Selected critical experiments were analyzed using ENDF/B-V data and results compared with measured parameters and with values obtained using ENDF/B-IV. The TRX-1 and -2 U-metal criticals were reanalyzed using ENDF/B-V with consistent multilevel processing of U-238 resonance data and increased spatial detail in the resonance slowing down calculations. The improved resonance treatment was applied in TRX cell calculations performed with the DIT code, and resulted in reduced predictions of U-238 capture by more than two percent relative to previous calculations. The results of the TRX analyses using ENDF/B-V indicate calculated rho 28 values 2 to 3% higher than measurements, and are found in overall agreement with results reported by other laboratories. Full core calculations for the TRX criticals were performed with the ANISN code using cross sections obtained from DIT core-reflector lattice calculations. An evaluation of core versus cell calculations for these criticals indicates differences corresponding to about one half percent in predicted reactivity

  3. ENDF-201: ENDF/B-VI summary documentation

    International Nuclear Information System (INIS)

    Rose, P.F.

    1991-10-01

    Responsibility for oversight of the ENDF/B Evaluated Nuclear Data file lies with the Cross Section Evaluation Working Group (CSEWG), which is comprised of representatives from various governmental and industrial laboratories in the United States. Individual evaluations are provided by scientists at several US laboratories, including significant contributions by scientists from all over the world. In addition, ENDF/B-VI includes for the first time complete evaluations for three materials that were provided from laboratories outside the US. All data are checked and reviewed by CSEWG, and the data file is maintained and issued by the National Nuclear Data Center at Brookhaven National Laboratory. The previous version of the library, ENDF/B-V, was issued in 1979, and two revisions to the data file were provided in subsequent years, the latest occurring in 1981. A total of 75 new or extensively modified neutron sublibrary evaluations are included in ENDF/B-VI, and are summarized in this document. One incident proton sublibrary is described for Fe 56 . The remaining evaluations in ENDF/B-VI have been carried over from earlier versions of ENDF, and have been updated to reflect the new formats. The release of ENDF/B-VI was carried out between January and June of 1990, with groups of materials being released on ''tapes.'' Table 1 is an index to the evaluation summaries, and includes the material identification or MAT number, the responsible laboratory, and the ''tape'' number. These evaluations have been released without restrictions on their distribution or use

  4. Neutron capture prompt gamma-ray activation analysis at the NIST cold neutron research facility

    Energy Technology Data Exchange (ETDEWEB)

    Lindstrom, R M; Zeisler, R; Vincent, D H; Greenberg, R R; Stone, C A; Mackey, E A [National Inst. of Standards and Technology, Gaithersburg, MD (United States); Anderson, D L [Food and Drug Administration, Washington, DC (United States); Clark, D D [Cornell Univ., Ithaca, NY (United States)

    1993-01-01

    An instrument for neutron capture prompt gamma-ray activation analysis (PGAA) has been constructed as part of the Cold Neutron Research Facility at the 20 MW National Institute of Standards and Technology Research Reactor. The neutron fluence rate (thermal equivalent) is 1.5*10[sup 8] n*cm[sup -2]*s[sup -] [sup 1], with negligible fast neutrons and gamma-rays. With compact geometry and hydrogen-free construction, the sensitivity is sevenfold better than an existing thermal instrument. Hydrogen background is thirtyfold lower. (author) 17 refs.; 2 figs.

  5. The 1989 ENDF pre-processing codes

    International Nuclear Information System (INIS)

    Cullen, D.E.; McLaughlin, P.K.

    1989-12-01

    This document summarizes the 1989 version of the ENDF pre-processing codes which are required for processing evaluated nuclear data coded in the format ENDF-4, ENDF-5, or ENDF-6. The codes are available from the IAEA Nuclear Data Section, free of charge upon request. (author)

  6. Detection efficiency for radionuclides decaying by electron capture and gamma-Ray

    International Nuclear Information System (INIS)

    Grau, A.; Fernandez, A.

    1985-01-01

    In this paper, the electron capture partial counting efficiency vs the figure of merit for electron-capture and gamma-ray emitters has been computed. The radionuclides tabulated are 48 c r, 54 M n, 57 C o 56 N i, 72 S e, 73 A s, 85 S r, 88 Z r, 92 N b, 103 P d, 111 l n, 119 S b, 125 I , 139 C e and 152 D y. It has been assumed that the liquid is a toluene based scintillator solution in standard glass vials containing 15 cm 3 . (Author) 17 refs

  7. Photoneutron cross sections measurements in 9Be, 13C e 17O with thermal neutron capture gamma-rays

    International Nuclear Information System (INIS)

    Semmler, Renato

    2006-01-01

    Photoneutron cross sections measurements of 9 Be, 13 C and 17 O have been obtained in the energy interval between 1,6 and 10,8 MeV, using neutron capture gamma-rays with high resolution in energy (3 a 21 eV), produced by 21 target materials, placed inside a tangential beam port, near the core of the IPEN/CNEN-SP IEA-R1 (5 MW) research reactor. The samples have been irradiated inside a 4π geometry neutron detector system 'Long Counter', 520,5 cm away from the capture target. The capture gamma-ray flux was determined by means of the analysis of the gamma spectrum obtained by using a Ge(Li) solid-state detector (EG and G ORTEC, 25 cm 3 , 5%), previously calibrated with capture gamma-rays from a standard target of Nitrogen (Melamine). The neutron photoproduction cross section has been measured for each target capture gamma-ray spectrum (compound cross section). A inversion matrix methodology to solve inversion problems for unfolding the set of experimental compound cross sections, was used in order to obtain the cross sections at specific excitation energy values (principal gamma line energies of the capture targets). The cross sections obtained at the energy values of the principal gamma lines were compared with experimental data reported by other authors, with have employed different gamma-ray sources. A good agreement was observed among the experimental data in this work with reported in the literature. (author)

  8. ENDF/X: an Extended ENDF Format (Evolution, not Revolution)

    International Nuclear Information System (INIS)

    Cullen, Dermott E.

    2012-01-01

    Recently there has been yet another round of complaints about the ENDF format not being modern and general enough to handle today's nuclear data. This has led to suggestions to abandon the current ENDF and move on to a new format. The complaints I hear I fear are based upon not understanding the primary purpose of ENDF and a lack of experience in using the ENDF format and not being flexible to enough to deal with the current format. Personally I do not think that any changes to the ENDF format are Necessary. But here I address the complaints that I have recently heard about the limitations of the ENDF format, and I suggest minor changes that will completely handle these complaints. In turn I would ask those who are complaining and feel that extensions are needed Please give us some examples where these extensions are Needed. Personally I am not aware of any such data, but I am keeping an open mind and I would love to see examples that really Require extensions. I propose a few fairly simple extensions to the current ENDF/B format; what I call ENDF/X. Compared to other suggested revolutionary changes, my evolutionary approach has the advantage that it maintains compatibility with the existing ENDF/B format that we have used successfully for almost fifty years, and still allows the format to be extended for use with other types of data. In addition to my suggested changes to the ENDF/B format I also include a brief history of ENDF/B, in the hope that the experience we have gained over the last almost 50 years will be of help to the present generation of nuclear data developers and users. Please let's not make the mistake of learning nothing from history. Lastly I finish by identifying what I see as the weak point in the current infrastructure that we use to handle evaluated nuclear data; it me it is not the format of the data. (author)

  9. Analysis of MCNP simulated gamma spectra of CdTe detectors for boron neutron capture therapy.

    Science.gov (United States)

    Winkler, Alexander; Koivunoro, Hanna; Savolainen, Sauli

    2017-06-01

    The next step in the boron neutron capture therapy (BNCT) is the real time imaging of the boron concentration in healthy and tumor tissue. Monte Carlo simulations are employed to predict the detector response required to realize single-photon emission computed tomography in BNCT, but have failed to correctly resemble measured data for cadmium telluride detectors. In this study we have tested the gamma production cross-section data tables of commonly used libraries in the Monte Carlo code MCNP in comparison to measurements. The cross section data table TENDL-2008-ACE is reproducing measured data best, whilst the commonly used ENDL92 and other studied libraries do not include correct tables for the gamma production from the cadmium neutron capture reaction that is occurring inside the detector. Furthermore, we have discussed the size of the annihilation peaks of spectra obtained by cadmium telluride and germanium detectors. Copyright © 2017 Elsevier Ltd. All rights reserved.

  10. A fast Monte Carlo program for pulsed-neutron capture-gamma tools

    International Nuclear Information System (INIS)

    Hovgaard, J.

    1992-02-01

    A fast model for the pulsed-neutron capture-gamma tool has been developed. It is believed that the program produce valid results even though some approximation have been introduced. A correct γ photon transport simulation, which is under preparation, has for instance not yet been included. Simulations performed so far has shown that the model, with respect to computing time and accuracy, fully lives up to expectations with respect to computing time and accuracy. (au)

  11. ENDF/X: An Extended ENDF Format (Evolution, not Revolution)

    International Nuclear Information System (INIS)

    Cullen, Dermott E.

    2014-04-01

    Recently there has been yet another round of complaints about the ENDF format not being modern and general enough to handle today’s nuclear data. This has led to suggestions to abandon the current ENDF and move on to a new format. The complaints I hear I fear are based upon not understanding the primary purpose of ENDF and a lack of experience in using the ENDF format and not being flexible enough to deal with the current format. Personally I don’t think that any changes to the ENDF format are NECESSARY. But here I address the complaints that I have recently heard about the limitations of the ENDF format, and I suggest minor changes that will completely handle these complaints. In turn I would ask those who are complaining and feel that extensions are needed PLEASE give us some examples where these extensions are NEEDED. Personally I am not aware of any such data, but I am keeping an open mind and I would love to see examples that really REQUIRE extensions.

  12. Study of gamma ray multiplicity spectra for radiative capture of neutrons in 113,115In

    International Nuclear Information System (INIS)

    Georgiev, G.P.; Fajkov-Stanchik, Kh.; Grigor'ev, Yu.V.; Muradyan, G.V.; Yaneva, N.B.

    1997-08-01

    Neutron radiative capture measurements were performed for the enriched isotopes 113 In and 115 In on the neutron spectrometer at the Neutron Physics Laboratory of the Joint Institute for Nuclear Research employing the gamma ray multiplicity technique and using a ''Romashka'' multi-sectional 4p detector on the 500 m time base of the IBR-30 booster. The gamma multiplicity spectra of resolved resonances were obtained for the 20-500 eV energy range. The mean gamma ray multiplicity was determined for each resonance. The dependence of the ratio S of the low-energy coincidence multiplicity spectrum to the high-energy coincidence multiplicity spectrum on resonance energy exhibits a non-statistical structure. This structure was found to correlate with the local neutron strength function. (author). 10 refs, 6 figs, 2 tabs

  13. ENDF/B summary documentation

    International Nuclear Information System (INIS)

    Kinsey, R.

    1979-07-01

    This publication provides a localized source of descriptions for the evaluations contained in the ENDF/B Library. The summary documentation presented is intended to be a more detailed description than the (File 1) comments contained in the computer readable data files, but not so detailed as the formal reports describing each ENDF/B evaluation. The summary documentations were written by the CSEWB (Cross Section Evaluation Working Group) evaluators and compiled by NNDC (National Nuclear Data Center). This edition includes documentation for materials found on ENDF/B Version V tapes 501 to 516 (General Purpose File) excluding tape 504. ENDF/B-V also includes tapes containing partial evaluations for the Special Purpose Actinide (521, 522), Dosimetry (531), Activation (532), Gas Production (533), and Fission Product (541-546) files. The materials found on these tapes are documented elsewhere. Some of the evaluation descriptions in this report contain cross sections or energy level information

  14. ENDF/B summary documentation

    Energy Technology Data Exchange (ETDEWEB)

    Kinsey, R. (comp.)

    1979-07-01

    This publication provides a localized source of descriptions for the evaluations contained in the ENDF/B Library. The summary documentation presented is intended to be a more detailed description than the (File 1) comments contained in the computer readable data files, but not so detailed as the formal reports describing each ENDF/B evaluation. The summary documentations were written by the CSEWB (Cross Section Evaluation Working Group) evaluators and compiled by NNDC (National Nuclear Data Center). This edition includes documentation for materials found on ENDF/B Version V tapes 501 to 516 (General Purpose File) excluding tape 504. ENDF/B-V also includes tapes containing partial evaluations for the Special Purpose Actinide (521, 522), Dosimetry (531), Activation (532), Gas Production (533), and Fission Product (541-546) files. The materials found on these tapes are documented elsewhere. Some of the evaluation descriptions in this report contain cross sections or energy level information. (RWR)

  15. ENDF/B summary documentation

    International Nuclear Information System (INIS)

    Garber, D.

    1975-10-01

    Descriptions of the evaluations contained in the ENDF/B library are given. The summary documentation presented is intended to be a more detailed description than the (File 1) comments contained in the computer-readable data files, but not so detailed as the formal reports describing each ENDF/B evaluation. The documentations were written by the CSEWG evaluators and compiled by NNCSC. Selected materials which comprise this volume include from 1 H to 244 Cm

  16. Self-absorption of neutron capture gamma-rays in gold samples

    International Nuclear Information System (INIS)

    Wisshak, K.; Walter, G.; Kaeppeler, F.

    1983-06-01

    The self absorption of neutron capture gamma rays in gold samples has been determined experimentally for two standard setups used in measurements of neutron capture cross sections. One makes use of an artificially collimated neutron beam and two C 6 D 6 detectors, the other of kinematically collimated neutrons and three Moxon-Rae detectors. Correction factors for an actual measurement of a neutron capture cross section using a gold standard of 1 mm thickness up to 12% were found for the first setup while they are only 4% for the second setup. The present data allow to determine the correction in an actual measurement with an accuracy of 0.5-1%. (orig.) [de

  17. Monte Carlo simulation of the scattered component of neutron capture prompt gamma-ray analyzer responses

    International Nuclear Information System (INIS)

    Jin, Y.; Verghese, K.; Gardner, R.P.

    1986-01-01

    This paper describes a major part of our efforts to simulate the entire spectral response of the neutron capture prompt gamma-ray analyzer for bulk media (or conveyor belt) samples by the Monte Carlo method. This would allow one to use such a model to augment or, in most cases, essentially replace experiments in the calibration and optimum design of these analyzers. In previous work, we simulated the unscattered gamma-ray intensities, but would like to simulate the entire spectral response as we did with the energy-dispersive x-ray fluorescence analyzers. To accomplish this, one must account for the scattered gamma rays as well as the unscattered and one must have available the detector response function to translate the incident gamma-ray spectrum calculated by the Monte Carlo simulation into the detected pulse-height spectrum. We recently completed our work on the germanium detector response function, and the present paper describes our efforts to simulate the entire spectral response by using it with Monte Carlo predicted unscattered and scattered gamma rays

  18. ENDF-201: ENDF/B-VI summary documentation

    Energy Technology Data Exchange (ETDEWEB)

    Rose, P.F. (comp.)

    1991-10-01

    Responsibility for oversight of the ENDF/B Evaluated Nuclear Data file lies with the Cross Section Evaluation Working Group (CSEWG), which is comprised of representatives from various governmental and industrial laboratories in the United States. Individual evaluations are provided by scientists at several US laboratories, including significant contributions by scientists from all over the world. In addition, ENDF/B-VI includes for the first time complete evaluations for three materials that were provided from laboratories outside the US. All data are checked and reviewed by CSEWG, and the data file is maintained and issued by the National Nuclear Data Center at Brookhaven National Laboratory. The previous version of the library, ENDF/B-V, was issued in 1979, and two revisions to the data file were provided in subsequent years, the latest occurring in 1981. A total of 75 new or extensively modified neutron sublibrary evaluations are included in ENDF/B-VI, and are summarized in this document. One incident proton sublibrary is described for Fe{sup 56}. The remaining evaluations in ENDF/B-VI have been carried over from earlier versions of ENDF, and have been updated to reflect the new formats. The release of ENDF/B-VI was carried out between January and June of 1990, with groups of materials being released on tapes.'' Table 1 is an index to the evaluation summaries, and includes the material identification or MAT number, the responsible laboratory, and the tape'' number. These evaluations have been released without restrictions on their distribution or use.

  19. ENDF-201: ENDF/B-VI summary documentation

    Energy Technology Data Exchange (ETDEWEB)

    Rose, P.F. [comp.

    1991-10-01

    Responsibility for oversight of the ENDF/B Evaluated Nuclear Data file lies with the Cross Section Evaluation Working Group (CSEWG), which is comprised of representatives from various governmental and industrial laboratories in the United States. Individual evaluations are provided by scientists at several US laboratories, including significant contributions by scientists from all over the world. In addition, ENDF/B-VI includes for the first time complete evaluations for three materials that were provided from laboratories outside the US. All data are checked and reviewed by CSEWG, and the data file is maintained and issued by the National Nuclear Data Center at Brookhaven National Laboratory. The previous version of the library, ENDF/B-V, was issued in 1979, and two revisions to the data file were provided in subsequent years, the latest occurring in 1981. A total of 75 new or extensively modified neutron sublibrary evaluations are included in ENDF/B-VI, and are summarized in this document. One incident proton sublibrary is described for Fe{sup 56}. The remaining evaluations in ENDF/B-VI have been carried over from earlier versions of ENDF, and have been updated to reflect the new formats. The release of ENDF/B-VI was carried out between January and June of 1990, with groups of materials being released on ``tapes.`` Table 1 is an index to the evaluation summaries, and includes the material identification or MAT number, the responsible laboratory, and the ``tape`` number. These evaluations have been released without restrictions on their distribution or use.

  20. Within the framework of the new fuel cycle {sup 232}Th/{sup 233}U, determination of the {sup 233}Pa(n.{gamma}) radiative capture cross section for neutron energies ranging between 0 and 1 MeV; Dans le cadre du nouveau cycle de combustible {sup 232}Th/{sup 233}U, determination de la section efficace de capture radiative {sup 233}Pa(n,{gamma}) pour des energies de neutrons comprises entre 0 et 1 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Boyer, S

    2004-10-15

    The Thorium cycle Th{sup 232}/U{sup 233} may face brilliant perspectives through advanced concepts like molten salt reactors or accelerator driven systems but it lacks accurate nuclear data concerning some nuclei. Pa{sup 233} is one of these nuclei, its high activity makes the direct measurement of its radiative neutron capture cross-section almost impossible. This difficulty has been evaded by considering the transfer reaction Th{sup 232}(He{sup 3},p)Pa{sup 234}* in which the Pa{sup 234} nucleus is produced in various excited states according to the amount of energy available in the reaction. The first chapter deals with the thorium cycle and its assets to contribute to the quenching of the fast growing world energy demand. The second chapter gives a detailed description of the experimental setting. A scintillation detector based on deuterated benzene (C{sub 6}D{sub 6}) has been used to counter gamma ray cascades. The third chapter is dedicated to data analysis. In the last chapter we compare our experimental results with ENDF and JENDL data and with computed values from 2 statistical models in the 0-1 MeV neutron energy range. Our results disagree clearly with evaluated data: our values are always above ENDF and JENDL data but tend to near computed values. We have also perform the measurement of the radiative neutron cross-section of Pa{sup 231} for a 110 keV neutron: {sigma}(n,{gamma}) 2.00 {+-} 0.14 barn. (A.C.)

  1. Measurement of salinity of fluids in earth formations by comparison of inelastic and capture gamma ray spectra

    International Nuclear Information System (INIS)

    1979-01-01

    A method of borehole logging by detecting and counting gamma rays from inelastic scattering of fast neutrons by carbon, oxygen, silicon and calcium, gamma rays from capture of thermal neutrons by calcium, chlorine and silicon and comparing the former with the latter thereby deriving an estimate of the salinity of the fluids in the borehole, is given (UK)

  2. Multiparameter data acquisition and analysis system for capture gamma-ray studies

    International Nuclear Information System (INIS)

    Hejja, I.; Belgya, T.; Molnar, G.L.; Szepesvary, A.

    1997-01-01

    A PC-based multiparameter data acquisition system has been built for the Budapest neutron capture gamma-ray spectrometer. The hardware consists of a homemade multiplexer accommodating up to ten ADC inputs, a 64 kword histogram memory board and a National Instruments 32-bit DIO card, used for data acquisition and control, as well as a timer/scaler TIO card of the same company. The multiplexer inputs can be flexibly configured by means of programmable XILINX logic chips. The system is driven by a Pentium PC connected to the local Ethernet. (author)

  3. Apparatus for parity-violation study via capture gamma-ray measurements

    CERN Document Server

    Seestrom, S J; Bowman, J D; Crawford, B C; Haseyama, T; Masaike, A; Matsuda, A; Penttilae, S I; Roberson, R N; Sharapov, E I; Stephenson, S L

    1999-01-01

    The Time Reversal and Parity at Low Energy (TRIPLE) Collaboration uses a short-pulsed longitudinally polarized epithermal neutron beam at the Los Alamos Neutron Science Center to study spatial parity violation (PV) in the compound nucleus. The typical PV experiment measures the longitudinal cross-section asymmetry by the neutron transmission method through thick samples. Neutron capture gamma-ray measurement provides an alternative method for the study of PV, which enables the use of smaller amounts of isotopically pure target material. In 1995 TRIPLE commissioned a new neutron-capture detector consisting of 24 pure CsI scintillators arranged in a cylindrical geometry around the neutron beam. The characteristics and the performance of the detector and spin transport are described.

  4. Cross Sections for High-Energy Gamma Transitions from MeV Neutron Capture in {sup 206}Pb

    Energy Technology Data Exchange (ETDEWEB)

    Bergqvist, I; Lundberg, B; Nilsson, L

    1970-03-15

    Gamma-ray spectra from neutron capture in Pb (radiogenic lead) in the energy range 1.5 to 8.5 MeV were recorded using time-of-flight techniques. The spectrometer was a Nal (Tl) crystal, 20.8 cm long and 22.6 cm in diameter. The spectra are dominated by gamma transitions to levels with large single-particle strength, in agreement with predictions of semi-direct capture theories. The theories predict enhancements of the direct capture cross section by a factor of 10 - 15 in the region of the giant dipole resonance. The observed enhancement is about 50.

  5. GPU-based prompt gamma ray imaging from boron neutron capture therapy

    International Nuclear Information System (INIS)

    Yoon, Do-Kun; Jung, Joo-Young; Suk Suh, Tae; Jo Hong, Key; Sil Lee, Keum

    2015-01-01

    Purpose: The purpose of this research is to perform the fast reconstruction of a prompt gamma ray image using a graphics processing unit (GPU) computation from boron neutron capture therapy (BNCT) simulations. Methods: To evaluate the accuracy of the reconstructed image, a phantom including four boron uptake regions (BURs) was used in the simulation. After the Monte Carlo simulation of the BNCT, the modified ordered subset expectation maximization reconstruction algorithm using the GPU computation was used to reconstruct the images with fewer projections. The computation times for image reconstruction were compared between the GPU and the central processing unit (CPU). Also, the accuracy of the reconstructed image was evaluated by a receiver operating characteristic (ROC) curve analysis. Results: The image reconstruction time using the GPU was 196 times faster than the conventional reconstruction time using the CPU. For the four BURs, the area under curve values from the ROC curve were 0.6726 (A-region), 0.6890 (B-region), 0.7384 (C-region), and 0.8009 (D-region). Conclusions: The tomographic image using the prompt gamma ray event from the BNCT simulation was acquired using the GPU computation in order to perform a fast reconstruction during treatment. The authors verified the feasibility of the prompt gamma ray image reconstruction using the GPU computation for BNCT simulations

  6. Database of prompt gamma rays from slow neutron capture for elemental analysis

    International Nuclear Information System (INIS)

    Firestone, R.B.; Choi, H.D.; Lindstrom, R.M.; Molnar, G.L.; Mughabghab, S.F.; Paviotti-Corcuera, R.; Revay, Zs; Trkov, A.; Zhou, C.M.; Zerkin, V.

    2004-01-01

    The increasing importance of Prompt Gamma-ray Activation Analysis (PGAA) in a broad range of applications is evident, and has been emphasized at many meetings related to this topic (e.g., Technical Consultants' Meeting, Use of neutron beams for low- and medium-flux research reactors: radiography and materials characterizations, IAEA Vienna, 4-7 May 1993, IAEA-TECDOC-837, 1993). Furthermore, an Advisory Group Meeting (AGM) for the Coordination of the Nuclear Structure and Decay Data Evaluators Network has stated that there is a need for a complete and consistent library of cold- and thermal neutron capture gamma ray and cross-section data (AGM held at Budapest,14-18 October 1996, INDC(NDS)-363); this AGM also recommended the organization of an IAEA CRP on the subject. The International Nuclear Data Committee (INDC) is the primary advisory body to the IAEA Nuclear Data Section on their nuclear data programs. At a biennial meeting in 1997, the INDC strongly recommended that the Nuclear Data Section support new measurements and update the database on Neutron-induced Prompt Gamma-ray Activation Analysis (21st INDC meeting, INDC/P(97)-20). As a consequence of the various recommendations, a CRP on ''Development of a Database for Prompt Gamma-ray Neutron Activation Analysis (PGAA)'' was initiated in 1999. Prior to this project, several consultants had defined the scope, objectives and tasks, as approved subsequently by the IAEA. Each CRP participant assumed responsibility for the execution of specific tasks. The results of their and other research work were discussed and approved by the participants in research co-ordination meetings (see Summary reports: INDC(NDS)-411, 2000; INDC(NDS)-424, 2001; and INDC(NDS)-443, 200). PGAA is a non-destructive radioanalytical method, capable of rapid or simultaneous ''in-situ'' multi-element analyses across the entire Periodic Table, from hydrogen to uranium. However, inaccurate and incomplete data were a significant hindrance in the

  7. ENDF utility codes version 6.4 for ENDF-5 and ENDF-6

    International Nuclear Information System (INIS)

    McLaughlin, P.K.

    1988-09-01

    Description and operating instructions are given for a package of utility codes operating on evaluated nuclear data files in the formats ENDF-5 and ENDF-6. Included are the data checking codes CHECKER, FIZCON, PSYCHE; the code INTER for retrieving thermal cross-sections and some other data; graphical plotting codes PLOTEF, GRABLIB, VERSAT; and the file maintenance and retrieval codes LISTEF, SETMDC, GETMAT, STANEF. This program package can be obtained on magnetic tape or floppy diskette, free of charge, from the IAEA Nuclear Data Section. (author)

  8. ENDF/B-V actinides

    International Nuclear Information System (INIS)

    Kocherov, N.; Lemmel, H.D.

    1981-01-01

    This document summarizes the contents of the actinides part of the ENDF/B-V nuclear data library released by the US National Nuclear Data Center. This library or selective retrievals of it, are available from the IAEA Nuclear Data Section. (author)

  9. Monitoring taconite process streams with thermal neutron capture-gamma ray analysis. Report of investigations/1980

    International Nuclear Information System (INIS)

    Woodbury, F.B.W.

    1980-12-01

    The Bureau of Mines is evaluating alternative technologies to treat oxidized taconites. Since process control is an essential element in the application of these process technologies, research was performed on a prototype monitoring system utilizing a californium-252 (252-Cf) neutron source and a thermal neutron capture-gamma ray spectra analysis method to measure the amount of iron and percent solids in process slurries. The prototype system was used to monitor the concentrate and tailing streams in a 900-lb/hr flotation pilot plant during continuous around-the-clock tests. The iron content of the process slurries was determined by measuring the total peak areas under the capture spectrum peaks at 7.626-7.632 MeV, the associated escape peaks at 7.136-7.122 and 6.626-6.612 MeV, and the iron doublets at 4.900 and 4.998 MeV. A potential method for determining the percent solids in process slurries using the 2.22 MeV hydrogen capture peak is discussed

  10. Activities of the Shielding Subcommittee of the ENDF/B Cross Section Evaluation Working Group

    International Nuclear Information System (INIS)

    Roussin, R.W.

    1977-01-01

    The Shielding Subcommittee of the Cross Section Evaluation Working Group (CSEWG) was established in 1967 to help ensure that the content of the ENDF/B cross section library was adequate for treating shielding problems. Early work of the subcommittee concentrated on devising formats for gamma-ray interaction and production data, as well as providing programs for testing the clerical and physics consistency of the files. The Radiation Shielding Information Center (RSIC) collaborated directly with evaluators on behalf of the National Neutron Cross Section Center (NNCSC) to begin testing and adding data sets to be fed into the official ENDF/B libraries. These efforts, which were sponsored by AEC-DRDT (now ERDA-DRDD), were augmented greatly through the Defense Nuclear Agency program of establishing a working cross section library in ENDF format. The effort concentrated on evaluation and testing of materials of interest to DNA programs and providing these for inclusion in the ENDF/B library. Shielding data testing efforts, as a part of the CSEWG Data Testing Program, are now also an integral part of the Shielding Subcommittee effort. Procedures for writing and approving the shielding benchmarks were devised by Shielding Subcommittee members. Data testing benchmark experiments have been documented and analyzed, and the most recent results for ENDF/B-IV are as reported as part of ENDF-230, ''Benchmark Testing of ENDF/B-IV.''

  11. PSYCHE, ENDF/B Data Consistency Check in ENDF Format

    International Nuclear Information System (INIS)

    Dunford, Charles L.

    2007-01-01

    1 - Description of program or function: PSYCHE is a program for checking the physics content of an evaluated data file. It can recognise the difference between ENDF-5 or ENDF-6 formats and performs its tests accordingly. Version 7.01 (April 2005): set success flag after return from begin, added potential scattering test formerly in Fizcon. Version 7.02 (May 2005): Fixed bug in calculation of L=2 penetrability. 2 - Method of solution: PSYCHE checks for energy conservation for emitted neutrons and photons, checks Wick's limit for elastic scattering, analyses resonance parameter statistics, calculates thermal cross sections and resonance integrals, examines continuity across resonance region boundaries and checks 'Q' values against mass tables

  12. Database of prompt gamma rays from slow neutron capture forelemental analysis

    Energy Technology Data Exchange (ETDEWEB)

    Firestone, R.B.; Choi, H.D.; Lindstrom, R.M.; Molnar, G.L.; Mughabghab, S.F.; Paviotti-Corcuera, R.; Revay, Zs; Trkov, A.; Zhou,C.M.; Zerkin, V.

    2004-12-31

    The increasing importance of Prompt Gamma-ray ActivationAnalysis (PGAA) in a broad range of applications is evident, and has beenemphasized at many meetings related to this topic (e.g., TechnicalConsultants' Meeting, Use of neutron beams for low- andmedium-fluxresearch reactors: radiography and materialscharacterizations, IAEA Vienna, 4-7 May 1993, IAEA-TECDOC-837, 1993).Furthermore, an Advisory Group Meeting (AGM) for the Coordination of theNuclear Structure and Decay Data Evaluators Network has stated that thereis a need for a complete and consistent library of cold- and thermalneutron capture gammaray and cross-section data (AGM held at Budapest,14-18 October 1996, INDC(NDS)-363); this AGM also recommended theorganization of an IAEA CRP on the subject. The International NuclearData Committee (INDC) is the primary advisory body to the IAEA NuclearData Section on their nuclear data programmes. At a biennial meeting in1997, the INDC strongly recommended that the Nuclear Data Section supportnew measurements andupdate the database on Neutron-induced PromptGamma-ray Activation Analysis (21st INDC meeting, INDC/P(97)-20). As aconsequence of the various recommendations, a CRP on "Development of aDatabase for Prompt Gamma-ray Neutron Activation Analysis (PGAA)" wasinitiated in 1999. Prior to this project, several consultants had definedthe scope, objectives and tasks, as approved subsequently by the IAEA.Each CRP participant assumed responsibility for the execution of specifictasks. The results of their and other research work were discussed andapproved by the participants in research co-ordination meetings (seeSummary reports: INDC(NDS)-411, 2000; INDC(NDS)-424, 2001; andINDC(NDS)-443, 200). PGAA is a non-destructive radioanalytical method,capable of rapid or simultaneous "in-situ" multi-element analyses acrossthe entire Periodic Table, from hydrogen to uranium. However, inaccurateand incomplete data were a significant hindrance in the qualitative andquantitative

  13. GAMSOURCE - WRS system module number 38474 for calculating gamma-ray sources produced by neutron capture

    International Nuclear Information System (INIS)

    Grimstone, M.J.

    1978-06-01

    The WRS Modular Programming System has been developed as a means by which programmes may be more efficiently constructed, maintained and modified. In this system a module is a self-contained unit typically composed of one or more Fortran routines, and a programme is constructed from a number of such modules. This report describes one WRS module, the function of which is to calculate the source strength of gamma-rays arising from neutron capture in a system represented in one-dimensional geometry. The information given in this manual is of use both to the programmer wishing to incorporate the module in a programme, and to the user of such a programme. (author)

  14. Measurement of keV-neutron capture cross sections and capture gamma-ray spectra of Er isotopes

    International Nuclear Information System (INIS)

    Harun-Ar-Rashid, A.K.M.; Igashira, Masayuki; Ohsaki, Toshiro

    2000-01-01

    Neutron capture cross sections and capture γ-ray spectra of 166,167, 168 Er were measured in the energy region of 10 to 550 keV. The measurements were performed with a pulsed 7 Li(p,n) 7 Be neutron source and a large anti-Compton NaI(Tl) γ-ray spectrometer. A pulse-height weighting technique and the standard capture cross sections of gold were used to derive the capture cross sections. The errors of the derived cross sections were about 5%. The present results were compared with other measurements and evaluations. The observed capture γ-ray pulse-height spectra were unfolded to obtain the corresponding γ-ray spectra. An anomalous shoulder was observed around 3 MeV in each of the capture γ-ray spectra. (author)

  15. Search for the radiative capture reaction d + d -> sup 4 He + gamma from the dd mu muonic molecule state

    CERN Document Server

    Bogdanova, L N; Eijk, C W E

    2002-01-01

    A search for the muon catalyzed fusion (MCF) reaction d + d -> sup 4 He + gamma in the dd mu muonic molecule was performed using the experimental MCF installation TRITON and NaI(Tl) detectors for gamma quanta. The high-pressure target filled with deuterium was exposed to the negative muon beam of the JINR phasotron to detect gamma quanta with energy 23.8 MeV. The first experimental estimation for the yield of the radiative deuteron capture from the dd mu state J = 1 was obtained at the level eta subgamma <= 2 x 10 sup - sup 5 per one fusion

  16. Partial neutron capture cross sections of actinides using cold neutron prompt gamma activation analysis

    International Nuclear Information System (INIS)

    Genreith, Christoph

    2015-01-01

    Nuclear waste needs to be characterized for its safe handling and storage. In particular long-lived actinides render the waste characterization challenging. The results described in this thesis demonstrate that Prompt Gamma Neutron Activation Analysis (PGAA) with cold neutrons is a reliable tool for the non-destructive analysis of actinides. Nuclear data required for an accurate identification and quantification of actinides was acquired. Therefore, a sample design suitable for accurate and precise measurements of prompt γ-ray energies and partial cross sections of long-lived actinides at existing PGAA facilities was presented. Using the developed sample design the fundamental prompt γ-ray data on 237 Np, 241 Am and 242 Pu were measured. The data were validated by repetitive analysis of different samples at two individual irradiation and counting facilities - the BRR in Budapest and the FRM II in Garching near Munich. Employing cold neutrons, resonance neutron capture by low energetic resonances was avoided during the experiments. This is an improvement over older neutron activation based works at thermal reactor neutron energies. 152 prompt γ-rays of 237 Np were identified, as well as 19 of 241 Am, and 127 prompt γ-rays of 242 Pu. In all cases, both high and lower energetic prompt γ-rays were identified. The most intense line of 237 Np was observed at an energy of E γ =182.82(10) keV associated with a partial capture cross section of σ γ =22.06(39) b. The most intense prompt γ-ray lines of 241 Am and of 242 Pu were observed at E γ =154.72(7) keV with σ γ =72.80(252) b and E γ =287.69(8) keV with σ γ =7.07(12) b, respectively. The measurements described in this thesis provide the first reported quantifications on partial radiative capture cross sections for 237 Np, 241 Am and 242 Pu measured simultaneously over the large energy range from 45 keV to 12 MeV. Detailed uncertainty assessments were performed and the validity of the given uncertainties was

  17. Experimental determination of nuclear reaction rates (n,γ) by the gamma-rays capture spectrometry technique

    International Nuclear Information System (INIS)

    Lucatero, M.A.

    1976-01-01

    The technique of the gamma-rays capture spectrometry was used in the experimental determination of nuclear reaction rates of the type (n,γ). This technique consists in the incidence of a thermal neutrons collimated beam upon a sample, detecting the capture spectrum of gamma rays emitted at a solid fixed angle. In the determination of the efficiency curve intrinsic to the detection electronic system the reactions 199 Hg(n,γ) 200 Hg, 56 Fe(n,γ) 57 Fe and 63 Cu(n,γ) 64 Cu were used with the energy of the gamma rays capture of 5.976, 7.635 and 7.915 Mev respectively, through the irradiation of standard samples of Hg(175.3g), Fe(110.4g) and Cu(108.5g) of cylindrical geometry the two former and parallelepiped the latter. The problem concerning the corrections due to the thermal neutrons flux depression, the gammas auto-attenuation, and the geometric factor due to the cylindrical and parallelepiped geometry are involved in the data process. The experimental determination of the reaction 35 Cl(n,γ) 36 Cl rate was made through the observation of the gamma caputre of 6.111 Mev when a sample of CaCl 2 of cylindrical geometry was irradiated. This rate can be favorably compared with the reaction rate determined theoretically. (author)

  18. Neutron Capture Gamma Ray Cross Sections for Ta, Ag, In and Au between 30 and 175 keV

    International Nuclear Information System (INIS)

    Hellstroem, J.; Beshai, S.

    1971-11-01

    A new detector has been used to determine neutron capture gamma ray cross sections for Ta, Ag, In and Au. The results are listed and discussed together with associated problems. The energy range from 30 keV to 175 keV is considered

  19. Radiative capture reaction {sup 7}Be(p,{gamma}){sup 8}B in the continuum shell model

    Energy Technology Data Exchange (ETDEWEB)

    Bennaceur, K; Ploszajczak, M [Grand Accelerateur National d` Ions Lourds (GANIL), Caen (France); Nowacki, F [Grand Accelerateur National d` Ions Lourds (GANIL), Caen (France); [Lab. de Physique Theorique Strasbourg, Strasbourg (France); Okolowicz, J [Grand Accelerateur National d` Ions Lourds (GANIL), Caen (France); [Inst. of Nuclear Physics, Krakow (Poland)

    1998-06-01

    We present here the first application of realistic shell model (SM) including coupling between many-particle (quasi-)bound states and the continuum of one-particle scattering states to the calculation of the total capture cross section and the astrophysical factor in the reaction {sup 7}Be(p,{gamma}){sup 8}B. (orig.)

  20. Neutron Capture Gamma Ray Cross Sections for Ta, Ag, In and Au between 30 and 175 keV

    Energy Technology Data Exchange (ETDEWEB)

    Hellstroem, J; Beshai, S

    1971-11-15

    A new detector has been used to determine neutron capture gamma ray cross sections for Ta, Ag, In and Au. The results are listed and discussed together with associated problems. The energy range from 30 keV to 175 keV is considered

  1. A Gamma Polarimeter for Neutron Polarization Measurement in a Liquid Deuterium Target for Parity Violation in Polarized Neutron Capture on Deuterium.

    Science.gov (United States)

    Komives, A; Sint, A K; Bowers, M; Snow, M

    2005-01-01

    A measurement of the parity-violating gamma asymmetry in n-D capture would yield information on N-N parity violation independent of the n-p system. Since cold neutrons will depolarize in a liquid deuterium target in which the scattering cross section is much larger than the absorption cross section, it will be necessary to quantify the loss of polarization before capture. One way to do this is to use the large circular polarization of the gamma from n-D capture and analyze the circular polarization of the gamma in a gamma polarimeter. We describe the design of this polarimeter.

  2. Comparison of the radiobiological effects of Boron neutron capture therapy (BNCT) and conventional Gamma Radiation

    International Nuclear Information System (INIS)

    Dagrosa, Maria A.; Carpano, Marina; Perona, Marina; Thomasz, Lisa; Juvenal, Guillermo J.; Pisarev, Mario; Pozzi, Emiliano; Thorp, Silvia

    2009-01-01

    BNCT is an experimental radiotherapeutic modality that uses the capacity of the isotope 10 B to capture thermal neutrons leading to the production of 4 He and 7 Li, particles with high linear energy transfer (LET). The aim was to evaluate and compare in vitro the mechanisms of response to the radiation arising of BNCT and conventional gamma therapy. We measured the survival cell fraction as a function of the total physical dose and analyzed the expression of p27/Kip1 and p53 by Western blotting in cells of colon cancer (ARO81-1). Exponentially growing cells were distributed into the following groups: 1) BPA (10 ppm 10 B) + neutrons; 2) BOPP (10 ppm 10 B) + neutrons; 3) neutrons alone; 4) gamma-rays. A control group without irradiation for each treatment was added. The cells were irradiated in the thermal neutron beam of the RA-3 (flux= 7.5 10 9 n/cm 2 sec) or with 60 Co (1Gy/min) during different times in order to obtain total physical dose between 1-5 Gy (±10 %). A decrease in the survival fraction as a function of the physical dose was observed for all the treatments. We also observed that neutrons and neutrons + BOPP did not differ significantly and that BPA was the more effective compound. Protein extracts of irradiated cells (3Gy) were isolated to 24 h and 48 h post radiation exposure. The irradiation with neutrons in presence of 10 BPA or 10 BOPP produced an increase of p53 at 24 h maintain until 48 h. On the contrary, in the groups irradiated with neutrons alone or gamma the peak was observed at 48 hr. The level of expression of p27/Kip1 showed a reduction of this protein in all the groups irradiated with neutrons (neutrons alone or neutrons plus boron compound), being more marked at 24 h. These preliminary results suggest different radiobiological response for high and low let radiation. Future studies will permit establish the role of cell cycle in the tumor radio sensibility to BNCT. (author)

  3. Nuclear models and data for gamma-ray production

    International Nuclear Information System (INIS)

    Young, P.G.

    1975-01-01

    The current Evaluated Nuclear Data File (ENDF/B, Version IV) contains information on prompt gamma-ray production from neutron-induced reactions for some 38 nuclides. In addition, there is a mass of fission product yield, capture, and radioactive decay data from which certain time-dependent gamma-ray results can be calculated. These data are needed in such applications as gamma-ray heating calculations for reactors, estimates of radiation levels near nuclear facilities and weapons, shielding design calculations, and materials damage estimates. The prompt results are comprised of production cross sections, multiplicities, angular distributions, and energy spectra for secondary gamma-rays from a variety of reactions up to an incident neutron energy of 20 MeV. These data are based in many instances on experimental measurements, but nuclear model calculations, generally of a statistical nature, are also frequently used to smooth data, to interpolate between measurements, and to calculate data in unmeasured regions. The techniques and data used in determining the ENDF/B evaluations are reviewed, and comparisons of model-code calculations and ENDF data with recent experimental results are given. 11 figures

  4. Boron analysis for neutron capture therapy using particle-induced gamma-ray emission

    International Nuclear Information System (INIS)

    Nakai, Kei; Yamamoto, Yohei; Okamoto, Emiko; Yamamoto, Tetsuya; Yoshida, Fumiyo; Matsumura, Akira; Yamada, Naoto; Kitamura, Akane; Koka, Masashi; Satoh, Takahiro

    2015-01-01

    The neutron source of BNCT is currently changing from reactor to accelerator, but peripheral facilities such as a dose-planning system and blood boron analysis have still not been established. To evaluate the potential application of particle-induced gamma-ray emission (PIGE) for boron measurement in clinical boron neutron capture therapy, boronophenylalanine dissolved within a cell culture medium was measured using PIGE. PIGE detected 18 μgB/mL f-BPA in the culture medium, and all measurements of any given sample were taken within 20 min. Two hours of f-BPA exposure was required to create a boron distribution image. However, even though boron remained in the cells, the boron on the cell membrane could not be distinguished from the boron in the cytoplasm. - Highlights: • PIGE was evaluated for measuring blood boron concentration during clinical BNCT. • PIGE detected 18 μgB/mL f-BPA in culture medium. • All measurements of any given sample were taken within 20 min. • Two hours of f-BPA exposure is required to create boron distribution image by PIGE. • Boron on the cell membrane could not be distinguished from boron in the cytoplasm.

  5. The 3D tomographic image reconstruction software for prompt-gamma measurement of the boron neutron capture therapy

    International Nuclear Information System (INIS)

    Morozov, Boris; Auterinen, Iiro; Kotiluoto, Petri; Kortesniemi, Mika

    2006-01-01

    A tomographic imaging system based on the spatial distribution measurement of the neutron capture reaction during Boron Neutron Capture Therapy (BNCT) would be very useful for clinical purpose. Using gamma-detectors in a 2D-panel, boron neutron capture and hydrogen neutron capture gamma-rays emitted by the neutron irradiated region can be detected, and an image of the neutron capture events can be reconstructed. A 3D reconstruction software package has been written to support the development of a 3D prompt-gamma tomographic system. The package consists of three independent modules: phantom generation, reconstruction and evaluation modules. The reconstruction modules are based on algebraic approach of the iterative reconstruction algorithm (ART), and on the maximum likelihood estimation method (ML-EM). In addition to that, two subsets of the ART, the simultaneous iterative reconstruction technique (SIRT) and the component averaging algorithms (CAV) have been included to the package employing parallel codes for multiprocessor architecture. All implemented algorithms use two different field functions for the reconstruction of the region. One is traditional voxel function, another is, so called, blob function, smooth spherically symmetric generalized Kaiser-Bessel function. The generation module provides the phantom and projections with background by tracing the prompt gamma-rays for a given scanner geometry. The evaluation module makes statistical comparisons between the generated and reconstructed images, and provides figure-of-merit (FOM) values for the applied reconstruction algorithms. The package has been written in C language and tested under Linux and Windows platforms. The simple graphical user interface (GUI) is used for command execution and visualization purposed. (author)

  6. Energy balance of ENDF/B-VI

    International Nuclear Information System (INIS)

    MacFarlane, R.E.

    1994-01-01

    ENDF/B-VI through Release 2 has been tested for neutron-photon energy balance using the Heater module of the NJOY nuclear data procesing system. The situation is much improved over ENDF/B-V, but there are still a number of maerials that show problems

  7. ENDF/B-5 formats manual 1984

    Energy Technology Data Exchange (ETDEWEB)

    Kinsey, R; Magurno, B A

    1986-09-01

    The ENDF-5 Format, originally the format of the US Evaluated Nuclear Data File ENDF/B-5, was internationally recommended for the computer storage, processing and exchange of evaluated neutron nuclear data. The present document contains the original Formats Manual of 1979, updated with revisions of Nov. 1983. (author) Figs, tabs

  8. ENDF/B-6 decay data library

    International Nuclear Information System (INIS)

    1992-01-01

    This document summarizes the contents of the nuclear decay data library of ENDF/B-6, the U.S. Evaluated Nuclear Data Library. The data are in ENDF-6 format. A copy of the library is available on magnetic tape from the IAEA Nuclear Data Section, free of charge, upon request. (author)

  9. Nuclear Data Resources for Capture gamma-Ray Spectroscopy and Related Topics

    International Nuclear Information System (INIS)

    Pritychenko, B.

    2011-01-01

    Nuclear reaction data play an important role in nuclear reactor, medical, and fundamental science and national security applications. The wealth of information is stored in internally adopted ENDF-6 and EXFOR formats. We present a complete calculation of resonance integrals, Westcott factors, thermal and Maxwellian-averaged cross sections for Z = 1-100 using evaluated nuclear reaction data. The addition of newly-evaluated neutron reaction libraries, and improvements in data processing techniques allows us to calculate nuclear industry and astrophysics parameters, and provide additional insights on all currently available neutron-induced reaction data. Nuclear reaction calculations will be discussed and an overview of the latest reaction data developments will be given.

  10. 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

  11. CRECT-J, Input Preparation of Evaluated Data in ENDF-4, ENDF-5 and ENDF-6 Formats

    International Nuclear Information System (INIS)

    Nakagawa, T.

    2000-01-01

    Description of program or function: In order to compile evaluated nuclear data in the ENDF format, the computer code CRECTJ has been developed. CRECTJ has two versions: CRECT-J5 treats the data in the ENDF/B-IV and ENDF/B-V format, and CRECTJ6 the data in the ENDF-6 format. These programs have been frequently used to make Japanese Evaluated Nuclear Data Library (JENDL). The program has functions for reading evaluated nuclear data and creating complete files from them. In addition, it also has functions such as arithmetic operations on cross section data, averaging of cross sections, correction of data, and construction of natural element data from its isotopes

  12. ENDF-6 formats manual. Version of Oct. 1991

    International Nuclear Information System (INIS)

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

    1992-01-01

    ENDF-6 is the international computer file format for evaluated nuclear data. In contrast to the earlier versions (ENDF-4 and ENDF-5) the new version ENDF-6 has been designed not only for neutron reaction data but also for photo-nuclear and charged-particle nuclear reaction data. This document gives a detailed description of the formats and procedures adopted for ENDF-6. The present version includes update pages dated Oct. 1991. (author). Refs, figs, and tabs

  13. Use of neutron capture gamma radiation for determining grade of iron ore in blast holes and exploration holes

    International Nuclear Information System (INIS)

    Eisler, P.L.; Huppert, P.; Mathew, P.J.; Wylie, A.W.; Youl, S.F.

    1977-01-01

    Neutron radiative capture and neutron-neutron logging have been applied to determining the grade of ore in dry blast holes and a dry exploration hole drilled into a layered iron deposit. Both thermal and epithermal neutron responses were measured as well as the gamma-ray responses due to neutron capture by iron and by hydrogen present in hydrated minerals. The results were fitted by a stepwise multiple linear regression technique to give expressions for mean grade of ore in the drill hole and 95% confidence intervals for estimation of this mean. For an overall range of ore grades of 20-68% Fe and a mean grade of 63% Fe, the confidence interval for prediction of mean grade for the neutron-gamma technique was 0.3% Fe for pooled data from all five blast holes and 0.8% Fe for a single hole. It was also shown that for this type of layered deposit a simpler neutron-neutron log incorporating simultaneous measurement of both thermal and epithermal neutron responses gave almost as good a grade prediction result for pooled results from five drill holes, namely 63+-0.4% Fe, as that obtained by the neutron-gamma technique. The results of both types of log are compared with those obtained by the spectral gamma-ray backscattering [Psub(z)] technique, or by logging of natural gamma radiations from the shale component of the ore. From this comparison conclusions are drawn regarding the most suitable technique to employ for determining grade of iron ore in various practical logging situations. (author)

  14. The file of evaluated decay data in ENDF/B

    International Nuclear Information System (INIS)

    Reich, C.W.

    1991-01-01

    One important application of nuclear decay data is the Evaluated Nuclear Data File/B (ENDF/B), the base of evaluated nuclear data used in reactor research and technology activities within the United States. The decay data in the Activation File (158 nuclides) and the Actinide File (108 nuclides) excellently represent the current status of this information. In particular, the half-lives and gamma and alpha emission probabilities, quantities that are so important for many applications, of the actinide nuclides represent a significant improvement over those in ENDF/B-V because of the inclusion of data produced by an International Atomic Energy Agency Coordinated Research Program. The Fission Product File contains experimental decay data on ∼510 nuclides, which is essentially all for which a meaningful number of data are available. For the first time, delayed-neutron spectra for the precursor nuclides are included. Some hint of problems in the fission product data base is provided by the gamma decay heat following a burst irradiation of 239 Pu

  15. ENDF/B-5 modifications 1986

    International Nuclear Information System (INIS)

    McLaughlin, P.K.

    1986-05-01

    This document summarizes the modifications made to the ENDF/B-5 evaluated neutron data files in 1986. The new versions of the files are available from the IAEA Nuclear Data Section upon request, costfree, on magnetic tape. (author)

  16. The 1992 ENDF Pre-processing codes

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1992-01-01

    This document summarizes the 1992 version of the ENDF pre-processing codes which are required for processing evaluated nuclear data coded in the format ENDF-4, ENDF-5, or ENDF-6. Included are the codes CONVERT, MERGER, LINEAR, RECENT, SIGMA1, LEGEND, FIXUP, GROUPIE, DICTION, MIXER, VIRGIN, COMPLOT, EVALPLOT, RELABEL. Some of the functions of these codes are: to calculate cross-sections from resonance parameters; to calculate angular distributions, group average, mixtures of cross-sections, etc; to produce graphical plottings and data comparisons. The codes are designed to operate on virtually any type of computer including PC's. They are available from the IAEA Nuclear Data Section, free of charge upon request, on magnetic tape or a set of HD diskettes. (author)

  17. Apparatus for reducing pulse pileup in an elemental analyzer measuring gamma rays arising from neutron capture in bulk substances

    International Nuclear Information System (INIS)

    Marshall, J.H. III.

    1979-01-01

    The active reduction of the number of analyzed events with pulse amplitudes which pileup has distorted improves measurement accuracy and response time in an apparatus for neutron-capture-based on-line elemental analysis of bulk substances. Within the apparatus, the analyzed bulk substance is exposed to neutrons, and neutron capture generates prompt gamma rays therefrom. A detector interacts with some of these gamma rays to produce electrical signals used to measure their energy spectrum by pulse-height analysis. Circuits associated with this pulse-height analysis also detect the pileup of the signals of two or more independent gamma rays using one or more of several techniques. These techniques include multiple outputs from a special amplifier-discriminator system, which has been optimized for low pulse-pair resolving time and may have adaptive thresholds, and the requirement that the relative amplitudes of the outputs of slow and fast amplifiers be consistent with a single event producing both outputs. Pulse-width measurements are also included in the pileup detection

  18. Baseline distortion effect on gamma-ray pulse-height spectra in neutron capture experiments

    International Nuclear Information System (INIS)

    Laptev, A.; Harada, H.; Nakamura, S.; Hori, J.; Igashira, M.; Ohsaki, T.; Ohgama, K.

    2005-01-01

    A baseline distortion effect due to gamma-flash at neutron time-of-flight measurement using a pulse neutron source has been investigated. Pulses from C 6 D 6 detectors accumulated by flash-ADC were processed with both standard analog-to-digital converter (ADC) and flash-ADC operational modes. A correction factor of gamma-ray yields, due to baseline shift, was quantitatively obtained by comparing the pulse height spectra of the two data-taking modes. The magnitude of the correction factor depends on the time after gamma-flash and has complex time dependence with a changing sign

  19. 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)

  20. Thermal neutron capture cross section for Fe-56(n,gamma)

    Czech Academy of Sciences Publication Activity Database

    Firestone, R. B.; Belgya, T.; Krtička, M.; Bečvář, F.; Szentmiklosi, L.; Tomandl, Ivo

    2017-01-01

    Roč. 95, č. 1 (2017), č. článku 014328. ISSN 2469-9985 R&D Projects: GA ČR GA13-07117S; GA MŠk LM2015056 Institutional support: RVO:61389005 Keywords : neutron cross section * gamma gamma-coincidence data Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders OBOR OECD: Nuclear physics Impact factor: 3.820, year: 2016

  1. Determination of contaminants in nuclear materials by measuring the capture gamma rays of thermal neutrons in a reactor internal geometry

    International Nuclear Information System (INIS)

    Suarez, A.A.

    1980-01-01

    A new method for analysis of impurities in nuclear fuel material was developed. Prompt gamma rays following thermal neutron capture, from a sample placed inside the research reactor were analyzed with a solid state high resolution detector. A number of improvements were introduced to improve the background-to-signal ratio, and the sensitivity of the method: use of collimeters for gamma rays and 6 Li 2 CO 3 filters to eliminate thermal neutrons from the beam were supplemented with the application of a pair spectrometer. Using a 42.5 cm 3 true coaxial Ge(Li) detector, and two optically separated NaI (Tl) scintillation detector, the sensitivity of the method for quantitative determination of impurities reached 30 p.p.m. The reproducibility of the results was better than 2%

  2. Elemental analysis of water and soil environmental samples in Tabuk area by neutron capture gamma-ray spectroscopy techniques

    International Nuclear Information System (INIS)

    Al-Aseery, Sh.M.; Alamoudi, Z.; Hassan, A.M.

    2006-01-01

    The prompt and delayed gamma-rays due to neutron capture in the nuclei of the constituent elements of three soil samples and one drinking water sample have been measured. The 252 Cf and 226 Ra/Be isotopic neutron sources are used for neutron irradiation. Also, the hyper pure germanium detection system is used. The soil samples were from Astra, Tadco and El-Gammaz farms, while the water sample was taken from Tabuk city. In case of prompt gamma-ray analysis, a total of 16 elements were identified and the concentration percentage values by weight were calculated for: C, Na, Mg, Al, Si, S, Cl,, Ca, Ti, Cr, Mn, Fe, Co, Zn, Sr ad Pb elements. A comparative study between the results obtained in this work and the results obtained by ICP-MS and EDX-Ray techniques for the same samples is given

  3. Investigation of dose distribution in mixed neutron-gamma field of boron neutron capture therapy using N isopropylacrylamide gel

    Energy Technology Data Exchange (ETDEWEB)

    Bavarmegin, Elham; Sadremomtaz, Alireza [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Khalafi, Hossein; Kasesaz, Yaser [Dept. of Physics, University of Guilan, Rasht (Iran, Islamic Republic of); Khajeali, Azim [Medical Education Research Center, Tabriz (Iran, Islamic Republic of)

    2017-02-15

    Gel dosimeters have unique advantages in comparison with other dosimeters. Until now, these gels have been used in different radiotherapy techniques as a reliable dosimetric tool. Because dose distribution measurement is an important factor for appropriate treatment planning in different radiotherapy techniques, in this study, we evaluated the ability of the N-isopropylacrylamide (NIPAM) polymer gel to record the dose distribution resulting from the mixed neutron-gamma field of boron neutron capture therapy (BNCT). In this regard, a head phantom containing NIPAM gel was irradiated using the Tehran Research Reactor BNCT beam line, and then by a magnetic resonance scanner. Eventually, the R2 maps were obtained in different slices of the phantom by analyzing T2-weighted images. The results show that NIPAM gel has a suitable potential for recording three-dimensional dose distribution in mixed neutron-gamma field dosimetry.

  4. Revisiting the U-238 thermal capture cross section and gamma-raymission probabilities from Np-239 decay

    Energy Technology Data Exchange (ETDEWEB)

    Trkov, A.; Molnar, G.L.; Revay, Zs.; Mughabghab, S.F.; Firestone,R.B.; Pronyaev, V.G.; Nichols, A.L.; Moxon, M.C.

    2005-03-03

    The precise value of the thermal capture cross section of238U is uncertain, and evaluated cross sections from various sourcesdiffer by more than their assigned uncertainties. A number of theoriginal publications have been reviewed to assess the discrepant data,corrections were made for more recent standard cross sections andotherconstants, and one new measurement was analyzed. Due to the strongcorrelations in activation measurements, the gamma-ray emissionprobabilities from the beta decay of 239Np were also analyzed. As aresult of the analysis, a value of 2.683 +- 0.012 barns was derived forthe thermal capture cross section of 238U. A new evaluation of thegamma-ray emission probabilities from 239Np decay was alsoundertaken.

  5. Characteristics of ENDF/B-V

    International Nuclear Information System (INIS)

    Pearlstein, S.; Kinsey, R.; Dunford, C.

    1978-01-01

    A primary source of microscopic nuclear data for processing into multigroup cross sections is the Evaluated Nuclear Data File (ENDF/B). This data file is maintained and distributed by the National Nuclear Data Center (NNDC) of Brookhaven National Laboratory. The File is based on nuclear data evaluations provided by members of the Cross Section Evaluation Working Group (CSEWG). A new version of the ENDF/B (ENDF/B-V) is in preparation for release in the first half of 1978. In order to improve the accuracy and reliability of ENDF/B-V, extensive improvements were made in the checking programs and the review kits. New evaluations are processed through three levels of checking codes that detect errors in formats, consistency, and physical information, in that order. Kits consisting of the results of checking codes, documentation, and plots are presented to designated reviewers for comments. Upon receiving CSEWG approval, evaluations are included in ENDF/B. The major materials in the General Purpose File are being revised for ENDF/B-V where new measurements indicate improvemens are required. The number of materials containing photon production data was increased. A revision of the Photon Interaction File is planned for the end of 1978. An extensive set of integral experiments was adopted as CSEWG Benchmarks to test ENDF/B data. Benchmark experiments were selected to test data for thermal- and fast-reactor, shielding, and dosimetry applications, and additional benchmark candidates are reviewed on a regular basis. CSEWG performs interlaboratory comparisons of the benchmark results. 5 figures

  6. Standardization of electron-capture and complex beta-gamma radionuclides by the efficiency extrapolation method

    International Nuclear Information System (INIS)

    Grigorescu, L.

    1976-07-01

    The efficiency extrapolation method was improved by establishing ''linearity conditions'' for the discrimination on the gamma channel of the coincidence equipment. These conditions were proved to eliminate the systematic error of the method. A control procedure for the fulfilment of linearity conditions and estimation of residual systematic error was given. For law-energy gamma transitions an ''equivalent scheme principle'' was established, which allow for a correct application of the method. Solutions of Cs-134, Co-57, Ba-133 and Zn-65 were standardized with an ''effective standard deviation'' of 0.3-0.7 per cent. For Zn-65 ''special linearity conditions'' were applied. (author)

  7. Atypical excitation-inhibition balance in autism captured by the gamma response to contextual modulation

    NARCIS (Netherlands)

    Snijders, T.M.; Milivojevic, B.; Kemner, C.

    2013-01-01

    Atypical visual perception in people with autism spectrum disorders (ASD) is hypothesized to stem from an imbalance in excitatory and inhibitory processes in the brain. We used neuronal oscillations in the gamma frequency range (30–90 Hz), which emerge from a balanced interaction of excitation and

  8. VIM: Initial ENDF/B-VI experience

    International Nuclear Information System (INIS)

    Blomquist, R.N.

    1997-01-01

    The VIM Monte Carlo particle transport code uses detailed continuous-energy cross sections produced from ENDF/B data by a set of specialized codes developed or adapted for use at Argonne National Laboratory. ENDF/B-IV data were used until about 1979, and Version V data since then. These VIM libraries were extensively benchmarked against the MC 2 -2 code and against ZPR and ZPPR criticals for fast spectrum calculations, as well as other fast and thermal experiments and calculations. Recently, the cross section processing codes have been upgraded to accommodate ENDF/B-VI files, and a small library has been tested. Several fundamental tasks comprise the construction of a faithful representation of ENDF data for VIM calculations: (1) The resolved resonance parameters are converted to Doppler-broadened continuous-energy cross sections with energy grids suitable for linear-linear interpolation. (2) The unresolved resonance parameter distributions are sampled to produce many (40-400) resonance ladders in each energy band. These are converted to Doppler-broadened continuous energy resonance cross sections that are then binned by cross section, accumulating ladders until statistical convergence, the result being probability tables of total cross sections and conditional mean scattering and fission cross sections. VIM samples these tables at run time, and File 3 back ground cross sections are added. (3) Anisotropic angular distribution data are converted to angular probability tables. All other ENDF data are unmodified, except for format

  9. TU-FG-BRB-07: GPU-Based Prompt Gamma Ray Imaging From Boron Neutron Capture Therapy

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S; Suh, T; Yoon, D; Jung, J; Shin, H; Kim, M [The catholic university of Korea, Seoul (Korea, Republic of)

    2016-06-15

    Purpose: The purpose of this research is to perform the fast reconstruction of a prompt gamma ray image using a graphics processing unit (GPU) computation from boron neutron capture therapy (BNCT) simulations. Methods: To evaluate the accuracy of the reconstructed image, a phantom including four boron uptake regions (BURs) was used in the simulation. After the Monte Carlo simulation of the BNCT, the modified ordered subset expectation maximization reconstruction algorithm using the GPU computation was used to reconstruct the images with fewer projections. The computation times for image reconstruction were compared between the GPU and the central processing unit (CPU). Also, the accuracy of the reconstructed image was evaluated by a receiver operating characteristic (ROC) curve analysis. Results: The image reconstruction time using the GPU was 196 times faster than the conventional reconstruction time using the CPU. For the four BURs, the area under curve values from the ROC curve were 0.6726 (A-region), 0.6890 (B-region), 0.7384 (C-region), and 0.8009 (D-region). Conclusion: The tomographic image using the prompt gamma ray event from the BNCT simulation was acquired using the GPU computation in order to perform a fast reconstruction during treatment. The authors verified the feasibility of the prompt gamma ray reconstruction using the GPU computation for BNCT simulations.

  10. Correcting the effects of the matrix using capture gamma-ray spectrometry: Application to measurement by Active Neutron Interrogation

    International Nuclear Information System (INIS)

    Baudry, G.

    2003-11-01

    In the field of the measurement of low masses of fissile material ( 235 U, 239 Pu, 241 Pu) in radioactive waste drums, the Active Neutron Interrogation is a non-destructive method achieving good results. It does however remain reliant upon uncertainties related to the matrix effects on interrogation and fission neutrons. The aim of this thesis is to develop a correction method able to take into account these matrix effects by quantifying the amount of absorbent materials (chlorine and hydrogen) in a 118- liter homogeneous matrix. The main idea is to use the gamma-ray spectrometry of gamma emitted by neutron captures to identify and quantify the composition of the matrix. An indicator from its chlorine content is then deduced in order to choose the calibration coefficient which best represents the real composition of the matrix. This document firstly presents the needs of control and characterization of radioactive objects, and the means used in the field of nuclear measurement. Emphases is put in particular on the Active Neutron Interrogation method. The matrices of interest are those made of light technological waste (density ≤ 0,4 g/cm 3 ) containing hydrogenated and chlorinated materials. The advantages of gamma-rays emitted by neutron captures for the determination of a chlorine content indicator of the matrices and the principles of the correction method are then explained. Measurements have been firstly realized with an existing Neutron Interrogation device (PROMETHEE 6). Such measurements have proven its inadequacy: no signal from the matrix hydrogen was detected, due to an intense signal from the polyethylene contained in some cell elements. Moreover, the matrix chlorine content appeared difficult to be measured. A new and specific device, named REGAIN and dedicated to active gamma-rays spectrometry, was defined with the Monte-Carlo N-Particle (MCNP) code. The experiments conducted with this new device made it possible to detect the hydrogen from the

  11. Analysis of benchmark critical experiments with ENDF/B-VI data sets

    International Nuclear Information System (INIS)

    Hardy, J. Jr.; Kahler, A.C.

    1991-01-01

    Several clean critical experiments were analyzed with ENDF/B-VI data to assess the adequacy of the data for U 235 , U 238 and oxygen. These experiments were (1) a set of homogeneous U 235 -H 2 O assemblies spanning a wide range of hydrogen/uranium ratio, and (2) TRX-1, a simple, H 2 O-moderated Bettis lattice of slightly-enriched uranium metal rods. The analyses used the Monte Carlo program RCP01, with explicit three-dimensional geometry and detailed representation of cross sections. For the homogeneous criticals, calculated k crit values for large, thermal assemblies show good agreement with experiment. This supports the evaluated thermal criticality parameters for U 235 . However, for assemblies with smaller H/U ratios, k crit values increase significantly with increasing leakage and flux-spectrum hardness. These trends suggest that leakage is underpredicted and that the resonance eta of the ENDF/B-VI U 235 is too large. For TRX-1, reasonably good agreement is found with measured lattice parameters (reaction-rate ratios). Of primary interest is rho28, the ratio of above-thermal to thermal U 238 capture. Calculated rho28 is 2.3 (± 1.7) % above measurement, suggesting that U 238 resonance capture remains slightly overpredicted with ENDF/B-VI. However, agreement is better than observed with earlier versions of ENDF/B

  12. Monitoring the sulfur content of coal streams by thermal-neutron-capture gamma-ray analysis

    International Nuclear Information System (INIS)

    Martin, J.W.; Hall, A.W.

    1976-07-01

    A theory was developed for evaluating a complex, prompt gamma ray spectrum to serve as the basis for an instrument to monitor continuously the sulfur content of tonnage streams of coal. Equations for the energies and intensities of prompt gamma rays emitted from 13 most significant elements in coal are combined into a single equation that defines the basic electronic design of the meter. The sulfur content of up to 10 tons per hour of coal was determined in pilot plant tests with a prototype meter. The precision of 0.04 percent sulfur substantiates the validity of the theory. In subsequent industrial plant tests the precision was determined to be a comparable 0.05 percent sulfur

  13. Uses of neutron capture gamma-rays in environmental pollution applications

    International Nuclear Information System (INIS)

    AbdAl-Samad, M.A.

    1998-01-01

    As a sensitive and accurate technique, the prompt gamma-rays neutron activation is used with success for elemental analysis. The advantages of this method over the other techniques are rapidity, usage of relatively large sample size and high reliability, beside the detection of the elements which have no gamma activity during the delayed neutron activation analysis or very short lived isotopes. Actually different techniques could be used for estimating the trace, minor and major elements of these environmental samples which are considered as complex samples. In the mean time the neutron activation analysis techniques have been improved and have become an excellent tool for elemental analysis of complex samples (Duffey et al., 1970; Senftle et al., 1971; Henkelmm and Born, 1973 ; Hassan et al., .; 1981, 1982, 1983; Clyton et al., 1983; Zaghloul et al., 1993) and the advantages of the prompt γ- ray neutron activation analysis over the other techniques put this technique in the fore front

  14. Interference of the two spin components of the capture state in the (n, [gamma]) reaction

    NARCIS (Netherlands)

    Kamp, A.M.F. op den; Kopecky, J.; Stecher-Rasmussen, F.; Abrahams, K.; Endt, P.M.

    1972-01-01

    Measurements of the circularγ-ray polarization for primary transitions in the 39K(n, γ)40K and 57Fe(n, γ)58Fe reactions induced by thermal neutrons give strong evidence for the interference of components in the capture state with different J=values.

  15. ENDF utility codes version 6.8

    International Nuclear Information System (INIS)

    McLaughlin, P.K.

    1992-01-01

    Description and operating instructions are given for a package of utility codes operating on evaluated nuclear data files in the formats ENDF-5 and ENDF-6. Included are the data checking codes CHECKER, FIZCON, PSYCHE; the code INTER for retrieving thermal cross-sections and some other data; graphical plotting codes PLOTEF, GRALIB, graphic device interface subroutine library INTLIB; and the file maintenance and retrieval codes LISTEF, SETMDC, GETMAT, STANEF. This program package which is designed for CDC, IBM, DEC and PC computers, can be obtained on magnetic tape or floppy diskette, free of charge, from the IAEA Nuclear Data Section. (author)

  16. Validating (d,p gamma) as a Surrogate for Neutron Capture

    Energy Technology Data Exchange (ETDEWEB)

    Ratkiewicz, A. [Rutgers University; Cizewski, J. A. [Rutgers University; Pain, S. [Oak Ridge National Laboratory (ORNL); Adekola, A. S. [Rutgers University; Burke, J. T. [Lawrence Livermore National Laboratory (LLNL); Casperson, R.J. [Lawrence Livermore National Laboratory (LLNL); Fotiades, N. [Los Alamos National Laboratory (LANL); McCleskey, M. [Texas A& M University; Burcher, S. [Rutgers University; Shand, C. M. [Rutgers Univ./Univ. of Surrey, UK; Austin, R. A. E. [Saint Mary’s University, Halifa, Canada; Baugher, T. [Rutgers University; Carpenter, M. P. [Argonne National Laboratory (ANL); Devlin, M. [Los Alamos National Laboratory (LANL); Escher, J. E. [Lawrence Livermore National Laboratory (LLNL); Hardy, S. [Rutgers Univ./Univ. of Surrey, UK; Hatarik, R. [Lawrence Livermore National Laboratory (LLNL); Howard, M. [Rutgers University; Hughes, R. [University of Richmond, VA; Jones, K. L. [University of Tennessee, Knoxville (UTK); Kozub, R. L. [Tennessee Technological University (TTU); Lister, C. J. [University of Massachusetts, Lowell; Manning, B. [Rutgers University; O' Donnell, J. M. [Los Alamos National Laboratory (LANL); Peters, W. A. [Oak Ridge Associated Universities (ORAU); Ross, T.J. [University of Richmond, VA; Scielzo, N.D. [Lawrence Livermore National Laboratory (LLNL); Seweryniak, D. [Argonne National Laboratory (ANL); Zhu, S. [Argonne National Laboratory (ANL)

    2015-01-01

    The r-process is responsible for creating roughly half of the elements heavier than iron. It has recently become understood that the rates at which neutron capture reactions proceed at late times in the rprocess may dramatically affect the final abundance pattern. However, direct measurements of neutron capture reaction rates on exotic nuclei are exceptionally difficult, necessitating the development of indirect approaches such as the surrogate technique. The (d,py) reaction at low energies was identified as a promising surrogate for the (n,y) reaction, as both reactions share many characteristics. We report on a program to validate (d,py) as a surrogate for (n,y) using 95Mo as a target. The experimental campaign includes direct measurements of the y-ray intensities from the decay of excited states populated in the 95Mo(n,y) and 95Mo(d,py) reactions.

  17. Fast neutron capture in 197Au and the gamma-ray strength function

    International Nuclear Information System (INIS)

    Earle, E.D.; Bergqvist, I.; Nilsson, L.

    1977-08-01

    Gamma-ray spectra from the reaction 197 Au(n,γ) 198 Au have been measured at several incident energies between 30 keV and 2.5 MeV. The γ-ray detector was a NaI(Tl) scintillation detector and time-of-flight techniques were utilized to suppress background. A γ-ray strength function was deduced from the spectra by a spectrum fitting method. The strength function indicates a resonance-like structure at E(γ) approximately 5.5 MeV. Comparison is made with the γ-ray strength derived from photonuclear work.(author)

  18. Monitoring of processes with gamma-rays of neutron capture and short-living radionuclides

    International Nuclear Information System (INIS)

    Aripov, G.A.; Kurbanov, B.I.; Allamuratova, G.

    2004-01-01

    Element content is a fundamental parameter of a substance, on which all its properties, and also character of physical, chemical, biological, technological and ecological processes depend. Therefore monitoring of element content (in the course of technological process - on line; in natural conditions - in site; or in living organisms - in vivo) becomes necessary for investigation of aforementioned processes. This problem can be successfully solved by using the methods of prompt gamma activation analysis (PGAA) and instrumental neutron activation analysis (INAA) on short-living radionuclides. These methods don't depend on type of substance (biological, geological, technological etc.), since the content is determined by gamma radiation of nuclei, and allows to meet such a serious requirement like the necessity of achieving minimal irradiation of the object and its minimal residual activity. In this work minimal determinable concentrations of various elements are estimated (based on experimental data) by the method of PGAA using radionuclide 252 Cf - source of neutrons with the yield of the oil of 10 8 neutron/sec on the experimental device with preliminary focusing of neutrons /1/, and also data of determination of elements by their isotopes with maximum time efficiency /2,3/ by the method of INAA. (author)

  19. Measurements of gamma rays from keV-neutron resonance capture by odd-Z nuclei in the 2s-1d shell region

    Energy Technology Data Exchange (ETDEWEB)

    Igashira, Masayuki; Lee, Sam Yol; Mizuno, Satoshi; Hori, Jun-ichi [Tokyo Inst. of Tech. (Japan). Research Lab. for Nuclear Reactors; Kitazawa, Hideo

    1998-03-01

    Measurements of gamma rays from keV-neutron resonance capture by {sup 19}F, {sup 23}Na, and {sup 27}Al, which are odd-Z nuclei in the 2s-1d shell region, were performed, using an anti-Compton HPGe spectrometer and a pulsed neutron source by the {sup 7}Li(p,n){sup 7}Be reaction. Capture gamma rays from the 27-, 49-, and 97-keV resonances of {sup 19}F, the 35- and 53-keV resonances of {sup 23}Na, and the 35-keV resonance of {sup 27}Al were observed. Some results are presented. (author)

  20. Thermal neutron capture cross sections resonance integrals and g-factors

    International Nuclear Information System (INIS)

    Mughabghab, S.F.

    2003-02-01

    The thermal radiative capture cross sections and resonance integrals of elements and isotopes with atomic numbers from 1 to 83 (as well as 232 Th and 238 U) have been re-evaluated by taking into consideration all known pertinent data published since 1979. This work has been undertaken as part of an IAEA co-ordinated research project on 'Prompt capture gamma-ray activation analysis'. Westcott g-factors for radiative capture cross sections at a temperature of 300K were computed by utilizing the INTER code and ENDF-B/VI (Release 8) library files. The temperature dependence of the Westcott g-factor is illustrated for 113 Cd, 124 Xe and 157 Gd at temperatures of 150, 294 and 400K. Comparisons have also been made of the newly evaluated capture cross sections of 6 Li, 7 Li, 12 C and 207 Pb with those determined by the k 0 method. (author)

  1. ENDF/B-5 Actinides (Rev. 86)

    International Nuclear Information System (INIS)

    Lemmel, H.D.

    1986-05-01

    This document summarizes the contents of the Actinides part of the ENDF/B-5 nuclear data library released by the US National Nuclear Data Center. This library or selective retrievals of it, are available costfree from the IAEA Nuclear Data Section upon request. The present version of the library is the Revision of 1986. (author). Refs, figs and tabs

  2. ENDF/B-V processing programs

    International Nuclear Information System (INIS)

    DayDay, N.

    1980-07-01

    A description and operating instructions are supplied for the following ENDF/B-V Processing Programs: CHECKER, CRECT, STNDRD, FIZCON, PSYCHE, RESEND, INTER, INTEND, SUMRIZ, PLOTEF, LSTFCV, RIGEL. These programs can be obtained on magnetic tape, free of charge, from the IAEA Nuclear Data Section. (author)

  3. ENDF/B Thermal Data Testing

    CERN Document Server

    McCrosson, F J

    2001-01-01

    The thermal data testing group is concerned with establishing the merit of ENDF/B cross sections for the analysis of thermal systems. The integral experiments used in the testing are designed to analyze each of the phenomena identified in the familiar four-factor formula. For brevity, only the testing of the cross sections in uranium systems is described in this report.

  4. LEGEND2007, Angular Distribution Table Calculations in ENDF Format

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of program or function: LEGEND calculates linearly interpolable tabulated angular distributions starting from data in the ENDF/B format. IAEA1310/11: This version include the updates up to January 30, 2007. Changes in ENDF/B-VII Format and procedures, as well as the evaluations themselves, make it impossible for versions of the ENDF/B pre-processing codes earlier than PREPRO 2007 (2007 Version) to accurately process current ENDF/B-VII evaluations. The present code can handle all existing ENDF/B-VI evaluations through release 8, which will be the last release of ENDF/B-VI. Modifications from previous versions: Legend VERS. 2007-1 (JAN. 2007): checked against all ENDF/B=VII; increased max. points from 60,000 to 240,000

  5. Study of gamma cascades and strength functions in the neutron capture reaction 77Se(n,γ)

    International Nuclear Information System (INIS)

    John, Robert

    2014-01-01

    One of the most important nuclear processes is the nuclear capture reaction. The cosmic nucleosynthesis (s-process) of heavy elements produces nuclei with mass numbers greater than 56 (Iron), which cannot be produced by nuclear fusion. A nucleus gets exited to the binding energy via capture of a neutron and afterwards deexcites to the groundstate by the emission of photons (gamma rays). The characteristics of the γ rays allow conclusions about the structure of the nucleus. In this work the photons, sent out by the excited 78 Se * , were analyzed. The experiment took place at the research reactor of the Institute Laue-Langevin in Grenoble, France. After a efficiency calibration and the addback procedure the multi detector setup allowed coincidence and directional correlation measurements. With the help of these measurements a level scheme was developed and the directional correlation measurements were used to assign spins to different levels. Furthermore the experimental acquired data were compared to results of a simulation (γDEX) and a photon scattering experiment carried out at the ELBE electron accelerator.

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

    International Nuclear Information System (INIS)

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

    2016-12-01

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

  7. Neutron cross sections for uranium-235 (ENDF/B-IV Release 3)

    International Nuclear Information System (INIS)

    Lubitz, C.R.

    1996-09-01

    The resonance parameters in ENDF6 (Release 2) U235 were adjusted to make the average capture and fission cross sections below 900 eV agree with selected differential capture and fission measurements. The measurements chosen were the higher of the credible capture measurements and the lower of the fission results, yielding a higher epithermal alpha. In addition, the 2200 m/s cross sections were adjusted to obtain agreement with the integral value of K1. As a result, criticality calculations for thermal benchmarks, and agreement with a variety of integral parameters, are improved

  8. MCNP calculations for criticality-safety benchmarks with ENDF/B-V and ENDF/B-VI libraries

    International Nuclear Information System (INIS)

    Iverson, J.L.; Mosteller, R.D.

    1995-01-01

    The MCNP Monte Carlo code, in conjunction with its continuous-energy ENDF/B-V and ENDF/B-VI cross-section libraries, has been benchmarked against results from 27 different critical experiments. The predicted values of k eff are in excellent agreement with the benchmarks, except for the ENDF/B-V results for solutions of plutonium nitrate and, to a lesser degree, for the ENDF/B-V and ENDF/B-VI results for a bare sphere of 233 U

  9. Feasibility of using prompt neutron capture gamma rays to detect mercury

    International Nuclear Information System (INIS)

    Bell, Z.W.

    1993-10-01

    This report describes a study to determine the feasibility to use neutrons to probe hidden spaces within buildings for the presence of mercury. The study was performed in four phases: First a search of the scientific literature was performed to ascertain the behavior of mercury subsequent to the capture of a thermal or near-thermal neutron. Second, a Monte Carlo investigation (using the code MCNP) of the effects of neutrons on materials expected to be found near and/or surrounding the mercury was undertaken. Third, a Monte Carlo study of the shielding and beam forming properties of various configurations of moderator material was started. Lastly, a Monte Carlo analysis of a likely field situation involving mercury behind 1 inch and 2 inch thicknesses of concrete was performed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  11. SB2. Experiment on secondary gamma-ray production cross sections arising from thermal-neutron capture in each of 14 different elements plus a stainless steel

    International Nuclear Information System (INIS)

    Maerker, R.E.

    1976-01-01

    The experimental and calculational details for a CSEWG integral data testing shielding experiment are presented. This particular experiment measured the secondary gamma-ray production cross sections arising from thermal-neutron capture in iron, nitrogen, sodium, aluminum, copper, titanium, calcium, potassium, chlorine, silicon, ickel, zinc, barium, sulfur and a type 321 stainless steel. 1 figure, 30 tables

  12. Comparison of decay and yield data between JNDC2 and ENDF/B-VI

    Energy Technology Data Exchange (ETDEWEB)

    Oyamatsu, K.; Sagosaka, M.; Miyazono, T. [Nagoya Univ. (Japan)

    1997-03-01

    This work is intended to be our first step to solve disagreements of the decay heat powers between measurements and summation calculations. We examine differences between nuclear data libraries to complement our uncertainty evaluation of the decay heat summation calculations only with ENDF/B-VI. The comparison is made mainly between JNDC2 and ENDF/B-VI while JEF2.2 decay data is also discussed. In this study, we propose and use a simple method which is an analogue of the overlap integral of two wave functions in quantum mechanics. As the first step, we compare the whole input nuclear data for the summation calculations as a whole. We find a slight difference of the fission yields especially for high-energy neutron induced fissions between JNDC2 and ENDF/B-VI. As for the decay energies, JNDC2, ENDF/B-VI are quite similar while JEF2.2 is found significantly different from these two libraries. We find substantial differences in the decay constant values among the three libraries. As the second step, we calculate the decay heat powers with FPGS90 using JNDC2 and ENDF/B-VI. The total decay heat powers with the two libraries differ by more than 10% at short cooling times while they agree well on the average at cooling times longer that 100 (s). We also point out nuclides whose contributions are significantly different between the two libraries even though the total decay heats agree well. These nuclides may cause some problems in predicting aggregate spectra of {beta} and {gamma} rays as well as delayed neutrons, and are to be reviewed in the future revision of decay and yield data. (author)

  13. Response of Moxon-Rae type gamma detectors for neutron capture cross section measurements

    International Nuclear Information System (INIS)

    Iyengar, K.V.K.; Lal, B.; Jhingan, M.L.

    1974-01-01

    A detector devised by Moxon and Rae for the absolute measurement of (n,γ) cross sections is briefly described. This detector is supposed to have an efficiency per MeV of γ-ray energy independent of the energy of the γ-rays. Such a detector consists of an electron converter placed before a thin plastic scintillator which detects the electron emitted by interaction of the γ-ray in the converter. The performance of this type of detector depends on the thickness and composition of the converter. Detailed Monte-Carlo calculations of the response for γ-ray energies from 0.2 to 12 MeV has been carried out for elements ranging from C to Bi and for a mixture of elements as well as for a mixture of an element plus compound, to find out the suitable material and thickness of the converter. Among the elements studied for the converter, Ni, Mo and Sn have a uniform response over the photon energy range 1-12 MeV. Out of these elements Mo has a low neutron capture cross section in the energy range 1-1000 keV and is thus to be preferred. A mixture of C + Bi 2 O 3 in the weight ratio 11.6 : 88.4 gives a uniform response over the photon energy range 1-12 MeV. (K.B.)

  14. ENDF-6 Formats Manual. Data Formats and Procedures for the Evaluated Nuclear Data File ENDF/B-VI and ENDF/B-VII

    International Nuclear Information System (INIS)

    Herman, M.

    2009-01-01

    In December 2006, the Cross Section Evaluation Working Group (CSEWG) of the United States released the new ENDF/B-VII.0 library. This represented considerable achievement as it was the 1st major release since 1990 when ENDF/B-VI has been made publicly available. The two libraries have been released in the same format, ENDF-6, which has been originally developed for the ENDF/B-VI library. In the early stage of work on the VII-th generation of the library CSEWG made important decision to use the same formats. This decision was adopted even though it was argued that it would be timely to modernize the formats and several interesting ideas were proposed. After careful deliberation CSEWG concluded that actual implementation would require considerable resources needed to modify processing codes and to guarantee high quality of the files processed by these codes. In view of this the idea of format modernization has been postponed and ENDF-6 format was adopted for the new ENDF/B-VII library. In several other areas related to ENDF we made our best to move beyond established tradition and achieve maximum modernization. Thus, the 'Big Paper' on ENDF/B-VII.0 has been published, also in December 2006, as the Special Issue of Nuclear Data Sheets 107 (1996) 2931-3060. The new web retrieval and plotting system for ENDF-6 formatted data, Sigma, was developed by the NNDC and released in 2007. Extensive paper has been published on the advanced tool for nuclear reaction data evaluation, EMPIRE, in 2007. This effort was complemented with release of updated set of ENDF checking codes in 2009. As the final item on this list, major revision of ENDF-6 Formats Manual was made. This work started in 2006 and came to fruition in 2009 as documented in the present report.

  15. ENDF-6 Formats Manual Data Formats and Procedures for the Evaluated Nuclear Data File ENDF/B-VI and ENDF/B-VII

    Energy Technology Data Exchange (ETDEWEB)

    Herman, M.; Members of the Cross Sections Evaluation Working Group

    2009-06-01

    In December 2006, the Cross Section Evaluation Working Group (CSEWG) of the United States released the new ENDF/B-VII.0 library. This represented considerable achievement as it was the 1st major release since 1990 when ENDF/B-VI has been made publicly available. The two libraries have been released in the same format, ENDF-6, which has been originally developed for the ENDF/B-VI library. In the early stage of work on the VII-th generation of the library CSEWG made important decision to use the same formats. This decision was adopted even though it was argued that it would be timely to modernize the formats and several interesting ideas were proposed. After careful deliberation CSEWG concluded that actual implementation would require considerable resources needed to modify processing codes and to guarantee high quality of the files processed by these codes. In view of this the idea of format modernization has been postponed and ENDF-6 format was adopted for the new ENDF/B-VII library. In several other areas related to ENDF we made our best to move beyond established tradition and achieve maximum modernization. Thus, the 'Big Paper' on ENDF/B-VII.0 has been published, also in December 2006, as the Special Issue of Nuclear Data Sheets 107 (1996) 2931-3060. The new web retrieval and plotting system for ENDF-6 formatted data, Sigma, was developed by the NNDC and released in 2007. Extensive paper has been published on the advanced tool for nuclear reaction data evaluation, EMPIRE, in 2007. This effort was complemented with release of updated set of ENDF checking codes in 2009. As the final item on this list, major revision of ENDF-6 Formats Manual was made. This work started in 2006 and came to fruition in 2009 as documented in the present report.

  16. Use of neutron-capture prompt gamma-ray activation analysis in environmental and energy research

    International Nuclear Information System (INIS)

    Gordon, G.E.

    1980-01-01

    Reactor-based neutron-capture prompt γ-ray activation analysis (PGAA) can be used for several trace elements (e.g., B, Cd, Sm, Gd), but it is mainly a major and minor element method. Among the 15 to 20 elements that can be determined in a particular type of sample, the major elements C, H, N and S can usually be measured in coals and biological samples and the elements of geochemical interest, Na, Al, Si, K, Ca, Ti, Fe and others in rocks, coal and other crustal samples. When used along with instrumental neutron activation analysis (INAA), the combined techniques can be used to analyze for 40 to 50 elements in crustal samples. Suggestions for further development of reactor-based PGAA include: the need to reduce low-energy γ-ray backgrounds for observations of species having intense lines only at low energy, more development of biological and special materials applications, the use of beam choppers to look for very short-lived species, and demonstrations of analyses of samples too large, precious or hazardous to be irradiated inside reactors. The greatest potential for applications of PGAA may be in the use of systems detached from reactors, i.e., those using 252 Cf or other isotopic neutron sources. This may be the only method for elemental analyses of large, inhomogeneous process streams in various industries, power plants, etc. Fluxes comparable to those of external beams from reactors are feasible with the use of mg quantitites of 252 Cf. Practical systems for use on coal have been developed. Many other applications would be possible: field prospecting for valuable ore deposits, studies of nutrients in soil and plants, in situ measurements in remote locations such as planetary surfaces, mines, bore holes, etc

  17. The 1996 ENDF pre-processing codes

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1996-01-01

    The codes are named 'the Pre-processing' codes, because they are designed to pre-process ENDF/B data, for later, further processing for use in applications. This is a modular set of computer codes, each of which reads and writes evaluated nuclear data in the ENDF/B format. Each code performs one or more independent operations on the data, as described below. These codes are designed to be computer independent, and are presently operational on every type of computer from large mainframe computer to small personal computers, such as IBM-PC and Power MAC. The codes are available from the IAEA Nuclear Data Section, free of charge upon request. (author)

  18. The status of ENDF/B-VI

    International Nuclear Information System (INIS)

    Roussin, R.; Dunford, C.; McKnight, R.; Young, P.

    1988-01-01

    A new version of the United States evaluated nuclear data file, ENDF/B-VI, is presently under development. Major emphasis is being placed on correcting some long-standing nuclear data problems that adversely affect applied calculations for both fission and fusion reactors. The paper reviews modifications to the formats and utility codes, outlines the evaluation activities, discusses the data testing programs, and projects a date for the unrestricted release of the new library. 27 refs., 2 tabs

  19. ENDF-201 Supplement 1. ENDF/B-V.2 summary documentation. Third edition

    International Nuclear Information System (INIS)

    Magurno, B.A.; Young, P.G.

    1985-01-01

    The purpose of the present publication is to provide the summary documentation for Revision 2 of the General Purpose File of ENDF/B-V. Revision 2 embodies a series of updates for important materials in the ENDF/B-V libraries in advance of the more comprehensive changes that will accompany a complete version change to ENDF/B, Verison VI. The summary documents for the evaluations that were completely or extensively modified during Revision 2 are provided. Less extensive changes made to other evaluations are also documented. Modifications made to the fission energy release parameters for fissionable nuclei are described. Finally, a series of modifications made to the covariance files of the major fissionable nuclei, including cross correlation information, is summarized

  20. STANEF, ENDF/B Book-keeping Operations for ENDF Format Files

    International Nuclear Information System (INIS)

    Dunford, Charles L.

    2007-01-01

    1 - Description of program or function: STANEF performs bookkeeping operations on a data file containing one or more material evaluations in ENDF format. Version 7.01 (Feb 2005): set success flag when done added new element roentgenium (rg) corrected symbol generation for second third metastable state (Kellett, NEA) added symbol nn for neutron in symbol generation and xx for unnamed elements corrections for fact that some compilers do not recognize the intrinsic function jfix 2 - Method of solution: STANEF operations include: Creation or modification of a 'tape ID' record, Creation or update of the directory in MT=451, Create or modify special Hollerith ID records in MT=451 (ENDF-6 only), Re-sequencing, Conversion of integer and floating point fields to standard format, Creation of a binary (ENDF alternate format) file

  1. ENDF/B-VII.1 versus ENDF/B-VII.0: What's Different?

    International Nuclear Information System (INIS)

    Cullen, D.E.

    2012-01-01

    Recently the new ENDF/B-VII.1 library was released; this completely replaces the earlier ENDF/B-VII.0 library. One of the first questions we ask about a new library is: What's Different? Here I attempt to at least partially answer this question. I present results in both tabulated form (so you can quickly determine if any evaluations of interest to you have changed), and graphic form (so that you can see how much evaluations have changed and in what energy ranges). For the table I have compared what I refer to as the ENDF neutron data, namely MF=1 through 6. Here I did a character-by-character comparison of the same sections (MF/MT) that appear I both ENDF/B-VII.0 and VII.1; here I found differences in 170 evaluations. For the plots I have only compared the total cross sections for all evaluations that are common to both libraries, and I found that of the 423 evaluations in ENDF/B-VII.1, 120 of these have total cross sections that differ by 1% or more from the evaluation of the same isotope in ENDF/B-VII.0. This should be considered only a preliminary comparison; obviously there can be more subtle important differences that do not effect of total cross sections. Here I present plots comparing the total cross section of these 120 isotopes. The plots are only broad overviews of the total cross sections over their entire energy range. If you have interest in more detailed plots for specific evaluations, you can download the evaluations (1,2) and the PREPRO (3) codes I used to prepare and view the data. This is all I needed to do my comparisons, and is all you should need to do any more detailed comparisons to meet your individual needs.

  2. Monitoring of blood-10B concentration for boron neutron capture therapy using prompt gamma-ray analysis

    International Nuclear Information System (INIS)

    Raaijmakers, C.P.J.; Konijnenberg, M.W.; Dewit, L.; Mijnheer, B.J.; Haritz, D.; Huiskamp, R.; Philipp, K.; Siefert, A.; Stecher-Rasmussen, F.

    1995-01-01

    The aim of the present study was to monitor the blood- 10 B concentration of laboratory dogs receiving boron neutron capture therapy, in order to obtain optimal agreement between prescribed and actual dose. A prompt gamma-ray analysis system was developed for this purpose at the High Flux Reactor in Petten. The technique was compared with inductively coupled plasma-atomic emission spectrometry and showed good agreement. A substantial variation in 10 B clearance pattern after administration of borocaptate sodium was found between the different dogs. Consequently, the irradiation commencement was adjusted to the individually determined boron elimination curve. Mean blood- 10 B concentratios during irradiation of 25.8±2.2 μg/g (1 SD, n=18) and 49.3±5.3 μg/g (1 SD, n=17) were obtained for intended concentrations of 25 μg/g and 50 μg/g, respectively. These variations are a factor of two smaller than irradiations performed at a uniform post-infusion irradiation starting time. Such a careful bolld- 10 B monitoring procedure is a prerequisite for accurately obtaining such steep dose-response curves as observed during the dog study. (orig.)

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  4. ENDF/B-5 formats manual. Revised update pages of Nov. 1983. Reprint of B.A. Magurno, BNL-NCS--50496 (ENDF-102) 2nd Edition

    Energy Technology Data Exchange (ETDEWEB)

    Magurno, B A

    1986-09-01

    The ENDF-5 Format, originally the format of the US Evaluated Nuclear Data File ENDF/B-5, was internationally recommended for the computer storage, processing and exchange of evaluated neutron nuclear data. The pages included in this document serve as an update to the original ENDF-5 Formats Manual BNL-NCS-50496 [ENDF-102] 2nd Edition, October 1979. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-12-01

    confirmed for selected actinide reaction rates such as {sup 236}U, {sup 238,242}Pu and {sup 241,243}Am capture in fast systems. Other deficiencies, such as the overprediction of Pu solution system critical eigenvalues and a decreasing trend in calculated eigenvalue for {sup 233}U fueled systems as a function of Above-Thermal Fission Fraction remain. The comprehensive nature of this critical benchmark suite and the generally accurate calculated eigenvalues obtained with ENDF/B-VII.1 neutron cross sections support the conclusion that this is the most accurate general purpose ENDF/B cross section library yet released to the technical community.

  6. FIXUP2007, ENDF Format Redundant Cross-Sections Check

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of program or function: FIXUP is designed to read evaluated data in the ENDF/B format, perform corrections and output the results in the ENDF/B format. One of the most important functions of this code is to redefine all redundant cross sections to be exactly equal to the sum of its parts. IAEA1309/11: This version includes the updates up to January 30, 2007. Changes in ENDF/B-VII Format and procedures, as well as the evaluations themselves, make it impossible for versions of the ENDF/B pre-processing codes earlier than PREPRO 2007 (2007 Version) to accurately process current ENDF/B-VII evaluations. The present code can handle all existing ENDF/B-VI evaluations through release 8, which will be the last release of ENDF/B-VI. Modifications from previous versions: - Fixup VERS. 2007-1 (Jan. 2007): checked against all ENDF/B-VII; increased page size from 60,000 to 600,000 data points 2 - Method of solution: FIXUP: All MAT numbers on an ENDF/B tape are processed; each MAT is treated separately. Within each MAT, each section before and after MF=3 is read, checked/corrected and output. When MF=3 is located, all cross sections are read, sections deleted, created, checked/corrected (based on user input) and after several intermediate stages written to output. 3 - Restrictions on the complexity of the problem: The program uses only the ENDF/B BCD format tape and copy all sections except File 3 as read. It is assumed that the data is correctly coded. No error checking is performed. Since File 3 data are in identical format for ENDF/B versions I through VI, the program can be used with all these versions. - All data in file 3 and 23 must be linearly interpolable

  7. ENDF-102 DATA FORMATS AND PROCEDURES FOR THE EVALUATED NUCLEAR DATA FILE ENDF-6

    International Nuclear Information System (INIS)

    MCLANE, V.

    2001-01-01

    The Evaluated Nuclear Data File (ENDF) formats and libraries are decided by the Cross Section Evaluation Working Group (CSEWG), a cooperative effort of national laboratories, industry, and universities in the U.S. and Canada, and are maintained by the National Nuclear Data Center (NNDC). Earlier versions of the ENDF format provided representations for neutron cross sections and distributions, photon production from neutron reactions, a limited amount of charged-particle production from neutron reactions, photo-atomic interaction data, thermal neutron scattering data, and radionuclide production and decay data (including fission products). Version 6 (ENDF-6) allows higher incident energies, adds more complete descriptions of the distributions of emitted particles, and provides for incident charged particles and photonuclear data by partitioning the ENDF library into sub-libraries. Decay data, fission product yield data, thermal scattering data, and photo-atomic data have also been formally placed in sub-libraries. In addition, this rewrite represents an extensive update to the Version V manual

  8. ENDF-102 DATA FORMATS AND PROCEDURES FOR THE EVALUATION NUCLEAR DATA FILE ENDF-6.

    Energy Technology Data Exchange (ETDEWEB)

    MCLANE,V.

    2001-05-15

    The Evaluated Nuclear Data File (ENDF) formats and libraries are decided by the Cross Section Evaluation Working Group (CSEWG), a cooperative effort of national laboratories, industry, and universities in the U.S. and Canada, and are maintained by the National Nuclear Data Center (NNDC). Earlier versions of the ENDF format provided representations for neutron cross sections and distributions, photon production from neutron reactions, a limited amount of charged-particle production from neutron reactions, photo-atomic interaction data, thermal neutron scattering data, and radionuclide production and decay data (including fission products). Version 6 (ENDF-6) allows higher incident energies, adds more complete descriptions of the distributions of emitted particles, and provides for incident charged particles and photonuclear data by partitioning the ENDF library into sub-libraries. Decay data, fission product yield data, thermal scattering data, and photo-atomic data have also been formally placed in sub-libraries. In addition, this rewrite represents an extensive update to the Version V manual.

  9. ENDF/B-III scattering law library

    International Nuclear Information System (INIS)

    Goulo, V.; Lemmel, H.D.

    1989-12-01

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

  10. Gamma-Ray Emission Spectra as a Constraint on Calculations of 234,236,238U Neutron-Capture Cross Sections

    Energy Technology Data Exchange (ETDEWEB)

    Ullmann, John Leonard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kawano, Toshihiko [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bredeweg, Todd Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Baramsai, Bayarbadrakh [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Couture, Aaron Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Haight, Robert Cameron [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Jandel, Marian [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mosby, Shea Morgan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); O' Donnell, John M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rundberg, Robert S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vieira, David J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wilhelmy, Jerry B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Becker, John A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wu, Ching-Yen [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Krticka, Milan [Charles Univ., Prague (Czech Republic)

    2015-05-28

    Neutron capture cross sections in the “continuum” region (>≈1 keV) and gamma-emission spectra are of importance to basic science and many applied fields. Careful measurements have been made on most common stable nuclides, but physicists must rely on calculations (or “surrogate” reactions) for rare or unstable nuclides. Calculations must be benchmarked against measurements (cross sections, gamma-ray spectra, and <Γγ>). Gamma-ray spectrum measurements from resolved resonances were made with 1 - 2 mg/cm2 thick targets; cross sections at >1 keV were measured using thicker targets. The results show that the shape of capture cross section vs neutron energy is not sensitive to the form of the strength function (although the magnitude is); the generalized Lorentzian E1 strength function is not sufficient to describe the shape of observed gamma-ray spectra; MGLO + “Oslo M1” parameters produces quantitative agreement with the measured 238U(n,γ) cross section; additional strength at low energies (~ 3 MeV) -- likely M1-- is required; and careful study of complementary results on low-lying giant resonance strength is needed to consistently describe observations.

  11. ENDF/B-VI data for MCNP trademark

    International Nuclear Information System (INIS)

    Hendricks, J.S.; Frankle, S.C.; Court, J.D.

    1994-12-01

    Nuclear and atomic data are the foundation upon which the radiation transport codes are built. For neutron transport the international standard is the Evaluated Nuclear Data File from Brookhaven National Laboratory. The latest version, ENDF/B-VI release 2, has recently become available for use in the Monte Carlo N-Particle (MCNP) radiation transport code. These neutron cross-section data are designated by ZAID identifiers ending in .60c and are referred to as the ENDF60 library. The ENDF60 data library was processed from the ENDF/B-VI evaluations using the NJOY code. Fifty-two percent of the data evaluations are translations from ENDF/B-V. The remaining 48% are new evaluations which have sometimes changed significantly. The RSIC release package contains the ENDF60 neutron library, a new photon library MCPLIB02, the electron library EL1, and an updated XSDIR file. The authors report here the work done by the LANL Radiation Transport Group (X-6) in testing and validating the ENDF60 data library and in developing the necessary new sampling and detector schemes. When the ENDF60 library should be used in preference to the previous libraries, is also considered. The development of the new photon library MCPLIB02 is also discussed

  12. Benchmark analysis of MCNP trademark ENDF/B-VI iron

    International Nuclear Information System (INIS)

    Court, J.D.; Hendricks, J.S.

    1994-12-01

    The MCNP ENDF/B-VI iron cross-section data was subjected to four benchmark studies as part of the Hiroshima/Nagasaki dose re-evaluation for the National Academy of Science and the Defense Nuclear Agency. The four benchmark studies were: (1) the iron sphere benchmarks from the Lawrence Livermore Pulsed Spheres; (2) the Oak Ridge National Laboratory Fusion Reactor Shielding Benchmark; (3) a 76-cm diameter iron sphere benchmark done at the University of Illinois; (4) the Oak Ridge National Laboratory Benchmark for Neutron Transport through Iron. MCNP4A was used to model each benchmark and computational results from the ENDF/B-VI iron evaluations were compared to ENDF/B-IV, ENDF/B-V, the MCNP Recommended Data Set (which includes Los Alamos National Laboratory Group T-2 evaluations), and experimental data. The results show that the ENDF/B-VI iron evaluations are as good as, or better than, previous data sets

  13. ENDF/B fission product decay data

    International Nuclear Information System (INIS)

    Rose, P.F.; Burrows, T.W.

    1976-08-01

    The fission product data have been organized by A-chains in order of ascending A from A = 72 to A = 167. The heading page is followed by more detailed information on the individual members of the chain in order of increasing Z and decreasing metastable state. The detailed information for each member includes the ENDF/B-IV File 1 comments and references if available and applicable to the decay data. Following the comments is a decay scheme of the nuclide tabulating the quantities T/sub 1 / 2 /, Q, branching ratio (BR), (E/sub γ/), (E/sub β/), and (E/sub α/). Uncertainties are given if available in the file. Independent fission yields are given, as well as thermal cross sections and resonance integrals as obtained from ENDF/B-IV. All energies listed in this publication are in keV, and all branching ratios (BR) sum to unity. If there are spectra in the decay data file, the decay scheme is followed by tables of photon, particle, and characteristic radiation. For cases in which the multipolarities could be obtained from the file the tables also contain information on x-rays, conversion electrons, and Auger electrons. Associated with the photon and particle radiation tables are the appropriate average energies per decay for each type of radiation, including neutrino radiation

  14. Measurements of neutron-induced capture and fission reactions on $^{235}$ U: cross sections and ${\\alpha}$ ratios, photon strength functions and prompt ${\\gamma}$-ray from fission

    CERN Multimedia

    We propose to measure the neutron-induced capture cross section of the fissile isotope $^{235}$U using a fission tagging set-up. This new set-up has been tested successfully in 2010 and combines the n_TOF 4${\\pi}$ Total Absorption Calorimeter (TAC) with MicroMegas (MGAS) fission detectors. It has been proven that such a combination of detectors allows distinguishing with very good reliability the electromagnetic cascades from the capture reactions from dominant ${\\gamma}$-ray background coming from the fission reactions. The accurate discrimination of the fission background is the main challenge in the neutron capture cross section measurements of fissile isotopes. The main results from the measurement will be the associated capture cross section and ${\\alpha}$ ratio in the resolved (0.3-2250 eV) and unresolved (2.25-30 keV) resonance regions. According to the international benchmarks and as it is mentioned in the NEA High Priority Request List (HPRL), the 235U(n,${\\gamma}$) cross section is of utmost impo...

  15. Benchmarking ENDF/B-VII.0

    International Nuclear Information System (INIS)

    Marck, Steven C. van der

    2006-01-01

    The new major release VII.0 of the ENDF/B nuclear data library has been tested extensively using benchmark calculations. These were based upon MCNP-4C3 continuous-energy Monte Carlo neutronics simulations, together with nuclear data processed using the code NJOY. Three types of benchmarks were used, viz., criticality safety benchmarks (fusion) shielding benchmarks, and reference systems for which the effective delayed neutron fraction is reported. For criticality safety, more than 700 benchmarks from the International Handbook of Criticality Safety Benchmark Experiments were used. Benchmarks from all categories were used, ranging from low-enriched uranium, compound fuel, thermal spectrum ones (LEU-COMP-THERM), to mixed uranium-plutonium, metallic fuel, fast spectrum ones (MIX-MET-FAST). For fusion shielding many benchmarks were based on IAEA specifications for the Oktavian experiments (for Al, Co, Cr, Cu, LiF, Mn, Mo, Si, Ti, W, Zr), Fusion Neutronics Source in Japan (for Be, C, N, O, Fe, Pb), and Pulsed Sphere experiments at Lawrence Livermore National Laboratory (for 6 Li, 7 Li, Be, C, N, O, Mg, Al, Ti, Fe, Pb, D 2 O, H 2 O, concrete, polyethylene and teflon). For testing delayed neutron data more than thirty measurements in widely varying systems were used. Among these were measurements in the Tank Critical Assembly (TCA in Japan) and IPEN/MB-01 (Brazil), both with a thermal spectrum, and two cores in Masurca (France) and three cores in the Fast Critical Assembly (FCA, Japan), all with fast spectra. In criticality safety, many benchmarks were chosen from the category with a thermal spectrum, low-enriched uranium, compound fuel (LEU-COMP-THERM), because this is typical of most current-day reactors, and because these benchmarks were previously underpredicted by as much as 0.5% by most nuclear data libraries (such as ENDF/B-VI.8, JEFF-3.0). The calculated results presented here show that this underprediction is no longer there for ENDF/B-VII.0. The average over 257

  16. Analysis of the ZPPR-15 Critical Experiments with ENDF/B-V.2 and ENDF/B-VII.0 Data

    International Nuclear Information System (INIS)

    Kim, Sang Ji; Yang, Won Sik; Lee, Changho

    2008-01-01

    This paper presents the analysis results for the ZPPR-15 critical experiments. Using the ENDF/B-V.2 and ENDF/B-VII.0 data, three loading configurations of the ZPPR-15 Phase A experiments were analyzed with the ANL code suite for a fast reactor neutronics analysis, including the recently updated MC 2 -2 code. For the VIM Monte Carlo analyses with 3-D as-built models, the ENDF/B-VII.0 data improved the core multiplication factors by 0.21 to 0.37 %Δk, relative to the ENDF/B-V.2 data. With the plate heterogeneity effects taken into account by the SDX 1-D unit cell calculations, the DIF3D nodal transport solutions with the ENDF/B-V.2 data showed a good agreement for the core multiplication factors with the VIM Monte Carlo results to within 0.12 %Δk, but those with the ENDF/B-VII.0 data showed relatively larger deviations. Sensitivity studies based on the RZ models with homogenized cells showed excellent agreement for the core multiplication factors between the deterministic and Monte Carlo calculations to within 0.1 %Δk for both ENDF/B data. These results indicate that the MC 2 -2 methods are adequate for generating the multigroup cross sections for a fast reactor analysis, but the SDX process to account for the heterogeneity effect needs to be improved for the ENDF/B-VII.0 data. (authors)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  18. Monte Carlo analysis of TRX lattices with ENDF/B version 3 data

    International Nuclear Information System (INIS)

    Hardy, J. Jr.

    1975-01-01

    Four TRX water-moderated lattices of slightly enriched uranium rods have been reanalyzed with consistent ENDF/B Version 3 data by means of the full-range Monte Carlo program RECAP. The following measured lattice parameters were studied: ratio of epithermal-to-thermal 238 U capture, ratio of epithermal-to-thermal 235 U fissions, ration of 238 U captures to 235 U fissions, ratio of 238 U fissions to 235 U fissions, and multiplication factor. In addition to the base calculations, some studies were done to find sensitivity of the TRX lattice parameters to selected variations of cross section data. Finally, additional experimental evidence is afforded by effective 238 U capture integrals for isolated rods. Shielded capture integrals were calculated for 238 U metal and oxide rods. These are compared with other measurements

  19. PREPRO2007, Data Preparation and Management, Subsidiary Calculations (ENDF Format)

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of program or function: The following programs are all part of the PREPRO2007 package. ACTIVATE: is designed to create file 10 activation cross sections by combining file 3 cross sections and file 9 multipliers COMPLOT: Compares ENDF/B formatted data from two separate input files. Results are in graphical form. CONVERT: Automatically converts a FORTRAN program for use on any one of a variety of: (1) computers; (2) compilers; (3) precision; (4) installations; (5) standard or non-standard file names. DICTIN: Creates a reaction index for each material. EVALPLOT: Plots data in the ENDF/B format. FIXUP: Reads evaluated data in the ENDF/B format; performs corrections and outputs the results in the ENDF/B format. GROUPIE: Calculates unshielded group averaged cross sections, Bondarenko self-shielded group averaged cross sections, and multiband parameters from data in the ENDF/B format. LEGEND: Calculates linearly interpolable tabulated angular distributions starting from data in the ENDF/B format. LINEAR: Converts cross sections in the ENDF/B format (File 3, 23, and 27) to linearly interpolable form (in energy and cross section) and outputs the result in the ENDF/B format. MERGER: Selectively retrieves data by MAT/MF/MT or ZA/MF/MT from up to 10 ENDF/B data tapes and merges the data into a single MAT/MF/MT ordered output file. MIXER: Calculates the energy dependent cross sections for a composite mixture. RECENT: Reconstructs energy-dependent cross sections from a combination of resonance parameters and tabulated background cross sections in the ENDF/B format. RELABEL: Re-labels a ENDF/B preprocessing program so that statement labels are in increasing order in increments of 10 within each routine. SIGMA-1: Doppler broadens evaluated cross sections in the linear-linear interpolation form of the ENDF/B format. SIXPAK: Checks all double-differential ENDF/B-VI format data (MF=6) and outputs equivalent uncorrelated data (MF=4, 5, 12, 14, and 15) VIRGIN

  20. RESEND, Infinitely Dilute Point Cross-Sections Calculation from ENDF/B Resonance Parameter. ADLER, ENDF/B Adler-Adler Resonance Parameter to Point Cross-Sections with Doppler Broadening

    International Nuclear Information System (INIS)

    Bhat, M.R.; Ozer, O.

    1982-01-01

    1 - Description of problem or function: RESEND generates infinitely- dilute, un-broadened, point cross sections in the ENDF format by combining ENDF File 3 background cross sections with points calculated from ENDF File 2 resonance parameter data. ADLER calculates total, capture, and fission cross sections from the corresponding Adler-Adler parameters in the ENDF/B File 2 Version II data and also Doppler-broadens cross sections. 2 - Method of solution: RESEND calculations are done in two steps by two separate sections of the program. The first section does the resonance calculation and stores the results on a scratch file. The second section combines the data from the scratch file with background cross sections and prints the results. ADLER uses the Adler-Adler formalism. 3 - Restrictions on the complexity of the problem: RESEND expects its input to be a standard mode BCD ENDF file (Version II/III). Since the output is also a standard mode BCD ENDF file, the program is limited by the six significant figure accuracy inherent in the ENDF formats. (If the cross section has been calculated at two points so close in energy that only their least significant figures differ, that interval is assumed to have converged, even if other convergence criteria may not be satisfied.) In the unresolved range the cross sections have been averaged over a Porter-Thomas distribution. In some regions the calculated resonance cross sections may be negative. In such cases the standard convergence criterion would cause an unnecessarily large number of points to be produced in the region where the cross section becomes zero. For this reason an additional input convergence criterion (AVERR) may be used. If the absolute value of the cross section at both ends of an interval is determined to be less than AVERR then the interval is assumed to have converged. There are no limitations on the total number of points generated. The present ENDF (Version II/III) formats restrict the total number of

  1. Photonuclear reactions of U-233 and Pu-239 near threshold induced by thermal neutron capture gamma rays

    International Nuclear Information System (INIS)

    Moraes, M.A.P. de.

    1990-01-01

    The photonuclear cross sections of U-293 and Pu-239 have been studied by using monochromatic and discrete photons, in the energy interval from 5.49 to 9.72 MeV, produced by thermal neutron capture. The gamma fluxes incident on the samples were measured using a ( 3 x 3 )'' NaI (TI) crystal. The photofission fragments were detected in Makrofol-Kg (SSNTD). A possible structure was observed in the U-233 cross sections, near 7.23 MeV. The relative fissionability of the nuclides was determined at each excitation energy and shown to be energy independent: ( 2.12 ± 0.25) for U-233 and ( 3.32 ± 0.41 ) for Pu-239. The angular distribution of photofission fragments of Pu-239 were measured at two mean excitation energies of 5.43 and 7.35 MeV. An anisotropic distribution of ( 12.2 ± 3.6 ) % was observed at 5.43 MeV. The total neutron cross sections were measured by using a long counter detector. The photoneutron cross sections were calculated by using energy dependent neutron multiplicities values, γ(E), obtained in the literature. The competition Γn/γf was also determined at each excitation energy, and shown to be energy independent: ( 0.54 ± 0.05 ) for U-233 and ( 0.44 ± 0.05 ) for Pu-239, and were correlated to the parameters Z sup(2)/A, ( Ef'-Bn'), A. According to the FUJIMOTO-YAMAGUCHI and CONSTANT NUCLEAR TEMPERATURE models, the nuclear temperatures were calculated. The total photoabsorption cross sections were also calculated as a sum of the photofission and photoneutron cross sections at each energy excitation. From these results the competition Γf/ΓA, called fission probability Pf, were obtained: ( 0.66 ± 0.02) for U-233 and ( 0.70 ± 0.02 ) for Pu-239. (author)

  2. On the ENDF Formats and Data Processing

    International Nuclear Information System (INIS)

    Trkov, Andrej

    2012-01-01

    The ENDF formats have served the community of nuclear data users from different fields of applications quite well for decades. Enormous effort has been devoted to the development and validation of the processing codes. Although there is no urgent need for a rapid transition to something completely new, there are signs that the current ENDF format is being pushed close to its limits. The time is right to look for a modern replacement, with due consideration for the following: - Development of data processing capabilities, starting from the data in the new format. - Backward compatibility through robust translation codes between the new and the old format until the majority of processing tools have been adequately validated. - Standardisation of the format features on the international level to maintain the possibility of easy data comparison and exchange. The NJOY Data Processing System is the most versatile and widely used code system for generating application libraries. The AMPX system is mainly used for generating libraries for codes from Oak Ridge. The Pre-Pro codes are found to be very robust, but their main purpose is data verification, validation and display. These codes do a good job for the present scope of applications, but current trend rely heavily on Monte Carlo simulations and sensitivity/uncertainty calculations. Further developments in the data processing tools should reflect these trends, focusing on the following: - Further verification and validation of covariance processing methods. - Development of a common tool for generating a global covariance matrix of nuclear data, including all available cross-reaction and cross-material correlations. - Consider if we can move away from histogram covariance representation into a piecewise linear domain. - Having a 'global' covariance matrix (that can include the covariance matrix of the resonance parameters), pursue the development of a common tool for statistical sampling of the cross sections and other

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-01

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

  4. RELABEL2007, Labels FORTRAN Statements in ENDF Format Processing Programs

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of program or function: RELABEL labels a ENDF/B pre-processing program so that statement labels are in increasing order in increments of 10 within each routine, and cards are identified in columns 73-80 by three alphanumeric characters in columns 73-75 and sequence numbers in columns 76-80 in increments of 10. IAEA1314/10: This version include the updates up to January 30, 2007. Changes in ENDF/B-VII Format and procedures, as well as the evaluations themselves, make it impossible for versions of the ENDF/B pre-processing codes earlier than PREPRO 2007 (2007 Version) to accurately process current ENDF/B-VII evaluations. The present code can handle all existing ENDF/B-VI evaluations through release 8, which will be the last release of ENDF/B-VI. Modifications from previous versions: Relabel VERS. 2007-1 (JAN. 2007): No change since March 2004 version 2 - Method of solution: 3 - Restrictions on the complexity of the problem: RELABEL is designed to maintain ENDF/B processing programs which use a restricted set of FORTRAN statements. As such, this program is not completely general

  5. Neutron cross section standards evaluations for ENDF/B-VI

    International Nuclear Information System (INIS)

    Carlson, A.D.; Poenitz, W.P.; Hale, G.M.; Peelle, R.W.

    1985-01-01

    The neutron cross section standards are now being evaluated as the initial phase in the development of the new ENDF/B-VI file. These standards evaluations are following a somewhat different process compared with that used for earlier versions of ENDF. The primary effort is concentrated on a simultaneous evaluation using a generalized least squares program, R-matrix evaluations, and a procedure for combining the results of these evaluations. The ENDF/B-VI standards evaluation procedure is outlined, and preliminary simultaneous evaluation and R-matrix results are presented. 16 refs., 7 figs

  6. 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)

  7. Neutron-capture gamma-ray study of levels in 135Ba and description of nuclear levels in the interacting-boson-fermion model

    International Nuclear Information System (INIS)

    Chrien, R.E.; Koene, B.K.S.; Stelts, M.L.; Meyer, R.A.; Brant, S.; Paar, V.; Lopac, V.

    1993-01-01

    We have performed neutron-capture gamma-ray studies on natural and enriched targets of 134 Ba in order to investigate the nuclear levels of 135 Ba. The low-energy level spectra were compared with the calculations using the interacting-boson-fermion model (IBFM) and the cluster-vibration model. The level densities up to 5 MeV that are calculated within the IBFM are in accordance with the constant temperature Fermi gas model. From the spin distribution we have determined the corresponding spin cutoff parameter σ and compared it to the prediction from nuclear systematics

  8. CAREN-4, ENDF/B Utility, Discontinuity Check at Resonance Region Boundary. CHECK-4, ENDF/B Utility, Structure Consistency Check and Format Check. CRECT, ENDF/B Utility, Data Correlation and Data Update. DICT-4, ENDF/B Utility, Section Table of Contents Generator. LISTF-4, ENDF/B Utility, Data Listing. RIGEL-4, ENDF/B Utility, Data Retrieval, BCD to BIN Conversion. SUMUP-4, ENDF/B Utility, Partial Cross-Sections Sum Check Against Tot Cross-Sections

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1975-08-01

    Description of problem or function: These programs are updated versions of ENDF/B checking, retrieval, and display codes capable of processing the Version-IV data files. RIGEL4 retrieves ENDF/B data and changes mode (BCD-BIN) or arrangement. Updated version processes decay data (mf=1, mt=457) and error files. CHECK4 checks the structure, consistency, and formats of ENDF data files. Updated version recognizes newly defined mt and mf numbers, checks mt=457 and the formats of the error files, contains RSIC photon file changes. SUMUP4 checks whether partial cross sections add up to the total. Updated version flags grid points present in the partial cross section, absent in the total. LISTF4 produces interpreted listings of ENDF/B data. Updated version lists mt=457, skips over error files, contains minor corrections. PLOTF4 produces plots of ENDF/B data. Updated version plots mt=457, skips over error files, contains minor corrections. RESEND processes ENDF materials with resonance parameters into a pointwise form. Capability of processing Adler-Adler parameters added. CRECT corrects ENDF/B data files. DICT4 generates a section table of contents (dictionary) for ENDF/B materials. CAREN4 tests for discontinuities across the limits of resonance ranges of an ENDF/B material. Updated version contains minor corrections.

  9. ENDF/B-5. Fission Product Yields File

    International Nuclear Information System (INIS)

    Schwerer, O.

    1985-10-01

    The ENDF/B-5 Fission Product Yields File contains a complete set of independent and cumulative fission product yields, representing the final data from ENDF/B-5 as received at the IAEA Nuclear Data Section in June 1985. Yields for 11 fissioning nuclides at one or more neutron incident energies are included. The data are available costfree on magnetic tape from the IAEA Nuclear Data Section. (author). 4 refs

  10. Program RECENT (version 79-1): reconstruction of energy-dependent neutron cross sections from resonance parameters in the ENDF/B format

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1979-01-01

    Program RECENT reconstructs energy-dependent neutron total, elastic, capture, and fission cross sections from a combination of resonance parameters and tabulated background cross sections in the ENDF/B format. Entire evaluations, not just cross sections, are written to the result file, which is in ENDF/B format. The output includes the original resonance parameters in a form that can be used in Doppler broadening and self-shielding calculations. A listing of the source deck is available on request. 5 figures, 5 tables

  11. ENDF-102 data formats and procedures for the evaluated nuclear data file ENDF-6. Revision November 1995

    International Nuclear Information System (INIS)

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

    1995-11-01

    The ENDF formats and libraries are decided by the Cross Section Evaluation Working Group (CSEWG), a cooperative effort of national laboratories, industry, and universities in the US and Canada, and are maintained by the National Nuclear Data Center (NNDC). Earlier versions of the ENDF format provided representations for neutron cross sections and distributions, photon production from neutron reactions, a limited amount of charged-particle production from neutron reactions, photo-atomic interaction data, thermal neutron scattering data, and radionuclide production and decay data (including fission products). Version 6 (ENDF-6) allows higher incident energies, adds more complete descriptions of the distributions of emitted particles, and provides for incident charged particles and photo-nuclear data by partitioning the ENDF library into sublibraries. Decay data, fission product yield data, thermal scattering data, and photo-atomic data have also been formally placed in sublibraries. In addition, this rewrite represents an extensive update to the Version V manual

  12. Data formats and procedures for the evaluated nuclear data format. ENDF-IV. Reprint of the report BNL-NCS-50496 (ENDF-102), revised

    Energy Technology Data Exchange (ETDEWEB)

    Garber, D; Dunford, C; Pearlstein, S

    1984-05-01

    These revisions to Data Formats and Procedures for the ENDF Neutron Cross Section Library, ENDF-102, pertain to the latest version of ENDF/B-IV.The descriptions of the formats have been brought up to date and important procedural matters have been explained. Three new appendices have been added

  13. Data formats and procedures for the evaluated nuclear data format. ENDF-IV. Reprint of the report BNL-NCS-50496 (ENDF-102), revised

    International Nuclear Information System (INIS)

    Garber, D.; Dunford, C.; Pearlstein, S.

    1984-01-01

    These revisions to Data Formats and Procedures for the ENDF Neutron Cross Section Library, ENDF-102, pertain to the latest version of ENDF/B-IV.The descriptions of the formats have been brought up to date and important procedural matters have been explained. Three new appendices have been added

  14. Detection efficiency for radionuclides decaying by electron capture and gamma-Ray; Calculo de la eficiencia de deteccion de nucleidos que se desintegran por captura elec- tronica y emision gamma

    Energy Technology Data Exchange (ETDEWEB)

    Grau, A; Fernandez, A

    1985-07-01

    In this paper, the electron capture partial counting efficiency vs the figure of merit for electron-capture and gamma-ray emitters has been computed. The radionuclides tabulated are 48{sup c}r, 54{sup M}n, 57{sup C}o 56{sup N}i, 72{sup S}e, 73{sup A}s, 85{sup S}r, 88{sup Z}r, 92{sup N}b, 103{sup P}d, 111{sup l}n, 119{sup S}b, 125{sup I}, 139{sup C}e and 152{sup D}y. It has been assumed that the liquid is a toluene based scintillator solution in standard glass vials containing 15 cm{sup 3}. (Author) 17 refs.

  15. Update of WIMS-D libraries using JENDL-3.2, ENDF/B-VI.5 and JEF-2.2

    International Nuclear Information System (INIS)

    Gil, Choong Sup; Kim, Jung Do; Chang, Jong Wha

    2001-01-01

    The WIMS-D5 Libraries based on JENDL-3.2, ENDF/B-VI.5, and JEF-2.2 have been prepared and are being tested against the benchmark problems. Several sensitivity calculations for stabililty confirmation of the libraries were carried out such as the fission spectrum dependency, the self shielding effects of the elastic scattering cross sections, the self shielding effects of Pu -240 and Pu -242 capture cross sections below 4.0 eV, etc. The results of benchmark calculations with the libraries based on JENDL-3.2, ENDF/B-VI.5, JEF-2.2, and the '1986 library were intercompared. The predictions of criticalities and isotopic compositions with the updated libraries show good agreements with the measurements or the reference results. The multiplication factors with the library based on JENDL-3.2 are slightly higher than those of ENDF/B-VI.5 and JEF-2.2

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  17. Gamma ray generator

    Science.gov (United States)

    Firestone, Richard B; Reijonen, Jani

    2014-05-27

    An embodiment of a gamma ray generator includes a neutron generator and a moderator. The moderator is coupled to the neutron generator. The moderator includes a neutron capture material. In operation, the neutron generator produces neutrons and the neutron capture material captures at least some of the neutrons to produces gamma rays. An application of the gamma ray generator is as a source of gamma rays for calibration of gamma ray detectors.

  18. Prediction of the Sodium Void Reactivity in the Metal-fueled SFR Using the ENDF/B-VII.0 Library

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Lim, Jae-Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The SVR (Sodium Void Reactivity) is one of the most important parameters in SFR (Sodium-cooled Fast Reactor) safety analysis. In this paper, to estimate the error of the SVR in metal-fueled SFR, three physics experiments named as BFS-75-1, BFS-109-2A, and BFS-84-1 were examined using recent cross-section library, ENDF/B-VII.0 and the MCNP code. In the MCNP6 calculation, two million histories/generation with 50 inactive/300 active generations are used with the continuous-energy ENDF/B-VII.0 library. We expect that accuracy of total cross-section of the sodium may play a dominant role in errors of SVRs at core peripheral and sodium plenum regions, whereas accuracy of capture cross-section of the sodium may play a dominant role for the results in errors of SVRs at core central region. In addition, capture cross-sections of the sodium in the ENDF/B-VII.0, the JEFF-3.2, and the JENDL-4.0 libraries show significant differences between each other, while total cross-sections of sodium in three libraries show good agreement.

  19. Dose Determination using alanine detectors in a Mixed Neutron and Gamma Field for Boron Neutron Capture Therapy of Liver Malignancies

    DEFF Research Database (Denmark)

    Schmitz, T.; Blaickner, M.; Ziegner, M.

    2011-01-01

    Introduction Boron Neutron Capture Therapy for liver malignancies is being investigated at the University of Mainz. One important aim is the set-up of a reliable dosimetry system. Alanine dosimeters have previously been applied for dosimetry of mixed radiation fields in antiproton therapy, and ma...

  20. ENDF/B-IV representation of the 238U total neutron cross section in the resolved resonance energy region

    International Nuclear Information System (INIS)

    de Saussure, G.; Olsen, D.K.; Perez, R.B.

    1976-01-01

    The ENDF/B-IV prescription fails to represent correctly the 238 U total (and scattering) cross section between the levels of the resolved range. It is shown how this representation can be improved by properly accounting for the contribution of levels outside the resolved region to the cross section at energies inside the resolved region, and by substituting the more precise multilevel Breit-Wigner formula for the presently used single-level formula. The importance of computing accurately the minima in the total cross section is illustrated by comparing values of the self-shielded capture resonance integral computed with ENDF/B-IV and with a more accurate cross section model

  1. ENDF-UTILITY-CODES, codes to check and standardize data in the Evaluated Nuclear Data File (ENDF)

    International Nuclear Information System (INIS)

    Dunford, Charles L.

    2007-01-01

    1 - Description of program or function: The ENDF Utility Codes include 9 codes to check and standardize data in the Evaluated Nuclear Data File (ENDF). Four programs of this release, GETMAT, LISTEF, PLOTEF and SETMDC are no more maintained since release 6.13. The suite of ENDF utility codes includes: - CHECKR (version 7.01) is a program for checking that an evaluated data file conforms to the ENDF format. - FIZCON (version 7.02) is a program for checking that an evaluated data file has valid data and conforms to recommended procedures. - GETMAT (version 6.13) is designed to retrieve one or more materials from an ENDF formatted data file. The output will contain only the selected materials. - INTER (version 7.01) calculates thermal cross sections, g-factors, resonance integrals, fission spectrum averaged cross sections and 14.0 MeV (or other energy) cross sections for major reactions in an ENDF-6 or ENDF-5 format data file. - LISTEF (version 6.13) is designed to produce summary and annotated listings of a data file in either ENDF-6 or ENDF-5 format. - PLOTEF (version 6.13) is designed to produce graphical displays of a data file in either ENDF-5 or ENDF-6 format. The form of graphical output depends on the graphical devices available at the installation where this code will be used. - PSYCHE (version 7.02) is a program for checking the physics content of an evaluated data file. It can recognise the difference between ENDF-5 or ENDF-6 formats and performs its tests accordingly. - SETMDC (version 6.13) is a utility program that converts the source decks of programs to different computers (DOS, UNIX, LINUX, VMS, Windows). - STANEF (version 7.01) performs bookkeeping operations on a data file containing one or more material evaluations in ENDF format. The version 7.02 of the ENDF Utility Codes corrects all bugs reported to NNDC as of April 1, 2005 and supersedes all previous releases. Three codes CHECKR, STANEF, and INTER were actually ported from the 7.01 release

  2. 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

  3. Evaluated Nuclear Data Covariances: The Journey From ENDF/B-VII.0 to ENDF/B-VII.1

    International Nuclear Information System (INIS)

    Smith, Donald L.

    2011-01-01

    Recent interest from data users on applications that utilize the uncertainties of evaluated nuclear reaction data has stimulated the data evaluation community to focus on producing covariance data to a far greater extent than ever before. Although some uncertainty information has been available in the ENDF/B libraries since the 1970's, this content has been fairly limited in scope, the quality quite variable, and the use of covariance data confined to only a few application areas. Today, covariance data are more widely and extensively utilized than ever before in neutron dosimetry, in advanced fission reactor design studies, in nuclear criticality safety assessments, in national security applications, and even in certain fusion energy applications. The main problem that now faces the ENDF/B evaluator community is that of providing covariances that are adequate both in quantity and quality to meet the requirements of contemporary nuclear data users in a timely manner. In broad terms, the approach pursued during the past several years has been to purge any legacy covariance information contained in ENDF/B-VI.8 that was judged to be subpar, to include in ENDF/B-VII.0 (released in 2006) only those covariance data deemed then to be of reasonable quality for contemporary applications, and to subsequently devote as much effort as the available time and resources allowed to producing additional covariance data of suitable scope and quality for inclusion in ENDF/B-VII.1. Considerable attention has also been devoted during the five years since the release of ENDF/B-VII.0 to examining and improving the methods used to produce covariance data from thermal energies up to the highest energies addressed in the ENDF/B library, to processing these data in a robust fashion so that they can be utilized readily in contemporary nuclear applications, and to developing convenient covariance data visualization capabilities. Other papers included in this issue discuss in considerable

  4. Presentation of a semiempirical method for the calculation of doses due to neutrons and capture gamma rays inside high energy accelerators rooms

    International Nuclear Information System (INIS)

    Larcher, A.M.; Bonet Duran, S.M.

    1998-01-01

    Full text: Medical electron accelerators operating above 10 MeV produce radiation beams that are contaminated with neutrons. Therefore, shielding design for high energy accelerator rooms must consider the neutron component of the radiation field. In this paper a semiempirical method is presented to calculate doses due to neutrons and capture gamma rays inside the room and the maze. The calculation method is based on the knowledge of the neutron yield Q (neutrons/Gy of photons at isocenter) and the average energy of the primary beam of neutrons Eo (MeV). The method constitutes an appropriate tool for shielding facilities evaluation. The accuracy of the method has been contrasted with data obtained from the literature and an excellent correlation among the calculations and the measured values was achieved. In addition, the method has been used in the verification of experimental data corresponding to a 15 MeV linear accelerator installed in the country with similar results. (author) [es

  5. ENDF/B yield evaluation for 1992: Methods and content

    International Nuclear Information System (INIS)

    England, T.R.; Rider, B.F.

    1992-01-01

    The basic evaluation process, completed in May 1992, for 60 independent, plus corresponding cumulative yield sets is described thirty-six fissioning nuclides at one-or-more neutron fission energies or spontaneous fission are included. The resulting recommended yields include approximately 1200 nuclides per set; these will be slightly extended to encompass all nuclides in the ENDF/B-VI decay files and issued as the second release of ENDF/B-VI yields. All current yield sets in ENDF/B-VI have been reevaluated using ∼3000 new measurements and model parameters for distribution along mass chains. Compiled measurements through 1992 will be included in the documentation of the recommended yields. This paper can only summarize the primary features of the evaluations

  6. New fission-neutron-spectrum representation for ENDF

    International Nuclear Information System (INIS)

    Madland, D.G.

    1982-04-01

    A new representation of the prompt fission neutron spectrum is proposed for use in the Evaluated Nuclear Data File (ENDF). The proposal is made because a new theory exists by which the spectrum can be accurately predicted as a function of the fissioning nucleus and its excitation energy. Thus, prompt fission neutron spectra can be calculated for cases where no measurements exist or where measurements are not possible. The mathematical formalism necessary for application of the new theory within ENDF is presented and discussed for neutron-induced fission and spontaneous fission. In the case of neutron-induced fission, expressions are given for the first-chance, second-chance, third-chance, and fourth-chance fission components of the spectrum together with that for the total spectrum. An ENDF format is proposed for the new fission spectrum representation, and an example of the use of the format is given

  7. INDXENDF, Preparation of Visual Catalogue of ENDF Format Data

    International Nuclear Information System (INIS)

    Silva, Orion de O.; Paviotti Corcuera, R.; De Moraes Cunha, M.; Ferreira, P.A.

    1996-01-01

    1 - Description of program or function: This program is a video catalogue for libraries in the ENDF-4, ENDF-5 or ENDF-6 format (Evaluated Nuclear Data File) which can be run on an IBM-PC or compatible computer. This user friendly catalogue is of interest to nuclear and reactor physics researchers. The input is the filename of ENDF data and the two output files contain: i. the list of materials with corresponding laboratory, author and date of evaluation; ii. information about the MF and MT numbers for each material. The program is written in the C language whose capability of providing windows and 'interrupts' along with speed and portability, has been greatly exploited. The system allows output of options (i) and (ii) either on screen, printer or hard disk. 2 - Method of solution: The source code of about 3000 lines was written in C. The routines for windowing were based on the following works: Hummel (1988), Stevens (1989), Lafore (1987), Borland International (1988a, 1988b) and Schildt (1988, 1989). 3 - Restrictions on the complexity of the problem: The executable program occupies about 52 Kb of memory. The extra hard disk space needed depends upon the size of the ENDF/B data file to be processed (e.g. the Activation file contains about 1.3 M-bytes, the General Purpose ENDF/B-VI has four parts, each containing about 12 M-bytes). To run the program the 'datafile' and the executable code '.EXE' file should be on the hard-drive. The program may be run on any IBM/PC or compatible with at least 640 Kb RAM

  8. Control of radioactive wastes and coupling of neutron/gamma measurements: use of radiative capture for the correction of matrix effects that penalize the fissile mass measurement by active neutron interrogation; Controle des dechets radioactifs et couplage de mesures neutron/gamma: exploitation de la capture radiative pour corriger les effets de matrice penalisant la mesure de la masse fissile par interrogation neutronique active

    Energy Technology Data Exchange (ETDEWEB)

    Loche, F

    2006-10-15

    In the framework of radioactive waste drums control, difficulties arise in the nondestructive measurement of fissile mass ({sup 235}U, {sup 239}Pu..) by Active Neutron Interrogation (ANI), when dealing with matrices containing materials (Cl, H...) influencing the neutron flux. The idea is to use the neutron capture reaction (n,{gamma}) to determine the matrix composition to adjust the ANI calibration coefficient value. This study, dealing with 118 litres, homogeneous drums of density less than 0,4 and composed of chlorinated and/or hydrogenated materials, leads to build abacus linking the {gamma} ray peak areas to the ANI calibration coefficient. Validation assays of these abacus show a very good agreement between the corrected and true fissile masses for hydrogenated matrices (max. relative standard deviation: 23 %) and quite good for chlorinated and hydrogenated matrices (58 %). The developed correction method improves the measured values. It may be extended to 0,45 density, heterogeneous drums. (author)

  9. ENDF/B-6 Photon Atomic Interaction Data Library

    International Nuclear Information System (INIS)

    Lemmel, H.D.

    1990-09-01

    The ENDF/B-6 version of the Photo-Atomic Interaction Data Library of the Livermore Evaluated Photon Data Library (EPDL) contains pair and triplet cross-sections, photoelectric cross-sections, atom form factors, coherent scattering cross-sections and some other data for all the elements from Z=1 to 100. The data library is available on magnetic tape costfree from the IAEA Nuclear Data Section. The library supersedes the earlier photo-atomic data library by the US Radiation Shielding Information Center RSIC that was included in the data libraries ENDF/B-5 and JEF-1. (author). Refs, figs and tabs

  10. ENDFIC - A program for indexing and intercomparison of ENDFs

    International Nuclear Information System (INIS)

    Gopalakrishnan, V.; Devan, K.

    1995-01-01

    The program ENDFIC, a nuclear data utility program in FORTRAN, was written under contract on 'Indexing and Intercomparison Programme of Evaluated Nuclear Data Files' (No. 7866/RB/TC), between Indira Gandhi Centre for Atomic Research, Kalpakkam, and the International Atomic Energy Agency, Vienna. The program can be used for the following two activities: i. INFO activity: to find from a given ENDF/B formatted evaluated nuclear data library, information regarding the contents of the library; ii. COMP activity: to do comparison of cross sections from several given ENDF libraries. 1 ref

  11. Evaluated nuclear data file ENDF/B-VI

    International Nuclear Information System (INIS)

    Dunford, C.L.

    1991-01-01

    For the past 25 years, the United States Department of Energy has sponsored a cooperative program among its laboratories, contractors and university research programs to produce an evaluated nuclear data library which would be application independent and universally accepted. The product of this cooperative activity is the ENDF/B evaluated nuclear data file. After approximately eight years of development, a new version of the data file, ENDF/B-VI has been released. The essential features of this evaluated data library are described in this paper. 7 refs

  12. New ENDF/B-7.0 library

    International Nuclear Information System (INIS)

    Oblozinsky, P.

    2008-01-01

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

  13. Analysis of benchmark lattices with endf/b-vi, jef-2.2 and jendl-3 data

    International Nuclear Information System (INIS)

    Saglam, M.

    1995-01-01

    The NJOY Nuclear Data Processing System has been used to process the ENDF/B-VI , JEF-2.2 and JENDL-3 Nuclear Cross Section Data Bases into multigroup form. A brief description of the data bases is given and the assumptions made in processing the data from evaluated nuclear data file format to multigroup format are presented. The differences and similarities of the Evaluated Nuclear Data Files have been investigated by producing four group cross sections by using the GROUPIE code and calculating thermal, fission spectrum averaged and 2200 m/s cross sections and resonance integrals using the INTER cale. It has been shown that the evaluated data for U238 in JEF and ENDF/B-VI are principally the same while in case of U235 the same is true for JENDL and ENDF/B-VI. The evaluations for U233 and Th232 are different for all three ENDF files. Several utility codes have been written to convert the multigroup library into a WIMS-D4 compatible binary library. The performance and suitability of the generated libraries have been tested with the use of metal tueled TRX lattices, uranium oxide fueled BAPL lattices and Th232-U233 fueled BNL lattices. The use ot a new thermal scattering matrix for Hydrogen from ENDF/B-VI increased keff for 0.5 o/ while the use of ENDF/B-VI U238 decreased it for 2.5 %. Although the original WIMS library performed well for Ihe effective multiplication factor of the lattices there is an improvement for the epithermal to thermal capture rate of U238 while using new data in the TRX and BAPL lattices. The effect of the fission spectrum is investigated for the BNL lattices and it is shown that using U233 fission spectrum instead of the original U235 spectrum gives a keff which agrees better with the experimental value. The results obtained by using new multigroup data are generally acceptable and in the experimental error range. They especially improve the prediction of the reaction rate dependent benchmark parameters

  14. Data formats and procedures for the Evaluated Nuclear Data File, ENDF

    International Nuclear Information System (INIS)

    Kinsey, R.

    1979-10-01

    These revisions to Data Formats and Procedures for the ENDF Neutron Cross Section Library, ENDF-102, pertain to the latest version of ENDF/B-V. The descriptions of the formats are brought up to date, and important procedural matters are explained

  15. Reactor benchmarks and integral data testing and feedback into ENDF/B-VI

    International Nuclear Information System (INIS)

    McKnight, R.D.; Williams, M.L.

    1992-01-01

    The role of integral data testing and its feedback into the ENDF/B evaluated nuclear data files are reviewed. The use of the CSEWG reactor benchmarks in the data testing process is discussed and selected results based on ENDF/B Version VI data are presented. Finally, recommendations are given to improve the implementation in future integral data testing of ENDF/B

  16. Polarized proton capture reaction /sup 7/Li(p,. gamma. )/sup 8/Be in the energy range from 380 to 960 keV

    Energy Technology Data Exchange (ETDEWEB)

    Ulbricht, J; Arnold, W; Berg, H; Huttel, E; Krause, H H; Clausnitzer, G [Giessen Univ. (Germany, F.R.). Abt. Grossgeraete (Angewandte Kernphysik)

    1977-09-05

    The polarized proton capture in /sup 7/Li was used to study the reaction mechanism and to obtain spectroscopic information on the /sup 8/Be nucleus. Gamma-ray angular distributions of the analyzing power were measured as a function of proton energy from Esub(p) = 380-960 keV with three Ge(Li) detectors simultaneously. The excitation functions of the cross section and the analyzing power are strongly energy dependent. The data were analyzed unambiguously and represented by three R-matrix elements, two M1 and one E1. The energy dependence of the two M1 matrix elements agrees with the well-known two 1/sup +/ resonances at Esub(x) = 17.642 and 18.157 MeV. The energy dependence of the E1 matrix element shows a smooth background presumably caused by a direct-capture mechanism, and furthermore, a resonant contribution, which is a significant suggestion of a new 1/sup -/ state in the /sup 8/Be system at Esub(x) = 17.70 MeV with a width of GAMMAsub(p) = 180 keV.

  17. Description of evaluations for 54,56,57,58Fe performed for ENDF/B-VI

    International Nuclear Information System (INIS)

    Larson, D.C.; Fu, C.Y.; Hetrick, D.M.

    1989-01-01

    Isotopic evaluations for 54,56,57,58 Fe performed for ENDF/B-VI are briefly reviewed. The evaluations are based on analysis of experimental data and results of model calculations which reproduce the experimental data. Evaluated data are given for neutron induced reaction cross sections, angular and energy distributions, and for gamma-ray production cross sections associated with the reactions. File 6 formats are used to represent energy-angle correlated data and recoil spectra. Uncertainty files are included for the major cross sections. A detailed evaluation is given for 56 Fe and results of calculations for the major reactions are used for evaluations of the minor isotopes, with particular attention paid to inelastic scattering to the low-lying levels in 57 Fe. (author). 26 refs, 5 figs

  18. ENDF/B-VI evaluations for isotopes of Cr, Fe, Ni, Cu, and Pb

    International Nuclear Information System (INIS)

    Hetrick, D.M.; Fu, C.Y.; Larson, D.C.

    1989-01-01

    Evaluations have been done for each of the stable isotopes of chromium, iron, nickel, copper, and lead. They are based on analysis of experimental data and results of nuclear model calculations which reproduce the experimental data. Evaluated data are given for neutron induced reaction cross sections, angular and energy distributions, and gamma-ray production cross sections associated with the reactions. The new file 6 formats are used to represent energy-angle correlated data and recoil spectra for the first time in ENDF. This paper reviews the structure of the evaluations, notes the major pieces of experimental data utilized, gives a summary of the model codes used, and compares calculations to measured data

  19. Evaluation of 28,29,30Si neutron induced cross sections for ENDF/B-VI

    International Nuclear Information System (INIS)

    Hetrick, D.M.; Larson, D.C.; Larson, N.M.; Leal, L.C.; Epperson, S.J.

    1997-04-01

    Separate evaluations have been done for the three stable isotopes of silicon for ENDF/B-VI. The evaluations are based on analysis of experimental data, supplemented by results of nuclear model calculations. The computational methods and the parameters required as input to the nuclear model codes are reviewed. Discussion of the evaluated data given for resonance parameters, neutron induced reaction cross sections, associated angular and energy distributions, and gamma-ray production cross sections is included. Extensive comparisons of the evaluated cross sections to measured data are shown in this report. The evaluations include all necessary data to allow KERMA (Kinetic Energy Released in MAterials) and displacement cross sections to be calculated directly. These quantities are fundamental to studies of neutron heating and radiation damage

  20. Description of evaluations for 50,52,53,54Cr performed for ENDF/B-VI

    International Nuclear Information System (INIS)

    Larson, D.C.; Hetrick, D.M.; Fu, C.Y.

    1989-01-01

    Isotopic evaluations for 50,52,53,54 Cr performed for ENDF/B-VI are briefly reviewed. The evaluations are based on analysis of experimental data and results of model calculations which reproduce the experimental data. Evaluated data are given for neutron induced reaction cross sections, angular and energy distributions, and for gamma-ray production cross sections associated with the reactions. File 6 formats are used to represent energy-angle correlated data and recoil spectra. Uncertainty files are included for the major cross sections. Detailed evaluations are given for 52,53 Cr, and results of calculations for reactions with large cross sections are used for evaluation of the minor isotopes. (author). 27 refs, 6 figs

  1. A new gamma-ray diagnostic for energetic ion distributions - The Compton tail on the neutron capture line

    International Nuclear Information System (INIS)

    Vestrand, W.T.

    1990-01-01

    This paper presents a new radiation diagnostic for assaying the energy spectrum and the angular distribution of energetic ions incident on thick hydrogen-rich thermal targets. This diagnostic compares the number of emergent photons in the narrow neutron capture line at 2.223 MeV to the number of Compton scattered photons that form a low-energy tail on the line. It is shown that the relative strength of the tail can be used as a measure of the hardness of the incident ion-energy spectrum. Application of this diagnostic to solar flare conditions is the main thrust of the work presented here. It is examined how the strength of the Compton tail varies with flare viewing angle and the angular distribution of the flare-accelerated particles. Application to compact X-ray binary systems is also briefly discussed. 39 refs

  2. Neutron capture cross section measurements of 109Ag, 186W and 158Gd on filtered neutron beams of 55 and 144 keV

    International Nuclear Information System (INIS)

    Vuong Huu Tan; Nguyen Canh Hai; Pham Ngoc Son; Tran Tuan Anh

    2004-12-01

    The neutron capture cross sections of the 109 Ag(n, γ) 110 mAg, 186 W(n, γ) 187 W and 158 Gd(n, γ) 159 Gd have been measured at 55 and 144 keV by the activation method with filtered neutron beams of the Dalat nuclear research reactor. The cross sections were determined relative to the standard capture cross sections of 197 Au using highly purity metallic foils of Ag, W, Gd and Au. The high efficient HPGe detector was used for the gamma rays measurement from the samples, and absolute efficiency calibration was performed by using a set of standard radioisotope sources and a multi-nuclides standard solution. The present results were compared with the previous measurements listed in EXFOR-CINDA, and the evaluated data of ENDF/B-VI. (author)

  3. UKE, Format Conversion from UKNDL to ENDF/B

    International Nuclear Information System (INIS)

    1973-01-01

    1 - Nature of physical problem solved: UKE reads a card image tape of data in the UK format and translates neutron cross sections, angular distributions, and secondary energy distributions to the ENDF/B card image format. 2 - Restrictions on the complexity of the problem: Maximum number of cross section data points allowed for each reaction type is 4000

  4. ENDF/B-6 fission-product yield sublibraries

    International Nuclear Information System (INIS)

    Lemmel, H.D.

    1994-01-01

    The contents and the documentation of the ENDF/B-6 fission-product yield sublibraries which were released in 1991 and updated in 1993, are summarized. Copies of the data libraries are available on magnetic tape of PC diskettes from the IAEA Nuclear Data Section, costfree upon request. (author). 1 tab

  5. ENDF/B-V utility programs: description and operating instructions

    International Nuclear Information System (INIS)

    McLaughlin, K.

    1984-03-01

    A description and operating instructions are supplied for the following ENDF/B-V Processing Programs: CHECKER, CRECT, STNDRD, FIZCON, PSYCHE, RESEND, INTER, INTEND, SUMRIZ, PLOTEF, LSTFCV, RIGEL. These programs can be obtained on magnetic tape, free of charge, from the IAEA Nuclear Data Section. (author)

  6. Correcting the effects of the matrix using capture gamma-ray spectrometry: Application to measurement by Active Neutron Interrogation; Correction des effets de matrice par spectrometrie des rayonnements gamma de capture: Application a la mesure par Interrogation Neutronique Active (I.N.A.)

    Energy Technology Data Exchange (ETDEWEB)

    Baudry, G.

    2003-11-15

    In the field of the measurement of low masses of fissile material ({sup 235}U, {sup 239}Pu, {sup 241}Pu) in radioactive waste drums, the Active Neutron Interrogation is a non-destructive method achieving good results. It does however remain reliant upon uncertainties related to the matrix effects on interrogation and fission neutrons. The aim of this thesis is to develop a correction method able to take into account these matrix effects by quantifying the amount of absorbent materials (chlorine and hydrogen) in a 118- liter homogeneous matrix. The main idea is to use the gamma-ray spectrometry of gamma emitted by neutron captures to identify and quantify the composition of the matrix. An indicator from its chlorine content is then deduced in order to choose the calibration coefficient which best represents the real composition of the matrix. This document firstly presents the needs of control and characterization of radioactive objects, and the means used in the field of nuclear measurement. Emphases is put in particular on the Active Neutron Interrogation method. The matrices of interest are those made of light technological waste (density {<=} 0,4 g/cm{sup 3}) containing hydrogenated and chlorinated materials. The advantages of gamma-rays emitted by neutron captures for the determination of a chlorine content indicator of the matrices and the principles of the correction method are then explained. Measurements have been firstly realized with an existing Neutron Interrogation device (PROMETHEE 6). Such measurements have proven its inadequacy: no signal from the matrix hydrogen was detected, due to an intense signal from the polyethylene contained in some cell elements. Moreover, the matrix chlorine content appeared difficult to be measured. A new and specific device, named REGAIN and dedicated to active gamma-rays spectrometry, was defined with the Monte-Carlo N-Particle (MCNP) code. The experiments conducted with this new device made it possible to detect the

  7. Actinide integral measurements in the CFRMF and integral tests for ENDF/B-V

    International Nuclear Information System (INIS)

    Anderl, R.A.

    1982-01-01

    Integral capture and/or fission rates have been reported earlier for several actinides irradiated in the fast neutron field of the Coupled Fast Reactivity Measurements Facility (CFRMF). These nuclides include 232 Th, 233 U, 235 U, 238 U, 237 Np, 239 Pu, 240 Pu, 242 Pu, 241 Am and 243 Am. This paper forucses on the utilization of these integral data for testing the respective cross sections on ENDF/B-V. Integral cross sections derived from the measured reaction rates are tabulated. Results are presented for cross-section data testing which includes integral testing based on a comparison of calculated and measured integral cross sections and testing based on least-squares-adjustment analyses

  8. The impact of ENDF/B-VI Rev. 3 data on thermal reactor lattices

    International Nuclear Information System (INIS)

    Trkov, A.

    1995-10-01

    The ENDF/B-VI Revision 3 files have been released through the International Atomic Energy Agency. The data for hydrogen, aluminium and uranium-235 were processed to prepare an updated WIMS-D library. Thermal benchmark lattices TRX, BAPL and DIMPLE were analyzed. The new data for the thermal scattering laws of hydrogen bound in water had no significant influence on the integral parameters. The effect of the new uranium-235 data was to reduce the lattice multiplication factor by up to 0.3% Δ k/k. The effect of the new aluminium data was also non-negligible. It was traced to the change in the interpolation law for the total and the capture cross sections, which seems incorrect. (author). 8 refs, 1 fig., 2 tabs

  9. Neutron capture measurements and resonance parameters of dysprosium

    Energy Technology Data Exchange (ETDEWEB)

    Shin, S.G.; Kye, Y.U.; Namkung, W.; Cho, M.H. [Pohang University of Science and Technology, Division of Advanced Nuclear Engineering, Pohang, Gyeongbuk (Korea, Republic of); Kang, Y.R.; Lee, M.W. [Dongnam Inst. of Radiological and Medical Sciences, Research Center, Busan (Korea, Republic of); Kim, G.N. [Kyungpook National University, Department of Physics, Daegu (Korea, Republic of); Ro, T.I. [Dong-A University, Department of Physics, Busan (Korea, Republic of); Danon, Y.; Williams, D. [Rensselaer Polytechnic Institute, Department of Mechanical, Aerospace, and Nuclear Engineering, Troy, NY (United States); Leinweber, G.; Block, R.C.; Barry, D.P.; Rapp, M.J. [Naval Nuclear Laboratory, Knolls Atomic Power Laboratory, Schenectady, NY (United States)

    2017-10-15

    Neutron capture yields of dysprosium isotopes ({sup 161}Dy, {sup 162}Dy, {sup 163}Dy, and {sup 164}Dy) were measured using the time-of-flight method with a 16 segment sodium iodide multiplicity detector. The measurements were made at the 25m flight station at the Gaerttner LINAC Center at Rensselaer Polytechnic Institute. Resonance parameters were obtained using the multilevel R-matrix Bayesian code SAMMY. The neutron capture data for four enriched dysprosium isotopes and one natural dysprosium sample were sequentially fitted. New resonances not listed in ENDF/B-VII.1 were observed. There were 29 and 17 new resonances from {sup 161}Dy and {sup 163}Dy isotopes, respectively. Six resonances from {sup 161}Dy isotope, two resonances from {sup 163}Dy, and four resonances from {sup 164}Dy were not observed. The capture resonance integrals of each isotope were calculated with the resulting resonance parameters and those of ENDF/B-VII.1 in the energy region from 0.5 eV to 20 MeV and were compared to the capture resonance integrals with the resonance parameters from ENDF/B-VII.1. A resonance integral value of the natural dysprosium calculated with present resonance parameters was 1405 ± 3.5 barn. The value is ∝ 0.3% higher than that obtained with the ENDF/B-VII.1 parameters. The distributions of the present and ENDF/B-VII.1 neutron widths were compared to a Porter-Thomas distribution. Neutron strength functions for {sup 161}Dy and {sup 163}Dy were calculated with the present resonance parameters and both values were in between the values of ''Atlas of Neutron Resonances'' and ENDF/B-VII.1. The present radiation width distributions of {sup 161}Dy and {sup 163}Dy were fitted with the χ{sup 2} distribution by varying the degrees of freedom. (orig.)

  10. Standard reference and other important nuclear data. Supplement 1 to the report BNL-NCS-51123 (Dec. 1979) = ENDF-300 = IAEA-NDS-15/300 (microfiche)

    Energy Technology Data Exchange (ETDEWEB)

    Bhat, M R [ed.

    1981-03-01

    The document contains inserts to be added to the report `Standard Reference and Other Important Nuclear Data` (BNL-NCS-51123, ENDF-300), including the following two articles: Fast Neutron Capture in {sup 238}U and {sup 232}Th by W.P. Poentiz (ANL), and {sup 239}Pu Decay Power Discrepancy by T.R. England and P.G. Young (LANL) Refs, figs

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

    International Nuclear Information System (INIS)

    Kim, Kang Seog

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-12

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

  13. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

    International Nuclear Information System (INIS)

    Palmiotti, G.

    2011-01-01

    range of MCNP simulations of criticality benchmarks, with improved performance coming from new structural material evaluations, especially for Ti, Mn, Cr, Zr and W. For Be we see some improvements although the fast assembly data appear to be mutually inconsistent. Actinide cross section updates are also assessed through comparisons of fission and capture reaction rate measurements in critical assemblies and fast reactors. We describe the cross section evaluations that have been updated for ENDF/B-VII.1 and the measured data and calculations that motivated the changes, and therefore this paper augments the ENDF/B-VII.0 publication (1).

  14. Status of the ENDF/B special applications files

    International Nuclear Information System (INIS)

    Stewart, L.

    1977-01-01

    The newly formed SAFE Subcommittee of the Cross Section Evaluation Working Group is charged with the responsibility for providing, reviewing, and testing several ENDF/B special purpose evaluated files. This responsibility currently encompasses dosimetry, activation, hydrogen and helium production, and radioactive decay data required by a variety of users. New formats have been approved by CSEWG for the inclusion of the activation and hydrogen and helium production cross-section libraries. The decay data will be in the same format as that already employed by the Fission Product and Actinide Subcommittee of CSEWG. While an extensive dosimetry file was available on the ENDF/B-IV library for fast reactor applications, other data are needed to extend the range of applications, especially to higher incident neutron energies. This Subcommittee has long-range plans to provide evaluated neutron interaction data that can be recommended for use in many specialized applications. 1 figure, 3 tables

  15. ENDF/B-5 Fission Products Library 1979

    International Nuclear Information System (INIS)

    Schwerer, O.; Lemmel, H.D.

    1981-10-01

    This document summarizes contents and documentation of the 1979 version of the Fission Products File of the ENDF/B Library maintained by the National Nuclear Data Center (NNDC) at the Brookhaven National Laboratory, USA. This file contains numerical neutron reaction data and decay data for 877 fission product nuclides. The entire file or selective retrievals from it can be obtained on magnetic tape from the IAEA Nuclear Data Section. (author)

  16. ENDF/B-5 Fission Products Library. Rev. 2

    International Nuclear Information System (INIS)

    Schwerer, O.; Pronyaev, V.G.; Lemmel, H.D.

    1984-07-01

    This document summarizes contents and documentation of the 1984 version of the Fission Products Nuclear Data File of the ENDF/B-5 Library (Rev. 2) maintained by the National Nuclear Data Center (NNDC) at the Brookhaven National Laboratory, USA. This file contains numerical neutron reaction data and decay data for 877 fission product nuclides. The entire file or selective retrievals from it can be obtained on magnetic tape from the IAEA Nuclear Data Section. (author)

  17. ENDF/B-4 General Purpose File 1974

    International Nuclear Information System (INIS)

    Schwerer, O.

    1980-04-01

    This document summarizes contents and documentation of the 1974 version of the General Purpose File of the ENDF/B Library maintained by the National Nuclear Data Center (NNDC) at the Brookhaven National Laboratory, USA. The Library contains numerical neutron reaction data for 90 isotopes or elements. The entire Library or selective retrievals from it can be obtained on magnetic tape from the IAEA Nuclear Data Section. (author)

  18. Summary of ENDF/B pre-processing codes

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1981-12-01

    This document contains the summary documentation for the ENDF/B pre-processing codes: LINEAR, RECENT, SIGMA1, GROUPIE, EVALPLOT, MERGER, DICTION, CONVERT. This summary documentation is merely a copy of the comment cards that appear at the beginning of each programme; these comment cards always reflect the latest status of input options, etc. For the latest published documentation on the methods used in these codes see UCRL-50400, Vol.17 parts A-E, Lawrence Livermore Laboratory (1979)

  19. Endf/B-VII.0 Based Library for Paragon - 313

    International Nuclear Information System (INIS)

    Huria, H.C.; Kucukboyaci, V.N.; Ouisloumen, M.

    2010-01-01

    A new 70-group library has been generated for the Westinghouse lattice physics code PARAGON using the ENDF/B-VII.0 nuclear data files. The new library retains the major features of the current library, including the number of energy groups and the reduction in the U-238 resonance integral. The upper bound for the up-scattering effects in the new library, however, has been moved to 4.0 eV from 2.1 eV for better MOX fuel predictions. The new library has been used to analyze standard benchmarks and also to compare the measured and predicted parameters for different types of Westinghouse and Combustion Engineering (CE) type operating reactor cores. Results indicate that the new library will not impact the reactivity, power distribution and the temperature coefficient predictions over a wide range of physics design parameters; however, will improve the MOX core predictions. In other words, the ENDF/B-VI.3 and ENDF/B-VII.0 produce similar results for reactor core calculations. (authors)

  20. ENDVER-ENDVER/GUI, The ENDF File Verification Support Package

    International Nuclear Information System (INIS)

    Trkov, Andrej; Zerkin, V.; Cullen, Dermott E.

    2005-01-01

    1 - Description of program or function: Experimental and evaluated nuclear reaction data are compiled world-wide in EXFOR and in ENDF format, respectively. The ENDVER package can be used to convert EXFOR data into computational C4 format, display them and compare graphically with the contents of a specified evaluated data file. The package also contains utilities to retrieve selected materials from a master library in ENDF format, extract cross sections (including differential and double differential data) and output them in two-column PLOTTAB 'curves' format. IAEA1402/03: The ENVER/GUI version contains in addition to ENDVER, also the EXFOR and CINDA Databases (version 1.70 of January 2005) and software providing data search and presentation, the PREPRO ENDF pre-processing codes by D.E. Cullen (PREPRO-2002) and the ZVVIEW interactive graphic data display package. A new GUI (Graphics User Interface) is presented. EndVer and EXFOR/CINDA are directly accessible by the user through the GUI. 2 - Restrictions on the complexity of the problem: None

  1. Monte Carlo analyses of simple U233 O2-ThO2 and U235 O2-ThO2 lattices with ENDF/B-IV data (AWBA development program)

    International Nuclear Information System (INIS)

    Hardy, J. Jr.; Ullo, J.J.

    1980-09-01

    A number of water-moderated Th-U235 and Th-U233 lattice integral experiments were analyzed in a consistent manner, with ENDF/B-IV data and detailed Monte Carlo methods. These experiments provide a consistent test of the nuclear data. The ENDF/B-IV data are found to perform reasonably well. Adequate agreement is found with integral measurements of thorium capture. Calculated K/sub eff/ values show a generally coherent pattern which is consistent with K/sub eff/ results obtained for homogeneous aqueous critical assemblies. Harder prompt fission spectra for U233 and U235 can correct the principal discrepancy observed with ENDF/B-IV, a bias trend in K/sub eff/ attributed to an underprediction of leakage

  2. On evaluated nuclear data for beta-delayed gamma rays following of special nuclear materials

    Energy Technology Data Exchange (ETDEWEB)

    Mencarini, Leonardo de H.; Caldeira, Alexandre D., E-mail: mencarini@ieav.cta.b, E-mail: alexdc@ieav.cta.b [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)

    2011-07-01

    In this paper, a new type of information available in ENDF is discussed. During a consistency check of the evaluated nuclear data library ENDF/B-VII.0 performed at the Nuclear Data Subdivision of the Institute for Advanced Studies, the size of the files for some materials drew the attention of one of the authors. Almost 94 % of all available information for these special nuclear materials is used to represent the beta-delayed gamma rays following fission. This is the first time this information is included in an ENDF version. (author)

  3. On evaluated nuclear data for beta-delayed gamma rays following of special nuclear materials

    International Nuclear Information System (INIS)

    Mencarini, Leonardo de H.; Caldeira, Alexandre D.

    2011-01-01

    In this paper, a new type of information available in ENDF is discussed. During a consistency check of the evaluated nuclear data library ENDF/B-VII.0 performed at the Nuclear Data Subdivision of the Institute for Advanced Studies, the size of the files for some materials drew the attention of one of the authors. Almost 94 % of all available information for these special nuclear materials is used to represent the beta-delayed gamma rays following fission. This is the first time this information is included in an ENDF version. (author)

  4. Measurements of neutron induced capture and fission reactions on $^{233}$ U (EAR1)

    CERN Multimedia

    The $^{233}$U plays the essential role of ssile nucleus in the Th-U fuel cycle, which has been proposed as a safer and cleaner alternative to the U-Pu fuel cycle. Considered the scarce data available to assess the capture cross section, a measurement was proposed and successfully performed at the n_TOF facility at CERN using the 4$\\pi$ Total Absorp- tion Calorimeter (TAC). The measurement was extremely dicult due to the need to accurately distinguish between capture and fission $\\gamma$-rays without any additional discrim-ination tool and the measured capture cross section showed a signicant disagreement in magnitude when compared with the ENDF/B-VII.1 library despite the agreement in shape. We propose a new measurement that is aimed at providing a higher level of dis-crimination between competing nuclear reactions, to extend the neutron energy range and to obtain more precise and accurate data, thus fullling the demands of the "NEA High Priority Nuclear Data Request List". The setup is envisaged as a combin...

  5. 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

  6. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

    Energy Technology Data Exchange (ETDEWEB)

    Chadwick, M.B.; Herman, M.; Author(s): Chadwick,M.B.; Herman,M.; Oblozinsky,P.; Dunn,M.E.; Danon,Y.; Kahler,A.C.; Smith,D.L.; Pritychenko,B.; Arbanas,G.; Arcilla,R.; Brewer,R.; Brown,D.A.; Capote,R.; Carlson,A.D.; Cho,Y.S.; Derrien,H.; Guber,K.; Hale,G.M.; Hoblit,S.; Holloway,S.: Johnson,T.D.; Kawano,T.; Kiedrowski,B.C.; Kim,H.; Kunieda,S.; Larson,N.M.; Leal,L.; Lestone,J.P.; Little,R.C.; McCutchan,E.A.; MacFarlane,R.E.; MacInnes,M.; Mattoon,C.M.; McKnight,R.D.; Mughabghab,S.F.; Nobre,G.P.A.; Palmiotti,G.; Palumbo,A.; Pigni,M.T.; Pronyaev,V.G.; Sayer,R.O.; Sonzogni,A.A.; Summers,N.C.; Talou,P.; Thompson,I.J.; Trkov,A.; Vogt,R.L.; van der Marck,S.C.; Wallner,A.; White,M.C.; Wiarda,D.; Young,P.G.

    2011-12-01

    for a wide range of MCNP simulations of criticality benchmarks, with improved performance coming from new structural material evaluations, especially for Ti, Mn, Cr, Zr and W. For Be we see some improvements although the fast assembly data appear to be mutually inconsistent. Actinide cross section updates are also assessed through comparisons of fission and capture reaction rate measurements in critical assemblies and fast reactors, and improvements are evident. Maxwellian-averaged capture cross sections at 30 keV are also provided for astrophysics applications. We describe the cross section evaluations that have been updated for ENDF/B-VII.1 and the measured data and calculations that motivated the changes, and therefore this paper augments the ENDF/B-VII.0 publication 'ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology,' Nuclear Data Sheets 107, 2931 (2006).

  7. The ENDF/B-VI photon interaction library

    International Nuclear Information System (INIS)

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

    1992-02-01

    The ENDF/B-VI photon interaction library includes data to describe the interaction of photons with the elements Z = 1 to 100 over the energy range 10 eV to 100 MeV. This library has been designed to meet the traditional needs of users to model the interaction and transport of primary photons. However, this library contains additional information which used in a combination with our other data libraries can be used to perform much more detailed calculations, e.g., emission of secondary fluorescence photons. This paper describes both traditional and more detailed uses of this library

  8. Phase 2 testing of ENDF/B-VI shielding data

    International Nuclear Information System (INIS)

    Ingersoll, D.T.; Wright, R.Q.; Slater, C.O.

    1992-01-01

    Version 6 of the US Evaluated Nuclear Data File (ENDF/B-VI) was released in early 1990 and is currently undergoing phase 2 testing. In Phase 2 testing, the evaluated data are approximately processed and used in an integral manner to predict the solution of previously specified benchmark experiments. Results are presented for the initial testing of several light elements and structural materials which are important for shielding applications. These initial tests indicate that the relatively subtle changes made to the iron data and the major modernization of the boron-11 data in Version 6 both represent significant and positive advancements in the quality of the evaluated data files

  9. Effects of the Application of the New Nuclear Data Library ENDF/B to the Criticality Analysis of AP1000

    Science.gov (United States)

    Kuntoro, Iman; Sembiring, T. M.; Susilo, Jati; Deswandri; Sunaryo, G. R.

    2018-02-01

    Calculations of criticality of the AP1000 core due to the use of new edition of nuclear data library namely ENDF/B-VII and ENDF/B-VII.1 have been done. This work is aimed to know the accuracy of ENDF/B-VII.1 compared to ENDF/B-VII and ENDF/B-VI.8. in determining the criticality parameter of AP1000. Analysis ws imposed to core at cold zero power (CZP) conditions. The calculations have been carried out by means of MCNP computer code for 3 dimension geometry. The results show that criticality parameter namely effective multiplication factor of the AP1000 core are higher than that ones resulted from ENDF/B-VI.8 with relative differences of 0.39% for application of ENDF/B-VII and of 0.34% for application of ENDF/B-VII.1.

  10. VIRGIN2007, Calculates Un-collided Neutron Flux and Neutron Reactions from Transmission in ENDF Format

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of program or function: VIRGIN calculates un-collided flux and reactions due to transmission of a mono-directional beam of neutrons through any thickness of material. In order to simulate an experimental measurement the results are given as integrals over energy tally groups (as opposed to point-wise in energy). IAEA0932/10: This version include the updates up to January 30, 2007. Changes in ENDF/B-VII Format and procedures, as well as the evaluations themselves, make it impossible for versions of the ENDF/B pre-processing codes earlier than PREPRO 2007 (2007 Version) to accurately process current ENDF/B-VII evaluations. The present code can handle all existing ENDF/B-VI evaluations through release 8, which will be the last release of ENDF/B-VI. Modifications from previous versions: Virgin VERS. 2007-1 (Jan. 2007): checked against all ENDF/B-VII; increased in-core page size from 60,000 to 240,000. 2 - Method of solution: By taking the ratio of reactions to flux in each group an equivalent spatially dependent group averaged cross section is calculated. 3 - Restrictions on the complexity of the problem: The evaluated data must be in the ENDF/B format. However it must be linear-linear interpolable in energy-cross section between tabulated points. Since only cross sections (file 3) are used, this program will work on any version of ENDF/B

  11. Beta and gamma decay heat evaluation for the thermal fission of 235U

    International Nuclear Information System (INIS)

    Schenter, G.K.; Schmittroth, F.

    1979-01-01

    Beta and gamma fission product decay heat curves are evaluated for the thermal fission of 235 U. Experimental data that include beta, gamma, and total measurements are combined with summation calculations based on ENDF/B in a consistent evaluation. Least-squares methods are used that take proper account of data uncertainties and correlations. 4 figures, 2 tables

  12. Status of CINDER and ENDF/B-V based libraries for transmutation calculations

    International Nuclear Information System (INIS)

    Wilson, W.B.; England, T.R.; LaBauve, R.J.; Battat, M.E.; Wessol, D.E.; Perry, R.T.

    1980-01-01

    The CINDER codes and their data libraries are described, and their range of calculational capabilities are described using documented applications. The importance of ENDF/B data and the features of the ENDF/B-IV and ENDF/B-V fission-product and actinide data files are emphasized. The actinide decay data of ENDF/B-V, augmented by additional data from available sources, are used to produce average decay energy values and neutron source values from sponteneous fission, (α,n) and delayed neutron emission for 144 actinide nuclides that are formed in reactor fuel. The status and characteristics of the CINDER-2 code is described, along with a brief description of more well known code versions; a review of the status of new ENDF/B-V based libraries for all versions is presented

  13. X4ECS, ENDF/B-4 and EXFOR Data Comparison

    International Nuclear Information System (INIS)

    Shen, L.X.

    1985-01-01

    1 - Description of program or function: Compares the evaluated nuclear cross section data (in ENDF/B-IV Format) with experimental data (in EXFOR Format) e.g. compare all the data curves for 92-U-238 (N,2N) from ENDF/B-IV, ENDF/B-V, and EXFOR master files. 2 - Method of solution: Combines the cross section data in EXFOR and/or ENDF/B-IV Format, and converts them to a uniform Format for further processing, especially for graphical comparison of curves. 3 - Restrictions on the complexity of the problem: Input data files must be correctly coded in ENDF/B-IV Format and/or EXFOR Format. In general they are retrieval results for the required data. In EXFOR data files every sub-entry includes cross section data and SUB 001 of the same entry. In evaluated data files the data of MF=1 and MF=3 must be included

  14. Verification and validation of ACE-format library created from ENDF/B-VII.0

    International Nuclear Information System (INIS)

    Chen Chaobin; Hu Zehua; Zhang Benai; Chen Yixue; Wu Jun

    2009-01-01

    ENDF/B-VII.0, released by the USA Cross Section Evaluation Working Group(CSEWG) in December 2006, was developed in five years since the previous release of ENDF/B-VI.8 and was demonstrated to contain much better physical representations of the data and to perform much better than previus ENDF evaluations over a broad range of applications. We generated ACE-format pointwise cross section library from the ENDF/B-VII.0 neutron reaction sublibrary with the processing code NJOY. The paper provides an overview of ENDF/B-VII.0, a summary of the ACE-format files producing process and a detail description of the validation of the ACE-format library. The conclusion is that the ACE-format library produced is correct. (authors)

  15. Visual system of recovering and combination of information for ENDF (Evaluated Nuclear Data File) format libraries

    International Nuclear Information System (INIS)

    Ferreira, Claudia A.S. Velloso; Corcuera, Raquel A. Paviotti

    1997-01-01

    This report presents a data information retrieval and merger system for ENDF (Evaluated Nuclear Data File) format libraries, which can be run on personal computers under the Windows TM environment. The input is the name of an ENDF/B library, which can be chosen in a proper window. The system has a display function which allows the user to visualize the reaction data of a specific nuclide and to produce a printed copy of these data. The system allows the user to retrieve and/or combine evaluated data to create a single file of data in ENDF format, from a number of different files, each of which is in the ENDF format. The user can also create a mini-library from an ENDF/B library. This interactive and easy-to-handle system is a useful tool for Nuclear Data Centers and it is also of interest to nuclear and reactor physics researchers. (author)

  16. Testing ENDF/B-V data for thermal reactors

    International Nuclear Information System (INIS)

    Craig, D.S.

    1982-10-01

    Lattice parameters have been calculated for some thermal reactor benchmark lattices using ENDF/B-V data. These lattices were TRX-1, -2; BAPL-UO 2 -1,-2,-3; BNL-ThO 2 - 233 UO 2 -H 2 0-1,-2,-3; MIT-4,-5,-6; and PNL-31,-33,-35 (infinite lattices). In addition, parameters were calculated for 3 ZEEP lattices, 3 High-Conversion U0 2 -H 2 0 lattices, and 7 BNL-Th0 2 - 233 U0 2 -D 2 0 lattices. These calculations were made using the integral transport cell code RAHAB with the resonance reaction rates obtained using the OZMA code operating in the discrete ordinate mode. This code calculates the resonance rates allowing for the interaction of all resonances. Four group reaction rates for use in method comparisons are given for several lattices. The author discusses the use of the OZMA code for these calculations, including the choice of options and the orders of the angular quadratures, and compares results obtained using the CRNL thermal scattering data with those obtained using ENDF/B data

  17. NJOY: a comprehensive ENDF/B processing system

    International Nuclear Information System (INIS)

    MacFarlane, R.E.; Barrett, R.J.; Muir, D.W.; Boicourt, R.M.

    1978-01-01

    NJOY is the successor to the MINX code. It provides an efficient and accurate capability for processing ENDF/B-IV and -V data for use in fast reactor, thermal reactor, fusion reactor, shielding, and weapons analysis. NJOY produces neutron cross sections and group-to-group scattering matrices, heat production cross sections, photon production matrices, photon interaction cross sections and group-to-group matrices, delayed neutron spectra, thermal scattering cross sections and matrices, and cross-section covariances. Detailed pointwise cross sections, heating KERMA factors, thermal cross sections, and energy-to-energy thermal matrices are also available for plotting and Monte Carlo applications. NJOY currently processes all types of data on ENDF/B except for the decay chain and fission product yield files. NJOY provides output in the CCCC ISOTXS, BRKOXS, and DLAYXS formats, in DTF/ANISN format, and in a new comprehensive format called MATXS. Other important features of NJOY include free-format input, efficient binary I/O, dynamic storage allocation, an extremely modular structure, an accurate center-of-mass Gaussian integration for two-body scattering, and a flux calculator that makes it possible to compute accurate self-shielded cross sections when wide and intermediate-width resonance effects are important. NJOY is a single, integrated, efficient system that produces almost all of the basic cross sections required for multigroup methods of nuclear analysis. 3 figures, 4 tables

  18. FIZCON, ENDF/B Cross-Sections Redundancy Check

    International Nuclear Information System (INIS)

    Dunford, Charles L.

    2007-01-01

    1 - Description of program or function: FIZCON is a program for checking that an evaluated data file has valid data and conforms to recommended procedures. Version 7.01 (April 2005): set success flag after return from beginning; fixed valid level check for an isomer; fixed subsection energy range test in ckf9; changed lower limit on potential scattering test; fixed error in j-value test when l=0 and i=0; added one more significant figure to union grid check and sum up output messages; partial fission cross sections mt=19,20,21 and 38 did not require secondary energy distributions in file 5; corrected product test for elastic scattering; moved potential scattering test to psyche. Version 7.02 (May 2005): Fixed resonance parameter sum test. 2 - Method of solution: FIZCON can recognise the difference between ENDF-6 and ENDF-5 formats and performs its tests accordingly. Some of the tests performed include: data arrays are in increasing energy order; resonance parameter widths add up to the total; Q-values are reasonable and consistent; no required sections are missing and all cover the proper energy range; secondary distributions are normalized to 1.0; energy conservation in decay spectra. Optional tests can be performed to check the redundant cross sections, and algorithms can be used to check for possible incorrect entry of data values (Deviant Point test)

  19. A Short History of ENDF/B Unresolved Resonance Parameters

    Energy Technology Data Exchange (ETDEWEB)

    Cullen, Dermott E. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2010-10-31

    This paper is designed to address two topics relating to ENDF/B data in the unresolved resonance region, Part 1: For years code users have pointed out and complained that various ENDF data processing codes, in particular PREPRO and NJOY, produce different answers from one another for the cross sections in unresolved resonance region. First I assure code users that NJOY has now been updated to agree with PREPRO, so that this problem has now been solved. Part 2: Next, this paper documents why we saw these differences; the emphasis here is on explaining what my own codes do, but I will also try to briefly outline what other codes do, so the reader can understand why we were producing different answers. The first topic should be of general interest to all readers, particularly users of our codes, whereas the second topic will be of more limited interest only to those readers who are interested in the details of our calculations in the unresolved resonance region. Now that our PREPRO and NJOY results agree we consider this problem solved and no further action is necessary.

  20. Analysis of OKTAVIAN Shielding Benchmark Experiments by ENDF/B-VII, JEFF-3.1, and JENDL-3.3

    International Nuclear Information System (INIS)

    Kim, Do-Heon; Gil, Choong-Sup; Lee, Young-Ouk

    2007-01-01

    International collaborations for the ITER Project and other fusion-related development projects have been conducted to create a reference fusion nuclear data library such as FENDL, which was a collection of the best cross section data from national nuclear data libraries. Recent release of newly evaluated nuclear data libraries requires an extensive and intensive benchmarking of the updated transport libraries to become a candidate for the future collection. In this study, the pulsed sphere experiments for leakage neutron and gamma-ray spectra at the D-T neutron source facility of Osaka University, OKTAVIAN were employed to test the ENDF/B-VII beta 1, JEFF-3.1, and JENDL-3.3 libraries. The continuous energy Monte Carlo transport code MCNPX-2.5 was used along with the ACE format libraries processed by a modified version of the NJOY99.90 code

  1. Description of evaluation for /sup 50,52,53,54/Cr performed for ENDF/B-VI

    International Nuclear Information System (INIS)

    Larson, D.C.; Hetrick, D.M.; Fu, C.Y.

    1987-01-01

    Isotopic evaluations for /sup 50,52,53,54/Cr performed for ENDF/B-VI are briefly reviewed. The evaluations are based on analysis of experimental data and results of model calculations which reproduce the experimental data. Evaluated data are given for neutron induced reaction cross sections, angular and energy distributions, and for gamma-ray production cross sections associated with the reactions. File 6 formats are used to represent energy-angle correlated data and recoil spectra. Uncertainty files are included for the major cross sections. Detailed evaluations are given for /sup 52,53/Cr, and results of calculations for reactions with large cross sections are used for evaluation of the minor isotopes. 27 refs., 6 figs

  2. Capturing Thoughts, Capturing Minds?

    DEFF Research Database (Denmark)

    Nielsen, Janni

    2004-01-01

    Think Aloud is cost effective, promises access to the user's mind and is the applied usability technique. But 'keep talking' is difficult, besides, the multimodal interface is visual not verbal. Eye-tracking seems to get around the verbalisation problem. It captures the visual focus of attention...

  3. Apparatus and method for identification of matrix materials in which transuranic elements are embedded using thermal neutron capture gamma-ray emission

    Science.gov (United States)

    Close, D.A.; Franks, L.A.; Kocimski, S.M.

    1984-08-16

    An invention is described that enables the quantitative simultaneous identification of the matrix materials in which fertile and fissile nuclides are embedded to be made along with the quantitative assay of the fertile and fissile materials. The invention also enables corrections for any absorption of neutrons by the matrix materials and by the measurement apparatus by the measurement of the prompt and delayed neutron flux emerging from a sample after the sample is interrogated by simultaneously applied neutrons and gamma radiation. High energy electrons are directed at a first target to produce gamma radiation. A second target receives the resulting pulsed gamma radiation and produces neutrons from the interaction with the gamma radiation. These neutrons are slowed by a moderator surrounding the sample and bathe the sample uniformly, generating second gamma radiation in the interaction. The gamma radiation is then resolved and quantitatively detected, providing a spectroscopic signature of the constituent elements contained in the matrix and in the materials within the vicinity of the sample. (LEW)

  4. ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data

    Energy Technology Data Exchange (ETDEWEB)

    Chadwick, M. B. [Los Alamos National Laboratory (LANL); Herman, Micheal W [Brookhaven National Laboratory (BNL); Oblozinsky, Pavel [Brookhaven National Laboratory (BNL); Dunn, Michael E [ORNL; Danon, Y. [Rensselaer Polytechnic Institute (RPI); Kahler, A. [Los Alamos National Laboratory (LANL); Smith, Donald L. [Argonne National Laboratory (ANL); Pritychenko, B [Brookhaven National Laboratory (BNL); Arbanas, Goran [ORNL; Arcilla, r [Brookhaven National Laboratory (BNL); Brewer, R [Los Alamos National Laboratory (LANL); Brown, D A [Brookhaven National Laboratory (BNL); Capote, R. [International Atomic Energy Agency (IAEA); Carlson, A. D. [National Institute of Standards and Technology (NIST); Cho, Y S [Korea Atomic Energy Research Institute; Derrien, Herve [ORNL; Guber, Klaus H [ORNL; Hale, G. M. [Los Alamos National Laboratory (LANL); Hoblit, S [Brookhaven National Laboratory (BNL); Holloway, Shannon T. [Los Alamos National Laboratory (LANL); Johnson, T D [Brookhaven National Laboratory (BNL); Kawano, T. [Los Alamos National Laboratory (LANL); Kiedrowski, B C [Los Alamos National Laboratory (LANL); Kim, H [Korea Atomic Energy Research Institute; Kunieda, S [Los Alamos National Laboratory (LANL); Larson, Nancy M [ORNL; Leal, Luiz C [ORNL; Lestone, J P [Los Alamos National Laboratory (LANL); Little, R C [Los Alamos National Laboratory (LANL); Mccutchan, E A [Brookhaven National Laboratory (BNL); Macfarlane, R E [Los Alamos National Laboratory (LANL); MacInnes, M [Los Alamos National Laboratory (LANL); Matton, C M [Lawrence Livermore National Laboratory (LLNL); Mcknight, R D [Argonne National Laboratory (ANL); Mughabghab, S F [Brookhaven National Laboratory (BNL); Nobre, G P [Brookhaven National Laboratory (BNL); Palmiotti, G [Idaho National Laboratory (INL); Palumbo, A [Brookhaven National Laboratory (BNL); Pigni, Marco T [ORNL; Pronyaev, V. G. [Institute of Physics and Power Engineering (IPPE), Obninsk, Russia; Sayer, Royce O [ORNL; Sonzogni, A A [Brookhaven National Laboratory (BNL); Summers, N C [Lawrence Livermore National Laboratory (LLNL); Talou, P [Los Alamos National Laboratory (LANL); Thompson, I J [Lawrence Livermore National Laboratory (LLNL); Trkov, A. [Jozef Stefan Institute, Slovenia; Vogt, R L [Lawrence Livermore National Laboratory (LLNL); Van der Marck, S S [Nucl Res & Consultancy Grp, Petten, Netherlands; Wallner, A [University of Vienna, Austria; White, M C [Los Alamos National Laboratory (LANL); Wiarda, Dorothea [ORNL; Young, P C [Los Alamos National Laboratory (LANL)

    2011-01-01

    range of MCNP simulations of criticality benchmarks, with improved performance coming from new structural material evaluations, especially for Ti, Mn, Cr, Zr and W. For Be we see some improvements although the fast assembly data appear to be mutually inconsistent. Actinide cross section updates are also assessed through comparisons of fission and capture reaction rate measurements in critical assemblies and fast reactors, and improvements are evident. Maxwellian-averaged capture cross sections at 30 keV are also provided for astrophysics applications. We describe the cross section evaluations that have been updated for ENDF/B-VII.1 and the measured data and calculations that motivated the changes, and therefore this paper augments the ENDF/B-VII.0 publication [H.

  5. A method to construct covariance files in ENDF/B format for criticality safety applications

    International Nuclear Information System (INIS)

    Naberejnev, D.G.; Smith, D.L.

    1999-01-01

    Argonne National Laboratory is providing support for a criticality safety analysis project that is being performed at Oak Ridge National Laboratory. The ANL role is to provide the covariance information needed by ORNL for this project. The ENDF/B-V evaluation is being used for this particular criticality analysis. In this evaluation, covariance information for several isotopes or elements of interest to this analysis is either not given or needs to be reconsidered. For some required materials, covariance information does not exist in ENDF/B-V: 233 U, 236 U, Zr, Mg, Gd, and Hf. For others, existing covariance information may need to be re-examined in light of the newer ENDF/B-V evaluation and recent experimental data. In this category are the following materials: 235 U, 238 U, 239 Pu, 240 Pu, 241 Pu, Fe, H, C, N, O, Al, Si, and B. A reasonable estimation of the fractional errors for various evaluated neutron cross sections from ENDF/B-V can be based on the comparisons between the major more recent evaluations including ENDF/B-VI, JENDL3.2, BROND2.2, and JEF2.2, as well as a careful examination of experimental data. A reasonable method to construct correlation matrices is proposed here. Coupling both of these considerations suggests a method to construct covariances files in ENDF/B format that can be used to express uncertainties for specific ENDF/B-V cross sections

  6. 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

  7. ENDF/B-VI nuclear data evaluations for fusion applications

    International Nuclear Information System (INIS)

    Dunford, C.L.; Larson, D.C.; Young, P.G.

    1988-01-01

    The next release of the ENDF/B data library planned for 1989 contains improved data evaluations of interest to the fusion neutronics community. New data formats permit inclusion of energy-angle correlated particle emission spectra and recoil nucleus energy spectra. Enhanced formats for covariance information have been developed. Many new isotopic evaluations will lead to improved energy conservation and kerma factor calculations. Improved nuclear model calculations will provide reliable particle emission data where experimental information is sparse. Improved Bayssian fitting codes will provide more accurate evaluations for data rich reactions such as Li(n,nt)α. All of the most important fusion material evaluations contain these new features. 32 refs., 8 figs

  8. Guidebook for the ENDF/B-V nuclear data files

    International Nuclear Information System (INIS)

    Magurno, B.A.; Kinsey, R.R.; Scheffel, F.M.

    1982-07-01

    The National Nuclear Data Center (NNDC) has provided the Electric Power Research Institute (EPRI) with a convenient reference/guidebook to nuclear data derived from the Evaluated Nuclear Data File, Version V (ENDF/B-V). The main part of the edition consists of plots of the major cross sections for each of the General Purpose Nuclides. These plots are reconstructed from the resonance parameters and background cross sections given in the library. The resolution and display format have been selected to show general trends in the data. Following the section for individual nuclides, an intercomparison of cross section ratios (plots of eta and α values) is provided for the major fissile nuclei. The final section contains a table of nuclide properties derived from the data files. Included are thermal (2200m/sec and maxwellian averaged) cross sections, g factors, infinitely dilute resonance integrals and fission spectrum averages

  9. Photoneutron cross sections measurements in {sup 9}Be, {sup 13}C e {sup 17}O with thermal neutron capture gamma-rays; Medidas das secoes de choque de fotoneutrons do {sup 9}Be, {sup 13}C e {sup 17}O com radiacao gama de captura de neutrons termicos

    Energy Technology Data Exchange (ETDEWEB)

    Semmler, Renato

    2006-07-01

    Photoneutron cross sections measurements of {sup 9}Be, {sup 13}C and {sup 17}O have been obtained in the energy interval between 1,6 and 10,8 MeV, using neutron capture gamma-rays with high resolution in energy (3 a 21 eV), produced by 21 target materials, placed inside a tangential beam port, near the core of the IPEN/CNEN-SP IEA-R1 (5 MW) research reactor. The samples have been irradiated inside a 4{pi} geometry neutron detector system 'Long Counter', 520,5 cm away from the capture target. The capture gamma-ray flux was determined by means of the analysis of the gamma spectrum obtained by using a Ge(Li) solid-state detector (EG and G ORTEC, 25 cm{sup 3}, 5%), previously calibrated with capture gamma-rays from a standard target of Nitrogen (Melamine). The neutron photoproduction cross section has been measured for each target capture gamma-ray spectrum (compound cross section). A inversion matrix methodology to solve inversion problems for unfolding the set of experimental compound cross sections, was used in order to obtain the cross sections at specific excitation energy values (principal gamma line energies of the capture targets). The cross sections obtained at the energy values of the principal gamma lines were compared with experimental data reported by other authors, with have employed different gamma-ray sources. A good agreement was observed among the experimental data in this work with reported in the literature. (author)

  10. Evaluation for ENDF/B-IV of the neutron cross sections for 235U from 82 eV to 25 keV

    International Nuclear Information System (INIS)

    Peelle, R.W.

    1976-05-01

    Capture and fission cross sections for 235 U in the ''unresolved resonance'' energy region were evaluated to permit determination of local-average resonance parameters for the ENDF/B-IV cross section file. Microscopic data were examined for infinitely dilute average fission and capture cross sections and also for intermediate structure unlikely to be reproduced by statistical fluctuations of resonance widths and spacings within known laws. Evaluated cross sections, averaged over lethargy intervals greater than 0.1, were obtained as an average over selected data sets after appropriate renormalization. Estimated uncertainties are given for these evaluated average cross sections. The ''intermediate'' structure fluctuations common to a few independent data sets were approximated by straight lines joining successive cross sections at 120 selected energy points; the cross sections at the vertices were adjusted to reproduce the evaluated average cross sections over the broad energy regions. Data sources and methods are reviewed, output values are tabulated, and some modified procedures are suggested for future evaluations. Evaluated fission and capture integrals for the resolved resonance region are also tabulated. These are not in agreement with integrals based on the resonance parameters of ENDF/B versions III and IV. 8 tables, 5 figures

  11. Review of ENDF/B-VI Fission-Product Cross Section

    Energy Technology Data Exchange (ETDEWEB)

    Wright, R.Q.

    1999-01-01

    the uncertainty in calculated results and provide a better interpretation of criticality safety margins. Thus, the thrust of the Nuclear Data Task is to obtain high-resolution data in the intermediate energy region and provide fits to the data that utilize the modern RM formalism and covariance information for subsequent use in criticality predictability applications. As a subtask of the Nuclear Data Task, this review of the fission-product cross sections has several objectives. The first objective is a general data status review at various levels for the some 200 fission products. The second objective is a more detailed investigation of the top 20 fission products with regard to thermal- and intermediate-energy capture and scatter cross sections. The third objective is to demonstrate the revision of ENDF/B evaluations utilizing new data and evaluation techniques for 13 fission products. The fourth objective is to make recommendations for improvements, both specific and general in nature.

  12. Specific activities and the relevant gamma ray dose rates at 1 meter from radioisotopes and isomers following thermal neutron capture reaction

    International Nuclear Information System (INIS)

    Eissa, E.A.; Aly, R.A.; Gomaa, M.A.; Hassan, A.M.

    1995-01-01

    Calculations were performed for the specific activity of 245 gamma-ray emitting radioisotopes and isomers produced in 48, 72 and 96 hour irradiation periods of the natural isotopic mixture of their 77 elements with thermal neutron flux 1.0 E + 13 n/cm 2 .5, at the core of the (ET-R R-1) reactor. The relevant gamma-ray dose rate at a point 1 meter apart from each radioisotope or isomer was evaluated whenever the specific gamma-ray dose rate constant is available. The irradiation time factor (ITF) for the irradiation periods 24, 48, 72 and 96 hours are reported for each of the 248 gamma-ray emitters. The average of (ITF) over these 248 radionuclides for each irradiation period is taken as a measure of the feasibility of the irradiation time. The results favour the increase of the irradiation period from the conventional 48 to 72 hours but not to 96 hours. A programme was established in the VAX computer to carry out the above mentioned calculations. Tables of the present work are very useful for isotope production and reactor safety. 1 fig., 2 tabs

  13. Within the framework of the new fuel cycle 232Th/233U, determination of the 233Pa(n.γ) radiative capture cross section for neutron energies ranging between 0 and 1 MeV

    International Nuclear Information System (INIS)

    Boyer, S.

    2004-10-01

    The Thorium cycle Th 232 /U 233 may face brilliant perspectives through advanced concepts like molten salt reactors or accelerator driven systems but it lacks accurate nuclear data concerning some nuclei. Pa 233 is one of these nuclei, its high activity makes the direct measurement of its radiative neutron capture cross-section almost impossible. This difficulty has been evaded by considering the transfer reaction Th 232 (He 3 ,p)Pa 234 * in which the Pa 234 nucleus is produced in various excited states according to the amount of energy available in the reaction. The first chapter deals with the thorium cycle and its assets to contribute to the quenching of the fast growing world energy demand. The second chapter gives a detailed description of the experimental setting. A scintillation detector based on deuterated benzene (C 6 D 6 ) has been used to counter gamma ray cascades. The third chapter is dedicated to data analysis. In the last chapter we compare our experimental results with ENDF and JENDL data and with computed values from 2 statistical models in the 0-1 MeV neutron energy range. Our results disagree clearly with evaluated data: our values are always above ENDF and JENDL data but tend to near computed values. We have also perform the measurement of the radiative neutron cross-section of Pa 231 for a 110 keV neutron: σ(n,γ) 2.00 ± 0.14 barn. (A.C.)

  14. MACK-IV, a new version of MACK: a program to calculate nuclear response functions from data in ENDF/B format

    International Nuclear Information System (INIS)

    Abdou, M.A.; Gohar, Y.; Wright, R.Q.

    1978-07-01

    MACK-IV calculates nuclear response functions important to the neutronics analysis of nuclear and fusion systems. A central part of the code deals with the calculation of the nuclear response function for nuclear heating more commonly known as the kerma factor. Pointwise and multigroup neutron kerma factors, individual reactions, helium, hydrogen, and tritium production response functions are calculated from any basic nuclear data library in ENDF/B format. The program processes all reactions in the energy range of 0 to 20 MeV for fissionable and nonfissionable materials. The program also calculates the gamma production cross sections and the gamma production energy matrix. A built-in computational capability permits the code to calculate the cross sections in the resolved and unresolved resonance regions from resonance parameters in ENDF/B with an option for Doppler broadening. All energy pointwise and multigroup data calculated by the code can be punched, printed and/or written on tape files. Multigroup response functions (e.g., kerma factors, reaction cross sections, gas production, atomic displacements, etc.) can be outputted in the format of MACK-ACTIVITY-Table suitable for direct use with current neutron (and photon) transport codes

  15. Fission product and actinide data evaluations for ENDF/B--V

    International Nuclear Information System (INIS)

    Schenter, R.E.

    1978-05-01

    The planned content and performance of fission product and actinide nuclide evaluations for the ENDF/B-V collection of data are reviewed. Representative values of parameters for a few nuclides are shown. 10 figures, 5 tables

  16. WinMerger. Visual merging and retrieval of information from ENDF-6 format libraries. Summary documentation

    International Nuclear Information System (INIS)

    Paviotti Corcuera, R.

    1998-01-01

    WinMerger is a PC code that will process any library in ENDF-6 format. The system has a display function which allows the user to visualize the reaction data of a specific nuclide and to produce a printed copy of these data. The system allows the user to retrieve and/or combine evaluated data to create a single file of data in the ENDF-6 format, from a number of different files, each of which is in the ENDF-6 format. The user can also create a mini-library from an ENDF-6 format library. The database was developed under a research contract with the IAEA (No. 302F4BRA88840) and is available on diskette from the IAEA Nuclear Data Section. (author)

  17. Overview of the contents of ENDF/B-VI [Evaluated Nuclear Data File

    International Nuclear Information System (INIS)

    Dunford, C.L.; Pearlstein, S.

    1989-01-01

    The sixth release of the Evaluated Nuclear Data File (ENDF/B-VI) is now being prepared for general distribution. This data file serves as the primary source of nuclear data for nuclear applications in the United States and Canada and in many other countries of the world. The data library is maintained and distributed by the National Nuclear Data Center at Brookhaven National Laboratory from evaluations provided by members of the Cross Section Evaluation Working Group (CSEWG). Unlike its predecessor, ENDF/B-V, this file will be available to all requesters without restrictions. Compared to ENDF/B-V, released more than 11 yr ago, the ENDF/B-VI data library contains significant improvements for both fission and fusion reaction design. Future work will continue with limited staffing and foreign cooperation to provide the data needed for future nuclear applications

  18. FDMXPC, ENDF/B Processing, with Reich-Moore and Adler-Adler Resonance Parameter Calculation

    International Nuclear Information System (INIS)

    Vertes, P.

    1995-01-01

    1 - Description of program or function: FDMXPC is a PC program to process evaluated data files in ENDF format. It computes transmission, self-indication and lumped average cross-section functionals for mixtures of isotopes

  19. Determination of gamma production from (n, gamma) reactions

    International Nuclear Information System (INIS)

    Kostal, M.

    2007-06-01

    Calculation of gamma production by interaction of neutrons with materials requires a reasonable accuracy of the nuclear libraries, i. e. effective cross sections, nuclear levels and probabilities of transitions between them. Accurate data enable accurate calculations to be performed, e.g. for PGNAA. First, gamma production in a thick 56 Fe target was examined. Appreciable discrepancies were found among the nuclear libraries available. Additional calculations were performed and compared with the observed data. The fluence of photons observed behind a thick iron target was investigated, the target being irradiated with neutrons from the front side. The results were evaluated for the various nuclear libraries. It is concluded that the libraries ENDF/B VI.2., i.e. data embedded in the MCNPX code, are sufficient for a number of applications. However, their accuracy is insufficient for prompt gamma neutron activation analysis. This is also true of data from the libraries JEFF 3.1. a JENDL 3.3, so that other libraries will have to be used for PGNAA. Specifically for 56 Fe, the data from the libraries ENDF/B VII.0 seem to be usable. (P.A.)

  20. Calculated K-effectives using ENDF/B-V data for U + Pu solution critical experiments

    International Nuclear Information System (INIS)

    Primm, R.T. III; Mincey, J.F.

    1981-01-01

    Effective multiplication factors for 12 critical experiments have been calculated using multigroup cross sections derived from the ENDF/B-V library. All 12 experiments contained mixed plutonium and uranium nitrate solutions. The range of hydrogen-to-fissile plutonium atom ratios spanned by these experiments was 200 to 2200. A comparison with K-effectives calculated with ENDF/B-IV data is presented

  1. Data formats and procedures for the Evaluated Nuclear Data File, ENDF

    International Nuclear Information System (INIS)

    Garber, D.; Dunford, C.; Pearlstein, S.

    1975-10-01

    This report describes the philosophy of the Evaluated Nuclear Data File (ENDF) and the data formats and procedures that have been developed for it. The ENDF system was designed for the storage and retrieval of the evaluated nuclear data that are required for neutronics, photonics and decay heat calculations. This system is composed of several parts that include a series of data processing codes and neutron and photon cross section nuclear structure libraries

  2. Data formats and procedures for the Evaluated Nuclear Data File, ENDF

    Energy Technology Data Exchange (ETDEWEB)

    Garber, D.; Dunford, C.; Pearlstein, S.

    1975-10-01

    This report describes the philosophy of the Evaluated Nuclear Data File (ENDF) and the data formats and procedures that have been developed for it. The ENDF system was designed for the storage and retrieval of the evaluated nuclear data that are required for neutronics, photonics and decay heat calculations. This system is composed of several parts that include a series of data processing codes and neutron and photon cross section nuclear structure libraries.

  3. INTER, ENDF/B Thermal Cross-Sections, Resonance Integrals, G-Factors Calculation

    International Nuclear Information System (INIS)

    Dunford, Charles L.

    2007-01-01

    1 - Description of program or function: INTER calculates thermal cross sections, g-factors, resonance integrals, fission spectrum averaged cross sections and 14.0 MeV (or other energy) cross sections for major reactions in an ENDF-6 or ENDF-5 format data file. Version 7.01 (Jan 2005): set success flag after return from beginning. 2 - Method of solution: INTER performs integrations by using the trapezoidal rule

  4. Important comments on KERMA factors and DPA cross-section data in ACE files of JENDL-4.0, JEFF-3.2 and ENDF/B-VII.1

    Science.gov (United States)

    Konno, Chikara; Tada, Kenichi; Kwon, Saerom; Ohta, Masayuki; Sato, Satoshi

    2017-09-01

    We have studied reasons of differences of KERMA factors and DPA cross-section data among nuclear data libraries. Here the KERMA factors and DPA cross-section data included in the official ACE files of JENDL-4.0, ENDF/B-VII.1 and JEFF-3.2 are examined in more detail. As a result, it is newly found out that the KERMA factors and DPA cross-section data of a lot of nuclei are different among JENDL-4.0, ENDF/B-VII.1 and JEFF-3.2 and reasons of the differences are the followings: 1) large secondary particle production yield, 2) no secondary gamma data, 3) secondary gamma data in files12-15 mt = 3, 4) mt = 103-107 data without mt = 600 s-800 s data in file6. The issue 1) is considered to be due to nuclear data, while the issues 2)-4) seem to be due to NJOY. The ACE files of JENDL-4.0, ENDF/B-VII.1 and JEFF-3.2 with these problems should be revised after correcting wrong nuclear data and NJOY problems.

  5. Evaluation of the neutron and gamma-ray production cross-sections for 55Mn

    International Nuclear Information System (INIS)

    Takahashi, H.

    1974-11-01

    The evaluation of neutron and gamma production cross sections for manganese-55 from 1.0 (10) -5 eV to 20.0 MeV for ENDF/ B-IV is summarized. Included are resonance parameters, neutron cross sections, angular and energy distribution of secondary neutrons, gamma multiplicities and transition probability array, gamma angular and energy distributions, nuclear model calculations, uncertainty estimates of cross sections, and evaluated cross sections. (U.S.)

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

    International Nuclear Information System (INIS)

    Weinman, J.P.

    1997-06-01

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

  7. Precision and some other problems in ENDF-6 format: thoughts of a programmer

    International Nuclear Information System (INIS)

    Zerkin, V.; )

    2012-01-01

    Problems of ENDF-6 format mentioned in presentations and private communications on several recent nuclear data meetings can be summarized as following: Rigid structure not allowing extensions; Fixed number of digits for data presentation (not enough precision for covariance data); Dropped 'E' in REAL numbers (allowed by FORTRAN, but not by other languages); Impossible to read by a human; Punch-card structure of information (repeating MAT/MF/MT); Short MAT (4 digits only); Short MT (3 digits)[, and long procedure to add new official MT]; Extensive usage of MT=5 for many reactions/products; Complex coding needed to read, no standard software support; Price to rewrite large codes dealing with ENDF-6 format in case it would be replaced by another format/s or modern language/s, etc.. Although it may look strange to start development of a universal FORTRAN library for ENDF-6 files after 20 years of existence of ENDF-6 format, and at the time when nuclear data community begins the development of a new (XML) format to replace ENDF-6 format, nevertheless, the steps described in this paper can help to solve some current problems of ENDF-6 format, to prepare software structure to make transition to XML format for existing codes: smooth, understandable, tested and confirmed at the end

  8. High Energy Neutron Induced Gamma Production

    International Nuclear Information System (INIS)

    Brown, D.A.; Johnson, M.; Navratil, P.

    2007-01-01

    N Division has an interest in improving the physics and accuracy of the gamma data it provides to its customers. It was asked to look into major gamma producing reactions for 14 MeV incident neutrons for several low-Z materials and determine whether LLNL's processed data files faithfully represent the current state of experimental and theoretical knowledge for these reactions. To address this, we surveyed the evaluations of the requested materials, made recommendations for the next ENDL release and noted isotopes that will require further experimental study. This process uncovered several major problems in our translation and processing of the ENDF formatted evaluations, most of which have been resolved

  9. SIGMA1-2007, Doppler Broadening ENDF Format Linear-Linear. Interpolated Point Cross Section

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of problem or function: SIGMA-1 Doppler broadens evaluated Cross sections given in the linear-linear interpolation form of the ENDF/B Format to one final temperature. The data is Doppler broadened, thinned, and output in the ENDF/B Format. IAEA0854/15: This version include the updates up to January 30, 2007. Changes in ENDF/B-VII Format and procedures, as well as the evaluations themselves, make it impossible for versions of the ENDF/B pre-processing codes earlier than PREPRO 2007 (2007 Version) to accurately process current ENDF/B-VII evaluations. The present code can handle all existing ENDF/B-VI evaluations through release 8, which will be the last release of ENDF/B-VI. 2 - Modifications from previous versions: Sigma-1 VERS. 2007-1 (Jan. 2007): checked against all ENDF/B-VII; increased page size from 60,000 to 360,000 energy points 3 - Method of solution: The energy grid is selected to ensure that the broadened data is linear-linear interpolable. SIGMA-1 starts from the free-atom Doppler broadening equations and adds the assumptions of linear data within the table and constant data outside the range of the table. If the Original data is not at zero Kelvin, the data is broadened by the effective temperature difference to the final temperature. If the data is already at a temperature higher than the final temperature, Doppler broadening is not performed. 4 - Restrictions on the complexity of the problem: The input to SIGMA-1 must be data which vary linearly in energy and cross section between tabulated points. The LINEAR program provides such data. LINEAR uses only the ENDF/B BCD Format tape and copies all sections except File 3 as read. Since File 3 data are in identical Format for ENDF/B Versions I through VI, the program can be used with all these versions. - The present version Doppler broadens only to one final temperature

  10. GROUPIE2007, Bondarenko Self-Shielded Cross sections from ENDF/B

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of problem or function - GROUPIE reads evaluated data in ENDF/B Format and uses these to calculate unshielded group averaged Cross sections, Bondarenko self-shielded Cross sections, and multiband parameters. The program allows the user to specify arbitrary energy groups and an arbitrary energy-dependent neutron spectrum (weighting function). IAEA0849/15: This version include the updates up to January 30, 2007. Changes in ENDF/B-VII Format and procedures, as well as the evaluations themselves, make it impossible for versions of the ENDF/B pre-processing codes earlier than PREPRO 2007 (2007 Version) to accurately process current ENDF/B-VII evaluations. The present code can handle all existing ENDF/B-VI evaluations through release 8, which will be the last release of ENDF/B-VI. 2 - Modifications from previous versions: Groupie VERS. 2007-1 (Jan. 2007): checked against all ENDF/B-VII; increased page size from 120,000 to 600,000 points. 3 - Method of solution: All integrals are performed analytically; in no case is iteration or any approximate form of integration used. GROUPIE reads either the 0 deg. Kelvin Cross sections or the Doppler broadened Cross sections to calculate the self-shielded Cross sections and multiband parameters for 25 values of the 'background' Cross sections (representing the combined effects of all other isotopes and of leakage). 4 - Restrictions on the complexity of the problem: GROUPIE requires that the energy-dependent neutron spectrum and all Cross sections be given in tabular form, with linear interpolation between tabulated values. There is no limit to the size of the table used to describe the spectrum, so the spectrum may be described in as much detail as required. - If only unshielded averages are calculated, the program can handle up to 3000 groups. If self-shielded averages and/or multiband parameters are calculated, the program can handle up to 175 groups. These limits can easily be extended. - The program only uses the

  11. ENDF/B-6 charged-particle sublibraries. Summary

    International Nuclear Information System (INIS)

    Lemmel, H.D.; McLaughlin, P.K.; Pronyaev, V.G.

    1999-01-01

    This document summarizes the ENDF/B-6 sublibraries for nuclear reaction data of protons, deuterons and tritons. Presently included are complete (double differential) evaluations for the interaction of protons with H-1 and He-3, and evaluated cross-sections for five fusion reactions between d, t, and He-3 particles. The 1998 and 1999 updates includes complete presentation of the nuclear data for H-1,H-2,He-3,C-12,N-14,O-16,Al-27,P-31,Ca40,Nb-93 and Isotopes of Si, Cr, Fe, Ni, Cu, W, Pb, needed for transport, damage, heating, radioactivity and shielding applications over the incident proton energy range from 1 to 150 MeV. The data library, which has a size of 40.0 Megabytes, is available upon request from the IAEA Nuclear Data Section, costfree, on floppy diskette, or CD-ROM. The library is available online within NDIS, the Nuclear Data Information System and also from the WWW pages of the Nuclear Data Section. (author)

  12. A neutron capture gamma-ray system using isotopic neutron sources and its application for elemental analysis of some saudi samples

    International Nuclear Information System (INIS)

    Bahareth, R.; Alamoudi, Z.; Hassan, A.M.

    2005-01-01

    The design and construction of a prompt gamma-ray spectrometer using 226 R a/Be and 252 C f isotopic neutron sources are described. The characteristic curves of the system for energy and efficiency are presented. Elemental investigations of three Saudi samples [Table salt, Glass and Lubrication oil] have been done. For sake of comparison, the data obtained is compared with the results obtained for the same samples using different techniques. A discussion on the results as well as on the comparative studies is given

  13. Asymptotic normalization coefficients from proton transfer reactions and astrophysical S factors for the CNO 13C(p,gamma)14N radiative capture prosess

    Czech Academy of Sciences Publication Activity Database

    Mukhamedzhanov, A. M.; Azhari, A.; Burjan, Václav; Gagliardi, C. A.; Kroha, Václav; Sattarov, A.; Tang, X.; Trache, L.; Tribble, R. E.

    2003-01-01

    Roč. 725, č. 725 (2003), s. 279-294 ISSN 0375-9474 R&D Projects: GA ČR GA202/01/0709; GA MŠk ME 385 Institutional research plan: CEZ:AV0Z1048901 Keywords : radiative capture reaction * asymptotic normalization coefficient Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.761, year: 2003

  14. Asymptotic normalization coefficients from proton transfer reactions and astrophysical S factors for the (CNOC)-C-13(p, gamma)N-14 radiative capture process

    Czech Academy of Sciences Publication Activity Database

    Mukhamedzhanov, A. M.; Azhari, A.; Burjan, Václav; Gagliardi, C. A.; Kroha, Václav; Sattarov, A.; Tang, X.; Trache, L.; Tribble, R. E.

    2003-01-01

    Roč. 725, č. 22 (2003), s. 279-294 ISSN 0375-9474 R&D Projects: GA ČR GA202/01/0709; GA MŠk ME 385 Institutional research plan: CEZ:AV0Z1048901 Keywords : radiative capture reaction * asymptotic normalization coefficient Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders Impact factor: 1.761, year: 2003

  15. Proton capture resonance studies

    Energy Technology Data Exchange (ETDEWEB)

    Mitchell, G.E. [North Carolina State University, Raleigh, North Carolina (United States) 27695]|[Triangle Universities Nuclear Laboratory, Durham, North Carolina (United States) 27708; Bilpuch, E.G. [Duke University, Durham, North Carolina (United States) 27708]|[Triangle Universities Nuclear Laboratory, Durham, North Carolina (United States) 27708; Bybee, C.R. [North Carolina State University, Raleigh, North Carolina (United States) 27695]|[Triangle Universities Nuclear Laboratory, Durham, North Carolina (United States) 27708; Cox, J.M.; Fittje, L.M. [Tennessee Technological University, Cookeville, Tennessee (United States) 38505]|[Triangle Universities Nuclear Laboratory, Durham, North Carolina (United States) 27708; Labonte, M.A.; Moore, E.F.; Shriner, J.D. [North Carolina State University, Raleigh, North Carolina (United States) 27695]|[Triangle Universities Nuclear Laboratory, Durham, North Carolina (United States) 27708; Shriner, J.F. Jr. [Tennessee Technological University, Cookeville, Tennessee (United States) 38505]|[Triangle Universities Nuclear Laboratory, Durham, North Carolina (United States) 27708; Vavrina, G.A. [North Carolina State University, Raleigh, North Carolina (United States) 27695]|[Triangle Universities Nuclear Laboratory, Durham, North Carolina (United States) 27708; Wallace, P.M. [Duke University, Durham, North Carolina (United States) 27708]|[Triangle Universities Nuclear Laboratory, Durham, North Carolina (United States) 27708

    1997-02-01

    The fluctuation properties of quantum systems now are used as a signature of quantum chaos. The analyses require data of extremely high quality. The {sup 29}Si(p,{gamma}) reaction is being used to establish a complete level scheme of {sup 30}P to study chaos and isospin breaking in this nuclide. Determination of the angular momentum J, the parity {pi}, and the isospin T from resonance capture data is considered. Special emphasis is placed on the capture angular distributions and on a geometric description of these angular distributions. {copyright} {ital 1997 American Institute of Physics.}

  16. ENDF-6 formats manual. Version of June 1997. Written by the members of the US cross section evaluation working group

    International Nuclear Information System (INIS)

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

    1997-01-01

    ENDF-6 is the international computer file format for evaluated nuclear data. This document gives a detailed description of the formats and procedures adopted for ENDF-6. It consists of the report BNL-NCS-44945 (Rev. 2/97) (=ENDF-201, Rev. 2/97) with an Interim Revision of June 1997 and a few front pages added by the IAEA Nuclear Data Section. (author)

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  18. 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)

  19. Release of the ENDF/B-VII.1 Evaluated Nuclear Data File

    Energy Technology Data Exchange (ETDEWEB)

    Brown, David

    2012-06-30

    The Cross Section Evaluation Working Group (CSEWG) released the ENDF/B-VII.1 library on December 22, 2011. The ENDF/B-VII.1 library is CSEWG's latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0, including: many new evaluation in the neutron sublibrary (423 in all and over 190 of these contain covariances), new fission product yields and a greatly improved decay data sublibrary. This summary barely touches on the five years worth of advances present in the ENDF/B-VII.1 library. We expect that these changes will lead to improved integral performance in reactors and other applications. Furthermore, the expansion of covariance data in this release will allow for better uncertainty quantification, reducing design margins and costs. The ENDF library is an ongoing and evolving effort. Currently, the ENDF data community embarking on several parallel efforts to improve library management: (1) The adoption of a continuous integration system to provide evaluators 'instant' feedback on the quality of their evaluations and to provide data users with working 'beta' quality libraries in between major releases. (2) The transition to new hierarchical data format - the Generalized Nuclear Data (GND) format. We expect GND to enable new kinds of evaluated data which cannot be accommodated in the legacy ENDF format. (3) The development of data assimilation and uncertainty propagation techniques to enable the consistent use of integral experimental data in the evaluation process.

  20. PUFF-III: A Code for Processing ENDF Uncertainty Data Into Multigroup Covariance Matrices

    International Nuclear Information System (INIS)

    Dunn, M.E.

    2000-01-01

    PUFF-III is an extension of the previous PUFF-II code that was developed in the 1970s and early 1980s. The PUFF codes process the Evaluated Nuclear Data File (ENDF) covariance data and generate multigroup covariance matrices on a user-specified energy grid structure. Unlike its predecessor, PUFF-III can process the new ENDF/B-VI data formats. In particular, PUFF-III has the capability to process the spontaneous fission covariances for fission neutron multiplicity. With regard to the covariance data in File 33 of the ENDF system, PUFF-III has the capability to process short-range variance formats, as well as the lumped reaction covariance data formats that were introduced in ENDF/B-V. In addition to the new ENDF formats, a new directory feature is now available that allows the user to obtain a detailed directory of the uncertainty information in the data files without visually inspecting the ENDF data. Following the correlation matrix calculation, PUFF-III also evaluates the eigenvalues of each correlation matrix and tests each matrix for positive definiteness. Additional new features are discussed in the manual. PUFF-III has been developed for implementation in the AMPX code system, and several modifications were incorporated to improve memory allocation tasks and input/output operations. Consequently, the resulting code has a structure that is similar to other modules in the AMPX code system. With the release of PUFF-III, a new and improved covariance processing code is available to process ENDF covariance formats through Version VI

  1. ZZ POINT-2007, linearly interpolable ENDF/B-VII.0 data for 14 temperatures

    International Nuclear Information System (INIS)

    Cullen, Dermott E.

    2007-01-01

    A - Description or function: The ENDF/B data library, ENDF/B-VII.0 was processed into the form of temperature dependent cross sections. The original evaluated data include 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, these ENDF/B-VII.0 data were processed into the form of temperature dependent cross sections at eight temperatures: 0, 300, 600, 900, 1200, 1500, 1800 and 2100 Kelvin. It has also been processed to six astrophysics like temperatures: 0.1, 1, 10, 100 eV, 1 and 10 keV. At each temperature the cross sections are tabulated and linearly interpolable in energy with a tolerance of 0.1 %. POINT 2007 contains all of the evaluations in the ENDF/B-VII general purpose library, which contains 78 new evaluations + 315 old ones: total 393 nuclides. It also includes 16 new elemental evaluations replaced by isotopic evaluations + 19 old ones. No special purpose ENDF/B-VII libraries, such as fission products, thermal scattering, photon interaction data are included. These evaluations include all cross sections over the energy range 10 e-5 eV to at least 20 MeV. The list of nuclides is indicated. B - Methods: The PREPRO 2007 code system was used to process the ENDF/B data. Listed below are the steps, including the PREPRO2007 codes, which were used to process the data in the order in which the codes were run. 1) Linearly interpolable, tabulated cross sections (LINEAR) 2) Including the resonance contribution (RECENT) 3) Doppler broaden all cross sections to temperature (SIGMA1) 4) Check data, define redundant cross sections by summation (FIXUP) 5) Update evaluation dictionary in MF/MT=1/451 (DICTIN) C - Restrictions: Due to recent changes in ENDF-6 Formats and Procedures only the latest version of the ENDF/B Pre-processing codes, namely PREPRO 2007, can be used to accurately process all current ENDF/B-VII evaluations. The use of

  2. How Accurate Are Our Processed ENDF Cross Sections?

    International Nuclear Information System (INIS)

    Cullen, Dermott E.

    2014-05-01

    Let me start by reassuring you that currently our nuclear data processing codes are very accurate in the calculations that they perform INSIDE COMNPUTERS. However, most of them drop the ball in what should be a trivial final step to output their results into the ENDF format. This is obviously a very important step, because without accurately outputting their results we would not be able to confidently use their results in our applications. This is indeed a very important step, but unfortunately it is one that is not given the attention it deserves; hence we come to the purpose of this paper. Here I document first the state of a number of nuclear data processing codes as of February 2012, when this comparison began, and then the current state, November 2013, of the same codes. I have delayed publishing results until now to give participants time to distribute updated codes and data. The codes compared include, in alphabetical order: AMPX, NJOY, PREPRO, and SAMMY/SAMRML. During this time we have seen considerable improvement in output results, and as a direct result of this study we now have four codes that produce high precision results, but this is still a long way from ensuring that all codes that handle nuclear data are maintaining the accuracy that we require today. In the first part of this report I consider the precision of our tabulated energies; here we see obvious flaws when less-precise output is used. In the second part I consider the precision of our cross sections; here we see more subtle flaws. The important point to stress is that once these flaws are recognized it is relatively easy to eliminate them and produce high precision energies and cross sections.

  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. New thermal neutron scattering files for ENDF/B-VI release 2

    International Nuclear Information System (INIS)

    MacFarlane, R.E.

    1994-03-01

    At thermal neutron energies, the binding of the scattering nucleus in a solid, liquid, or gas affects the cross section and the distribution of secondary neutrons. These effects are described in the thermal sub-library of Version VI of the Evaluated Nuclear Data Files (ENDF/B-VI) using the File 7 format. In the original release of the ENDF/B-VI library, the data in File 7 were obtained by converting the thermal scattering evaluations of ENDF/B-III to the ENDF-6 format. These original evaluations were prepared at General Atomics (GA) in the late sixties, and they suffer from accuracy limitations imposed by the computers of the day. This report describes new evaluations for six of the thermal moderator materials and six new cold moderator materials. The calculations were made with the LEAPR module of NJOY, which uses methods based on the British code LEAP, together with the original GA physics models, to obtain new ENDF files that are accurate over a wider range of energy and momentum transfer than the existing files. The new materials are H in H 2 O, Be metal, Be in BeO, C in graphite, H in ZrH, Zr in ZrH, liquid ortho-hydrogen, liquid para-hydrogen, liquid ortho-deuterium, liquid para-deuterium liquid methane, and solid methane

  5. S/sub n/ analysis of the TRX metal lattices with ENDF/B version III data

    International Nuclear Information System (INIS)

    Wheeler, F.J.

    1975-01-01

    Two critical assemblies, designated as thermal-reactor benchmarks TRX-1 and TRX-2 for ENDF/B data testing, were analyzed using the one-dimensional S/sub n/-theory code SCAMP. The two assemblies were simple lattices of aluminum-clad, uranium-metal fuel rods in triangular arrays with D 2 O as moderator and reflector. The fuel was low-enriched (1.3 percent 235 U), 0.387-inch in diameter and had an active height of 48 inches. The volume ratio of water to uranium was 2.35 for the TRX-1 lattice and 4.02 for TRX-2. Full-core S/sub n/ calculations based on Version III data were performed for these assemblies and the results obtained were compared with the measured values of the multiplication factors, the ratio of epithermal-to-thermal neutron capture in 238 U, the ratio of epithermal-to-thermal fission in 235 U, the ratio of 238 U fission to 235 U fission, and the ratio of capture in 238 U to fission in 235 U. Reaction rates were obtained from a central region of the full-core problems. Multigroup cross sections for the reactor calculation were obtained from S/sub n/ cell calculations with resonance self-shielding calculated using the RABBLE treatment. The results of the analyses are generally consistent with results obtained by other investigators

  6. Consistency of neutron and proton capture intensity standards new relative intensities for 56Co, 66Ga decay and 35Cl(n,γ) reaction gamma rays

    International Nuclear Information System (INIS)

    Molnar, G.L.; Revay, Z.; Belgya, T.

    2000-01-01

    The equivalence of efficiency determination procedures based on neutron and proton capture lines has been verified and the deviation of high-energy efficiency from linearity confirmed. The new, accurate relative intensities for 56 Co and 66 Ga extend the range of secondary radioactive standards up to 4.8 MeV. Extreme care has to be taken with any high-energy intensity value obtained in the past with the help of 56 Co and 66 Ga calibration sources, and corrections have to be made using the present data of high accuracy. Relative intensities have also been improved for the 35 Cl(n,γ) reaction, a useful secondary standard in a wide energy range, between 0.3-8.5 MeV. The new data are supported by other most recent measurements of a slightly lower precision

  7. The use of averages and other summation quantities in the testing of evaluated fission product yield and decay data. Applications to ENDF/B(IV)

    International Nuclear Information System (INIS)

    Walker, W.H.

    1976-01-01

    Averages of some fission product properties can be obtained by multiplying the fission product yield for each fission product by the value of the property (e.g. mass, atomic number, mass defect) for that fission product and summing all significant contributions. These averages can be used to test the reliability of the yield set or provide useful data for reactor calculations. The report gives the derivation of these averages and discusses their application using the ENDF/B(IV) fission product library. The following quantities are treated here: the number of fission products per fission ΣYsub(i); the average mass number and the average number of neutrons per fission; the average atomic number of the stable fission products and the average number of β-decays per fission; the average mass defect of the stable fission products and the total energy release per fission; the average decay energy per fission (beta, gamma and anti-neutrino); the average β-decay energy per fission; individual and group-averaged delayed neutron emission; the total yield for each fission product element. Wherever it is meaningful to do so, a sum is subdivided into its light and heavy mass components. The most significant differences between calculated values based on ENDF/B(IV) and measurements are the β and γ decay energies for 235 U thermal fission and delayed neutron yields for other fissile nuclides, most notably 238 U. (author)

  8. Microscopic beta and gamma data for decay-heat needs

    International Nuclear Information System (INIS)

    Dickens, J.K.

    1983-01-01

    Microscopic beta and gamma data for decay-heat needs are defined as absolute-intensity spectral distributions of beta and gamma rays following radioactive decay of radionuclides created by, or following, the fission process. Four well-known evaluated data files, namely the US ENDF/B-V, the UK UKFPDD-2, the French BDN (for fission products), and the Japanese JNDC Nuclear Data Library, are reviewed. Comments regarding the analyses of experimental data (particularly gamma-ray data) are given; the need for complete beta-ray spectral measurements is emphasized. Suggestions on goals for near-term future experimental measurements are presented. 34 references

  9. Gravitational capture

    International Nuclear Information System (INIS)

    Bondi, H.

    1979-01-01

    In spite of the strength of gravitational focres between celestial bodies, gravitational capture is not a simple concept. The principles of conservation of linear momentum and of conservation of angular momentum, always impose severe constraints, while conservation of energy and the vital distinction between dissipative and non-dissipative systems allows one to rule out capture in a wide variety of cases. In complex systems especially those without dissipation, long dwell time is a more significant concept than permanent capture. (author)

  10. Plans for use of ENDF/B in reactor research in Indonesia

    International Nuclear Information System (INIS)

    Santoso, B.; Syaukat, A.; Subki, I.; Ganesan, S.

    1989-07-01

    Nuclear data are numerical constants of nature which quantify the nuclear behaviour of all elements and isotopes which make up the reactor medium and its environment, and which are needed as input for performing design calculations for safe and reliable operation of nuclear reactors. The nuclear data are available in the form of recommended values in specially formatted computerized files such as the Evaluated Nuclear Data File-B, known as ENDF/B. The development of base technology in the scheme of original reactor design calculations involves the mastering of the art of ENDF/B data processing. This paper briefly discusses the current status of this activity in Jakarta and gives an account of the future plans, with emphasis on the role of ENDF/B in reactor calculations. (author). 15 refs, 9 figs

  11. 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

  12. A video automated system for nuclear data in the ENDF format

    International Nuclear Information System (INIS)

    Oliveira Silva, O. de; Corcuera, R.P.; Ferreira, P.A.; Moraes Cunha, M. de

    1992-01-01

    This paper presents a video catalogue for libraries in the ENDF-5 or ENDF-6 format (Evaluated Nuclear Data File) which can be run on an IBM-PC computer. This user friendly catalogue is of interest to nuclear and reactor physics researchers. The input is the filename of ENDF data and the output two files giving: a) the list of materials with corresponding laboratory, author and date of evaluation: b) uncorresponding about the MF and MT numbers for each material. The program is written in the C language whose capability of providing windows and interrupts along with speed and portability, has been greatly exploited. The system allows output of options (a) and (b) either on screen, printer or hard disk. (author)

  13. NJOY91, General ENDF/B Processing System for Reactor Design Problems

    International Nuclear Information System (INIS)

    MacFarlane, R.E.; Barrett, R.J.; Muir, D.W.; Boicourt, R.M.

    1997-01-01

    1 - Description of problem or function: The NJOY nuclear data processing system is a comprehensive computer code package for producing pointwise and multigroup neutron, photon, and charged particle cross sections from ENDF/B evaluated nuclear data. NJOY-89 is a substantial upgrade of the previous release. It includes photon production and photon interaction capabilities, heating calculations, covariance processing, and thermal scattering capabilities. It is capable of processing data in ENDF/B-4, ENDF/B-5, and ENDF/B-6 formats for evaluated data (to the extent that the latter have been frozen at the time of this release). NJOY-91.118: This is the last in the NJOY-91 series. It uses the same module structure as the earlier versions and its graphics options depend on DISSPLA. NJOY91.118 includes bug fixes, improvements in several modules, and some new capabilities. Information on the changes is included in the README file. A new test problem was added to test some ENDF/B-6 features, including Reich-Moore resonance reconstruction, energy-angle matrices in GROUPR, and energy-angle distributions in ACER. The 91.118 release is basically configured for UNIX. Short descriptions of the different modules follow: RECONR Reconstructs pointwise (energy-dependent) cross sections from ENDF/B resonance parameters and interpolation schemes. BROADR Doppler broadens and thins pointwise cross sections. UNRESR Computes effective self-shielded pointwise cross sections in the unresolved-resonance region. HEATR Generates pointwise heat production cross sections (KERMA factors) and radiation-damage-energy production cross sections. THERMR Produces incoherent inelastic energy-to-energy matrices for free or bound scatterers, coherent elastic cross sections for hexagonal materials, and incoherent elastic cross sections. GROUPR Generates self-shielded multigroup cross sections, group- to-group neutron scattering matrices, and photon production matrices from pointwise input. GAMINR Calculates

  14. 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

  15. 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).

  16. ZZ POINT-2004, Linearly Interpolable ENDF/B-VI.8 Data for 13 Temperatures

    International Nuclear Information System (INIS)

    Cullen, Dermott E.

    2004-01-01

    A - Description or function: The ENDF/B data library, ENDF/B-VI, Release 8 was processed into the form of temperature dependent cross sections. The original evaluated data include 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, these ENDF/B-VI, Release 8 data were processed into the form of temperature dependent cross sections at eight 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. At each temperature the cross sections are tabulated and linearly interpolable in energy with a tolerance of 0.1 %. POINT2004 contains all of the evaluations in the ENDF/B-VI general purpose library, which contains evaluations for 328 materials (isotopes or naturally occurring elemental mixtures of isotopes). No special purpose ENDF/B-VI libraries, such as fission products, thermal scattering, photon interaction data are included. The majority of these evaluations are complete, in the sense that they include all cross sections over the energy range 10 e-5 eV to at least 20 MeV. B - Methods: The PREPRO2002 code system was used to process the ENDF/B data. Listed below are the steps, including the PREPRO2002 codes, which were used to process the data in the order in which the codes were run. 1) Linearly interpolable, tabulated cross sections (LINEAR); 2) Including the resonance contribution (RECENT); 3) Doppler broaden all cross sections to temperature (SIGMA1); 4) Check data, define redundant cross sections by summation (FIXUP)

  17. ENDF6-transformation of the MENDL-2 and WIND transmutation libraries

    International Nuclear Information System (INIS)

    Koning, A.J.

    1997-09-01

    The MENDL (Medium Energy Nuclear Data library) and WIND (Waste Incineration Nuclear Data library) transmutation libraries have been transformed to the ENDF6-format. A drawback of the original form of these libraries was that they were not processable due to an alternative method to store the residual production cross sections. The new representation of the data is outlined. The transformed library has been checked with the ENDF6 preprocessing tools CHECKR, FIZCON and PSYCHE. In the process of transformation, several errors have been corrected. 5 refs., 3 appendices

  18. TNG calculations and evaluations of photon production data for some ENDF/B-VI materials

    International Nuclear Information System (INIS)

    Fu, C.Y.

    1994-01-01

    Among the new evaluations in the ENDF/B-VI general purpose files, 25 were based on calculations using TNG, a consistent Hauser-Feshbach pre-equilibrium nuclear model code. The photon production cross sections and spectra were calculated simultaneously with the particle emission cross sections and spectra, assuring energy balance for each reaction. The theories used in TNG for these calculations are summarized. Several examples of photon production data, taken from the ENDF/B-VI files, are compared with the available experimental data

  19. Thermal reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    Liu Guisheng

    1996-01-01

    In order to test CENDL-2, ten homogeneous and eight heterogeneous thermal assemblies were used. Both of 123 group cross section libraries based on CENDL-2 and ENDF/B-6 were generated by a nuclear data processing system NSLINK, respectively. The calculations of resonance self-shielding, cell spectra, cell reaction rate ratios and effective multiplication factors (K eff ) of these assemblies have been performed by the modified PASC-1 code system. The calculated results using CENDL-2 show an excellent agreement with corresponding experimental values. However, for some assemblies the K eff values calculated by ENDF/B-6 data are underestimated. (7 tabs.)

  20. Thermal reactor benchmark testing of CENDL-2 and ENDF/B-6

    Energy Technology Data Exchange (ETDEWEB)

    Guisheng, Liu [Chinese Nuclear Data Center, Beijing, BJ (China)

    1996-06-01

    In order to test CENDL-2, ten homogeneous and eight heterogeneous thermal assemblies were used. Both of 123 group cross section libraries based on CENDL-2 and ENDF/B-6 were generated by a nuclear data processing system NSLINK, respectively. The calculations of resonance self-shielding, cell spectra, cell reaction rate ratios and effective multiplication factors (K{sub eff}) of these assemblies have been performed by the modified PASC-1 code system. The calculated results using CENDL-2 show an excellent agreement with corresponding experimental values. However, for some assemblies the K{sub eff} values calculated by ENDF/B-6 data are underestimated. (7 tabs.).

  1. ENDF6-transformation of the MENDL-2 and WIND transmutation libraries

    Energy Technology Data Exchange (ETDEWEB)

    Koning, A.J.

    1997-09-01

    The MENDL (Medium Energy Nuclear Data library) and WIND (Waste Incineration Nuclear Data library) transmutation libraries have been transformed to the ENDF6-format. A drawback of the original form of these libraries was that they were not processable due to an alternative method to store the residual production cross sections. The new representation of the data is outlined. The transformed library has been checked with the ENDF6 preprocessing tools CHECKR, FIZCON and PSYCHE. In the process of transformation, several errors have been corrected. 5 refs., 3 appendices.

  2. ENDF utility codes release 6.10. Description and operating instructions

    International Nuclear Information System (INIS)

    Dunford, C.L.

    1995-01-01

    Description and operating instructions are given for a package of utility codes operating on evaluated nuclear data files in the formats ENDF-6 (and ENDF-5). Included are the data checking codes CHECKER, FIZCON, PSYCHE; the code INTER for retrieving thermal cross-sections and some other data; graphical plotting subroutines PLOTEF, GRALIB, INTLIB; and the file maintenance and retrieval codes LISTEF, SETMDC, GETMAT, STANEF. This program package which is designed for CDC, IBM, DEC and PC computers, can be obtained on magnetic tape or floppy diskette, free of charge, from the IAEA Nuclear Data Section. (author)

  3. The performance of ENDF/B-V.2 nuclear data for fast reactor calculations

    International Nuclear Information System (INIS)

    Atkinson, C.A.; Collins, P.J.

    1987-01-01

    Calculations with ENDF/B-V.2 data have been made for twenty-five fast-spectrum integral assemblies covering a wide range of sizes and compositions. Analysis was done by transport codes with refined cross section processing methods and detailed reactor modelling. The predictions of fission rate distributions and control rod worths were emphasized for the more prototypic benchmark cores. The results show considerable improvements in agreement with experiment compared with analysis using ENDF/B-IV data, but it is apparent that significant errors remain for fast reactor design calculations

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

    International Nuclear Information System (INIS)

    Chiba, Go

    2007-01-01

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

  5. Comparison of results from the MCNP criticality validation suite using ENDF/B-VI and preliminary ENDF/B-VII nuclear data

    Energy Technology Data Exchange (ETDEWEB)

    Mosteller, R. D. (Russell D.)

    2004-01-01

    The MCNP Criticality Validation Suite is a collection of 31 benchmarks taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments. MCNP5 calculations clearly demonstrate that, overall, nuclear data for a preliminary version of ENDFB-VII produce better agreement with the benchmarks in the suite than do corresponding data from ENDF/B-VI. Additional calculations identify areas where improvements in the data still are needed. Based on results for the MCNP Criticality Validation Suite, the Pre-ENDF/B-VII nuclear data produce substantially better overall results than do their ENDF/B-VI counterparts. The calculated values for k{sub eff} for bare metal spheres and for an IEU cylinder reflected by normal uranium are in much better agreement with the benchmark values. In addition, the values of k{sub eff} for the bare metal spheres are much more consistent with those for corresponding metal spheres reflected by normal uranium or water. In addition, a long-standing controversy about the need for an ad hoc adjustment to the {sup 238}U resonance integral for thermal systems may finally be resolved. On the other hand, improvements still are needed in a number of areas. Those areas include intermediate-energy cross sections for {sup 235}U, angular distributions for elastic scattering in deuterium, and fast cross sections for {sup 237}Np.

  6. SU-E-J-100: Reconstruction of Prompt Gamma Ray Three Dimensional SPECT Image From Boron Neutron Capture Therapy(BNCT)

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, D; Jung, J; Suh, T [The Catholic University of Korea, College of medicine, Department of biomedical engineering (Korea, Republic of)

    2014-06-01

    Purpose: Purpose of paper is to confirm the feasibility of acquisition of three dimensional single photon emission computed tomography (SPECT) image from boron neutron capture therapy (BNCT) using Monte Carlo simulation. Methods: In case of simulation, the pixelated SPECT detector, collimator and phantom were simulated using Monte Carlo n particle extended (MCNPX) simulation tool. A thermal neutron source (<1 eV) was used to react with the boron uptake region (BUR) in the phantom. Each geometry had a spherical pattern, and three different BURs (A, B and C region, density: 2.08 g/cm3) were located in the middle of the brain phantom. The data from 128 projections for each sorting process were used to achieve image reconstruction. The ordered subset expectation maximization (OSEM) reconstruction algorithm was used to obtain a tomographic image with eight subsets and five iterations. The receiver operating characteristic (ROC) curve analysis was used to evaluate the geometric accuracy of reconstructed image. Results: The OSEM image was compared with the original phantom pattern image. The area under the curve (AUC) was calculated as the gross area under each ROC curve. The three calculated AUC values were 0.738 (A region), 0.623 (B region), and 0.817 (C region). The differences between length of centers of two boron regions and distance of maximum count points were 0.3 cm, 1.6 cm and 1.4 cm. Conclusion: The possibility of extracting a 3D BNCT SPECT image was confirmed using the Monte Carlo simulation and OSEM algorithm. The prospects for obtaining an actual BNCT SPECT image were estimated from the quality of the simulated image and the simulation conditions. When multiple tumor region should be treated using the BNCT, a reasonable model to determine how many useful images can be obtained from the SPECT could be provided to the BNCT facilities. This research was supported by the Leading Foreign Research Institute Recruitment Program through the National Research

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  8. Study on neutron capture cross sections using the filtered neutron beams of 55 keV and 144 keV at the Dalat reactor and related applications

    International Nuclear Information System (INIS)

    Vuong Huu Tan; Nguyen Canh Hai; Pham Ngoc Son; Tran Tuan Anh

    2007-01-01

    In this fundamental research project on nuclear physics in period of 2006, the neutron capture cross sections for the reactions of 139 La (n,γ) 140 La, 152 Sm (n,γ) 153 Sm, 191 Ir (n,γ) 192 Ir and 193 Ir (n,γ) 194 Ir have been measured at 55 keV and 144 keV by the activation method using the filtered neutron beams of the Dalat nuclear research reactor. The cross sections were determined relative to the standard capture cross sections of 197 Au. The samples and standard were prepaid from high purity (99.99%) foil of Au and natural oxide powders of La 2 O 3 , Sm 2 O 3 and IrO 2 . A high efficient HPGe detector (58%) was used to detect the gamma rays, emitted from the activated samples. The absolute efficiency curve of the detector has been precisely calibrated thanks to a set of standard radioisotope sources and a multi-nuclide standard solution, supported by IAEA. The present results were compared with the previous measurements from EXFOR-2003, and the evaluated values of JENDL 3.3 and ENDF/B-6.8. (author)

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  10. Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    Liu Guisheng

    1995-11-01

    How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testing of CENDL-2 and ENDF/B-6. (author). 8 refs, 2 figs, 4 tabs

  11. INDXENDF: A PC code for indexing nuclear data files in ENDF-6 format

    International Nuclear Information System (INIS)

    Silva, O.O. de; Corcuera, R.P.; Ferreira, P.A.; Moraes Cunha, M. de.

    1992-01-01

    The PC code INDXENDF which creates visual or printed indexes of nuclear data files in ENDF-6 format, is available from the IAEA Nuclear Data Section on a PC diskette, free of charge upon request. The present document describes the features of this code. (author). 11 refs, 9 figs

  12. Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

    International Nuclear Information System (INIS)

    Liu Guisheng

    1995-01-01

    How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testings of CENDL-2 and ENDF/B-6. (4 tabs., 2 figs.)

  13. An evaluation of the ENDF/GASKET model for thermal neutron scattering in heavy water

    International Nuclear Information System (INIS)

    Abbate, M.J.; Antunez, H.M.

    1977-06-01

    The ENDF/GASKET model for computing thermal neutron scattering was selected for studies undertaken with the purpose of getting thoroughly acquainted with the behavior of the heavy water as a moderator. As a first step in its evaluation, the scattering law S(α,β) was computed with ENDF/GASKET. A comparison of the values so obtained with others previously measured or computed showed that the model is not completely satisfactory in this respect. This is attributed to coherent scattering not included in the model and to the need of improving its frequency spectrum. Any way, the experimental values show serious descrepancies and it is difficult to reach definitive conclusions. The Legendre moments of the double differential cross section and its microscopic values were also computed. As it was found by other authors, the incoherent approximation of ENDF/GASKET results in a drastic departure from the measured total cross section below 0,006 eV. In addition, the discrepancies between measured and calculated average μ, might also imply that the coherence effects are appreciable at higher energies. Also decay constance and diffusion parameters were computed for D 2 O (100%), and these agree well with values of other sources. The measurement and computation of neutron spectra in heavy water is presently intented for the sake of completing evaluation. So far two alternatives are foreseen for further work: the improvement of ENDF/GASKET, or the evaluation of the more recent Jarvis model. (author) [es

  14. 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

  15. Criticality benchmark results for the ENDF60 library with MCNP trademark

    International Nuclear Information System (INIS)

    Keen, N.D.; Frankle, S.C.; MacFarlane, R.E.

    1995-01-01

    The continuous-energy neutron data library ENDF60, for use with the Monte Carlo N-Particle radiation transport code MCNP4A, was released in the fall of 1994. The ENDF60 library is comprised of 124 nuclide data files based on the ENDF/B-VI (B-VI) evaluations through Release 2. Fifty-two percent of these B-VI evaluations are translations from ENDF/B-V (B-V). The remaining forty-eight percent are new evaluations which have sometimes changed significantly. Among these changes are greatly increased use of isotopic evaluations, more extensive resonance-parameter evaluations, and energy-angle correlated distributions for secondary particles. In particular, the upper energy limit for the resolved resonance region of 235 U, 238 U and 239 Pu has been extended from 0.082, 4.0, and 0.301 keV to 2..25, 10.0, and 2.5 keV respectively. As regulatory oversight has advanced and performing critical experiments has become more difficult, there has been an increased reliance on computational methods. For the criticality safety community, the performance of the combined transport code and data library is of interest. The purpose of this abstract is to provide benchmarking results to aid the user in determining the best data library for their application

  16. 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.

  17. Summary of ENDF/B Pre-Processing Codes June 1983

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1983-06-01

    This is the summary documentation for the 1983 version of the ENDF/B Pre-Processing Codes LINEAR, RECENT, SIGMA1, GROUPIE, EVALPLOT, MERGER, DICTION, COMPLOT, CONVERT. This summary documentation is merely a copy of the comment cards that appear at the beginning of each programme; these comment cards always reflect the latest status of input options, etc

  18. KTOE, KEDAK to ENDF/B Format Conversion with Linear Linear Interpolation

    International Nuclear Information System (INIS)

    Panini, Gian Carlo

    1985-01-01

    1 - Nature of physical problem solved: This code performs a fully automated translation from KEDAK into ENDF-4 or -5 format. Output is on tape in card image format. 2 - Method of solution: Before translation the reactions are sorted in the ENDF format order. Linear-linear interpolation rule is preserved. The resonance parameters for both resolved and unresolved, could also be translated and a background cross section is formed as the difference of the calculated contribution from the parameters and the point-wise data given in the original file. Elastic angular distributions originally given in tabulated form are converted into Legendre polynomial coefficients. Energy distributions are calculated using a simple evaporation model with the temperature expressed as a function of the incident mass. 3 - Restrictions on the complexity of the problem: The existing restrictions both on KEDAK and ENDF have been applied to the array sizes used in the code, except for the number of points in a section which in the ENDF format are limited to 5000 points. The code only translates one material at a time

  19. ENDF/B-5 Dosimetry Files, mod. 2 1979/81

    International Nuclear Information System (INIS)

    DayDay, N.; Lemmel, H.D.

    1981-09-01

    This document summarizes the contents and documentation of the ENDF/B-5 Dosimetry Files (Point or Group Data) released in October 1979 and modified in August 1981. The files contain data for 36 neutron reactions of 26 isotopes. The entire libraries or selective retrievals from them can be obtained free of charge from the IAEA Nuclear Data Section. (author)

  20. Approximate methods for generation of covariance data for the structural materials of ENDF/B-VI

    International Nuclear Information System (INIS)

    Hetrick, D.M.; Larson, D.C.; Fu, C.Y.

    1992-01-01

    The considerations that governed the development of cross section uncertainty files for the isotopes of Cr, Fe, Ni, Cu, and Pb in ENDF/B-VI are summarized. Four different approaches were used in providing the uncertainty information. Illustrative examples are given which show the resulting standard deviations as a function of incident energy and the corresponding correlation matrices

  1. Library generation and tests for the HAMMER system from the ENDF/B-IV data

    International Nuclear Information System (INIS)

    Queiroz Bogado Leite, S. de; Chalhoub, E.S.; Moraes, Marisa de.

    1983-05-01

    The modifications made to a number of programs belonging to the HAMMER system in order to process data from ENDF/B-IV are presented. An alternate scheme of representing the resonance region by means of extensive tabulation of profiles is made available for selected materials. Numerical examples illustrate comparisons of the results obtained with these libraries against those from the literature. (Author) [pt

  2. 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

  3. LINEAR2007, Linear-Linear Interpolation of ENDF Format Cross-Sections

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of program or function: LINEAR converts evaluated cross sections in the ENDF/B format into a tabular form that is subject to linear-linear interpolation in energy and cross section. The code also thins tables of cross sections already in that form. Codes used subsequently need thus to consider only linear-linear data. IAEA1311/15: This version include the updates up to January 30, 2007. Changes in ENDF/B-VII Format and procedures, as well as the evaluations themselves, make it impossible for versions of the ENDF/B pre-processing codes earlier than PREPRO 2007 (2007 Version) to accurately process current ENDF/B-VII evaluations. The present code can handle all existing ENDF/B-VI evaluations through release 8, which will be the last release of ENDF/B-VI. Modifications from previous versions: - Linear VERS. 2007-1 (JAN. 2007): checked against all ENDF/B-VII; increased page size from 60,000 to 600,000 points 2 - Method of solution: Each section of data is considered separately. Each section of File 3, 23, and 27 data consists of a table of cross section versus energy with any of five interpolation laws. LINEAR will replace each section with a new table of energy versus cross section data in which the interpolation law is always linear in energy and cross section. The histogram (constant cross section between two energies) interpolation law is converted to linear-linear by substituting two points for each initial point. The linear-linear is not altered. For the log-linear, linear-log and log- log laws, the cross section data are converted to linear by an interval halving algorithm. Each interval is divided in half until the value at the middle of the interval can be approximated by linear-linear interpolation to within a given accuracy. The LINEAR program uses a multipoint fractional error thinning algorithm to minimize the size of each cross section table

  4. NJOY-94, General ENDF/B Processing System for Reactor Design Problems

    International Nuclear Information System (INIS)

    1997-01-01

    1 - Description of program or function: The NJOY nuclear data processing system is a comprehensive computer code system for producing pointwise and multigroup cross sections and related quantities from ENDF/B evaluated nuclear data in the ENDF format, including the latest US library, ENDF/B-VI. The NJOY code works with neutrons, photons, and charged particles and produces libraries for a wide variety of particle transport and reactor analysis codes. It is capable of processing data in ENDF/B-4, ENDF/B-5, and ENDF/B-6 formats for evaluated data. Short descriptions of the different modules follow: RECONR Reconstructs pointwise cross sections from ENDF/B resonance parameters and interpolation schemes. BROADR Doppler broadens and thins pointwise cross sections. UNRESR Computes effective self-shielded pointwise cross sections in the unresolved-resonance region. HEATR Generates pointwise heat production cross sections and radiation-damage-energy production cross sections. THERMR Produces incoherent inelastic energy-to-energy matrices for free or bound scatterers, coherent elastic cross sections for hexagonal materials, and incoherent elastic cross sections. GROUPR Generates self-shielded multigroup cross sections, group- to-group neutron scattering matrices, and photon production matrices from pointwise input. GAMINR Calculates multigroup photon interaction cross sections and KERMA factors and group-to-group photon scattering matrices. ERRORR Produces multigroup covariance matrices from ENDF/B uncertainties. COVR Reads the output of ERRORR and performs covariance plotting and output-formatting operations. DTFR Formats multigroup data for transport codes such as DTF-IV and ANISN. CCCCR Formats multigroup data for the CCCC standard interface files ISOTXS, BRKOXS, and DLAYXS. MATXSR Formats multigroup data for the MATXS cross section interface file. ACER Prepares libraries for the Los Alamos continuous-energy Monte Carlo code MCNP. POWR Prepares libraries for the EPRI

  5. Improvement of non-destructive fissile mass assays in α low-level waste drums: A matrix correction method based on neutron capture gamma-rays and a neutron generator

    Science.gov (United States)

    Jallu, F.; Loche, F.

    2008-08-01

    the matrix components by using prompt gamma-rays following neutron capture. The method aims to refine the value of the adequate calibration coefficient used for ANI analysis. This paper presents the final results obtained for 118 l waste drums with low α-activity and low density. This paper discusses the experimental and modelling studies and describes the development of correction abacuses based on gamma-ray spectrometry signals.

  6. Improvement of non-destructive fissile mass assays in α low-level waste drums: A matrix correction method based on neutron capture gamma-rays and a neutron generator

    International Nuclear Information System (INIS)

    Jallu, F.; Loche, F.

    2008-01-01

    quantifying the matrix components by using prompt gamma-rays following neutron capture. The method aims to refine the value of the adequate calibration coefficient used for ANI analysis. This paper presents the final results obtained for 118 l waste drums with low α-activity and low density. This paper discusses the experimental and modelling studies and describes the development of correction abacuses based on gamma-ray spectrometry signals

  7. Improvement of non-destructive fissile mass assays in {alpha} low-level waste drums: A matrix correction method based on neutron capture gamma-rays and a neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Jallu, F. [Commissariat a l' Energie Atomique, CEA, DEN, Nuclear Measurement Laboratory, Bat. 224, 13108 Saint Paul lez Durance (France)], E-mail: fanny.jallu@cea.fr; Loche, F. [Commissariat a l' Energie Atomique, CEA, DEN, Nuclear Measurement Laboratory, Bat. 224, 13108 Saint Paul lez Durance (France)

    2008-08-15

    method, which consists in identifying and quantifying the matrix components by using prompt gamma-rays following neutron capture. The method aims to refine the value of the adequate calibration coefficient used for ANI analysis. This paper presents the final results obtained for 118 l waste drums with low {alpha}-activity and low density. This paper discusses the experimental and modelling studies and describes the development of correction abacuses based on gamma-ray spectrometry signals.

  8. Prompt gamma neutron activation analysis of toxic elements in radioactive waste packages

    Energy Technology Data Exchange (ETDEWEB)

    Ma, J.-L. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 St Paul-lez-Durance (France); Carasco, C., E-mail: cedric.carasco@cea.fr [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 St Paul-lez-Durance (France); Perot, B. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 St Paul-lez-Durance (France); Mauerhofer, E.; Kettler, J.; Havenith, A. [Institute of Energy and Climate Research - Nuclear Waste Management and Reactor Safety, Forschungszentrum Juelich GmbH (Germany)

    2012-07-15

    The French Alternative Energies and Atomic Energy Commission (CEA) and National Radioactive Waste Management Agency (ANDRA) are conducting an R and D program to improve the characterization of long-lived and medium activity (LL-MA) radioactive waste packages. In particular, the amount of toxic elements present in radioactive waste packages must be assessed before they can be accepted in repository facilities in order to avoid pollution of underground water reserves. To this aim, the Nuclear Measurement Laboratory of CEA-Cadarache has started to study the performances of Prompt Gamma Neutron Activation Analysis (PGNAA) for elements showing large capture cross sections such as mercury, cadmium, boron, and chromium. This paper reports a comparison between Monte Carlo calculations performed with the MCNPX computer code using the ENDF/B-VII.0 library and experimental gamma rays measured in the REGAIN PGNAA cell with small samples of nickel, lead, cadmium, arsenic, antimony, chromium, magnesium, zinc, boron, and lithium to verify the validity of a numerical model and gamma-ray production data. The measurement of a {approx}20 kg test sample of concrete containing toxic elements has also been performed, in collaboration with Forschungszentrum Juelich, to validate the model in view of future performance studies for dense and large LL-MA waste packages. - Highlights: Black-Right-Pointing-Pointer Comparison between measurements and MCNP calculation has been performed for a PGNAA system. Black-Right-Pointing-Pointer The system aims at controlling the amount of toxic elements in nuclear waste. Black-Right-Pointing-Pointer Simple samples and a concrete cylinder in which impurities have been added are used. Black-Right-Pointing-Pointer Calculations agree within a factor 2 with measurements. Black-Right-Pointing-Pointer The system can be improved with a better neutron flux monitoring and the use of boron-free graphite.

  9. Neutron Capture and Transmission Measurements and Resonance Parameter Analysis of Niobium

    International Nuclear Information System (INIS)

    NJ Drindak; JA Burke; G Leinweber; JA Helm; JG Hoole; RC Block; Y Danon; RE Slovacek; BE Moretti; CJ Werner; ME Overberg; SA Kolda; MJ Trbovich; DP Barry

    2005-01-01

    Epithermal neutron capture and transmission measurements were performed using the time-of-flight method at the RPI linac using metallic Nb samples. The capture measurements were made at the 25-meter flight station with a 16-section sodium iodide multiplicity detector and the transmission measurements at the 25-meter flight station with a Li-6 glass scintillation detector. Resonance parameters were determined for all resonances up to 500eV with a combined analysis of capture and transmission data using the multi-level R-matrix Bayesian code SAMMY. The present results are compared to those presented in ENDF/B-VI, updated through Release 3

  10. Study on keV-neutron capture cross sections and capture γ-ray spectra of 117,119Sn

    International Nuclear Information System (INIS)

    Nishiyama, J.; Igashira, M.; Ohsaki, T.; Kim, G.N.; Chung, W.C.; Ro, T.I.

    2006-01-01

    The capture cross sections and capture γ-ray spectra of 117,119 Sn were measured in an incident neutron energy region from 10 to 100 keV and at 570 keV, using a 1.5-ns pulsed neutron source by the 7 Li(p,n) 7 Be reaction and a large anti-Compton NaI(Tl) γ-ray spectrometer. A pulse-height weighting technique was applied to observed capture γ-ray pulse-height spectra to derive capture yields. The capture cross sections of 117,119 Sn were obtained with the error of about 5% by using the standard capture cross sections of 197 Au. The present cross sections were compared with previous experimental data and the evaluated values in JENDL-3.3 and ENDF/B-VI. The capture γ-ray spectra of 117,119 Sn were derived by unfolding the observed capture γ-ray pulse-height spectra. The calculations of capture cross sections and capture γ-ray spectra of 117,119 Sn were performed with the EMPIRE-II code. The calculated results were compared with the present experimental ones. (author)

  11. DELTA - a computer program to analyze gamma-gamma angular correlations from unaligned states

    International Nuclear Information System (INIS)

    Ekstroem, L.P.

    1983-10-01

    A computer program to analyze gamma-gamma angular correlations from radioactive decay and from thermal-neutron capture is described. The program can, in addition to correlation data, handle mixing ratio and conversion coefficient data. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-10-02

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  14. A brief description of ENDF/B-IV format data for inventory and decay heating calculations

    International Nuclear Information System (INIS)

    Tobias, A.

    1976-07-01

    In recent years there has been considerable effort directed towards establishing an international standard format for computerised nuclear data files. At the recent conference on Fission Product Nuclear Data (Bologna, 1973) it was agreed that the ENDF/B format, with certain modifications, be adopted as the standard format for the exchange of such data. A brief description of the basic ENDF/B-IV format of nuclear data files for inventory and decay heat calculations is presented. Although data exchange and inter-comparison will be simple for all files using this format, the data is not generally in a form which can be used directly by inventory codes. One solution to this problem may be for each code to possess a 'translating' routine for rearranging the data into its own format. (author)

  15. Thermal and fast reactor benchmark testing of ENDF/B-6.4

    International Nuclear Information System (INIS)

    Liu Guisheng

    1999-01-01

    The benchmark testing for B-6.4 was done with the same benchmark experiments and calculating method as for B-6.2. The effective multiplication factors k eff , central reaction rate ratios of fast assemblies and lattice cell reaction rate ratios of thermal lattice cell assemblies were calculated and compared with testing results of B-6.2 and CENDL-2. It is obvious that 238 U data files are most important for the calculations of large fast reactors and lattice thermal reactors. However, 238 U data in the new version of ENDF/B-6 have not been renewed. Only data of 235 U, 27 Al, 14 N and 2 D have been renewed in ENDF/B-6.4. Therefor, it will be shown that the thermal reactor benchmark testing results are remarkably improved and the fast reactor benchmark testing results are not improved

  16. 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

  17. Extended covariance data formats for the ENDF/B-VI differential data evaluation

    International Nuclear Information System (INIS)

    Peelle, R.W.; Muir, D.W.

    1988-01-01

    The ENDF/B-V included cross section covariance data, but covariances could not be encoded for all the important data types. New ENDF-6 covariance formats are outlined including those for cross-file (MF) covariances, resonance parameters over the whole range, and secondary energy and angle distributions. One ''late entry'' format encodes covariance data for cross sections that are output from model or fitting codes in terms of the model parameter covariance matrix and the tabulated derivatives of cross sections with respect to the model parameters. Another new format yields multigroup cross section variances that increase as the group width decreases. When evaluators use the new formats, the files can be processed and used for improved uncertainty propagation and data combination. 22 refs

  18. 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

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

    International Nuclear Information System (INIS)

    Pritychenko, B.

    2012-01-01

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

  20. ENDF/B Pre-Processing Codes: Implementing and testing on a Personal Computer

    International Nuclear Information System (INIS)

    McLaughlin, P.K.

    1987-05-01

    This document describes the contents of the diskettes containing the ENDF/B Pre-Processing codes by D.E. Cullen, and example data for use in implementing and testing these codes on a Personal Computer of the type IBM-PC/AT. Upon request the codes are available from the IAEA Nuclear Data Section, free of charge, on a series of 7 diskettes. (author)

  1. NJOY. A comprehensive system for the processing of ENDF formatted nuclear data

    International Nuclear Information System (INIS)

    Muir, D.W.

    1990-07-01

    An introduction to the program system NJOY is given which processes data files of evaluated neutron nuclear data coded in ENDF format. NJOY is primarily used for neutron and photon transport calculations for nuclear power reactor design. The NJOY code is not available from the IAEA Nuclear Data Section but may be obtained from the Reactor Shielding Information Center, Oak Ridge, USA. (author). 10 refs, 1 fig

  2. Processing of evaluated neutron data files in ENDF format on personal computers

    International Nuclear Information System (INIS)

    Vertes, P.

    1991-11-01

    A computer code package - FDMXPC - has been developed for processing evaluated data files in ENDF format. The earlier version of this package is supplemented with modules performing calculations using Reich-Moore and Adler-Adler resonance parameters. The processing of evaluated neutron data files by personal computers requires special programming considerations outlined in this report. The scope of the FDMXPC program system is demonstrated by means of numerical examples. (author). 5 refs, 4 figs, 4 tabs

  3. WIMSTAR-4: a computer program for generating WIMS library data from ENDF/B

    International Nuclear Information System (INIS)

    Wilkin, G.B.

    1981-08-01

    WIMSTAR (Version 4) is a FORTRAN-IV computer program developed to generate data files for the WIMS lattice code library from the ENDF/B data base. The program must be used in conjunction with the AMPX-II system and has been designed for implementation as a module of that system. This report describes the structure, implementation and use of the AMPX/WIMSTAR system

  4. Potential impacts of ENDF/B-V on critical experiment analysis based on ZEBRA-8 criticals

    Energy Technology Data Exchange (ETDEWEB)

    Choong, T S

    1982-06-01

    The ZEBRA-8 series of null-zone measurements featured a different neutron spectrum for each assembly. The experiments were designed for the purpose of basic data testing. The series cover a range of spectra both harder and softer than that for the LMFBR. The potential impacts of the newly released ENDF/BV cross section library on LMFBR critical exeriment analysis are discussed based on analysis of ZEBRA-8 series.

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

    International Nuclear Information System (INIS)

    Greene, N.M.

    1975-01-01

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

  6. Extensions to COGEND for ENDF/B-V output of spontaneous fission decay data

    International Nuclear Information System (INIS)

    Tobias, A.

    1978-06-01

    The computer code COGEND, used to produce ENDF/B-IV or -V format nuclear decay scheme data, has been modified in order to extend its range of application. Details are given of the additional facilities which permit the handling of spontaneous fission decay data including any associated continuous spectra. In order to accommodate these additional features it is necessary to increase the core region by 4 kilobytes. (author)

  7. Status of data testing of ENDF/B-V reactor dosimetry file

    International Nuclear Information System (INIS)

    Magurno, B.A.

    1979-01-01

    The ENDF/B-V Reactor Dosimetry File was released August 1979, and Phase II data testing started. The results presented here are from Brookhaven National Laboratory only, and are considered preliminary. The tests include calculated spectrum-averaged cross sections using 235 U fission spectrum (Watt), 252 Cf spontaneous fission spectrum (Watt and Maxwellian), and the Coupled Fast Reactor Measurement Facility (CFRMF) spectrum. 6 tables

  8. Monte Carlo analyses of TRX slightly enriched uranium-H2O critical experiments with ENDF/B-IV and related data sets (AWBA Development Program)

    International Nuclear Information System (INIS)

    Hardy, J. Jr.

    1977-12-01

    Four H 2 O-moderated, slightly-enriched-uranium critical experiments were analyzed by Monte Carlo methods with ENDF/B-IV data. These were simple metal-rod lattices comprising Cross Section Evaluation Working Group thermal reactor benchmarks TRX-1 through TRX-4. Generally good agreement with experiment was obtained for calculated integral parameters: the epi-thermal/thermal ratio of U238 capture (rho 28 ) and of U235 fission (delta 25 ), the ratio of U238 capture to U235 fission (CR*), and the ratio of U238 fission to U235 fission (delta 28 ). Full-core Monte Carlo calculations for two lattices showed good agreement with cell Monte Carlo-plus-multigroup P/sub l/ leakage corrections. Newly measured parameters for the low energy resonances of U238 significantly improved rho 28 . In comparison with other CSEWG analyses, the strong correlation between K/sub eff/ and rho 28 suggests that U238 resonance capture is the major problem encountered in analyzing these lattices

  9. 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.

  10. 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.

  11. 16-APR-03 Final Release of ENDF/B-V for use with LLNL Codes

    International Nuclear Information System (INIS)

    Hill, T S; McNabb, D P; Hedstrom, G W; Beck, B; Hagmann, C A

    2003-01-01

    The new data files were prepared in two steps. First, the ENDF/B-V database was translated to an ENDL-format ascii database. The ENDL ascii format is a point-wise tabular storage scheme where intermediate values are extracted via interpolation. Sufficient point-wise information was generated in the translation to insure an extraction tolerance of 0.1% for most of the data. The only exception is along the incident neutron energy axis of the outgoing particle energy probability density function where a 0.5% tolerance was maintained. Second, processed files were generated from the translated database. Since the translated ENDF/B-V data is in ENDL-format, the standard processing codes were used to generate the new processed data files. To the best of our knowledge, these processed data files are accurate representations of the ENDF/B-V database to within the stated tolerances. However, there are several issues that users must be aware of and they are listed in this report

  12. Integral decay-heat measurements and comparisons to ENDF/B--IV and V

    International Nuclear Information System (INIS)

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

    Results from recent integral decay-power experiments are presented and compared with summation calculations. The experiments include the decay power following thermal fission of 233 U, 235 U, and 239 Pu. The summation calculations use ENDF/B-IV decay data and yields from Versions IV and V. Limited comparisons of experimental β and γ spectra with summation calculations using ENDF/B-IV are included. Generalized least-squares methods are applied to the recent 235 U and 239 Pu decay-power experiments and summation calculations to arrive at evaluated values and uncertainties. Results for 235 U imply uncertainties less than 2% (1 sigma) for the ''infinite'' exposure case for all cooling times greater than 10 seconds. The uncertainties for 239 Pu are larger. Accurate analytical representations of the decay power are presented for 235 , 238 U, and 239 Pu for use in light-water reactors and as the nominal values in the new ANS 5.1 Draft Standard (1978). Comparisons of the nominal values with ENDF/B-IV and the 1973 ANS Draft Standard in current use are included. Gas content, important to decay-heat experiments, and absorption effects on decay power are reviewed. 37 figures, 8 tables

  13. Comparative sensitivity study of some criticality safety benchmark experiments using JEFF-3.1.2, JEFF-3.2T and ENDF/B-VII.1

    International Nuclear Information System (INIS)

    Kooyman, Timothee; Messaoudia, Nadia

    2014-01-01

    A sensitivity study on a set of evaluated criticality benchmarks with two versions of the JEFF nuclear data library, namely JEFF-3.1.2 and JEFF-3.2T, and ENDF/B-VII.1 was performed using MNCP(X) 2.6.0. As these benchmarks serve to estimate the upper safety limit for criticality risk analysis at SCK.CEN the sensitivity of their results to nuclear data is an important parameter to asses. Several nuclides were identified as being responsible for an evident change in the effective multiplication factor k eff : 235 U, 239 Pu, 240 Pu, 54 Fe, 56 Fe, 57 Fe and 208 Pb. A high sensitivity was found to the fission cross-section of all the fissile material in the study. Additionally, a smaller sensitivity to inelastic and capture cross-section of 235 U and 240 Pu was also found. Sensitivity to the scattering law for non-fissile material was postulated. The biggest change in the k eff due to non-fissile material was due to 208 Pb evaluation (±700 pcm), followed by 56 Fe (±360 pcm) for both versions of the JEFF library. Changes due to 235 U (±300 pcm) and Pu isotopes (±120 pcm for 239 Pu and ±80 pcm for 240 Pu) were found only with JEFF-3.1.2. 238 U was found to have no effect on the k eff . Significant improvements were identified between the two versions of the JEFF library. No further differences were found between the JEFF-3.2T and the ENDF/B-VII.1 calculations involving 235 U or Pu. (authors)

  14. Secondary gamma-ray data for shielding calculation

    International Nuclear Information System (INIS)

    Miyasaka, Sunichi

    1979-01-01

    In deep penetration transport calculations, the integral design parameters is determined mainly by secondary particles which are produced by interactions of the primary radiation with materials. The shield thickness and the biological dose rate at a given point of a bulk shield are determined from the contribution from secondary gamma rays. The heat generation and the radiation damage in the structural and shield materials depend strongly on the secondary gamma rays. In this paper, the status of the secondary gamma ray data and its further problems are described from the viewpoint of shield design. The secondary gamma-ray data in ENDF/B-IV and POPOP4 are also discussed based on the test calculations made for several shield assemblies. (author)

  15. NJOY99, Data Processing System of Evaluated Nuclear Data Files ENDF Format

    International Nuclear Information System (INIS)

    2000-01-01

    1 - Description of program or function: The NJOY nuclear data processing system is a modular computer code used for converting evaluated nuclear data in the ENDF format into libraries useful for applications calculations. Because the Evaluated Nuclear Data File (ENDF) format is used all around the world (e.g., ENDF/B-VI in the US, JEF-2.2 in Europe, JENDL-3.2 in Japan, BROND-2.2 in Russia), NJOY gives its users access to a wide variety of the most up-to-date nuclear data. NJOY provides comprehensive capabilities for processing evaluated data, and it can serve applications ranging from continuous-energy Monte Carlo (MCNP), through deterministic transport codes (DANT, ANISN, DORT), to reactor lattice codes (WIMS, EPRI). NJOY handles a wide variety of nuclear effects, including resonances, Doppler broadening, heating (KERMA), radiation damage, thermal scattering (even cold moderators), gas production, neutrons and charged particles, photo-atomic interactions, self shielding, probability tables, photon production, and high-energy interactions (to 150 MeV). Output can include printed listings, special library files for applications, and Postscript graphics (plus color). More information on NJOY is available from the developer's home page at http://t2.lanl.gov/tour/tourbus.html. Follow the Tourbus section of the Tour area to find notes from the ICTP lectures held at Trieste in March 2000 on the ENDF format and on the NJOY code. NJOY contains the following modules: NJOY directs the flow of data through the other modules and contains a library of common functions and subroutines used by the other modules. RECONR reconstructs pointwise (energy-dependent) cross sections from ENDF resonance parameters and interpolation schemes. BROADR Doppler broadens and thins pointwise cross sections. UNRESR computes effective self-shielded pointwise cross sections in the unresolved energy range. HEATR generates pointwise heat production cross sections (KERMA coefficients) and radiation

  16. Beyond the ENDF format: A modern nuclear database structure. SG38 meeting, NEA Headquarters, 29-30 November 2012

    International Nuclear Information System (INIS)

    McNabb, D.; Zerkin, V.; Mattoon, C.; Koning, A.; Brown, D.; Leal, L.; Sublet, J.C.; Coste-Delclaux, M.; Capote, R.; Forrest, R.; Kodeli, I.; Trkov, A.; Beck, B.; Haeck, W.; Fukahori, T.; Mills, R.W.; White, M.C.; Cullen, D.E.

    2012-11-01

    WPEC subgroup 38 (SG38) was formed to develop a new structure for storing nuclear reaction data, that is meant to eventually replace ENDF-6 as the standard way to store and share evaluations. The work of SG38 covers the following tasks: Designing flexible, general-purpose data containers; Determining a logical and easy-to-understand top-level hierarchy for storing evaluated nuclear reaction data; Creating a particle database for storing particles, masses and level schemes; Specifying the infrastructure (plotting, processing, etc.) that must accompany the new structure; Developing an Application Programming Interface or API to allow other codes to access data stored in the new structure; Specifying what tests need to be implemented for quality assurance of the new structure and associated infrastructure; Ensuring documentation and governance of the structure and associated infrastructure. This document is the proceedings of the SG38 meeting, held at the NEA Headquarters on 29-30 November 2012. It comprises all the available presentations (slides) given by the participants as well as 3 reports: A - Welcome and Introduction: - Purpose and goals for SG38 (D. McNabb); - Lessons from ENDF, EXFOR and other formats (V. Zerkin); - Lessons from first LLNL attempt at defining a new nuclear data structure (C. Mattoon); - Example of "2"3"9Pu data B - Purpose of the new data structure: - GND: Purpose of the new data structure (A. Koning); - Purpose of the new data structure: Dave's Perspective (D. Brown); C - Nuclear Data System Overview: - ENDF File uses in AMPX (L. Leal); D - Benefits and requirements for data evaluation and processing: - Benefits and requirements (J.C. Sublet); - CEA/DEN contribution (M. Coste-Delclaux); - Proposals from the IAEA-NDS (V. Zerkin, R. Capote, R. Forrest); - User View on the ENDF Formats and Data Processing (I. Kodeli); - On the ENDF Formats and Data Processing - report (A. Trkov); E - Format perspective, organization and requirements for basic

  17. Comparison of measured and calculated 238U capture self-indication ratios from 4 to 10 keV

    International Nuclear Information System (INIS)

    Perez, R.B.; de Saussure, G.; Yang, J.T.; Munoz-Cobos, J.L.; Todd, J.H.

    1983-01-01

    From 4 keV to 149 keV the 238 U cross sections are represented in ENDF/B-V by unresolved-resonance parameters (URP). The purpose of this representation is to enable the calculation of resonance self-protection as a function of temperature and dilution. Since the URPs are not defined unambiguously by the cross-section data, it is important that the unresolved representation be tested with appropriate experiments, such as capture self-indication ratio (SIR) measurements. In this paper we compare 238 U capture SIR measurements in the 4- to 10-keV energy range with calculations done with ENDF/B-V and with recently published resolved resonance parameters

  18. Calculation of neutron and gamma-ray energy spectra in liquid air and liquid nitrogen due to 14-MeV neutron and californium-252 sources

    International Nuclear Information System (INIS)

    Straker, E.A.; Gritzner, M.L.; Harris, L. Jr.

    1978-01-01

    Calculations of neutron and gamma-ray fluences from 14-MeV neutron and 252 Cf sources in liquid air and liquid nitrogen have been performed. These calculations were made specifically for comparison with experimental data measured at Stohl, Federal Republic of Germany. The discrete-ordinates method was utilized with neutron and gamma-ray cross sections from ENDF/B-IV. One-dimensional calculational models were developed for the sources and tank. Limited comparisons are made with experimental data

  19. ENDF/B-IV fission-product files: summary of major nuclide data

    International Nuclear Information System (INIS)

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

    1975-09-01

    The major fission-product parameters [sigma/sub th/, RI, tau/sub 1/2/, E-bar/sub β/, E-bar/sub γ/, E-bar/sub α/, decay and (n,γ) branching, Q, and AWR] abstracted from ENDF/B-IV files for 824 nuclides are summarized. These data are most often requested by users concerned with reactor design, reactor safety, dose, and other sundry studies. The few known file errors are corrected to date. Tabular data are listed by increasing mass number

  20. ENDF/B-V 7 Standards Data File (EN5-ST Library)

    International Nuclear Information System (INIS)

    DayDay, N.; Lemmel, H.D.

    1980-10-01

    This document summarizes the contents and documentation of the ENDF/B-V 7 Standards Data File (EN5-ST Library) released in September 1979. The library contains complete evaluations for all significant neutron reactions in the energy range 10 -5 eV to 20 MeV for H-1, He-3, Li-6, B-10, C-12, Au-197 and U-235 isotopes. The entire library or selective retrievals from it can be obtained free of charge from the IAEA Nuclear Data Section. (author)

  1. 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

  2. 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

  3. NJOY-97, General ENDF/B Processing System for Reactor Design Problems

    International Nuclear Information System (INIS)

    1999-01-01

    1 - Description of program or function: The NJOY nuclear data processing system is a modular computer code used for converting evaluated nuclear data in the ENDF format into libraries useful for applications calculations. Because the Evaluated Nuclear Data File (ENDF) format is used all around the world (e.g., ENDF/B-VI in the US, JEF-2.2 in Europe, JENDL-3.2 in Japan, BROND-2.2 in Russia), NJOY gives its users access to a wide variety of the most up-to-date nuclear data. NJOY provides comprehensive capabilities for processing evaluated data, and it can serve applications ranging from continuous-energy Monte Carlo (MCNP), through deterministic transport codes (DANT, ANISN, DORT), to reactor lattice codes (WIMS, EPRI). NJOY handles a wide variety of nuclear effects, including resonances, Doppler broadening, heating (KERMA), radiation-damage, thermal scattering (even cold moderators), gas production, neutrons and charged particles, photo-atomic interactions, self shielding, probability tables, photon production, and high-energy interactions (to 150 MeV). Output can include printed listings, special library files for applications, and Postscript graphics (plus colour). More information on NJOY is available from the developer's home page at http://t2.lanl.gov. Follow the Tourbus section of the Tour area to find notes from the ICTP lectures held at Trieste in March 1998 on the ENDF format and on the NJOY code. 2 - Methods: NJOY97 consists of a set of modules, each performing a well-defined processing task. Each of these modules is essentially a separate computer program linked together by input and output files and a few common constants. The methods and instructions on how to use them are documented in the LA-12740-M report on NJOY91 and in the 'README' file. No new published document is yet available. NJOY97 is a cleaned up version of NJOY94.105 that features compatibility with a wider variety of compilers and machines, explicit double precision for 32-bit systems, a

  4. Fast-reactor-data testing of ENDF/B-V at ORNL

    International Nuclear Information System (INIS)

    Wright, R.Q.; Ford, W.E. III; Lucius, J.L.; Webster, C.C.; Marable, J.H.

    1982-01-01

    The Cross Section Evaluation Working Group (CSEWG) is coordinating a program to assess the adequacy of ENDF/B-V cross sections for both fast- and thermal-reactor design applications. A secondary goal is to evaluate cross-section processing codes, cross-section libraries, and radiation-transport codes. Fast reactor data testing (FRDT) goals are accomplished, in part, by comparison of calculated results with documented performance parameters of CSEWG fast reactor benchmarks and with results obtained by other data testers. The purpose of this paper is to describe the results of FRDT at Oak Ridge National Laboratory

  5. Neutron capture reactions on Lu isotopes at DANCE

    CERN Document Server

    Roig, O

    2010-01-01

    The DANCE (Detector for Advanced Neutron Capture Experiments) array located at the Los Alamos national laboratory has been used to obtain the neutron capture cross sections for 175Lu and 176Lu with neutron energies from thermal up to 100 keV. Both isotopes are of current interest for the nucleosynthesis s-process in astrophysics and for applications as in reactor physics or in nuclear medicine. Three targets were used to perform these measurements. One was natLu foil and the other two were isotope-enriched targets of 175Lu and 176Lu. The cross sections are obtained for now through a precise neutron flux determination and a normalization at the thermal neutron cross section value. A comparison with the recent experimental data and the evaluated data of ENDF/B-VII.0 will be presented. In addition, resonances parameters and spin assignments for some resonances will be featured.

  6. 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

  7. Comparison of {sup 235}U fission cross sections in JENDL-3.3 and ENDF/B-VI

    Energy Technology Data Exchange (ETDEWEB)

    Kawano, Toshihiko [Kyushu Univ., Fukuoka (Japan); Carlson, Allan D. [National Institute of Standards and Technology (United States); Matsunobu, Hiroyuki [Data Engineering, Inc., Fujisawa, Kanagawa (Japan); Nakagawa, Tsuneo; Shibata, Keiichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Talou, Patrick; Young, Philip G.; Chadwick, Mark B. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2002-01-01

    Comparisons of evaluated fission cross sections for {sup 235}U in JENDL-3.3 and ENDF/B-VI are carried out. The comparisons are made for both the differential and integral data. The fission cross sections as well as the fission ratios are compared with the experimental data in detail. Spectrum averaged cross sections are calculated and compared with the measurements. The employed spectra are the {sup 235}U prompt fission neutron spectrum, the {sup 252}Cf spontaneous fission neutron spectrum, and the neutron spectrum produced by a {sup 9}Be(d, xn) reaction. For {sup 235}U prompt fission neutron spectrum, the ENDF/B-VI evaluation reproduces experimental averaged cross sections. For {sup 252}Cf and {sup 9}Be(d, xn) neutron spectra, the JENDL-3.3 evaluation gives better results than ENDF/B-VI. (author)

  8. Preliminary proposals for extending the ENDF format to allow incident charged particles and energy-angle correlation for emitted particles

    International Nuclear Information System (INIS)

    MacFarlane, R.E.; Stewart, L.; Hale, G.M.; Dunford, C.L.

    1984-04-01

    This rewrite of Data Formats and Procedures for the Evaluated Nuclear Data File, ENDF pertains to the latest version, ENDF/B-VI. Earlier versions provided representations for neutron cross sections and distributions, photon production from neutron reactions, a limited amount of charged-particle production from neutron reactions, photo-atomic interaction data, thermal neutron scattering data, and radionuclide production and decay data (including fission products). This version allows higher incident energies, adds more complete descriptions of the distributions of emitted particles, and provides for incident charged particles and photo-nuclear data by partitioning the ENDF library into sublibraries. Decay data, fission product yield data, thermal scattering data, and photo-atomic data have also been formally placed in sublibraries. In addition, this rewrite represents an extensive update to the Version V manual

  9. Comparison of 235U fission cross sections in JENDL-3.3 and ENDF/B-VI

    International Nuclear Information System (INIS)

    Kawano, Toshihiko; Carlson, Allan D.; Matsunobu, Hiroyuki; Nakagawa, Tsuneo; Shibata, Keiichi

    2002-01-01

    Comparisons of evaluated fission cross sections for 235 U in JENDL-3.3 and ENDF/B-VI are carried out. The comparisons are made for both the differential and integral data. The fission cross sections as well as the fission ratios are compared with the experimental data in detail. Spectrum averaged cross sections are calculated and compared with the measurements. The employed spectra are the 235 U prompt fission neutron spectrum, the 252 Cf spontaneous fission neutron spectrum, and the neutron spectrum produced by a 9 Be(d, xn) reaction. For 235 U prompt fission neutron spectrum, the ENDF/B-VI evaluation reproduces experimental averaged cross sections. For 252 Cf and 9 Be(d, xn) neutron spectra, the JENDL-3.3 evaluation gives better results than ENDF/B-VI. (author)

  10. Qualification and improvement of iron ENDF/B-VI and JEF-2 evaluations by interpretation of the Aspis Benchmark

    International Nuclear Information System (INIS)

    Zheng, S.H.; Kodeli, I.; Raepsaet, C.; Diop, C.M.; Nimal, J.C.; Monnier, A.

    1992-01-01

    The aim of the present study is to contribute to the validation of the new evaluated nuclear data files like ENDF/B-VI or JEF-2.2. The new cross-section evaluation for iron isotopes are of particular interest for the nuclear community, since it is well known that the ENDF/B-IV data underestimate the neutron flux on deep penetration problems. The performances of the new nuclear data libraries are compared with those of ENDF-B-IV. The ASPIS benchmark, where the neutron transports through more than one meter iron plate, was chosen for this study. The cross-section libraries were produced by the THEMIS/NJOY (ref 1) processing system and the transport calculations were carried out using the 3D Monte-Carlo code TRIPOLI. The influence of different multigroup cross-section representations was investigated. Finally, sensitivity, uncertainty and data adjustment analyses were carried out to obtain some additional informations about the quality of the cross-section data in ENDF/B-VI files. The analyses were performed using the code package set up of different modules, either developed at CEA or obtained from the NEA Data Bank. The adjustment indicated that some modifications have to be introduced to the neutron cross-sections of iron and the whole calculations were repeated with the adjusted set of cross sections. The comparison of the results of the uncertainty and the adjustment analyses applied to ENDF/B-IV and ENDF/B-VI iron data permits to establish the progress made and gives some indications about the state-of-the-art of the cross-section data

  11. Update of ENDF/B-V Mod 3 iron: neutron-producing reaction cross sections and energy-angle correlations

    International Nuclear Information System (INIS)

    Fu, C.Y.; Hetrick, D.M.

    1986-07-01

    An update of the ENDF/B-V Mod-3 evaluation for natural iron is described. The cross sections of (n,n') and (n,2n) reactions are revised. Energy-angle correlations in the secondary (n,n') neutrons are introduced in the ENDF/B-V formats. Anisotropic angular distributions are provided for the secondary neutrons in (n,2n), (n,np), and (n,nα) reactions. Revelant integral results, microscopic data, and nuclear model calculations that influence the revised results are summarized. 54 refs., 9 figs., 2 tabs

  12. Calculation of 235U(n,n') cross sections for ENDF/B-VI

    International Nuclear Information System (INIS)

    Young, P.G.; Arthur, E.D.

    1988-01-01

    Cross sections for neutron-induced reactions on 235 U between 0.01 and 20 MeV have been calculated in a preliminary analysis for the ENDF/B-VI evaluation with particular emphasis on neutron inelastic scattering. A deformed optical model potential that fits total, elastic, inelastic, and low-energy average resonance data is used to calculate direct (n,n') cross sections and transmission coefficients for a Hauser-Feshbach statistical theory analysis using a multiple fission barrier representation. Direct cross sections for higher-lying vibrational states are provided from DWBA calculations, normalized using B(E/ital l/) values determined from (d,d') and Coulomb excitation data. Initial fission barrier parameters and transition state density enhancements appropriate to the compound systems involved were obtained from previous analyses, especially fits to charged-particle fission probability data. Further modifications to fit 235 U(n,f) data were small, and the final fission parameters are generally consistent with published values. The results from this preliminary analysis are compared with the ENDF/B-V evaluation as well as with experimental data. 26 refs., 5 figs., 3 tabs

  13. A comparison of reaction rate calculations using Endf/B-VII with critical assembly measurements

    International Nuclear Information System (INIS)

    Wilkerson, C.; Mac Innes, M.; Barr, D.; Trellue, H.; MacFarlane, R.; Chadwick, M.

    2008-01-01

    We present critical assembly reaction rate data, and modeling of the same using the recently released Endf/B-VII library. While some of the experimental measurements were performed as long as 50 years ago, the results have not been widely used/available outside of Los Alamos. Over the years, a variety of target foils were fabricated and placed in differing neutron spectrum/fluence environments within critical assemblies. Neutron-induced reactions such as (n,γ), (n,2n), and (n,f) on these targets were measured, typically referenced to 235 U(n,f) or 239 Pu(n,f). Because the cross section for the latter reactions are now well known, these experiments provide a rich data set for testing evaluated cross sections. Due to the large variety of critical assemblies that were historically available at Los Alamos, it was possible to make measurements in spectral environments ranging from hard (Pu Jezebel, center of Pu Flattop) through intermediate (Big Ten) to degraded (reflector region of Flattop). This broad range of configurations allows us to test both the cross section magnitudes and their energy dependencies. We will present data, along with reaction rate predictions using primarily MCNP5 in conjunction with Endf/B-VII, for a number of target nuclei, including iridium, isotopes of uranium (e.g., 233, 235, 237, 238), neptunium (237), plutonium (239), and americium (241). (authors)

  14. Testing of ENDF/B cross section data in the Californium-252 neutron benchmark field

    International Nuclear Information System (INIS)

    Mannhart, W.

    1979-01-01

    The fission neutron field of 252 Cf presently represents one of the most well-known neutron benchmark fields. For 13 neutron reactions which are of importance in reactor metrology, measurements of spectrum-averaged cross sections, [sigma], performed in this neutron field were compared with calculated average cross sections. This comparison allows one to draw conclusions as to the quality of different sigma(E) data taken from ENDF/B-IV, from ENDF/B-V, and from recent experiments and used in the calculation of average cross sections. The comparison includes an uncertainty analysis regarding the different uncertainty contributions of [sigma], of sigma(E), and of the spectral distribution of 252 Cf fission neutrons. Additionally, in a few examples, sensitivity studies were carried out. The sensitivity of the spectrum-averaged cross sections to individual characteristics of the sigma(E) data, such as normalization factors or shifts in the energy scale, was investigated. Similarly, the sensitivity of [sigma] to the spectral distribution of 252 Cf was determined. 4 figures, 2 tables

  15. Secondary gamma-ray skyshine from 14 MeV Neutron Source Facility (OKTAVIAN). Comparison of measurement with its simulation

    Energy Technology Data Exchange (ETDEWEB)

    Morotomi, Ryutaro; Kondo, Tetsuo; Murata, Isao; Yoshida, Shigeo; Takahashi, Akito [Osaka Univ., Department of Nuclear Engineering, Suita, Osaka (Japan); Yamamoto, Takayoshi [Osaka Univ., Radio Isotope Research Center, Suita, Osaka (Japan)

    2000-03-01

    Measurement of secondary gamma-ray skyshine was performed at the Intense 14 MeV Neutron Source Facility (OKTAVIAN) of Osaka University with NaI and Hp-Ge detectors. From the result of measurements, some mechanism of secondary gamma-ray skyshine from 14 MeV neutron source facility was found out. The analysis of the measured result were carried out with MCNP-4B for four nuclear data files of JENDL-3.2, JENDL-F.F., FENDL-2, and ENDF/B-VI. It was confirmed that all the nuclear data are fairly reliable for calculations of secondary gamma-ray skyshine. (author)

  16. Evaluation of the total gamma-ray production cross-sections for nonelastic interaction of fast neutrons with iron nuclei

    International Nuclear Information System (INIS)

    Savin, M.V.; Nefedov, Yu.Ya; Livke, A.V.; Zvenigorodskij, A.G.

    2001-01-01

    Experimental data on the total gamma-ray production cross-sections for inelastic interaction of fast neutrons with iron nuclei were analysed. The total gamma-ray production cross-sections, grouped according to E γ , were evaluated in the neutron energy range 0.5-19 MeV. The statistical spline approximation method was used to evaluate the experimental data. Evaluated data stored in the ENDF, JENDL, BROND, and other libraries on gamma-ray production spectra and cross-sections for inelastic interaction of fast neutrons with iron nuclei, were analysed. (author)

  17. Gamma-gamma angular correlation measurement in the 100 Ru

    International Nuclear Information System (INIS)

    Kenchian, G.

    1990-01-01

    An angular correlation automatic spectrometer with two Ge(Li) detectors has been developed. The spectrometer moves automatically, controlled by a microcomputer. The gamma-gamma directional angular correlations of coincidence transitions have been measured in 100 Ru nuclide, following the β + and electron capture of 100 Rh. The 100 Rh source has been produced with 100 Ru(p,n) 100 Rh reaction, using the proton beam of the Cyclotron Accelerator insiding in 100 Ru isotope. (author)

  18. Validation of actinide nuclear data from ENDF/B-V, INDL/A-83 and JENDL-2

    International Nuclear Information System (INIS)

    Paviotti Corcuera, R.; Moraes, M. de

    1988-11-01

    Resonance integrals and fission spectrum averaged cross sections are calculated for the actinides of ENDF/B-V, INDL/A-83 and JENDL-2. The results are compared with each other and with experimental data when available. The experimental data are scarce and there exist large differences among data from different libraries. (author). 16 refs, 4 tabs

  19. Evaluation of the 56Fe(n,p)56Mn cross sections for ENDF/B-III

    International Nuclear Information System (INIS)

    Dudey, N.D.; Kennerley, R.

    1975-01-01

    The literature examined in this review includes all references in CINDA 71 and its supplements plus some very recent measurements near the reaction threshold. All reference cross sections have been renormalized to ENDF/B-III cross sections and weighted least-squared fitting routines were used to systematize the evaluations

  20. The calculational VVER burnup Credit Benchmark No.3 results with the ENDF/B-VI rev.5 (1999)

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez Gual, Maritza [Centro de Tecnologia Nuclear, La Habana (Cuba). E-mail: mrgual@ctn.isctn.edu.cu

    2000-07-01

    The purpose of this papers to present the results of CB3 phase of the VVER calculational benchmark with the recent evaluated nuclear data library ENDF/B-VI Rev.5 (1999). This results are compared with the obtained from the other participants in the calculations (Czech Republic, Finland, Hungary, Slovaquia, Spain and the United Kingdom). The phase (CB3) of the VVER calculation benchmark is similar to the Phase II-A of the OECD/NEA/INSC BUC Working Group benchmark for PWR. The cases without burnup profile (BP) were performed with the WIMS/D-4 code. The rest of the cases have been carried with DOTIII discrete ordinates code. The neutron library used was the ENDF/B-VI rev. 5 (1999). The WIMS/D-4 (69 groups) is used to collapse cross sections from the ENDF/B-VI Rev. 5 (1999) to 36 groups working library for 2-D calculations. This work also comprises the results of CB1 (obtained with ENDF/B-VI rev. 5 (1999), too) and CB3 for cases with Burnup of 30 MWd/TU and cooling time of 1 and 5 years and for case with Burnup of 40 MWd/TU and cooling time of 1 year. (author)

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

    International Nuclear Information System (INIS)

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

    1985-06-01

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

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

    International Nuclear Information System (INIS)

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

    1982-05-01

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

  3. Generation of covariance files for the isotopes of Cr, Fe, Ni, Cu, and Pb in ENDF/B-VI

    International Nuclear Information System (INIS)

    Hetrick, D.M.; Larson, D.C.; Fu, C.Y.

    1991-02-01

    The considerations that governed the development of the uncertainty files for the isotopes of Cr, Fe, Ni, Cu, and Pb in ENDF/B-VI are summarized. Four different approaches were used in providing the covariance information. Some examples are given which show the standard deviations as a function of incident energy and the corresponding correlation matrices. 11 refs., 5 tabs

  4. The measurement of gamma ray induced heating in a mixed neutron and gamma ray environment

    International Nuclear Information System (INIS)

    Chiu, H.K.

    1991-10-01

    The problem of measuring the gamma heating in a mixed DT neutron and gamma ray environment was explored. A new detector technique was developed to make this measurement. Gamma heating measurements were made in a low-Z assembly irradiated with 14-Mev neutrons and (n, n') gammas produced by a Texas Nuclear Model 9400 neutron generator. Heating measurements were made in the mid-line of the lattice using a proportional counter operating in the Continuously-varied Bias-voltage Acquisition mode. The neutron-induced signal was separated from the gamma-induced signal by exploiting the signal rise-time differences inherent to radiations of different linear energy transfer coefficient, which are observable in a proportional counter. The operating limits of this measurement technique were explored by varying the counter position in the low-Z lattice, hence changing the irradiation spectrum observed. The experiment was modelled numerically to help interpret the measured results. The transport of neutrons and gamma rays in the assembly was modelled using the one- dimensional radiation transport code ANISN/PC. The cross-section set used for these calculations was derived from the ENDF/B-V library using the code MC 2 -2 for the case of DT neutrons slowing down in a low-Z material. The calculated neutron and gamma spectra in the slab and the relevant mass-stopping powers were used to construct weighting factors which relate the energy deposition in the counter fill-gas to that in the counter wall and in the surrounding material. The gamma energy deposition at various positions in the lattice is estimated by applying these weighting factors to the measured gamma energy deposition in the counter at those locations

  5. Response matrix calculation of a Bonner Sphere Spectrometer using ENDF/B-VII libraries

    Energy Technology Data Exchange (ETDEWEB)

    Morató, Sergio; Juste, Belén; Miró, Rafael; Verdú, Gumersindo [Instituto de Seguridad Industrial, Radiofísica y Medioambiental (ISIRYM), Universitat Politècnica de València (Spain); Guardia, Vicent, E-mail: bejusvi@iqn.upv.es [GD Energy Services, Valencia (Spain). Grupo dominguis

    2017-07-01

    The present work is focused on the reconstruction of a neutron spectra using a multisphere spectrometer also called Bonner Spheres System (BSS). To that, the determination of the response detector curves is necessary therefore we have obtained the response matrix of a neutron detector by Monte Carlo (MC) simulation with MCNP6 where the use of unstructured mesh geometries is introduced as a novelty. The aim of these curves was to study the theoretical response of a widespread neutron spectrometer exposed to neutron radiation. A neutron detector device has been used in this work which is formed by a multispheres spectrometer (BSS) that uses 6 high density polyethylene spheres with different diameters. The BSS consists of a set of 0.95 g/cm{sup 3} high density polyethylene spheres. The detector is composed of a lithium iodide 6LiI cylindrical scintillator crystal 4mm x 4mm size LUDLUM Model 42 coupled to a photomultiplier tube. Thermal tables are required to include polyethylene cross section in the simulation. These data are essential to get correct and accurate results in problems involving neutron thermalization. Nowadays available literature present the response matrix calculated with ENDF.B.V cross section libraries (V.Mares et al 1993) or with ENDF.B.VI (R.Vega Carrillo et al 2007). This work uses two novelties to calculate the response matrix. On the one hand the use of unstructured meshes to simulate the geometry of the detector and the Bonner Spheres and on the other hand the use of the updated ENDF.B.VII cross sections libraries. A set of simulations have been performed to obtain the detector response matrix. 29 mono energetic neutron beams between 10 KeV to 20 MeV were used as source for each moderator sphere up to a total of 174 simulations. Each mono energetic source was defined with the same diameter as the moderating sphere used in its corresponding simulation and the spheres were uniformly irradiated from the top of the photomultiplier tube. Some

  6. Monte Carlo analysis of Pu-H2O and UO2-PuO2-H2O critical assemblies with ENDF/B-IV data

    International Nuclear Information System (INIS)

    Hardy, J. Jr.; Ullo, J.J.

    1981-04-01

    A set of critical experiments, comprising thirteen homogeneous Pu-H 2 O assemblies and twelve UO 2 -PuO 2 lattices, was analyzed with ENDF/B-IV data and the RCPO1 Monte Carlo program, which modeled the experiments explicitly. Some major data sensitivities were also evaluated. For the Pu-H 2 O assemblies, calculated K/sub eff/ averaged 1.011. The large (2.7%) scatter of K/sub eff/ values for these assemblies was attributed mostly to uncertainties in physical specifications since no clear trends of K/sub eff/ were evident and data sensitivities were insignificant. The UO 2 -PuO 2 lattices showed just one trend of K/sub eff/, which indicated an overprediction of U238 capture consistent with that observed for uranium-H 2 O experiments. There was however a approx. 1% discrepancy in calculated K/sub eff/ between the two sets of UO 2 -PuO 2 lattices studied

  7. Gamma Knife

    Science.gov (United States)

    ... News Physician Resources Professions Site Index A-Z Gamma Knife Gamma Knife® is a radiation therapy that uses computerized ... If you're scheduled for radiation therapy using Gamma Knife®, a treatment team consisting of a radiation ...

  8. Neutron capture therapy

    International Nuclear Information System (INIS)

    Jun, B. J.

    1998-11-01

    The overall state of the art related with neutron capture therapy(NCT) is surveyed. Since the field related with NCT is very wide, it is not intended to survey all related subjects in depth. The primary objective of this report is to help those working for the installation of a NCT facility and a PGNAA(prompt gamma ray neutron activation analysis) system for the boron analysis understand overall NCT at Hanaro. Therefore, while the parts of reactor neutron source and PGNAA are dealt in detail, other parts are limited to the level necessary to understand related fields. For example, the subject of chemical compound which requires intensive knowledge on chemistry, is not dealt as a separated item. However, the requirement of a compound for NCT, currently available compounds, their characteristics, etc. could be understood through this report. Although the subject of cancer treated by NCT is out of the capability of the author, it is dealt focussing its characteristics related with the success of NCT. Each detailed subject is expected to be dealt more detail by specialists in future. This report would be helpful for the researchers working for the NCT to understand related fields. (author). 128 refs., 3 tabs., 12 figs

  9. Evaluation of Neutron Reactions on Iron Isotopes for CIELO and ENDF/B-VIII.0

    Science.gov (United States)

    Herman, M.; Trkov, A.; Capote, R.; Nobre, G. P. A.; Brown, D. A.; Arcilla, R.; Danon, Y.; Plompen, A.; Mughabghab, S. F.; Jing, Q.; Zhigang, G.; Tingjin, L.; Hanlin, L.; Xichao, R.; Leal, L.; Carlson, B. V.; Kawano, T.; Sin, M.; Simakov, S. P.; Guber, K.

    2018-02-01

    A new suite of evaluations for 54,56,57,58Fe has been developed in the framework of the CIELO international collaboration. New resolved resonance ranges were evaluated for 54Fe and 57Fe, while modifications were applied to resonances in 56Fe. The low energy part of the 56Fe file is almost totally based on measurements. At higher energies in 56Fe and in the whole fast neutron range for minor isotopes the evaluation consists of model predictions carefully adjusted to available experimental data. We also make use of the high quality and well experimentally-constrained dosimetry evaluations from the IRDFF library. Special attention was dedicated to the elastic angular distributions, which were found to affect results of the integral benchmarking. The new set of iron evaluations was developed in concert with other CIELO evaluations and they were tested together in the integral experiments before being adopted for the ENDF/B-VIII.0 library.

  10. POLLA/IECTA, ENDF/B Reich-Moore to Adler-Adler Resonance Parameter Conversion

    International Nuclear Information System (INIS)

    Carlson, B.V.; Chalhoub, E.S.; Melnikoff, M.

    1987-01-01

    1 - Description of program or function: POLLA1 transforms Reich-Moore resolved resonance parameters to the corresponding positive momentum Adler-Adler ones. It is designed to run directly on a file in the ENDF/B format, creating a new file in which the Reich-Moore parameterization has been replaced by the Adler-Adler one. 2 - Method of solution: The Adler-Adler poles are obtained by applying Newton's method to the inverse of the determinant of the Reich-Moore pole matrix. The perturbative solution of R.B. Perez and G. de Saussure, Phys. Rev. C10 (1974)187, is used as a first guess. The residues at the poles are calculated using a simple numerical difference method. 3 - Restrictions on the complexity of the problem: As currently dimensioned, the program permits a maximum of 400 coherent resonances. The transformation itself, which neglects all negative momentum poles, is only accurate for heavy nuclei

  11. Analysis of Np-237 ENDF for the theortical interpretation of critical assembly experiments.

    Energy Technology Data Exchange (ETDEWEB)

    Mihaila, B. (Bogdan); Chadwick, M. B. (Mark B.); MacFarlane, R. E. (Robert E.); Kawano, T. (Toshihiko)

    2004-01-01

    We report on the present status of our effort toward an improved Np-237 evaluated nuclear data file (ENDF). The aim here is to bridge the gap between calculated and observed k-eff values, as measured at the Np-U critical assembly at LANL, TA-18. As such, we perform a critical analysis of the existing body of experimental data and recommended evaluations. We are targeting in principal the fission nu-bar and cross section in Np-237, as well as the inelastic scattering which is particularly important since Np-237 is a threshold fissioner. This analysis will be employed in a future sensitivity study of the calculated k-eff with respect to variations of the afore mentioned nuclear data.

  12. Benchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1

    International Nuclear Information System (INIS)

    Van Der Marck, S. C.

    2012-01-01

    Three nuclear data libraries have been tested extensively using criticality safety benchmark calculations. The three libraries are the new release of the US library ENDF/B-VII.1 (2011), the new release of the Japanese library JENDL-4.0 (2011), and the OECD/NEA library JEFF-3.1 (2006). All calculations were performed with the continuous-energy Monte Carlo code MCNP (version 4C3, as well as version 6-beta1). Around 2000 benchmark cases from the International Handbook of Criticality Safety Benchmark Experiments (ICSBEP) were used. The results were analyzed per ICSBEP category, and per element. Overall, the three libraries show similar performance on most criticality safety benchmarks. The largest differences are probably caused by elements such as Be, C, Fe, Zr, W. (authors)

  13. Monte Carlo; based validation of the ENDF/MC2-II/SDX cell homogenization path

    International Nuclear Information System (INIS)

    Wade, D.C.

    1978-11-01

    The results are summarized of a program of validation of the unit cell homogenization prescriptions and codes used for the analysis of Zero Power Reactor (ZPR) fast breeder reactor critical experiments. The ZPR drawer loading patterns comprise both plate type and pin-calandria type unit cells. A prescription is used to convert the three dimensional physical geometry of the drawer loadings into one dimensional calculational models. The ETOE-II/MC 2 -II/SDX code sequence is used to transform ENDF/B basic nuclear data into unit cell average broad group cross sections based on the 1D models. Cell average, broad group anisotropic diffusion coefficients are generated using the methods of Benoist or of Gelbard. The resulting broad (approx. 10 to 30) group parameters are used in multigroup diffusion and S/sub n/ transport calculations of full core XY or RZ models which employ smeared atom densities to represent the contents of the unit cells

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

    International Nuclear Information System (INIS)

    Nishimura, Hideo

    1977-01-01

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

  16. ENDF/B-6 charged-particle sub-libraries. Summary

    International Nuclear Information System (INIS)

    Lemmel, H.D.; McLaughlin, P.K.; Pronyaev, V.G.

    1998-01-01

    This document summarizes the ENDF/B-6 sublibraries for nuclear reaction data of protons, deuterons and tritons. Presently included are complete (double differential) evaluations for the interaction of protons with H-1 and He-3, and evaluated cross-sections for five fusion reactions between d, t, and He-3 particles. The 1998 update includes complete presentation of the nuclear data for C-12 and O-16 needed for transport, damage, heating, radioactivity and shielding applications over the incident proton energy range from 1 to 150 MeV. The data library, which has a size of 1.60 Megabytes, is available upon request from the IAEA Nuclear Data Section, costfree, on floppy diskette, or CD-ROM. The library is available online within NDIS, the Nuclear Data Information System and also from the WWW pages of the Nuclear Data Section. (author)

  17. 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)

  18. EXFOR-CINDA-ENDF: Migration of Databases to Give Higher-Quality Nuclear Data Services

    International Nuclear Information System (INIS)

    Zerkin, V.V.; McLane, V.; Herman, M.W.; Dunford, C.L.

    2005-01-01

    Extensive work began in 1999 to migrate the EXFOR, CINDA, and ENDF nuclear reaction databases, and convert the available nuclear data services from VMS to a modern computing environment. This work has been performed through co-operative efforts between the IAEA Nuclear Data Section (IAEA-NDS) and the National Nuclear Data Center (NNDC), Brookhaven National Laboratory. The project also afforded the opportunity to make general revisions and improvements to the nuclear reaction data services by taking account of past experience with the old system and users' feedback. A main goal of the project was to implement databases in a relational form that provides full functionality for maintenance by data centre staff and improved retrieval capability for external users. As a result, the quality of our nuclear service has significantly improved, with better functionality of the system, accessibility of data, and improved data retrieval functions for users involved in a wide range of applications

  19. 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

  20. Observation of two photons in n-p capture

    International Nuclear Information System (INIS)

    Dress, W.B. Jr.

    1972-01-01

    The observation of two gamma rays in coincidence following the capture of subthermal neutrons by protons in water is reported. The measured branching ratio of two-gamma events to single-gamma events in the energy range of 600 keV to 1620 keV was found to be 0.0011 +- 0.0002. Possible sources of systematic error are also considered. (8 figures) (U.S.)

  1. 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

  2. Europium resonance parameters from neutron capture and transmission measurements in the energy range 0.01–200 eV

    International Nuclear Information System (INIS)

    Leinweber, G.; Barry, D.P.; Burke, J.A.; Rapp, M.J.; Block, R.C.; Danon, Y.; Geuther, J.A.; Saglime III, F.J.

    2014-01-01

    Highlights: • Metal samples were sealed and imaged with X-rays to determine sample uniformity. • Eleven new resonances were identified below 100 eV. • The resonance regions of 151 Eu and 153 Eu have been extended from 100 to 200 eV. • The thermal total cross section for 151 Eu was measured, up (9 ± 3)% from ENDF/B-VII.1. • Radiation widths were assigned for all resonances from experimental data. - Abstract: Europium is a good absorber of neutrons suitable for use as a nuclear reactor control material. It is also a fission product in the low-yield tail at the high end of the fission fragment mass distribution. Measurements have been made of the stable isotopes with natural and enriched samples. The linear electron accelerator center (LINAC) at the Rensselaer Polytechnic Institute (RPI) was used to explore neutron interactions with europium in the energy region from 0.01 to 200 eV. Neutron capture and transmission measurements were performed by the time-of-flight technique. Two transmission measurements were performed at flight paths of 15 and 25 m with 6 Li glass scintillation detectors. The neutron capture measurements were performed at a flight path of 25 m with a 16-segment sodium iodide multiplicity detector. Resonance parameters were extracted from the data using the multilevel R-matrix Bayesian code SAMMY. A table of resonance parameters and their uncertainties is presented. To prevent air oxidation metal samples were sealed in airtight aluminum cans in an inert environment. Metal samples of natural europium, 47.8 atom% 151 Eu, 52.2 atom% 153 Eu, as well as metal samples enriched to 98.77 atom% 153 Eu were measured. The measured neutron capture resonance integral for 153 Eu is (9.9 ± 0.4)% larger than ENDF/B-VII.1. The capture resonance integral for 151 Eu is (7 ± 1)% larger than ENDF/B-VII.1. Another significant finding from these measurements was a significant increase in thermal total cross section for 151 Eu, up (9 ± 3)% from ENDF/B-VII.1

  3. Measurement of the 232Th capture cross section in the energy region 5 keV-150 keV

    International Nuclear Information System (INIS)

    Lobo, G.; Schillebeeckx, P.; Brusegan, A.; Borella, A.; Corvi, F.; Janeva, N.; Volev, K.

    2003-01-01

    The 232 Th(n,γ) neutron capture cross-section is of great importance for accelerator driven reactor (ADS) systems based on the Thorium-Uranium fuel cycle. An analysis of the required nuclear data, reveals that the status of the 232 Th capture data is far from the requested 2 % uncertainty level. Recently 232 Th average capture measurements, between 5-200 keV neutron energy, were performed at the FzK Karlsruhe (DE). A comparison of the measured averaged capture cross section with the evaluated data files shows a reasonable agreement in the neutron energy range above 15 keV. However, discrepancies of up to 40 % at lower neutron energies are observed. The same order of discrepancies is observed when comparing their results with the results obtained by Macklin et al. at ORELA. To clarify these discrepancies we measured at IRMM the average capture cross-section at the GEel LINear Accelerator (GELINA). The measurements were performed at a 14.37 m flight-path using the Time-Of-Flight (TOF) method. The gamma rays, originating from the 232 Th(n,γ) reaction, were detected by a pair of C 6 D 6 -based liquid scintillators applying a pulse-height weighting method. The neutron flux was measured with an ionisation chamber placed at 80 cm before the Thorium sample. This chamber has a cathode loaded with two back-to-back layers of about 40 μg/cm 2 10 B. The sample consisted of a metallic natural thorium disc of 8 cm diameter and 0.5 mm thick, corresponding to a thickness of 1.588 10 -3 at/b. The background for the capture measurements consists of a time independent and time dependent component. The former, mainly produced by the radioactive decay of the sample, was deduced from measurements with a closed beam. The latter was measured by replacing the thorium sample with a 0.5 mm thick 208 Pb sample of the same size. Such a Pb sample has practically the same scattering probability as the thorium sample and has a negligible capture yield. Therefore, the 208 Pb run provides a good

  4. CAB models for water: A new evaluation of the thermal neutron scattering laws for light and heavy water in ENDF-6 format

    International Nuclear Information System (INIS)

    Márquez Damián, J.I.; Granada, J.R.; Malaspina, D.C.

    2014-01-01

    Highlights: • We present a new evaluation of the thermal scattering laws for light and heavy water. • This evaluation is based on molecular and experimental data, with no free parameters. • Calculations with these libraries compare well with experimental values. • Libraries result in an improvement over existing ENDF scattering law files. - Abstract: In this work we present the CAB models for water: a set of new models for the evaluation of the thermal neutron scattering laws for light and heavy water in ENDF-6 format, using the LEAPR module of NJOY. These models are based on experimental structure data and frequency spectra computed from molecular dynamics simulations. The calculations show a significant improvement over ENDF/B-VI and ENDF/B-VII when compared with measurements of differential and integral scattering data

  5. Verification OFENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 nuclear data libraries for criticality calculations using NEA/NSC benchmarks

    International Nuclear Information System (INIS)

    Bouhaddane, A.; Farkas, G.; Hascik, J.; Slugen, V.

    2015-01-01

    The paper presents verification of selected nuclear data libraries with the aim to apply them to fast reactor calculations. More precise results were achieved for thermal neutrons calculations. This corresponds with the demand for more precise nuclear data for fast reactors. However, fast neutron calculations show some consistency, in particular between ENDF-B/VII.1 and JENDL-4.0 nuclear data libraries. The results support the idea to prefer using newer ENDF-B/VII.1 instead of the previous version ENDF-B/VII.0. Certainly, there are still some issues to be addressed and there is potential to gain more conclusive results. Although, application of ENDF-B/VII.1 and JENDL-4.0 is expected for further calculations. (authors)

  6. Thermal neutron capture cross sections of tellurium isotopes

    International Nuclear Information System (INIS)

    Tomandl, I.; Honzatko, J.; Egidy, T. von; Wirth, H.-F.; Belgya, T.; Lakatos, M.; Szentmiklosi, L.; Revay, Zs.; Molnar, G.L.; Firestone, R.B.; Bondarenko, V.

    2004-01-01

    New values for thermal neutron capture cross sections of the tellurium isotopes 122Te, 124Te, 125Te, 126Te, 128Te, and 130Te are reported. These values are based on a combination of newly determined partial g-ray cross sections obtained from experiments on targets contained natural Te and gamma intensities per capture of individual Te isotopes. Isomeric ratios for the thermal neutron capture on the even tellurium isotopes are also given

  7. Thermal neutron capture cross sections of tellurium isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Tomandl, I.; Honzatko, J.; von Egidy, T.; Wirth, H.-F.; Belgya, T.; Lakatos, M.; Szentmiklosi, L.; Revay, Zs.; Molnar, G.L.; Firestone, R.B.; Bondarenko, V.

    2004-03-01

    New values for thermal neutron capture cross sections of the tellurium isotopes 122Te, 124Te, 125Te, 126Te, 128Te, and 130Te are reported. These values are based on a combination of newly determined partial g-ray cross sections obtained from experiments on targets contained natural Te and gamma intensities per capture of individual Te isotopes. Isomeric ratios for the thermal neutron capture on the even tellurium isotopes are also given.

  8. Program SIGMA1 (version 79-1): Doppler broaden evaluated cross sections in the evaluated nuclear data file/version B (ENDF/B) format

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1979-01-01

    Program SIGMA1 Doppler-broadens evaluated cross sections in the ENDF/B format. The program requires that input cross sections be tabulated as linearly interpolable functions of energy in ENDF/B File 3; broadened cross sections, in this same form, replace the original values in the output tape. This report describes the methods used in the code and serves as a user's guide. A listing of the source deck is available on request

  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. Differential cross sections for gamma-ray production by 14 MeV neutrons with several elements in structural materials

    International Nuclear Information System (INIS)

    Murata, Isao; Yamamoto, Junji; Takahashi, Akito

    1988-01-01

    Energy differential cross sections for the gamma-rays produced from the (n,xγ) reactions by 14 MeV neutrons were measured in the gamma-ray energy range from 700 keV to 10 MeV using an NaI spectrometer. Results were obtained for the 8 natural elements; C, Al, Si, Cr, Fe, Ni, Cu and Mo. For prominent discrete gamma-rays in the differential cross sections, the production cross sections were determined by measuring angular distributions with a Ge detector. The gamma-ray energy covered the range between 500 and 3000 keV. The energy distributions have been compared with the differential cross sections evaluated in the nuclear data files of JENDL-3T, ENDL and ENDF/B-IV. The evaluations in JENDL-3T agreed fairly well with the measurements concerning the continuum energy spectra for secondary photons. Discrepancies appeared, however, for Si, Cr and Ni at the energies where the discrete gamma-rays were dominant. The ENDL evaluations were largely deviated from the experimental data. The production cross sections for the discrete gamma-rays in ENDL and ENDF/B-IV were available for the comparison with some of the measured cross sections. Results are presented for C, Al and Si. (author)

  11. Comparison of Hansen--Roach and ENDF/B-IV cross sections for 233U criticality calculations

    International Nuclear Information System (INIS)

    McNeany, S.R.; Jenkins, J.D.

    1976-01-01

    A comparison is made between criticality calculations performed using ENDF/B-IV cross sections and the 16-group Hansen-- Roach library at ORNL. The area investigated is homogeneous systems of highly enriched 233 U in simple geometries. Calculations are compared with experimental data for a wide range of H/ 233 U ratios. Results show that calculations of k/sub eff/ made with the Hansen--Roach cross sections agree within 1.5 percent for the experiments considered. Results using ENDF/B-IV cross sections were in good agreement for well-thermalized systems, but discrepancies up to 7 percent in k/sub eff/ were observed in fast and epithermal systems

  12. Description of the DLC-99/HUGO package of photon interaction data in ENDF/B-V format

    International Nuclear Information System (INIS)

    Roussin, R.W.; Knight, J.R.; Hubbell, J.H.; Howerton, R.J.

    1983-12-01

    A new photon interaction data library, DLC,-99/HUGO, is described. The library was prepared by incorporating newly evaluated data from the National Bureau of Standards with that from an existing data library, DLC-7F/HPICE, which is the ENDF/B-IV photon interaction data. It contains pair and triplet cross sections, photoelectric cross sections, and atomic form factors and the corresponding coherent scattering cross sections. Evaluated data in INDF/B-V format are provided for elements Z=1 to 100. The data package, available from the Radiation Shielding Information Center (RSIC) at Oak Ridge National Laboratory, will be submitted to CSEWG for consideration as the ENDF/B-V Photon Interaction Library. Two computer codes, EDPHOT for selectively printing the data and COMP23 for comparing two photon interaction libraries, are also provided

  13. Validation of DRAGON code in connection with WIMS-AECL/RFSP code system based on ENDF/B-VI library and two group model

    International Nuclear Information System (INIS)

    Hong, In Seob; Suk, Ho Chun; Kim, Soon Young; Jo, Chang Keun

    2002-06-01

    The major objective of this research is to validate the incremental cross section property of DRAGON code in connection with WIMS-AECL/DRAGON/RFSP code system with ENDF/B-VI library and full 2G calculation model. The direct comparison between the incremental cross section results calculated by DRAGON with ENDF/B-VI and ENDF/B-V and MULTICELL with ENDF/B-V indicate that there are not much differences between the incremental cross sections of DRAGON with ENDF/B-V and ENDF/B-VI, but there exists large discrepancies between the results of DRAGON and those of MULTICELL. In the analysis of the difference between calculated and measured reactivity worths of various types of control devices during Phase-B Post-Simulation of Wolsong Units 2, 3 and 4, WIMS-AECL/DRAGON/RFSP analysis well agrees with those of previous WIMS-AECL /MULTICELL/RFSP analysis within very small differences. From those results, we can conclude that DRAGON code can be used as a general purpose incremental cross section generation tool for not only the natural uranium fuel but also slightly enriched fuel such as RU or SEU, to cover the shortcomings of natural uranium based MULTICELL code

  14. ENDL-84. The Evaluated Nuclear Data Library of the Lawrence Livermore National Laboratory in the ENDF-5 format

    International Nuclear Information System (INIS)

    Cullen, D.E.; McLaughlin, P.K.; Lemmel, H.D.

    1990-09-01

    This document summarizes the contents of the evaluated nuclear data library (ENDL) by the Lawrence Livermore National Laboratory, USA, converted to ENDF-5 format. The library contains evaluated data for all significant neutron reactions in the energy range from 10 -4 eV to 20 MeV for 94 elements or isotopes. The entire library or selective retrievals from it can be obtained on magnetic tape, free of charge, from the IAEA Nuclear Data Section. (author)

  15. Intercomparison of delayed neutron summation calculations among JEF2.2, ENDF/B-VI and JNDC-V2

    Energy Technology Data Exchange (ETDEWEB)

    Sagisaka, Mitsuyuki [Nagoya Univ. (Japan); Oyamatsu, K.; Kukita, Y.

    1998-03-01

    We perform intercomparison of delayed neutron activities calculated with JEF2.2, ENDF/B-VI and JNDC-V2 with a simple new method. Significant differences are found at t < 20 (s) for major fissioning systems. The differences are found to stem from fission yields or decay data of several nuclides. The list of these nuclides are also given for the future experimental determination of these nuclear data. (author)

  16. BRIGITTE-KA, ENDF/B to KEDAK Data Conversion with Resonance Cross-Sections Tables Generator

    International Nuclear Information System (INIS)

    Stein, Eckhard; Schepers, J.C.; Vandeplas, P.

    1976-01-01

    1 - Nature of physical problem solved: The program translates evaluated nuclear data from the ENDF representation (3) into the KEDAK representation (5). Nearly all nuclear data desired by the user to be present on KEDAK will be produced. 2 - Method of solution: The retrieval and processing codes of ENDF (4) have been used, but some have been modified. Point-wise cross sections are calculated from resonance parameters. In the resolved resonance region all resonances are taken into account for each energy point. In order to guarantee linear interpolation with an error less than eps in the resolved resonance region, an energy mesh constructed by using the UNICORN code (6) is refined by adding points, if a cross section value calculated from the resonance parameters differs appreciably from the value calculated by interpolation. The various ENDF interpolation rules are reduced to the linear-linear rule used by KEDAK. Pointwise cross sections are calculated from the given parameters (e.g. the angular distributions). Some data of ENDF/B MF=5 (energy distributions of secondary neutrons) are also converted. 3 - Restrictions on the complexity of the problem: Because variable dimensioning is used for nearly all arrays, there are only few restrictions. These are the following: - One (natural) element may have up to 10 isotopes. - Five different L-states (L=0,1,2,3,4) are allowed in the resolved Breit-Wigner resonance parameter set. - Three different L-states and 5 different J-states for each L-state are allowed in the unresolved Breit-Wigner resonance parameter set. - One hundred points are allowed as primary energy grid for energy distributions of secondary neutrons

  17. Calculation of transmission and other functionals from evaluated data in ENDF format by means of personal computers

    International Nuclear Information System (INIS)

    Vertes, P.

    1991-04-01

    The FDMXPC program package was developed on the basis of the program system FEDMIX written for mainframe computers. The new program package for personal computers was developed for the interpretation of neutron transmission experiments and for producing group averaged infinite diluted and self-shielded cross sections, starting from evaluated data in ENDF format. The package was written for different FORTRAN compilers residing in personal computers under MS-DOS. (R.P.) 12 refs

  18. 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

  19. 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

  20. Partial radiative capture of resonance neutrons; Capture radiative partielle des neutrons de resonance

    Energy Technology Data Exchange (ETDEWEB)

    Samour, C. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1969-07-01

    The radiative capture of resonance neutrons has been studied near the Saclay linac between 0.5 and 700 eV with time-of-flight method and a Ge(Li) detector. {sup 195}Pt + n and {sup 183}W + n allow the study of the distribution of partial radiative widths and their eventual correlation and also the variation of < {gamma}{sub {gamma}{sub i}} > with E{sub {gamma}}. The mean values of Ml and El transition intensities are compared in several tin isotopes. Interference effects, either between resonances or between direct capture and resonant capture are found in {sup 195}Pt + n, {sup 197}Au + n and {sup 59}Co + n. The excited level schemes of a great deal of nuclei are obtained and compared with theoretical predictions. This study has been completed by an analysis of thermal spectrum. (author) [French] La capture radiative des neutrons de resonance a ete etudiee pres de l'accelerateur lineaire de Saclay entre 0,5 et 700 eV a l'aide de la methode du temps-de-vol et d'un detecteur Ge(Li). Les noyaux {sup 195}Pt + n et {sup 183}W + n permettent l'analyse de la distribution de resonance en resonance des largeurs radiatives partielles {gamma}{sub {gamma}{sub i}} et de leur eventuelle correlation, ainsi que l'etude de la variation de < {gamma}{sub {gamma}{sub i}} > en fonction de E{sub {gamma}}. Les intensites moyennes des transitions Ml et El sont comparees pour quelques isotopes de l'etain. Des effets d'interference, soit entre resonances, soit entre capture directe et capture resonnante sont mis en evidence dans {sup 195}Pt + n, {sup 197}Au + n et {sup 59}Co + n. Enfin les schemas des etats excites d'un grand nombre de noyaux sont obtenus et compares avec les predictions theoriques. Cette etude a ete completee par une analyse des spectres thermiques. (auteur)

  1. Partial radiative capture of resonance neutrons; Capture radiative partielle des neutrons de resonance

    Energy Technology Data Exchange (ETDEWEB)

    Samour, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1969-07-01

    The radiative capture of resonance neutrons has been studied near the Saclay linac between 0.5 and 700 eV with time-of-flight method and a Ge(Li) detector. {sup 195}Pt + n and {sup 183}W + n allow the study of the distribution of partial radiative widths and their eventual correlation and also the variation of < {gamma}{sub {gamma}{sub i}} > with E{sub {gamma}}. The mean values of Ml and El transition intensities are compared in several tin isotopes. Interference effects, either between resonances or between direct capture and resonant capture are found in {sup 195}Pt + n, {sup 197}Au + n and {sup 59}Co + n. The excited level schemes of a great deal of nuclei are obtained and compared with theoretical predictions. This study has been completed by an analysis of thermal spectrum. (author) [French] La capture radiative des neutrons de resonance a ete etudiee pres de l'accelerateur lineaire de Saclay entre 0,5 et 700 eV a l'aide de la methode du temps-de-vol et d'un detecteur Ge(Li). Les noyaux {sup 195}Pt + n et {sup 183}W + n permettent l'analyse de la distribution de resonance en resonance des largeurs radiatives partielles {gamma}{sub {gamma}{sub i}} et de leur eventuelle correlation, ainsi que l'etude de la variation de < {gamma}{sub {gamma}{sub i}} > en fonction de E{sub {gamma}}. Les intensites moyennes des transitions Ml et El sont comparees pour quelques isotopes de l'etain. Des effets d'interference, soit entre resonances, soit entre capture directe et capture resonnante sont mis en evidence dans {sup 195}Pt + n, {sup 197}Au + n et {sup 59}Co + n. Enfin les schemas des etats excites d'un grand nombre de noyaux sont obtenus et compares avec les predictions theoriques. Cette etude a ete completee par une analyse des spectres thermiques. (auteur)

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

    International Nuclear Information System (INIS)

    Aboustta, Mohamed Ali Bashir.

    1994-01-01

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

  3. Validation of Nuclear Criticality Safety Software and 27 energy group ENDF/B-IV cross sections

    International Nuclear Information System (INIS)

    Lee, B.L. Jr.

    1994-08-01

    The validation documented in this report is based on calculations that were executed during June through August 1992, and was completed in June 1993. The statistical analyses in Appendix C and Appendix D were completed in October 1993. This validation gives Portsmouth NCS personnel a basis for performing computerized KENO V.a calculations using the Martin Marietta Nuclear Criticality Safety Software. The first portion of the document outlines basic information in regard to validation of NCSS using ENDF/B-IV 27-group cross sections on the IBM 3090 at ORNL. A basic discussion of the NCSS system is provided, some discussion on the validation database and validation in general. Then follows a detailed description of the statistical analysis which was applied. The results of this validation indicate that the NCSS software may be used with confidence for criticality calculations at the Portsmouth Gaseous Diffusion Plant. When the validation results are treated as a single group, there is 95% confidence that 99.9% of future calculations of similar critical systems will have a calculated K eff > 0.9616. Based on this result the Portsmouth Nuclear Criticality Safety Department has adopted the calculational acceptance criteria that a k eff + 2σ ≤ 0.95 is safety subcritical. The validation of NCSS on the IBM 3090 at ORNL was extended to include NCSS on the IBM 3090 at K-25

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

    International Nuclear Information System (INIS)

    Alonso V, G.

    1991-01-01

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

  5. Validation of SCALE-4 criticality sequences using ENDF/B-V data

    International Nuclear Information System (INIS)

    Bowman, S.M.; Wright, R.Q.; DeHart, M.D.; Taniuchi, H.

    1993-01-01

    The SCALE code system developed at Oak Ridge National Laboratory contains criticality safety analysis sequences that include the KENO V.a Monte Carlo code for calculation of the effective multiplication factor. These sequences are widely used for criticality safety analyses performed both in the United States and abroad. The purpose of the current work is to validate the SCALE-4 criticality sequences with an ENDF/B-V cross-section library for future distribution with SCALE-4. The library used for this validation is a broad-group library (44 groups) collapsed from the 238-group SCALE library. Extensive data testing of both the 238-group and the 44-group libraries included 10 fast and 18 thermal CSEWG benchmarks and 5 other fast benchmarks. Both libraries contain approximately 300 nuclides and are, therefore, capable of modeling most systems, including those containing spent fuel or radioactive waste. The validation of the broad-group library used 93 critical experiments as benchmarks. The range of experiments included 60 light-water-reactor fuel rod lattices, 13 mixed-oxide fuel rod lattice, and 15 other low- and high-enriched uranium critical assemblies

  6. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-16

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

  8. Compared performances of ENDF/B-VI and JEF-2.2 for MOX core physics

    International Nuclear Information System (INIS)

    Finck, P. J.

    1998-01-01

    The US is currently evaluating the use of MOX fuel in commercial LWR's for reducing weapons grade Pu stockpiles. The design and licensing processes will require that the validity of the nuclear data libraries and codes used in the effort be demonstrated. Unfortunately, there are only a very limited number of relatively old and non representative integral experiments' freely available to the US programs. This lack of adequate experimental data can be partially remediated by comparing the results of well validated European codes with the results of candidate US codes. The demonstration can actually be divided in two components: a code to code (Monte Carlo) comparison can easily demonstrate the validity and limits of the proposed algorithms; and the performances of nuclear data libraries should be compared, major trends should be observed, and their origins should be explained in terms of differences in evaluated nuclear data; In this paper, we have compared the performances of the JEF-2.2 and ENDF/B-VI.4 libraries for a series of benchmarks for k eff , void worth, and pin power distributions. Note that JEF-2.2 has been extensively validated for MOX applications

  9. Control of radioactive wastes and coupling of neutron/gamma measurements: use of radiative capture for the correction of matrix effects that penalize the fissile mass measurement by active neutron interrogation

    International Nuclear Information System (INIS)

    Loche, F.

    2006-10-01

    In the framework of radioactive waste drums control, difficulties arise in the nondestructive measurement of fissile mass ( 235 U, 239 Pu..) by Active Neutron Interrogation (ANI), when dealing with matrices containing materials (Cl, H...) influencing the neutron flux. The idea is to use the neutron capture reaction (n,γ) to determine the matrix composition to adjust the ANI calibration coefficient value. This study, dealing with 118 litres, homogeneous drums of density less than 0,4 and composed of chlorinated and/or hydrogenated materials, leads to build abacus linking the γ ray peak areas to the ANI calibration coefficient. Validation assays of these abacus show a very good agreement between the corrected and true fissile masses for hydrogenated matrices (max. relative standard deviation: 23 %) and quite good for chlorinated and hydrogenated matrices (58 %). The developed correction method improves the measured values. It may be extended to 0,45 density, heterogeneous drums. (author)

  10. Radiative electron capture

    International Nuclear Information System (INIS)

    Biggerstaff, J.A.; Appleton, B.R.; Datz, S.; Moak, C.D.; Neelavathi, V.N.; Noggle, T.S.; Ritchie, R.H.; VerBeek, H.

    1975-01-01

    Some data are presented for radiative electron capture by fast moving ions. The radiative electron capture spectrum is shown for O 8+ in Ag, along with the energy dependence of the capture cross-section. A discrepancy between earlier data, theoretical prediction, and the present data is pointed out. (3 figs) (U.S.)

  11. Application of a gamma spectroscopy system to the measurement of neutron cross sections necessary to the development of nuclear energy; Mise au point d'un systeme de spectroscopie pour mesurer des sections efficaces neutroniques applicables a un possible developpement du nucleaire comme source d'energie

    Energy Technology Data Exchange (ETDEWEB)

    Deruelle, O

    2002-09-01

    This work concerns the development of nuclear energy and nuclear waste management in particular. Two parts of this study can be distinguished. In the first part (theoretical), a thorium-plutonium fuel based on MOX and dedicated for PWR was investigated in order to transmute plutonium in a potentially low waste fuel cycle. It was shown that this type of fuel is not regenerative but could be used for a transition to the industrial thorium fuel cycle without building new reactors. Thanks to moderated neutron spectra and high loaded actinide mass in the core, U-233 is quickly created ({approx}300 kg/y) for a loss of about {approx}1200 kg of fissile plutonium. In the second part (experimental), we have developed and built a new reaction chamber to measure neutron cross sections of actinides by alpha-gamma spectroscopy. This experimental device (in principle transportable) was commissioned in the high flux reactor of ILL Grenoble. Neutron flux was measured by gamma spectroscopy of irradiated Al and Co samples and was found to be of the order of 6,0. 10{sup 14} n.cm{sup -2}.s{sup -1} (4%). By the irradiation of 11{mu}g of Am-243 and Pu-242, corresponding capture cross sections were measured in the thermal neutron flux at 50 deg C. These are the results: {sup 243}Am(n,{gamma}) {sup 244fond.}Am = 4,72{+-}1,42b; {sup 243}Am(n,{gamma}) {sup 244total}Am = 74,8{+-}3,25b; {sup 242}Pu (n,{gamma}){sup 243}Pu = 22,7{+-}1,09b. Uncertainties of the measurements are mostly due to the determination of the neutron flux, efficiency of the electronics and ambiguities related to the definition of the area under {alpha}-{gamma} spectra. Although our measured cross sections deviate (by 10-30%) from the corresponding values widely used in evaluated data libraries such as ENDF, JEF and JENDL, in this work we have demonstrated the feasibility and principle of our experimental method. Furthermore, the value for the 243-americium capture cross-section is in very good agreement with the last two

  12. Application of a gamma spectroscopy system to the measurement of neutron cross sections necessary to the development of nuclear energy; Mise au point d'un systeme de spectroscopie pour mesurer des sections efficaces neutroniques applicables a un possible developpement du nucleaire comme source d'energie

    Energy Technology Data Exchange (ETDEWEB)

    Deruelle, O

    2002-09-01

    This work concerns the development of nuclear energy and nuclear waste management in particular. Two parts of this study can be distinguished. In the first part (theoretical), a thorium-plutonium fuel based on MOX and dedicated for PWR was investigated in order to transmute plutonium in a potentially low waste fuel cycle. It was shown that this type of fuel is not regenerative but could be used for a transition to the industrial thorium fuel cycle without building new reactors. Thanks to moderated neutron spectra and high loaded actinide mass in the core, U-233 is quickly created ({approx}300 kg/y) for a loss of about {approx}1200 kg of fissile plutonium. In the second part (experimental), we have developed and built a new reaction chamber to measure neutron cross sections of actinides by alpha-gamma spectroscopy. This experimental device (in principle transportable) was commissioned in the high flux reactor of ILL Grenoble. Neutron flux was measured by gamma spectroscopy of irradiated Al and Co samples and was found to be of the order of 6,0. 10{sup 14} n.cm{sup -2}.s{sup -1} (4%). By the irradiation of 11{mu}g of Am-243 and Pu-242, corresponding capture cross sections were measured in the thermal neutron flux at 50 deg C. These are the results: {sup 243}Am(n,{gamma}) {sup 244fond.}Am = 4,72{+-}1,42b; {sup 243}Am(n,{gamma}) {sup 244total}Am = 74,8{+-}3,25b; {sup 242}Pu (n,{gamma}){sup 243}Pu = 22,7{+-}1,09b. Uncertainties of the measurements are mostly due to the determination of the neutron flux, efficiency of the electronics and ambiguities related to the definition of the area under {alpha}-{gamma} spectra. Although our measured cross sections deviate (by 10-30%) from the corresponding values widely used in evaluated data libraries such as ENDF, JEF and JENDL, in this work we have demonstrated the feasibility and principle of our experimental method. Furthermore, the value for the 243-americium capture cross-section is in very good agreement with the last two

  13. Gamma astronomy

    International Nuclear Information System (INIS)

    Cesarsky, C.; Cesarsky, J.P.

    1986-01-01

    This article overviews the gamma astronomy research. Sources already observed, and what causes to give to them; the galactic radiation and its interpretation; techniques already used and current projects [fr

  14. Gamma Spectroscopy

    NARCIS (Netherlands)

    Niemantsverdriet, J.W.; Butz, Tilman; Ertl, G.; Knözinger, H.; Schüth, F.

    2008-01-01

    No abstract. The sections in this article are 1 Introduction 2 Mössbauer Spectroscopy 3 Time-Differential Perturbed Angular Correlations (TDPAC) 4 Conclusions and Outlook Keywords: Mössbauer spectroscopy; gamma spectroscopy; perturbed angular correlation; TDPAC

  15. Application of a gamma spectroscopy system to the measurement of neutron cross sections necessary to the development of nuclear energy

    International Nuclear Information System (INIS)

    Deruelle, O.

    2002-09-01

    This work concerns the development of nuclear energy and nuclear waste management in particular. Two parts of this study can be distinguished. In the first part (theoretical), a thorium-plutonium fuel based on MOX and dedicated for PWR was investigated in order to transmute plutonium in a potentially low waste fuel cycle. It was shown that this type of fuel is not regenerative but could be used for a transition to the industrial thorium fuel cycle without building new reactors. Thanks to moderated neutron spectra and high loaded actinide mass in the core, U-233 is quickly created (∼300 kg/y) for a loss of about ∼1200 kg of fissile plutonium. In the second part (experimental), we have developed and built a new reaction chamber to measure neutron cross sections of actinides by alpha-gamma spectroscopy. This experimental device (in principle transportable) was commissioned in the high flux reactor of ILL Grenoble. Neutron flux was measured by gamma spectroscopy of irradiated Al and Co samples and was found to be of the order of 6,0. 10 14 n.cm -2 .s -1 (4%). By the irradiation of 11μg of Am-243 and Pu-242, corresponding capture cross sections were measured in the thermal neutron flux at 50 deg C. These are the results: 243 Am(n,γ) 244fond. Am = 4,72±1,42b; 243 Am(n,γ) 244total Am = 74,8±3,25b; 242 Pu (n,γ) 243 Pu = 22,7±1,09b. Uncertainties of the measurements are mostly due to the determination of the neutron flux, efficiency of the electronics and ambiguities related to the definition of the area under α-γ spectra. Although our measured cross sections deviate (by 10-30%) from the corresponding values widely used in evaluated data libraries such as ENDF, JEF and JENDL, in this work we have demonstrated the feasibility and principle of our experimental method. Furthermore, the value for the 243-americium capture cross-section is in very good agreement with the last two measurements done in 1975 and 1997. These facts allowed us to think of new experiments

  16. Capture and fission with DANCE and NEUANCE

    Energy Technology Data Exchange (ETDEWEB)

    Jandel, M.; Baramsai, B.; Bond, E.; Rusev, G.; Walker, C.; Bredeweg, T.A.; Chadwick, M.B.; Couture, A.; Fowler, M.M.; Hayes, A.; Kawano, T.; Mosby, S.; Stetcu, I.; Taddeucci, T.N.; Talou, P.; Ullmann, J.L.; Vieira, D.J.; Wilhelmy, J.B. [Los Alamos National Laboratory, Los Alamos, New Mexico (United States)

    2015-12-15

    A summary of the current and future experimental program at DANCE is presented. Measurements of neutron capture cross sections are planned for many actinide isotopes with the goal to reduce the present uncertainties in nuclear data libraries. Detailed studies of capture gamma rays in the neutron resonance region will be performed in order to derive correlated data on the de-excitation of the compound nucleus. New approaches on how to remove the DANCE detector response from experimental data and retain the correlations between the cascade gamma rays are presented. Studies on {sup 235}U are focused on quantifying the population of short-lived isomeric states in {sup 236}U after neutron capture. For this purpose, a new neutron detector array NEUANCE is under construction. It will be installed in the central cavity of the DANCE array and enable the highly efficient tagging of fission and capture events. In addition, developments of fission fragment detectors are also underway to expand DANCE capabilities to measurements of fully correlated data on fission observables. (orig.)

  17. Stimulus-driven capture and contingent capture

    NARCIS (Netherlands)

    Theeuwes, J.; Olivers, C.N.L.; Belopolsky, A.V.

    2010-01-01

    Whether or not certain physical events can capture attention has been one of the most debated issues in the study of attention. This discussion is concerned with how goal-directed and stimulus-driven processes interact in perception and cognition. On one extreme of the spectrum is the idea that

  18. An evaluation of the Nb-93(n,n')Nb-93m dosimeter reaction for ENDF/B-VI

    International Nuclear Information System (INIS)

    Smith, D.L.; Geraldo, L.P.

    1990-01-01

    The Nb-93(n,n')Nb-93m reaction plays an important role in nuclear energy applications. Because of its low threshold energy and relatively long half-life, it is a desirable reaction for long-term neutron fluence dosimetry in nuclear fission reactors. An evaluation of the differential cross section for this reaction was completed in 1985 by this laboratory as part of a comprehensive effort involving all neutron cross sections for niobium. The objective was to provide input for ENDF/B-VI. It was difficult to produce a reliable evaluation for this reaction in 1985 because the information available then was sparse and quite uncertain. In fact, that evaluation was based entirely on nuclear model calculations. The evaluated cross sections below 0.7 MeV were derived from calculations carried out in this laboratory, while the higher energy values were obtained from the work of Strohmaier and co-workers. In 1985 there was only one published experimental differential cross section value to consider for this reaction. Even the half-life of Nb-93m was in serious doubt. During the five years between the completion of the earlier evaluation and the finalization of ENDF/B-VI there have been some significant improvements and additions to the experimental database for this reaction. Also, new model calculations have been performed. Therefore, it was considered worthwhile to produce a new evaluation of Nb-93(n,n')Nb-93m for ENDF/B-VI which would supplant the one that had been completed in 1985

  19. Gold standard capture cross section from 100 keV to 15 MeV

    International Nuclear Information System (INIS)

    Ryves, T.B.

    1982-01-01

    The capture cross section of gold is now generally accepted as the principal reference standard, and therefore in this review only gold is considered. Recent measurements of the gold capture cross section in the unresolved region are discussed and compared with the ENDF/B-V evaluation. It is concluded that in the energy interval 100 to 2000 keV the present uncertainty in the evaluation is +-8%, in the interval 2 to 3.5 MeV the uncertainty is +-4%, in ther interval 3.5 to 14 MeV more measurements are needed before a realistic error can be assigned, and from 14 to 15 MeV the uncertainty is +-10%. Several recommendations for future work have been made

  20. Gamma-ray emission spectra from spheres with 14 MeV neutron source

    International Nuclear Information System (INIS)

    Yamamoto, Junji; Kanaoka, Takeshi; Murata, Isao; Takahashi, Akito; Sumita, Kenji

    1989-01-01

    Energy spectra of neutron-induced gamma-rays emitted from spherical samples were measured using a 14 MeV neutron source. The samples in use were LiF, Teflon:(CF 2 ) n , Si, Cr, Mn, Co, Cu, Nb, Mo, W and Pb. A diameter of the sphere was either 40 or 60 cm. The gamma-ray energy in the emission spectra covered the range from 500 keV to 10 MeV. Measured spectra were compared with transport calculations using the nuclear data files of JENDL-3T and ENDF/B-IV. The agreements between the measurements and the JENDL-3T calculations were good in the emission spectra for the low energy gamma-rays from inelastic scattering. (author)

  1. ENDF/B-5 Standards Data Library (including modifications made in 1986). Summary of contents and documentation

    International Nuclear Information System (INIS)

    DayDay, N.; Lemmel, H.D.

    1986-01-01

    This document summarizes the contents and documentation of the ENDF/B-5 Standards Data Library (EN5-ST) released in September 1979. The library contains complete evaluations for all significant neutron reactions in the energy range 10 -5 eV to 20 MeV for H-1, He-3, Li-6, B-10, C-12, Au-197 and U-235 isotopes. In 1986 the files for C-12, Au-197 and U-235 were slightly modified. The entire library or selective retrievals from it can be obtained free of charge from the IAEA Nuclear Data Section. (author)

  2. ENDF/B VI iron validation onpca-replica (H2O/FE) shielding benchmark experiment

    Energy Technology Data Exchange (ETDEWEB)

    Pescarini, M. [ENEA, Bologna (Italy). Centro Ricerche Energia `E. Clementel` - Area Energia e Innovazione

    1994-05-01

    The PCA-REPLICA (H2O/Fe) neutron shielding benchmark experiment is analysed using the SN 2-D DOT 3.5 code and the 3-D-equivalent flux synthesis method. This engineering benchmark reproduces the ex-core radial geometry of a PWR, including a mild steel reactor pressure vessel (RPV) simulator, and is dsigned to test the accuracy of the calculation of the in-vessel neutron exposure parameters (fast fluence and iron displacement rates). This accuracy is strongly dependent on the quality of the iron neutron cross section used to describe the nuclear reactions within the RPV simulator. In particular, in this report, the cross sections based on the ENDF/B VI iron data files are tested, through a comparison of the calculated integral and spectral results with the corresponding experimental data. In addition, the present results are compared, on the same benchmark experiment, with those of a preceding ENEA (Italian Agency for Energy, New Technologies and Environment)-Bologna validation of the JEF-2.1 iron cross sections. The integral result comparison indicates that, for all the thresold detectors considered (Rh-103 (n,n) Rh-103m, In-115 (n,n) In-115 (n,n) In-115m and S-32 (n.p) P-32), the ENDF/B VI iron data produce better results than the JEF-2.1 iron data. In particular, in the ENDF/B VI calcultaions, an improvement of the in-vessel C/E (Calculated/Experimental) activity ratios for the lower energy threshold detectors, Rh-103 and In-115, is observed. This improvement becomes more evident with increasing neutron penetration depth in the vessel. This is probably attributable to the fact that the inelastic scattering cross section values of the ENDF/B VI Fe-56 data file, approximately in the 0.86 - 1.5 MeV energy range, are lower then the corresponding values of the JEF-2.1 data file.

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

    International Nuclear Information System (INIS)

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

    1981-09-01

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

  4. ENDF/B-5 Standards Data Library (including modifications made in 1986). Summary of contents and documentation

    Energy Technology Data Exchange (ETDEWEB)

    DayDay, N; Lemmel, H D

    1986-05-01

    This document summarizes the contents and documentation of the ENDF/B-5 Standards Data Library (EN5-ST) released in September 1979. The library contains complete evaluations for all significant neutron reactions in the energy range 10{sup -5}eV to 20 MeV for H-1, He-3, Li-6, B-10, C-12, Au-197 and U-235 isotopes. In 1986 the files for C-12, Au-197 and U-235 were slightly modified. The entire library or selective retrievals from it can be obtained free of charge from the IAEA Nuclear Data Section. (author) Refs, figs, tabs

  5. Capture reactions on C-14 in nonstandard big bang nucleosynthesis

    Science.gov (United States)

    Wiescher, Michael; Gorres, Joachim; Thielemann, Friedrich-Karl

    1990-01-01

    Nonstandard big bang nucleosynthesis leads to the production of C-14. The further reaction path depends on the depletion of C-14 by either photon, alpha, or neutron capture reactions. The nucleus C-14 is of particular importance in these scenarios because it forms a bottleneck for the production of heavier nuclei A greater than 14. The reaction rates of all three capture reactions at big bang conditions are discussed, and it is shown that the resulting reaction path, leading to the production of heavier elements, is dominated by the (p, gamma) and (n, gamma) rates, contrary to earlier suggestions.

  6. Capture ready study

    Energy Technology Data Exchange (ETDEWEB)

    Minchener, A.

    2007-07-15

    There are a large number of ways in which the capture of carbon as carbon dioxide (CO{sub 2}) can be integrated into fossil fuel power stations, most being applicable for both gas and coal feedstocks. To add to the choice of technology is the question of whether an existing plant should be retrofitted for capture, or whether it is more attractive to build totally new. This miscellany of choices adds considerably to the commercial risk of investing in a large power station. An intermediate stage between the non-capture and full capture state would be advantageous in helping to determine the best way forward and hence reduce those risks. In recent years the term 'carbon capture ready' or 'capture ready' has been coined to describe such an intermediate stage plant and is now widely used. However a detailed and all-encompassing definition of this term has never been published. All fossil fuel consuming plant produce a carbon dioxide gas byproduct. There is a possibility of scrubbing it with an appropriate CO{sub 2} solvent. Hence it could be said that all fossil fuel plant is in a condition for removal of its CO{sub 2} effluent and therefore already in a 'capture ready' state. Evidently, the practical reality of solvent scrubbing could cost more than the rewards offered by such as the ETS (European Trading Scheme). In which case, it can be said that although the possibility exists of capturing CO{sub 2}, it is not a commercially viable option and therefore the plant could not be described as ready for CO{sub 2} capture. The boundary between a capture ready and a non-capture ready condition using this definition cannot be determined in an objective and therefore universally acceptable way and criteria must be found which are less onerous and less potentially contentious to assess. 16 refs., 2 annexes.

  7. Use of integral experiments for the assessment of the 235U capture cross section within the CIELO Project

    Directory of Open Access Journals (Sweden)

    Ichou Raphaelle

    2016-01-01

    Full Text Available A new 235U capture cross-section evaluation, evaluated by ORNL and the CEA Bruyères-le-Châtel (BRC has been proposed within the CIELO project. IRSN, who participates in the CIELO project, contributes with data testing and has carried out benchmark calculations using few benchmarks, extracted from the ICSBEP database, for testing the new 235U evaluation. The benchmarks have been selected by privileging the experiments showing small experimental uncertainties and a significant sensitivity to 235U capture cross-section. The keff calculations were performed with both the MCNP 6 code and the 5.C.1 release of the MORET 5 code, using the ENDF/B-VII.1 library for all isotopes except 235U, for which both the ENDF/B-VII.1 and the new 235U evaluation was used. The benchmark selection allowed highlighting a significant effect on keff of the new 235U capture cross-section. The results of this data testing, provided as input for the evaluators, are presented here.

  8. 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.

  9. 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.

  10. Carbon Capture and Storage

    NARCIS (Netherlands)

    Benson, S.M.; Bennaceur, K.; Cook, P.; Davison, J.; Coninck, H. de; Farhat, K.; Ramirez, C.A.; Simbeck, D.; Surles, T.; Verma, P.; Wright, I.

    2012-01-01

    Emissions of carbon dioxide, the most important long-lived anthropogenic greenhouse gas, can be reduced by Carbon Capture and Storage (CCS). CCS involves the integration of four elements: CO 2 capture, compression of the CO2 from a gas to a liquid or a denser gas, transportation of pressurized CO 2

  11. CAPTURED India Country Evaluation

    NARCIS (Netherlands)

    O'Donoghue, R.; Brouwers, J.H.A.M.

    2012-01-01

    This report provides the findings of the India Country Evaluation and is produced as part of the overall CAPTURED End Evaluation. After five years of support by the CAPTURED project the End Evaluation has assessed that results are commendable. I-AIM was able to design an approach in which health

  12. Interatomic Coulombic electron capture

    International Nuclear Information System (INIS)

    Gokhberg, K.; Cederbaum, L. S.

    2010-01-01

    In a previous publication [K. Gokhberg and L. S. Cederbaum, J. Phys. B 42, 231001 (2009)] we presented the interatomic Coulombic electron capture process--an efficient electron capture mechanism by atoms and ions in the presence of an environment. In the present work we derive and discuss the mechanism in detail. We demonstrate thereby that this mechanism belongs to a family of interatomic electron capture processes driven by electron correlation. In these processes the excess energy released in the capture event is transferred to the environment and used to ionize (or to excite) it. This family includes the processes where the capture is into the lowest or into an excited unoccupied orbital of an atom or ion and proceeds in step with the ionization (or excitation) of the environment, as well as the process where an intermediate autoionizing excited resonance state is formed in the capturing center which subsequently deexcites to a stable state transferring its excess energy to the environment. Detailed derivation of the asymptotic cross sections of these processes is presented. The derived expressions make clear that the environment assisted capture processes can be important for many systems. Illustrative examples are presented for a number of model systems for which the data needed to construct the various capture cross sections are available in the literature.

  13. Computational analysis of Bangladesh 3 MW TRIGA research reactor using MCNP4C, JENDL-3.3 and ENDF/B-Vl data libraries

    International Nuclear Information System (INIS)

    Huda, M.Q.

    2006-01-01

    The three-dimensional continuous energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the JENDL-3.3 and ENDF/BVI continuous energy cross-section data for MCNP4C was performed against some well-known benchmark lattices. For TRIGA analysis, data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for nat Zr, nat Mo, nat Cr, nat Fe, nat Ni, nat Si, and nat Mg) at 300 K evaluations were used. Full S(α, β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH and water molecule, and for graphite were used in both cases. The validation of the model was performed against the criticality and reactivity benchmark experiments of the TRIGA reactor. There is ∼20.0% decrease of thermal neutron flux occurs when the thermal library is removed during the calculation. Effect of erbium isotope that is present in the TRIGA fuel was also studied. In addition to the effective multiplication values, the well-known integral parameters: δ 28 , δ 25 , ρ 25 , and C * were calculated and compared for both JENDL3.3 and ENDF/B-VI libraries and were found to be in very good agreement. Results are also reported for most of the analyses performed by JENDL-3.2 and ENDF/B-V data libraries

  14. Analyses of iron and concrete shielding experiments at JAEA/TIARA with JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0

    International Nuclear Information System (INIS)

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

    2015-01-01

    IAEA released a new Fusion Evaluated Nuclear Data Library, FENDL-3.0, in 2012. FENDL-3.0 extends the neutron energy range from 20 MeV to greater than 60 MeV. Now there is increasing interest in nuclear data above 20 MeV. Thus we have analyzed the iron and concrete shielding experiments with the 40 and 65 MeV neutron sources at TIARA in Japan Atomic Energy Agency with the latest high-energy nuclear data libraries, JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0. The Monte Carlo code MCNP-5 and ACE files of JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0, which are supplied from JAEA, LANL and IAEA, respectively, were used for this analysis. The collimated neutron beam and test shields were modeled in the analysis. The measured source neutron data were adopted in the analysis. The followings are found out from the results; (1) Iron experiments: The calculation result with FENDL-3.0 agrees with the measured one best. That with JENDL/HE-2007 fairly agrees with the measured one. On the contrary that with ENDF/B-VII.1 drastically overestimates the measured one. It is confirmed that this overestimation is due to the smaller non-elastic scattering data of "5"6Fe in ENDF/B-VII.1. (2) Concrete experiments: The calculation result with ENDL/HE-2007 agrees with the measured one best, while those with FENDL-3.0 and ENDF/B-VII.1 drastically overestimate the measured one. It is confirmed that this overestimation is due to both the larger elastic and smaller non-elastic scattering data of "1"6O in FENDL-3.0 and ENDF/B-VII.1.

  15. Gamma camera

    International Nuclear Information System (INIS)

    Berninger, W.H.

    1975-01-01

    The light pulse output of a scintillator, on which incident collimated gamma rays impinge, is detected by an array of photoelectric tubes each having a convexly curved photocathode disposed in close proximity to the scintillator. Electronic circuitry connected to outputs of the phototubes develops the scintillation event position coordinate electrical signals with good linearity and with substantial independence of the spacing between the scintillator and photocathodes so that the phototubes can be positioned as close to the scintillator as is possible to obtain less distortion in the field of view and improved spatial resolution as compared to conventional planar photocathode gamma cameras

  16. Shape isomer excitation by mu-minus capture

    International Nuclear Information System (INIS)

    Kaplan, S.N.; Monard, J.A.; Nagamiya, S.

    1975-06-01

    In a search for back-decay gamma rays from the shape isomer in 238 U following mu-minus capture, no candidates have been found with yields greater than 2 percent of the muon stoppings. The intensities of the gamma rays are insufficient to permit definitive lifetime measurements of individual peaks; however, for 500-keV energy ranges of gamma ray pulses, lifetimes have been determined that give results consistent with recent electron lifetime measurements. (6 figures, 2 tables) (U.S.)

  17. Neutron Capture and Transmission Measurements and Resonance Parameter Analysis of Samarium

    International Nuclear Information System (INIS)

    Leinweber, G.; Burke, J.A.; Knox, H.D.; Drindak, N.J.; Mesh, D.W.; Haines, W.T.; Ballad, R.V.; Block, R.C.; Slovacek, R.E.; Werner, C.J.; Trbovich, M.J.; Barry, D.P.; Sato, T.

    2001-01-01

    The purpose of the present work is to accurately measure the neutron cross sections of samarium. The most significant isotope is 149 Sm, which has a large neutron absorption cross section at thermal energies and is a 235 U fission product with a 1% yield. Its cross sections are thus of concern to reactor neutronics. Neutron capture and transmission measurements were performed by the time-of-flight technique at the Rensselaer Polytechnic institute (RPI) LINAC facility using metallic and liquid Sm samples. The capture measurements were made at the 25 meter flight station with a multiplicity-type capture detector, and the transmission total cross-section measurements were performed at 15- and 25-meter flight stations with 6 Li glass scintillation detectors. Resonance parameters were determined by a combined analysis of six experiments (three capture and three transmission) using the multi-level R-matrix Bayesian code SAMMY version M2. The significant features of this work are as follows. Dilute samples of samarium nitrate in deuterated water (D 2 O) were prepared to measure the strong resonances at 0.1 and 8 eV without saturation. Disk-shaped spectroscopic quartz cells were obtained with parallel inner surfaces to provide a uniform thickness of solution. The diluent feature of the SAMMY program was used to analyze these data. The SAMMY program also includes multiple scattering corrections to capture yield data and resolution functions specific to the RPI facility. Resonance parameters for all stable isotopes of samarium were deduced for all resonances up to 30 eV. Thermal capture cross-section and capture resonance integral calculations were made using the resultant resonance parameters and were compared to results obtained using resonance parameters from ENDF/B-VI updated through release 3. Extending the definition of the capture resonance integral to include the strong 0.1 eV resonance in 149 Sm, present measurements agree within estimated uncertainties with En

  18. Measuring parity violation using the neutron capture reaction

    International Nuclear Information System (INIS)

    Frankle, C.M.; Bowman, J.D.; Seestrom, S.J.; Roberson, N.R.; Sharapov, E.I.

    1993-01-01

    Measuring parity violation using the total capture reaction has certain advantages over neutron transmission experiments. Very much less material is required for targets, a necessity when dealing with separated isotopes. The capture reaction is also quite sensitive to very weak resonances. These advantages indicated the need to construct a near 4π gamma ray detector for use at LANSCE. A design for such a detector has been completed. Issues influencing the design and the final design parameters will be discussed in detail

  19. Comparisons of the MCNP criticality benchmark suite with ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0

    International Nuclear Information System (INIS)

    Kim, Do Heon; Gil, Choong-Sup; Kim, Jung-Do; Chang, Jonghwa

    2003-01-01

    A comparative study has been performed with the latest evaluated nuclear data libraries ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0. The study has been conducted through the benchmark calculations for 91 criticality problems with the libraries processed for MCNP4C. The calculation results have been compared with those of the ENDF60 library. The self-shielding effects of the unresolved-resonance (UR) probability tables have also been estimated for each library. The χ 2 differences between the MCNP results and experimental data were calculated for the libraries. (author)

  20. 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)

  1. Optional carbon capture

    Energy Technology Data Exchange (ETDEWEB)

    Alderson, T.; Scott, S.; Griffiths, J. [Jacobs Engineering, London (United Kingdom)

    2007-07-01

    In the case of IGCC power plants, carbon capture can be carried out before combustion. The carbon monoxide in the syngas is catalytically shifted to carbon dioxide and then captured in a standard gas absorption system. However, the insertion of a shift converter into an existing IGCC plant with no shift would mean a near total rebuild of the gasification waste heat recovery, gas treatment system and HRSG, with only the gasifier and gas turbine retaining most of their original features. To reduce the extent, cost and time taken for the revamping, the original plant could incorporate the shift, and the plant would then be operated without capture to advantage, and converted to capture mode of operation when commercially appropriate. This paper examines this concept of placing a shift converter into an IGCC plant before capture is required, and operating the same plant first without and then later with CO{sub 2} capture in a European context. The advantages and disadvantages of this 'capture ready' option are discussed. 6 refs., 2 figs., 4 tabs.

  2. Gamma camera

    International Nuclear Information System (INIS)

    Tschunt, E.; Platz, W.; Baer, Ul; Heinz, L.

    1978-01-01

    A gamma camera has a plurality of exchangeable collimators, one of which is replaceably mounted in the ray inlet opening of the camera, while the others are placed on separate supports. Supports are swingably mounted upon a column one above the other

  3. Gamma camera

    International Nuclear Information System (INIS)

    Schlosser, P.A.; Steidley, J.W.

    1980-01-01

    The design of a collimation system for a gamma camera for use in nuclear medicine is described. When used with a 2-dimensional position sensitive radiation detector, the novel system can produce superior images than conventional cameras. The optimal thickness and positions of the collimators are derived mathematically. (U.K.)

  4. Study of the fluctuations of the partial and total radiative widths by neutron capture resonance method; Etude des fluctuations des largeurs radiatives partielles et totales par la capture des neutrons de resonance

    Energy Technology Data Exchange (ETDEWEB)

    Huynh, V D [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-06-01

    Radiative capture experiments by neutron time-of-flight methods have been made for following studies: distribution of partial radiative widths, effects of correlation between different radiative transitions, fluctuations of total radiative widths {gamma}{sub {gamma}} from resonance to resonance, variation of {gamma}{sub {gamma}} with number of mass and the search for the existence of potential capture. Also, some other experiments with the use of neutron capture gamma-rays spectra have been investigated. (author) [French] Par la capture des neutrons de resonance dont les energies sont selectionnees a l'aide de la technique du temps de vol, differents types d'experiences ont ete realisees concernant les etudes des distributions des largeurs radiatives partielles, des effets de correlation entre differentes voies de desexcitation, de la fluctuation des largeurs radiatives totales {gamma}{sub {gamma}} de resonance a resonance, de la variation de la quantite {gamma}{sub {gamma}} en fonction du nombre de masse et de la mise en evidence de l'existence du processus de capture potentielle. Quelques autres applications de l'emploi du spectre de rayons gamma ont egalement ete presentees. (auteur)

  5. Study of the fluctuations of the partial and total radiative widths by neutron capture resonance method; Etude des fluctuations des largeurs radiatives partielles et totales par la capture des neutrons de resonance

    Energy Technology Data Exchange (ETDEWEB)

    Huynh, V.D. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1965-06-01

    Radiative capture experiments by neutron time-of-flight methods have been made for following studies: distribution of partial radiative widths, effects of correlation between different radiative transitions, fluctuations of total radiative widths {gamma}{sub {gamma}} from resonance to resonance, variation of {gamma}{sub {gamma}} with number of mass and the search for the existence of potential capture. Also, some other experiments with the use of neutron capture gamma-rays spectra have been investigated. (author) [French] Par la capture des neutrons de resonance dont les energies sont selectionnees a l'aide de la technique du temps de vol, differents types d'experiences ont ete realisees concernant les etudes des distributions des largeurs radiatives partielles, des effets de correlation entre differentes voies de desexcitation, de la fluctuation des largeurs radiatives totales {gamma}{sub {gamma}} de resonance a resonance, de la variation de la quantite {gamma}{sub {gamma}} en fonction du nombre de masse et de la mise en evidence de l'existence du processus de capture potentielle. Quelques autres applications de l'emploi du spectre de rayons gamma ont egalement ete presentees. (auteur)

  6. Neutron capture studies of {sup 206}Pb at a cold neutron beam

    Energy Technology Data Exchange (ETDEWEB)

    Schillebeeckx, P.; Kopecky, S.; Quetel, C.R.; Tresl, I.; Wynants, R. [Institute for Reference Materials and Measurements, European Commission, Joint Research Centre, Geel (Belgium); Belgya, T.; Szentmiklosi, L. [Institute for Energy Security and Environmental Safety, Centre for Energy Research, Budapest (Hungary); Borella, A. [Institute for Reference Materials and Measurements, European Commission, Joint Research Centre, Geel (Belgium); SCK CEN, Mol (Belgium); Mengoni, A. [Nuclear Data Section, International Atomic Energy Agency (IAEA), Wagramerstrasse 5, PO Box 100, Vienna (Austria); Agenzia Nazionale per le Nuove Tecnologie, l' Energia e lo Sviluppo Economico Sostenibile (ENEA), Bologna (Italy)

    2013-11-15

    Gamma-ray transitions following neutron capture in {sup 206}Pb have been studied at the cold neutron beam facility of the Budapest Neutron Centre using a metallic sample enriched in {sup 206}Pb and a natural lead nitrate powder pellet. The measurements were performed using a coaxial HPGe detector with Compton suppression. The observed {gamma} -rays have been incorporated into a decay scheme for neutron capture in {sup 206}Pb. Partial capture cross sections for {sup 206}Pb(n, {gamma}) at thermal energy have been derived relative to the cross section for the 1884 keV transition after neutron capture in {sup 14}N. From the average crossing sum a total thermal neutron capture cross section of 29{sup +2}{sub -1} mb was derived for the {sup 206}Pb(n, {gamma}) reaction. The thermal neutron capture cross section for {sup 206}Pb has been compared with contributions due to both direct capture and distant unbound s-wave resonances. From the same measurements a thermal neutron-induced capture cross section of (649 {+-} 14) mb was determined for the {sup 207}Pb(n, {gamma}) reaction. (orig.)

  7. Gamma-Gompertz life expectancy at birth

    Directory of Open Access Journals (Sweden)

    Trifon I. Missov

    2013-02-01

    Full Text Available BACKGROUND The gamma-Gompertz multiplicative frailty model is the most common parametric modelapplied to human mortality data at adult and old ages. The resulting life expectancy hasbeen calculated so far only numerically. OBJECTIVE Properties of the gamma-Gompertz distribution have not been thoroughly studied. The focusof the paper is to shed light onto its first moment or, demographically speaking, characterizelife expectancy resulting from a gamma-Gompertz force of mortality. The paperprovides an exact formula for gamma-Gompertz life expectancy at birth and a simplerhigh-accuracy approximation that can be used in practice for computational convenience.In addition, the article compares actual (life-table to model-based (gamma-Gompertzlife expectancy to assess on aggregate how many years of life expectancy are not captured(or overestimated by the gamma-Gompertz mortality mechanism. COMMENTS A closed-form expression for gamma-Gomeprtz life expectancy at birth contains a special(the hypergeometric function. It aids assessing the impact of gamma-Gompertz parameterson life expectancy values. The paper shows that a high-accuracy approximation canbe constructed by assuming an integer value for the shape parameter of the gamma distribution.A historical comparison between model-based and actual life expectancy forSwedish females reveals a gap that is decreasing to around 2 years from 1950 onwards.Looking at remaining life expectancies at ages 30 and 50, we see this gap almost disappearing.

  8. Resonance capture reactions with a total energy detector

    International Nuclear Information System (INIS)

    Macklin, R.L.

    1978-01-01

    The determination of nuclear reaction rates is considered; the Moxon--Rae detector and pulse height weighting are reviewed. This method has been especially useful in measuring (n,γ) cross sections. Strength functions and level spacing can be derived from (n,γ) yields. The relevance of neutron capture data to astrophysical nucleosynthesis is pointed out. The total gamma energy detection method has been applied successfully to radiative neutron capture cross section measurements. A bibliography of most of the published papers reporting neutron capture cross sections measured by the pulse height weighting technique is included. 55 references

  9. A SHORT HISTORY CSISRS - AT THE CUTTING EDGE OF NUCLEAR DATA INFORMATION STORAGE AND RETRIEVAL SYSTEMS AND ITS RELATIONSHIP TO CINDA, EXFOR AND ENDF.

    Energy Technology Data Exchange (ETDEWEB)

    HOLDEN, N.E.

    2005-12-01

    A short history of CSISRS, pronounced ''scissors'' and standing for the Cross Section Information Storage and Retrieval System, is given. The relationship of CSISRS to CINDA, to the neutron nuclear data four-centers, to EXFOR and to ENDF, the evaluated neutron nuclear data file, is briefly explained.

  10. Program PLOTC4 (Version 86-1). Plot evaluated data from the ENDF/B format and/or experimental data which is in a computation format

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1986-09-01

    Experimental and evaluated nuclear reaction data are world-wide compiled in EXFOR format and ENDF format, respectively. The computer program PLOTC4 described in the present document plots data from both formats; EXFOR data must be converted first to a ''computation format''. The program is available costfree from the IAEA Nuclear Data Section, upon request. (author)

  11. 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

  12. Program RESENDD (version 84-07): a program for reconstruction of resonance cross sections from evaluated nuclear data in the ENDF/B format

    International Nuclear Information System (INIS)

    Nakagawa, Tsuneo

    1984-10-01

    RESENDD is a computer program to calculate resonance cross sections from evaluated resonance parameters in the ENDF/B format. This program was improved from RESEND by modifying the multi-level Breit-Wigner formula, adding a function of Doppler broadening, and so on. This report explains functions of RESENDD and describes input data. Some examples are also given. (author)

  13. Program EVALPLOT (Version 79-1): plot data in the Evaluated-Nuclear-Data File/Version B (ENDF/B) format

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1979-01-01

    Program EVALPLOT is designed to plot evaluated cross sections in the ENDF/B format. The program plots cross sections, angular distributions, energy distributions, and parameters (e.g., μ-bar, xi-bar, and ν-bar). 16 figures, 2 tables

  14. 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

  15. Gamma camera

    International Nuclear Information System (INIS)

    Reiss, K.H.; Kotschak, O.; Conrad, B.

    1976-01-01

    A gamma camera with a simplified setup as compared with the state of engineering is described permitting, apart from good localization, also energy discrimination. Behind the usual vacuum image amplifier a multiwire proportional chamber filled with trifluorine bromium methane is connected in series. Localizing of the signals is achieved by a delay line, energy determination by means of a pulse height discriminator. With the aid of drawings and circuit diagrams, the setup and mode of operation are explained. (ORU) [de

  16. Gamma irradiator

    International Nuclear Information System (INIS)

    Simonet, G.

    1986-09-01

    Fiability of devices set around reactors depends on material resistance under irradiation noticeably joints, insulators, which belongs to composition of technical, safety or physical incasurement devices. The irradiated fuel elements, during their desactivation in a pool, are an interesting gamma irradiation device to simulate damages created in a nuclear environment. The existing facility at Osiris allows to generate an homogeneous rate dose in an important volume. The control of the element distances to irradiation box allows to control this dose rate [fr

  17. Thermal capture cross section for 58Ni (n,γ)59 Ni reaction

    International Nuclear Information System (INIS)

    Carbonari, A.W.; Pecequilo, B.R.S.

    1989-01-01

    The 58 Ni total thermal capture cross section was determined by suming the partial cross sections calculated for the primary transitions of the reaction 58 Ni (n,γ) 59 Ni. The primary transitions energies and intensities were determined from the 58 Ni thermal neutrons prompt gamma capture gamma rays spectrum in the 3.7 to 9.3 MeV region. The obtained value for the total cross section was 4.52 + 0.10b. (author) [pt

  18. Gamma teletopography

    International Nuclear Information System (INIS)

    Simonet, G.

    1987-06-01

    The mapping of gamma sources radiation emission in a nuclear plant is an important safety point. A remote gamma ray mapping process was developed in SPS/CEA/SACLAY. It uses the ''pinhole camera'' principle, precursor of photography. It mainly consists of a radiation proof box, with a small orifice, containing sensitive emulsions at the opposite. A first conventional photographic type emulsion photographs the area. A second photographic emulsion shows up the gamma radiations. The superim position of the two shots gives immediate informations of the precise location of each source of radiation in the observed area. To make easier the presentation and to improve the accuracy of the results for radiation levels mapping, the obtained films are digitally processed. The processing assigns a colours scale to the various levels of observed radiations. Taking account physical data and standard parameters, it gets possible to estimate the dose rate. The device is portable. Its compactness and fully independent nature make it suitable for use anywhere. It can be adapted to a remote automatic handling system, robot... so as to avoid all operator exposure when the local dose rate is too high [fr

  19. 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

  20. Measurement of the 232Th neutron capture cross section in the region 5 keV-150 keV

    International Nuclear Information System (INIS)

    Lobo, Georges; Corvi, Franco; Schillebeeckx, Peter; Brusegan, Antonio; Mutti, Paolo; Janeva, Natalia

    2002-01-01

    The average capture cross-section of 232 Th has been measured at the 14.37 m flight path of GELINA, IRMM-Geel, in the energy range from 5 to 150 keV. The capture events were detected by two C 6 D 6 liquid scintillators and the neutron flux was measured with a 10 B-loaded ionisation chamber. The data, corrected with the pulse-height weighting technique, have been normalised to the well-isolated and nearly saturated 232 Th (n, γ) resonances at 21.8 eV and 23.5 eV. Below 15 keV neutron energy, we do not observe the discrepancies, up to 40%, with the evaluated ENDF/B-VI data as reported by Wisshak et al.. Between 5 and 80 keV our results are about 10% systematically above the ENDF/B-VI data and approach the evaluated data between 80 and 100 keV. (author)

  1. Current status of fast-neutron-capture calculations

    International Nuclear Information System (INIS)

    Gardner, D.G.

    1982-01-01

    This work is primarily concerned with the calculation of neutron capture cross sections and capture gamma-ray spectra, in the framework of the Hauser-Feshbach statistical model and for neutrons from the resonance region up to several MeV. An argument is made that, for applied purposes such as constructing evaluated cross-section libraries, nonstatistical capture mechanisms may be completely neglected at low energies and adequately approximated at high energies in a simple way. The use of gamma-ray strength functions to obtain radiation widths is emphasized. Using the reaction 89 Y + n as an example, the problems encountered in trying to construct a case that could be run equivalently on two different nuclear reaction codes are illustrated, and the effects produced by certain parameter variations are discussed

  2. Comparison of yield and decay data among JNDC2, ENDF/B-VI and JEF2.2

    Energy Technology Data Exchange (ETDEWEB)

    Oyamatsu, Kazuhiro; Sagisaka, Mitsuyuki; Miyazono, Toshimitsu [Nagoya Univ. (Japan)

    1997-03-01

    Fission yields and decay data for fission product summation calculations are compared among JNDC2 and ENDF/B-VI and JEF2.2. Special attention is paid to the summation calculation of the total delayed neutrons per fission because it requires the data of the most unstable nuclides among all fission products. The cumulative fission yields of delayed neutron precursors are found to be appreciably different among the libraries even though values of the independent fission yields and the total number of delayed neutrons are chosen to be in fair agreement with each other. This suggests that there still exist large uncertainties in delayed neutron emission probabilities (or decay chains) for the precursors far from the stability line. (author)

  3. Monte Carlo-based validation of the ENDF/MC2-II/SDX cell homogenization path

    International Nuclear Information System (INIS)

    Wade, D.C.

    1979-04-01

    The results are presented of a program of validation of the unit cell homogenization prescriptions and codes used for the analysis of Zero Power Reactor (ZPR) fast breeder reactor critical experiments. The ZPR drawer loading patterns comprise both plate type and pin-calandria type unit cells. A prescription is used to convert the three dimensional physical geometry of the drawer loadings into one dimensional calculational models. The ETOE-II/MC 2 -II/SDX code sequence is used to transform ENDF/B basic nuclear data into unit cell average broad group cross sections based on the 1D models. Cell average, broad group anisotropic diffusion coefficients are generated using the methods of Benoist or of Gelbard. The resulting broad (approx. 10 to 30) group parameters are used in multigroup diffusion and S/sub n/ transport calculations of full core XY or RZ models which employ smeared atom densities to represent the contents of the unit cells

  4. 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

  5. FAIR-DDX, Double Diffusion Cross-Sections Scattering Matrix Generated from ENDF/B-4 or JENDL-2

    International Nuclear Information System (INIS)

    Minami, Kazuyoshi; Yamano, Naoki

    2001-01-01

    1 - Description of program or function: FAIR-DDX produces double differential (energy and angle) cross sections (DDX) in the form of group-to-group scattering matrices using the evaluated nuclear data libraries JENDL-2 or ENDF/B-IV. The DDX form is useful for verification of the evaluated data, such as the inelastic scattering, through comparison with the experimental DDX values. 2 - Method of solution: DDX uses the file 4 data (angular distribution of secondary neutrons) and the energy and momentum conservation laws. For continuum region reactions, file 5 (energy spectrum of secondary neutrons) is used. To express the angular distribution of secondary neutrons in group-to-group scattering matrices FAIR-DDX adopts a direct angular representation method. 3 - Restrictions on the complexity of the problem: The maximum number of energy groups is 200

  6. US Spacesuit Knowledge Capture

    Science.gov (United States)

    Chullen, Cinda; Thomas, Ken; McMann, Joe; Dolan, Kristi; Bitterly, Rose; Lewis, Cathleen

    2011-01-01

    The ability to learn from both the mistakes and successes of the past is vital to assuring success in the future. Due to the close physical interaction between spacesuit systems and human beings as users, spacesuit technology and usage lends itself rather uniquely to the benefits realized from the skillful organization of historical information; its dissemination; the collection and identification of artifacts; and the education of those in the field. The National Aeronautics and Space Administration (NASA), other organizations and individuals have been performing United States (U.S.) Spacesuit Knowledge Capture since the beginning of space exploration. Avenues used to capture the knowledge have included publication of reports; conference presentations; specialized seminars; and classes usually given by veterans in the field. More recently the effort has been more concentrated and formalized whereby a new avenue of spacesuit knowledge capture has been added to the archives in which videotaping occurs engaging both current and retired specialists in the field presenting technical scope specifically for education and preservation of knowledge. With video archiving, all these avenues of learning can now be brought to life with the real experts presenting their wealth of knowledge on screen for future learners to enjoy. Scope and topics of U.S. spacesuit knowledge capture have included lessons learned in spacesuit technology, experience from the Gemini, Apollo, Skylab and Shuttle programs, hardware certification, design, development and other program components, spacesuit evolution and experience, failure analysis and resolution, and aspects of program management. Concurrently, U.S. spacesuit knowledge capture activities have progressed to a level where NASA, the National Air and Space Museum (NASM), Hamilton Sundstrand (HS) and the spacesuit community are now working together to provide a comprehensive closed-looped spacesuit knowledge capture system which includes

  7. Adiabatic capture and debunching

    International Nuclear Information System (INIS)

    Ng, K.Y.

    2012-01-01

    In the study of beam preparation for the g-2 experiment, adiabatic debunching and adiabatic capture are revisited. The voltage programs for these adiabbatic processes are derived and their properties discussed. Comparison is made with some other form of adiabatic capture program. The muon g-2 experiment at Fermilab calls for intense proton bunches for the creation of muons. A booster batch of 84 bunches is injected into the Recycler Ring, where it is debunched and captured into 4 intense bunches with the 2.5-MHz rf. The experiment requires short bunches with total width less than 100 ns. The transport line from the Recycler to the muon-production target has a low momentum aperture of ∼ ±22 MeV. Thus each of the 4 intense proton bunches required to have an emittance less than ∼ 3.46 eVs. The incoming booster bunches have total emittance ∼ 8.4 eVs, or each one with an emittance ∼ 0.1 eVs. However, there is always emittance increase when the 84 booster bunches are debunched. There will be even larger emittance increase during adiabatic capture into the buckets of the 2.5-MHz rf. In addition, the incoming booster bunches may have emittances larger than 0.1 eVs. In this article, we will concentrate on the analysis of the adiabatic capture process with the intention of preserving the beam emittance as much as possible. At this moment, beam preparation experiment is being performed at the Main Injector. Since the Main Injector and the Recycler Ring have roughly the same lattice properties, we are referring to adiabatic capture in the Main Injector instead in our discussions.

  8. Neutron capture cross section of ^243Am

    Science.gov (United States)

    Jandel, M.

    2009-10-01

    The Detector for Advanced Neutron Capture Experiments (DANCE) at Los Alamos National Laboratory (LANL) was used for neutron capture cross section measurement on ^243Am. The high granularity of DANCE (160 BaF2 detectors in a 4π geometry) enables the efficient detection of prompt gamma-rays following neutron capture. DANCE is located on the 20.26 m neutron flight path 14 (FP14) at the Manuel Lujan Jr. Neutron Scattering Center at the Los Alamos Neutron Science Center (LANSCE). The methods and techniques established in [1] were used for the determination of the ^243Am neutron capture cross section. The cross sections were obtained in the range of neutron energies from 0.02 eV to 400 keV. The resonance region was analyzed using SAMMY7 and resonance parameters were extracted. The results will be compared to existing evaluations and calculations. Work was performed under the auspices of the U.S. Department of Energy at Los Alamos National Laboratory by the Los Alamos National Security, LLC under Contract No. DE-AC52-06NA25396 and at Lawrence Livermore National Laboratory by the Lawrence Livermore National Security, LLC under Contract No. DE-AC52-07NA27344. [4pt] [1] M. Jandel et al., Phys. Rev. C78, 034609 (2008)

  9. Direct neutron capture and related mechanisms

    International Nuclear Information System (INIS)

    Lynn, J.E.; Raman, S.

    1990-01-01

    We consider the evidence for the role of direct and related mechanisms in neutron capture at low and medium energies. Firstly, we compare the experimental data on the thermal neutron cross sections for El transitions in light nuclei with careful estimates of direct capture. Over the full range of light nuclei with small cross sections direct capture is found to be the predominant mechanism, in some cases being remarkable accurate, but in a few showing evidence for collective effects. When resonance effects become substantial there is evidence for an important contribution from the closely related valence mechanism, but full agreement with the data in such cases appears to require the introduction of a more generalised valence model. The possibility of direct and valence mechanisms playing a role in M1 capture is studied, and it is concluded that in light nuclei at relatively low gamma ray energies, it does indeed play some role. In heavier nuclei it appears that the evidence, especially from the correlations between E1 and M1 transitions to the same final states, favours the hypothesis that the main transition strength is governed by the M1 giant resonance. 31 refs., 2 tabs

  10. Motion Capturing Emotions

    OpenAIRE

    Wood Karen; Cisneros Rosemary E.; Whatley Sarah

    2017-01-01

    The paper explores the activities conducted as part of WhoLoDancE: Whole Body Interaction Learning for Dance Education which is an EU-funded Horizon 2020 project. In particular, we discuss the motion capture sessions that took place at Motek, Amsterdam as well as the dancers’ experience of being captured and watching themselves or others as varying visual representations through the HoloLens. HoloLens is Microsoft’s first holographic computer that you wear as you would a pair of glasses. The ...

  11. Nuclear muon capture

    CERN Document Server

    Mukhopadhyay, N C

    1977-01-01

    Our present knowledge of the nuclear muon capture reactions is surveyed. Starting from the formation of the muonic atom, various phenomena, having a bearing on the nuclear capture, are reviewed. The nuclear reactions are then studied from two angles-to learn about the basic muon+nucleon weak interaction process, and to obtain new insights on the nuclear dynamics. Future experimental prospects with the newer generation muon 'factories' are critically examined. Possible modification of the muon+nucleon weak interaction in complex nuclei remains the most important open problem in this field. (380 refs).

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

    International Nuclear Information System (INIS)

    Pazirandeh, Ali; Tabesh, Alireza

    2002-01-01

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

  13. Beyond the ENDF format: A modern nuclear database structure. SG38 meeting, OECD Conference Centre, 9-11 May 2016

    International Nuclear Information System (INIS)

    Herman, Michal Wladyslaw; Cabellos De Francisco, Oscar; Beck, Bret; Trkov, Andrej; Conlin, Jeremy Lloyd; Mcnabb, Dennis P.; Malvagi, Fausto; Grudzevich, Oleg T.; Mattoon, Caleb; Wiarda, Dorothea; Brown, David; Chadwick, Mark; Roubtsov, Danila; Iwamoto, Osamu; Yokoyama, Kenji; White, Morgan C.; Coste-Delclaux, Mireille; Fiorito, Luca; Haeck, Wim; Dunn, Michael; Jouanne, Cedric

    2016-05-01

    WPEC subgroup 38 (SG38) was formed to develop a new structure for storing nuclear reaction data, that is meant to eventually replace ENDF-6 as the standard way to store and share evaluations. The work of SG38 covers the following tasks: Designing flexible, general-purpose data containers; Determining a logical and easy-to-understand top-level hierarchy for storing evaluated nuclear reaction data; Creating a particle database for storing particles, masses and level schemes; Specifying the infrastructure (plotting, processing, etc.) that must accompany the new structure; Developing an Application Programming Interface or API to allow other codes to access data stored in the new structure; Specifying what tests need to be implemented for quality assurance of the new structure and associated infrastructure; Ensuring documentation and governance of the structure and associated infrastructure. This document is the proceedings of the SG38 meeting which took place at the OECD Conference Centre, on 9-11 May 2016. It comprises all the available presentations (slides) given by the participants: - 1) Beyond the ENDF format: Working toward the first specifications (Dennis P. McNabb); - 2) Summary of LLNL/LANL/ORNL/BNL discussions on SG38 (Caleb M. Mattoon); - 3) Status of Top Level Hierarchy Requirements Document (D.A. Brown); - 4) GND: Storing multiple representations of a quantity using forms, components and styles (Bret Beck); - 5) Particle Database update (Caleb M. Mattoon); - 6) General Purpose Data Containers (Jeremy Lloyd Conlin); - 7) Functional data containers (Bret Beck); - 8) Top-level hierarchy specifications (D.A. Brown)

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  15. Gamma teletopography

    International Nuclear Information System (INIS)

    Simonet, G.

    1986-09-01

    To set the gamma activity cartography is an important element of safety in numerous cases: intervention in hot cell, search of a radioactive source, examination of radioactive waste circuit followed by a reprocessing definition of decontamination and decommissioning processes and for all other accidents. The device presented here is like a ''black box'' with an aperture and an emulsion photosensitive to the opposite; a classical film takes photography of the place; a X-ray type emulsion gives a spot more or less contrasted and extensive corresponding to each source. Images can be processed with a microprocessor [fr

  16. Testing FLUKA on neutron activation of Si and Ge at nuclear research reactor using gamma spectroscopy

    Science.gov (United States)

    Bazo, J.; Rojas, J. M.; Best, S.; Bruna, R.; Endress, E.; Mendoza, P.; Poma, V.; Gago, A. M.

    2018-03-01

    Samples of two characteristic semiconductor sensor materials, silicon and germanium, have been irradiated with neutrons produced at the RP-10 Nuclear Research Reactor at 4.5 MW. Their radionuclides photon spectra have been measured with high resolution gamma spectroscopy, quantifying four radioisotopes (28Al, 29Al for Si and 75Ge and 77Ge for Ge). We have compared the radionuclides production and their emission spectrum data with Monte Carlo simulation results from FLUKA. Thus we have tested FLUKA's low energy neutron library (ENDF/B-VIIR) and decay photon scoring with respect to the activation of these semiconductors. We conclude that FLUKA is capable of predicting relative photon peak amplitudes, with gamma intensities greater than 1%, of produced radionuclides with an average uncertainty of 13%. This work allows us to estimate the corresponding systematic error on neutron activation simulation studies of these sensor materials.

  17. Muon capture in deuterium

    Czech Academy of Sciences Publication Activity Database

    Ricci, P.; Truhlík, Emil; Mosconi, B.; Smejkal, J.

    2010-01-01

    Roč. 837, - (2010), s. 110-144 ISSN 0375-9474 Institutional research plan: CEZ:AV0Z10480505 Keywords : Negative muon capture * Deuteron * Potential models Subject RIV: BE - Theoretical Physics Impact factor: 1.986, year: 2010

  18. Capture Matrices Handbook

    Science.gov (United States)

    2014-04-01

    materials, the affinity ligand would need identification , as well as chemistries that graft the affinity ligand onto the surface of magnetic...ACTIVE CAPTURE MATRICES FOR THE DETECTION/ IDENTIFICATION OF PHARMACEUTICALS...6 As shown in Figure 2.3-1a, the spectra exhibit similar baselines and the spectral peaks lineup . Under these circumstances, the spectral

  19. Capacitance for carbon capture

    International Nuclear Information System (INIS)

    Landskron, Kai

    2018-01-01

    Metal recycling: A sustainable, capacitance-assisted carbon capture and sequestration method (Supercapacitive Swing Adsorption) can turn scrap metal and CO 2 into metal carbonates at an attractive energy cost. (copyright 2018 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

  20. Capacitance for carbon capture

    Energy Technology Data Exchange (ETDEWEB)

    Landskron, Kai [Department of Chemistry, Lehigh University, Bethlehem, PA (United States)

    2018-03-26

    Metal recycling: A sustainable, capacitance-assisted carbon capture and sequestration method (Supercapacitive Swing Adsorption) can turn scrap metal and CO{sub 2} into metal carbonates at an attractive energy cost. (copyright 2018 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)